The documents contained herein were retrieved from Google's cache on October 23, 2001. These are some of the files that were censored on the Canadian Nuclear Safety Commission's website. The entire website was pulled for a while, it is now back up but some files, including these, are no longer available.
In order to make them easier(?) for you to handle, I have merged them all into this html file. I have added an index at the top with anchor links to each document, each document's original url and rules to separate one document from the next. I removed the Google header from each page. Some of the formatting may be off, this is because the files were originally in PDF format and Google performs an automated transformation of the file into a basic html text file before storing them. Tables in the documents are corrupted, figures are missing. I also removed several thousand blank lines to help keep the file size down. Otherwise they are as found. They are ordered simply by document number.
I was unable to retrieve the following two documents:
- General Nuclear Safety and Control Regulations
http://www.nuclearsafety.gc.ca/pubs_catalogue/uploads/sor202.pdf
- Nuclear Security Regulations http://www.nuclearsafety.gc.ca/pubs_catalogue/uploads/sor209.pdf
they are listed on this page of the CNSC site: http://www.nuclearsafety.gc.ca/eng/licensees/proposed_regulations.htm.
C-091 Ascertaining and Recording Radiation Doses to Individuals
C-099 Reporting Requirements For Operating Nuclear Power Plants
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
C-200 Radiation Safety Training For Radioisotope, Medical Accelerator And Transportation Workers
C-204 Certification of Persons Working at Nuclear Power Plants
C-205 Access Control for Protected and Inner Areas of Nuclear Facilities
C-208 Transport Security for Category I, II and III Nuclear Material
C-210 Maintenance Programs for Nuclear Power Plants
C-218 (E) Preparing Codes Of Practice To Control Radiation Doses At Uranium Mines And Mills
C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
C-273 (E) Making, Reviewing And Receiving Orders Under The Nuclear Safety And Control Act
C-274 Preparing a Security Report for Licence Applications
C-276 Human Factors Engineering Program Plans
C-278 Guide to Human Factors Verification and Validation Plans
G-149 Computer Programs Used in Design and Safety Analyses of Nuclear Power Plants and Research Reactors
G-225 Emergency Planning at Class I Nuclear Facilities and Uranium Mines and Mills
G-228 Developing and Using Action Levels
P-119 Policy on Human Factors
DRAFT
REGULATORY
GUIDE
Ascertaining and
Recording Radiation
Doses to Individuals
C-091/Rev. 1
Issued for public comments by the
Canadian Nuclear Safety Commission
March 2001
DRAFT REGULATORY GUIDE
Ascertaining and Recording
Radiation Doses to Individuals
C-091 Rev. 1
Issued for public comments by the
Canadian Nuclear Safety Commission
March 2001
Regulatory Documents
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in regulatory programs.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY GUIDE
Ascertaining and Recording Radiation Doses to Individuals
C-091 Rev. 1
March 2001
About this document
This Draft Regulatory Guide describes approaches that may be used by CNSC licensees to
ascertain and record exposures and doses under the Nuclear Safety and Control (NSC) Act and its
regulations. The document discusses related requirements, including obligations on licensees to
use licensed dosimetry services and to make information on radiation doses available to workers.
Comments
The CNSC invites interested persons to assist in the further development of this draft regulatory
document by commenting in writing on the document's content and potential usefulness. Please
respond by May 31, 2001. Direct your comments to the postal or e-mail address below,
referencing file 1-8-8-91.
The CNSC will take the comments received on this draft into account when developing it further.
These comments will be subject to the provisions of the federal Access to Information Act.
Document availability
This document can be viewed on the CNSC Web site (www.nuclearsafety.gc.ca). To order a
printed copy of the document in English or French, please contact:
Operations Assistant
Corporate Documents Section
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 996-9505
Facsimile: (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
i
Ascertaining and Recording Radiation Doses to Individuals C-091 Rev. 1
Contents
About this document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.1 Regulatory framework . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.2 Licensing process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2.3 Legislative basis for this document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3 Ascertaining Exposures and Doses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.1 Methods to measure exposures and doses directly . . . . . . . . . . . . . . . . . . . . . . . 5
3.2 Methods to estimate exposures and doses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.3 Direct measurement versus the estimating of exposures and doses . . . . . . . . . . 6
4 Recording Radiation Doses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
5 Handling Radiation Dose Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
6 Requirement to Use a Licensed Dosimetry Service . . . . . . . . . . . . . . . . . . . . . . . . . . 8
ii
C-091 Rev. 1 Ascertaining and Recording Radiation Doses to Individuals
Purpose
This Regulatory Guide is intended to help applicants for Canadian Nuclear Safety Commission
(CNSC) licences and holders of CNSC licences to develop programs to ascertain and record
doses of radiation in accordance with section 27 of the Nuclear Safety and Control (NSC) Act ,
section 3 of the General Nuclear Safety and Control Regulations, and sections 5, 7 and 8 of the
Radiation Protection Regulations.
Scope
This document describes approaches that may be used by CNSC licensees to ascertain and record
exposures and doses under the NSC Act and regulations. It discusses related requirements,
including obligations on licensees to use licensed dosimetry services and to make information on
radiation doses available to workers.
1 Definitions
Within this document, the meanings of the terms "nuclear energy worker", "nuclear substance",
"nuclear facility", and "prescribed" are as defined in section 2 of the NSC Act. These definitions
are repeated below for the convenience of readers:
* "nuclear energy worker" means a person who is required, in the course of the person's
business or occupation in connection with a nuclear substance or nuclear facility, to
perform duties in such circumstances that there is a reasonable probability that the
person may receive a dose of radiation than is greater than the prescribed limit for the
general public.
* "nuclear substance" means
(a) deuterium, thorium, uranium or an element with an atomic number greater
than 92;
(b) a derivative or compound of deuterium, thorium, uranium or of an element
with an atomic number greater than 92;
(c) a radioactive nuclide;
(d) a substance that is prescribed as being capable of releasing nuclear energy
or as being required for the production or use of nuclear energy;
(e) a radioactive by-product of the development, production or use of nuclear
energy; and
(f) a radioactive substance or radioactive thing that was used for the
development or production, or in connection with the use, of nuclear
energy.
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Ascertaining and Recording Radiation Doses to Individuals C-091 Rev. 1
* "nuclear facility" means any of the following facilities, namely,
(a) a nuclear fission or fusion reactor or subcritical nuclear assembly,
(b) a particle accelerator,
(c) a uranium or thorium mine or mill,
(d) a plant for the processing, reprocessing or separation of an isotope of
uranium, thorium or plutonium,
(e) a plant for the manufacture of a product from uranium, thorium or
plutonium,
(f) a plant for the processing or use, in a quantity greater than 1015 Bq per
calendar year, of nuclear substances other than uranium, thorium or
plutonium,
(g) a facility for the disposal of a nuclear substance generated at another
nuclear facility,
(h) a vehicle that is equipped with a nuclear reactor, and
(i) any other facility that is prescribed for the development, production or use
of nuclear energy or the production, possession or use of a nuclear
substance, prescribed equipment or prescribed information, and includes,
where applicable, the land on which the facility is located, a building that
forms part of, or equipment used in conjunction with, the facility and any
system for the management, storage or disposal of a nuclear substance.
* "prescribed" means prescribed by regulation of the Commission.
2 Background
2.1 Regulatory framework
The Canadian Nuclear Safety Commission is the federal agency that regulates the use of
nuclear energy and materials to protect health, safety, security and the environment, and to
respect Canada's international commitments on the peaceful uses of nuclear energy.
The NSC Act requires persons or organizations to be licensed by the CNSC for carrying out
the activities referred to in section 26 of the Act, unless otherwise exempted. The associated
regulations stipulate prerequisites for CNSC licensing, and the obligations of licensees and
workers.
2
C-091 Rev. 1 Ascertaining and Recording Radiation Doses to Individuals
2.2 Licensing process
The CNSC typically applies a phased process to its licensing of nuclear facilities and
activities. For major facilities, this process begins with a consideration of the environmental
impacts of the proposed project, and proceeds progressively through site preparation,
construction, operation, decommissioning and abandonment phases.
The Nuclear Safety and Control Act and regulations require licence applicants to provide
certain information at each licensing stage. The type and level of detail of this information
will vary to accommodate the licensing stage and specific circumstances.
At all licensing stages, applications may incorporate (directly or by reference) new or
previously submitted information, in accordance with legislated requirements and the best
judgement of the applicant. An application that is submitted at one licensing stage can
become a building block for the next stage.
Upon receipt of an application that is complete, the CNSC reviews it to determine whether
the applicant is qualified to carry on the proposed activity, and has made adequate provision
for the protection of the environment, the health and safety of persons, and the maintenance
of national security and measures required to implement international obligations to which
Canada has agreed. If satisfied, the CNSC may issue, renew, amend or replace a licence that
contains relevant conditions. Typically, this licence will incorporate the applicant's
undertakings, and will contain other conditions that the CNSC considers necessary.
2.3 Legislative basis for this document
Section 27 of the NSC Act, section 3 of the General Nuclear Safety and Control
Regulations and sections 5, 7 and 8 of the Radiation Protection Regulations are relevant to
the ascertaining and recording of radiation exposures and doses and to an understanding of
this guide.
Section 27 of the Act states:
"Every licensee and every prescribed person shall
(a) keep the prescribed records, including a record of the dose of radiation received by
or committed to each person who performs duties in connection with any activity
that is authorized by this Act or who is present at a place where that activity is
carried on, retain those records for the prescribed time and disclose them under the
prescribed circumstances; and
(b) make the prescribed reports and file them in the prescribed manner, including a
report on
(i) any theft or loss of a nuclear substance, prescribed equipment or prescribed
information that is used in carrying on any activity that is authorized by this
Act, and
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Ascertaining and Recording Radiation Doses to Individuals C-091 Rev. 1
(ii) any contravention of this Act in relation to an activity that is authorized by
this Act and any measure that have been taken in respect of the
contravention."
Section 2 of the NSC Act defines the term "prescribed" to mean "prescribed by regulation of
the Commission". Accordingly, to understand the related obligations of licensees pursuant
to section 27 of the Act, refer to sections 5, 7, and 8 of the Radiation Protection
Regulations. These sections contain requirements that pertain directly or indirectly to the
ascertaining and recording of radiation exposures and doses. These requirements are:
* "5.(1) For the purpose of keeping a record of doses of radiation in accordance
with section 27 of the Act, every licensee shall ascertain and record the amount of
exposure to radon progeny of each person referred to in that section, as well as
the effective dose and equivalent dose received by and committed to that person.
(2) A licensee shall ascertain the amount of exposure to radon progeny and the
effective dose and equivalent dose (a) by direct measurement as a result of
monitoring; or (b) if the time and resources required for direct measurement as a
result of monitoring outweigh the usefulness of ascertaining the amount and doses
using that method, by estimating them".
* "7.(1)(d) Every licensee shall inform each nuclear energy worker, in writing, of
the worker's radiation dose levels".
* "8. Every licensee shall use a licensed dosimetry service to measure and monitor
the doses of radiation received by and committed to nuclear energy workers who
have a reasonable probability of receiving an effective dose greater than 5 mSv in
a one-year dosimetry period".
The Radiation Protection Regulations do not tell CNSC licensees how to meet the above
requirements, nor how they should make the corresponding determinations that are
necessary. In particular, the regulations do not describe:
* how licensees are to determine when the time and resources for monitoring
exposures and doses outweigh the usefulness of monitoring;
* how licensees are to "ascertain" exposures and doses by "direct measurement as a
result of monitoring" or by "estimating" them; nor
* how licensees are to determine when nuclear energy workers have "a reasonable
probability" of receiving an effective dose greater than 5 mSv in a one-year
dosimetry period.
4
C-091 Rev. 1 Ascertaining and Recording Radiation Doses to Individuals
Paragraph 3(1)(e) of the General Nuclear Safety and Control Regulations stipulates that an
application for a CNSC licence shall contain the proposed measures to ensure compliance
with the Radiation Protection Regulations. Accordingly, any application for a CNSC
licence must include a description of how the applicant proposes to meet the above
requirements to ascertain and record exposures and doses, including making or arriving at
any associated determinations. If these proposals are accepted by the CNSC and
incorporated into a corresponding licence, the licensee will be required to meet the resulting
obligations.
Thus, license applicants, whether renewing a CNSC license or applying for the licence for
the first time, should address the above requirements in their respective applications.
3 Ascertaining Exposures and Doses
3.1 Methods to measure exposures and doses directly
Radiation exposures and doses can be ascertained by direct measurement as a result of
monitoring. Direct measurements typically involve the use or application of personal
monitoring equipment and techniques. In each situation involving direct measurement as a
result of monitoring, the choice of the most appropriate equipment and techniques will
depend upon case-specific factors. Such factors include whether the source of the radiation
that is to be measured is external to the subject's body, or whether it could be incorporated
into the body (e.g., by inhalation or ingestion).
Personal monitoring devices that are carried in close proximity to the exterior of the body
(i.e. dosimeters) can be used to measure radiation exposures due to sources that remain
outside the human body. Or, in the case of radiation from a source within a person's body,
the person's exposure may be ascertained by direct measurements on the body (e.g., in vivo
measurements, thermoluminescent dosimeters), or by direct measurements on material
excreted, exhaled or otherwise sampled from the body (i.e., in vitro measurements).
Typically, the radiation doses measured directly by personal monitoring devices and
techniques are reasonably representative of the actual doses received from radon progeny
and other sources.
3.2 Methods to estimate exposures and doses
Exposures and doses may be estimated pursuant to section 5 of the Radiation Protection
Regulations by indirect methods that take into account non-personal monitoring results, and
work place and other data.
For example, if a person occupies an area that has a known concentration of airborne
radioactivity or a known radiation field for a known period of time, this knowledge can be
used, in conjunction with other information, to estimate the person's radiation exposure
over that occupancy. This approach is often used where an airborne radioactive substance is
the source of exposure. In such instances, the concentration in air of radon progeny or
5
Ascertaining and Recording Radiation Doses to Individuals C-091 Rev. 1
other radionuclides in an area might be measured by air sampling or another method, and
the time spent in the area by a person or persons recorded. The measured concentrations of
airborne radioactivity, the recorded period of occupancy, representative metabolic data,
and air-inhalation rates could then be used to estimate the exposures of the person or
persons to airborne radiation.
In some situations it may be possible to estimate exposures and doses by applying statistical
methods to representative exposure or dose data.
3.3 Direct measurement versus the estimating of exposures and doses
If an application under the NSC Act and regulations for a CNSC licence:
* demonstrates to the satisfaction of CNSC that the time and resources required to
ascertain nuclear energy worker exposures and doses by "direct measurement as a
result of monitoring" will outweigh the usefulness of the results thus obtained, and
* proposes an acceptable method of estimating those exposures and doses, the
CNSC will typically incorporate the proposed estimation approach in any licence
that it issues in response to the application. Upon incorporation of the proposed
approach, implementation will become a licence requirement.
Accordingly, proposals for estimating exposures and doses in lieu of direct measurement as
a result of monitoring should be well-founded, should take into account good practice and
available techniques, and should be adequately explained and substantiated. In deciding
whether to measure or estimate doses to persons involved in licensed activities, licensees
should take case-specific factors into account, including the numbers of workers involved,
the nature of their duties, the associated work processes, the activity of the associated
radiation sources, and the magnitude, distribution and range of the doses anticipated. These
judgements should involve input from experts, such as radiation safety officers or the
members of radiation safety committees. For situations that involve potential radiation
exposures from multiple sources or via different pathways, this input should include an
evaluation as to whether direct measurement as a result of monitoring is warranted for all or
any of the contributing components.
All proposed approaches to ascertaining doses (whether by "direct measurement as a result
of monitoring" or by "estimating") require any corresponding CNSC approvals that are
implicit in its licensing process pursuant to the NSC Act and regulations. The following
examples illustrate possible licensee responses to postulated exposure scenarios. These
scenarios and responses are examples only. They are not meant to constrain applicants for
CNSC licences from proposing other approaches for CNSC consideration within the
context of their preferred radiation protection programs.
6
C-091 Rev. 1 Ascertaining and Recording Radiation Doses to Individuals
Exposure Scenario Possible Response
(a) A reasonable probability that the effective (a) Ascertain dose by direct measurement as a
dose to a nuclear energy worker who is result of monitoring.
exposed to a single component of radiation
will exceed 5 mSv/a.
(b) A reasonable probability that the effective (b) Ascertain doses from all components that
dose to a nuclear energy worker will exceed 5 contribute more than 5mSv/a by direct
mSv/a, and that this dose will be made up of measurement as a result of monitoring;
multiple components, of which at least one estimate the respective doses for components
will contribute more than 5 mSv/a. that contribute more than 2 mSv/a and less
than 5 mSv/a.
(c) A reasonable probability that the effective (c) Ascertain the dose by estimation.
dose to a nuclear energy worker will be more
than 2 mSv/a and less than 5 mSv/a.
(d) A reasonable probability that the (d) Ascertain the equivalent dose by direct
equivalent dose to a nuclear energy worker measurement as a result of monitoring.
will exceed 1/10 of the limits for organs or
tissues in the table of section 14 of the
Radiation Protection Regulations.
(e) A reasonable probability that a worker will (e) Ascertain the dose by estimation.
not accumulate an effective dose of more than
2 mSv/a or more than 1/10 of the equivalent
dose limits for organs or tissues in the table of
section 14 of the Radiation Protection
Regulations.
4 Recording Radiation Doses
Under paragraph 27(a) the NSC Act , every licensee is required to keep any records prescribed by
the regulations under the Act, including a record of the dose received by or committed to each
person who performs duties in connection with any activity that is authorized by the Act or who is
present at a place where that activity is carried on.
Accordingly, CNSC licensees should keep the following dose-related records to satisfy regulatory
requirements, or to facilitate regulatory review:
* a record of the name and job category of each Nuclear Energy Worker, as defined in
section 2 of the NSC Act [Section 24 of the Radiation Protection Regulations];
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Ascertaining and Recording Radiation Doses to Individuals C-091 Rev. 1
* a record of the effective dose received by or committed to each person who performs
duties in connection with any activity that is authorized by the NSC Act or who is
present at a place where that activity is carried on [Paragraph 27(a) of NSC Act] ;
* a record of the time period over which the above dose was accumulated;
* a description of the dosimetric model that was used to obtain the dose from measured
data; and
* any other dosimetry record or information required by a condition of the licence, the
NSC Act, or the CNSC pursuant to paragraphs 3(1)(n) and 3(1)(m) of the General
Nuclear Safety and Control Regulations.
5 Handling Radiation Dose Records
Under paragraph 7(1)(d) of the Radiation Protection Regulations, every CNSC licensee is
required to inform each nuclear energy worker, in writing, of the worker's radiation dose levels.
Section 19 of the Radiation Protection Regulations requires licensed dosimetry services to file
assigned doses and related information for nuclear energy workers with the National Dose
Registry (NDR) maintained by Health Canada in a format that meets the registry's needs. The
information needed by the NDR is described in Appendix D of Regulatory Standard, S-106,
"Technical and Quality Assurance Requirements for Dosimetry Services in Canada".
In practice, a CNSC licensee may choose to voluntarily measure the radiation doses received by
its staff or workers, even when such monitoring is not required by law. For example, a licensee
might include in its monitoring program those staff who have little probability of being exposed to
radiation or radioactive materials, in order to confirm or demonstrate that these persons have not
received significant exposures. If voluntarily submitted to the NDR, the results of such non-
obligatory monitoring programs will add to the NDR data base on radiation exposures of
Canadian workers. Such data could subsequently prove useful to resolve related enquiries,
compensation claims or litigation.
6 Requirement to Use a Licensed Dosimetry Service
Under section 8 of the Radiation Protection Regulations, every CNSC licensee is required to use
a licensed dosimetry service to measure and monitor the doses of radiation received by and
committed to nuclear energy workers who have a reasonable probability of receiving an effective
dose greater than 5 mSv in a one-year dosimetry period.
Since the NSC Act and its regulations do not define what constitutes "reasonable probability", the
use of the term in section 8 creates a need for decisions by licence applicants and licensees as to
when the use of a licensed dosimetry service is required or not required. Accordingly, under the
Act and new regulations, licence applicants will be responsible for determining and proposing
8
C-091 Rev. 1 Ascertaining and Recording Radiation Doses to Individuals
what, if anything, constitutes evidence of "reasonable probability" for the purposes of their
planned operations and section 8 of the Radiation Protection Regulations.
In arriving at their determinations of reasonable probability, licensees should take all relevant
case-specific factors into account, including the numbers of workers involved, the nature of their
duties, the associated work processes, the inventory of radio nuclides, and the potential
magnitude, distribution and range of the resultant doses. These judgements should involve
professional input from experts, such as radiation safety officers or the members of radiation
safety committees.
9
DRAFT
REGULATORY
GUIDE
C-099 (REV. 1) (E)
REPORTING REQUIREMENTS
FOR OPERATING NUCLEAR
POWER PLANTS
Issued for public comments by the
Atomic Energy Control Board
September 1999
Atomic Energy Commission de contrôle
Control Board de l'énergie atomique
DRAFT REGULATORY GUIDE
Reporting Requirements
for
Operating Nuclear Power Plants
C-099 (Rev. 1)(E)
Issued for public comments by the
Atomic Energy Control Board
September 1999
AECB Regulatory Documents
The AECB operates within a legal framework that includes law and supporting regulatory
documents. Law includes such legally enforceable instruments as acts, regulations, licences and
directives. Regulatory documents such as policies, standards, guides, notices, procedures and
information documents support and provide further information on these legally enforceable
instruments. Together, law and regulatory documents form the legal framework for the regulatory
activities of the AECB.
The main classes of regulatory documents developed by the AECB are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the AECB in its regulatory program.
Regulatory Standard: a document that is suitable for use for compliance assessment and
describes rules, characteristics or practices which the AECB accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the AECB recommends for meeting the intent of regulatory requirements or
improving administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to
alert licensees and others about significant health, safety or compliance issues that should
be acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the AECB
follows to administer the regulatory requirements for which it is responsible.
Document types like regulatory policies, standards, guides, notices and procedures, do not create
legally enforceable requirements. They support regulatory requirements found in regulations,
licences and other legally enforceable requirements. However, if a regulatory document is suitable
for use for compliance assessment, the AECB may make it into a legally enforceable requirement
by incorporation in an AECB regulation or in a licence made pursuant to the AEC Act.
DRAFT REGULATORY STANDARD
Reporting Requirements
for
Operating Nuclear Power Plants
C-099(Rev. 1)(E)
September 1999
NOTICE
On March 20, 1997, Bill C-23, the Nuclear Safety and Control Act (NSC Act), received Royal Assent. New
regulations that are derived from this Act will become law and replace the existing regulations. Draft Regulatory
Standard C-099(Rev. 1) references the NSC Act and new regulations, which will come into force in 2000 on a date
to be fixed by order of the Governor in Council.
About this Document
Comments
This document C-099(Rev.1), Reporting Requirements for Operating Nuclear Power Plants, is
proposed to replace R-99, Reporting Requirements for Operating Nuclear Power Facilities, dated
January 1, 1995. In order for interested persons to determine this document's impact and value,
public comments are being solicited. At the end of the 60-day comment period, comments will be
studied to determine how best to improve the document. Unless otherwise requested, a copy of all
comments received will be placed in the Atomic Energy Control Board (AECB) Library in Ottawa.
Comments on this standard will be most helpful if received in writing by December 15, 1999.
Reference our file number 1-8-8-99, and direct enquiries and/or comments to the address below.
Document availability
The document can be viewed on the AECB internet website at www.aecb.ccea.gc.ca. A copy of
C-099(Rev.1) may be ordered in English or French using the contact information below.
Operations Assistant
Corporate Documents Section
Atomic Energy Control Board
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone (613) 996-9505
i
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
Facsimile (613) 995-5086
E-mail via Internet: reg@atomcon.gc.ca
ii
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
Contents
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2. Event Reports: Prompt, Detailed, Additional Reports. . . . . . . . . . . . . . . . . . . . . . . . 2
Reportable Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
Timing of Event Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
Content: prompt event report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
Content: detailed event report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
Content: Additional Event Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
3. Operation Quarterly Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
4. Security Quarterly Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
5. Safety Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
6. Environmental Monitoring Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
7. Research and Development Progress Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
8. Problem Reports: Analysis or Research and Development . . . . . . . . . . . . . . . . . . . 14
Reportable Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
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Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
9. Periodic Inspection Program Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
10. Reliability Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
11. Fissionable and Fertile Substances Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
12. Certified Personnel Notification Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
Reportable Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
Timing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
APPENDIX A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1
APPENDIX B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.1
* Appendices A and B to this standard have been in trial use at nuclear power plants during the past
year, and are attached for completeness only. No additional comment is expected from licensees.
Following the consultation period, the standard and the appendices will be edited as a complete
document.
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C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
Purpose
This Regulatory Standard presents the reports that are mandatory for operating nuclear power
plants, as a term or condition of the operating licence. Based on pertinent sections of the Nuclear
Safety and Control Act (NSC Act, also called "the Act") and the Class I Nuclear Facilities
Regulations, the information provided herein includes frequency of submission and content
requirements.
Scope
This standard applies to any operating nuclear power plant, where it is specifically made
applicable by the operating licence.
1. Introduction
1.1 Background
The Atomic Energy Control Board (AECB) is a federal regulatory
agency with a mandate to assure that nuclear facilities and nuclear
materials do not pose undue risk to health, safety, security and the
environment.
At present, the AECB operates under the authority of the Atomic
Energy Control Act and related regulations. However, these laws are to
be replaced by the new Nuclear Safety and Control Act and new
regulations in 1999. This legislation will come into force on a date to be
fixed by order of the Governor in Council.
Under the NSC Act, the AECB will become the Canadian Nuclear
Safety Commission (CNSC), with continuing responsibilities for
regulation of the nuclear industry. Accordingly, this Consultative
Document pertains to activities to be conducted under the authority of
the NSC Act and the proposed Class I Nuclear Facilities Regulations.
These regulations are currently under development.
This draft Regulatory Standard explains certain regulatory requirements
that the CNSC proposes to incorporate in the operating licences that it
issues - pursuant to the NSC Act and the Class I Nuclear Facilities
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Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
Regulations - to the operators of nuclear fission reactor installations
that have been constructed for purposes of generating electricity or
thermal energy on a commercial scale (hereinafter referred to as
"nuclear power plants").
The AECB will take into account the public's comments on this
Consultative Document when it finalizes the Standard for
implementation under the new legislation.
Note: The Glossary in the Appendices of this document defines many of the terms
used herein.
2. Event Reports: Prompt, Detailed, Additional
Reports
Every licensee who operates a nuclear power plant shall make the
following event reports, for any event described in 2.1 to 2.32, at the
time specified in 2.33 to 2.37, and containing the information specified
in this section:
! a Prompt Event Report,
! a written Detailed Event Report and,
! where the Detailed Event Report has been submitted
incomplete, an Additional Event Report.
All event reports shall be submitted to the Project Officer or to the
Director of the Power Reactor Division, or if neither individual can be
contacted within the allotted times, the reports shall be submitted to the
Duty Officer. Such reports are in addition to the requirements set out in
the General Nuclear Safety and Control Regulations with respect to
the Bankruptcy and Insolvency Act, the Companies' Creditors
Arrangement Act, the Winding-up Act and in the Radiation Protection
Regulations concerning the reporting of an occurrence resulting, or
likely to result, in a dose of ionizing radiation in excess of any dose
specified in the Radiation Protection Regulations.
REPORTABLE EVENTS
Non-compliance 2.1 Any contravention of, or failure to comply with, the Act;
regulations made pursuant to the Act; or an order of the
Commission, a designated officer or an inspector
2.2 Any contravention of, or failure to comply with, a licence
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C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
condition
2.3 Failure to comply with any document specifically referenced in
Health and safety the operating licence, such as the Operating Policies and
Principles
2.4 The death, serious illness or serious injury of any person at the
nuclear power plant as a result of the licensed activity
2.5 An event that has resulted, or is likely to result, in the exposure
of a person to ionizing radiation in excess of the applicable
radiation dose limits referred to in the Radiation Protection
Regulations
2.6 An event that could have caused a reportable dose of ionizing
radiation under the Radiation Protection Regulations, but did
not, due to fortuitous circumstances rather than to approved
procedures
Release 2.7 A release of a radioactive nuclear substance, or other hazardous
substance, into the environment in a quantity, or at a rate,
exceeding that which is authorized by the licence
2.8 An unmonitored emission, where the upper limit of the release
cannot be estimated or where the upper limit cannot be shown
to be below the limits authorized in the licensing documents
Process systems 2.9 A serious process failure
2.10 An initiating event
2.11 A potential serious process failure
2.12 An event requiring a reactor shutdown in accordance with a
condition of the licence or the Operating Policies and Principles
2.13 An event that results in an acute and unrecoverable loss of
heavy water greater than 100 kg
Safety systems 2.14 An actuation of one or both shutdown systems from any power
level, except:
(a) a reactor trip that occurs while the unit is in guaranteed
shutdown state and where the actuation is not indicative
of a failure or potential failure of the shutdown
guarantee, or
(b) a reactor trip that was part of a pre-planned sequence
2.15 An actuation of the emergency core cooling system or
subsystems, resulting from the initiating parameter exceeding
the set point or any spurious operation or failure of the final
controlled device separating the heat transport system from the
emergency core cooling system circuit
2.16 An actuation of the containment system or subsystems resulting
from the initiating parameter exceeding the set point
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Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
2.17 A degradation of a special safety system or a relevant
safety-related system that is hazardous to the radiological health
and safety of any person, or that prevents a special safety
system or a safety-related system from meeting its defined
specifications
Pressure boundary 2.18 A discovery, in a registered system, of a pressure boundary
degradation that exceeds a limit specified in the applicable
design analysis, design codes or standards, or inspection codes
or standards, and includes:
(a) a pressure boundary deformation, crack or rupture, or a
leakage in excess of a limit specified in the Operating
Policies and Principles;
(b) the occurrence of an abnormal loading transient that
exceeds a pressure boundary design condition, or a
Service Level B condition, for any nuclear component
designed in accordance with the rules of the American
Society of Mechanical Engineers (ASME) III subsection
NB;
(c) a change to the size, rating or material property of the
pressure boundary beyond that allowed for in the
design;
(d) a repair or modification that changes the strength of a
component of the pressure boundary, which did not
receive the prior authorization required by Canadian
Standards Association (CSA) Standard N285.0;
(e) a local or general reduction of the wall thickness beyond
that allowed in the design by the applicable pressure
vessel code, standard or Act under which the pressure
boundary was registered;
(f) a degradation of the overpressure protection equipment
for the pressure boundary, which contravenes a limit of
the overpressure protection report or any other licensing
document; or
(g) confirmation, resulting from analysis of a registered
pressure boundary degradation, that an applicable limit
has been exceeded
Reactor and turbine control 2.19 A reduction of the effectiveness of any of the following systems,
below the defined specifications (whether caused by failure,
equipment inadequacy, improper procedures or inappropriate
human action):
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C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
(a) reactor power control;
(b) the primary heat transport system pressure and
inventory control; or
(c) turbine protection
Safeguards 2.20 An interference with, or an interruption in, the operation of
safeguards equipment or the alteration, defacement or breakage
of a safeguards seal, other than in accordance with the
safeguards agreement, the Act, the regulations made under the
Act or the licence
2.21 Theft, loss or sabotage of safeguards equipment or samples
collected for the purpose of a safeguards inspection; damage to
such equipment or samples; or the illegal use, possession,
operation or removal of such equipment or samples
Security 2.22 Theft or loss of a nuclear substance, prescribed equipment or
prescribed information
2.23 An actual, attempted or threatenedact of sabotage
2.24 A deficiency of the procedures or malfunction of the security
system that results in a failure to comply with the Nuclear
Security Regulations or with the power reactor operating
licence
2.25 An actual or threatened work disruption by workers, or an
impending walkout, slowdown or strike that could affect the
safety or security of nuclear power plant operations or the
capability to maintain minimum staff complements, as described
in NSC Act s.49
Emergency 2.26 A situation or an event that requires the implementation of a
contingency plan defined in the licence
2.27 Declaration of an alert or emergency, within the nuclear power
plant, where personnel or resources are mobilized by the
licensee in response to an unexpected occurrence of a
radiological condition, chemical spill, fire, or potentially
explosive mixture of gases that creates an actual hazard to the
safe operation of the plant or to the safety of persons.
External event 2.28 The occurrence of an earthquake that exceeds, at the site, the
maximum free-field seismic instrumentation triggering level
specified by Standard CAN/CSA N289.5 or, where appropriate
instrumentation is not available, the occurrence of an
earthquake that is greater than magnitude 5 on the Richter scale
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Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
within 500 kilometres of the site
2.29 The occurrence of any unusual external conditions at the site,
including fire, flood, plane crash, gas explosion, gas release,
high winds, missile or ice conditions, which results in
operational transients
Testing and 2.30 Failure to perform a test required by a licence condition,
monitoring including any routine test of a safety-related system required in
the licensing documents, except in accordance with approved
procedures
2.31 Failure to monitor or control a release path of radioactive
material that is required to be continuously monitored and
controlled, except in accordance with approved procedures
Documentation 2.32 Discovery of a problem arising from nuclear power plant
deficiency operating experience, that reveals a hazard to the health and
safety of persons, national security or the environment, that is
different in nature, greater in probability, or greater in
magnitude than was previously represented to the CNSC in the
licensing documents, and includes the discovery of:
(a) a special safety system that does not meet its defined
specifications;
(b) a reactor that is operating in a state that was not
considered in the safety analysis;
(c) the occurrence of an event of a type that was not
considered in the safety analysis;
(d) an unexplained and unexpected reactor core behaviour;
(e) an event where two or more systems or components,
which were assumed in the safety analysis to be mutually
independent, are in fact interdependent;
(f) a mistake in a licensing document that, if relied upon or
acted upon, would increase the risk to health, nuclear
safety and/or the environment; or
(g) releases of radioactive nuclear substances and/or other
hazardous substances that are found to be greater than
those predicted in the safety analysis.
PROMPT EVENT
R TIMING OF EVENT REPORTS
EPORT:
reportable as soon 2.33 A licensee shall make a Prompt Event Report, as soon as
as possible possible after initiating the required response actions and
alerting the required provincial and/or municipal authorities or
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C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
station staff responsible for responding to an event, for the
following reportable events:
(a) an ongoing emergency;
(b) a security incident, where a hazard to the health and
safety of persons, national security or the environment
continues to exist; or
(c) the loss or theft of a nuclear substance.
PROMPT EVENT 2.34 A licensee shall make a Prompt Event Report, within 24 hours,
REPORT: for the following reportable events:
reportable within (a) an actual or potential dose of ionizing radiation, as
24 hours described in 2.5;
(b) release of radioactive nuclear substances in excess of the
limits, as described in 2.5 and 2.7; or
(c) any theft, loss or interference with the operation of
equipment installed by, or on behalf of, the International
Atomic Energy Agency (IAEA), or of samples taken by,
or on behalf of, the IAEA, as described in 2.20 and 2.21
PROMPT EVENT REPORT: 2.35 A licensee shall make a Prompt Event Report, by the first
reportable by the first busi- business day following an event, for all reportable events
ness day following the event described by 2.1 to 2.33 that are not covered by 2.34 and 2.35.
WRITTEN DETAILED EVENT 2.36 Following up a Prompt Event Report, a licensee shall make a
REPORT: reportable with- written Detailed Event Report for each reportable event
in 45 calendar days after described in 2.1 to 2.33, within 45 calendar days after the
occurrence of the event. occurrence of the event.
ADDITIONAL EVENT 2.37 A licensee shall make an Additional Event Report when a
REPORT: reportable as Detailed Event Report has been submitted incomplete due to
soon as the required any of the following reasons:
information becomes (a) unavailability of the relevant information,
available. (b) ongoing investigation, or
(c) the discovery of new information.
The Additional Event Report shall be made, in writing, as soon
as the required information becomes available.
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Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
CONTENT: PROMPT EVENT REPORT
PROMPT EVENT 2.38 The date and time of the event, if known, or a statement of the
REPORT: best estimate thereof
brief description 2.39 Identification of the plant and any reactor unit(s) affected
of event 2.40 Identification of any systems, components, functions or
personnel affected
2.41 A brief description of the event
CONTENT: DETAILED EVENT REPORT
Background information 2.42 The date and time of the event, if known, or a statement of the
best estimate of date and time of the event
2.43 Identification of the plant and any reactor unit(s) affected
2.44 Any systems, components, functions or personnel affected
2.45 Reference to the primary provisions of reportable events 2.1 to
2.32, the licence condition(s), or the regulations under which
the event is reportable
Event description and 2.46 A brief description of how the event occurred
evaluation 2.47 If applicable, the municipal, provincial and federal authorities
that were notified of the event
2.48 A detailed account of the event
2.49 A description of the condition of the event site and, if relevant,
the operating conditions of the power reactor unit(s) involved in
the event, including operational reactor power level(s) prior to
the event
2.50 A description of any actions taken in immediate response to the
event
2.51 A detailed account of the causes of the event and its
consequences, including, where relevant, any that have been
established by investigative process
2.52 For an event with human performance implications, the root-
cause analysis of the event
2.53 A statement of the safety significance of the event, including, if
the event is an actuation of either shutdown system, a statement
as to whether the event was or was not a serious process failure
2.54 Any resulting doses, or dose estimates, to the facility personnel
or to the public
2.55 Any evaluation of the degree of impairment of special safety
systems or of a safety-related system
2.56 A statement as to whether a review has been carried out and
account has been taken of similar related events
2.57 Any resulting impact on the environment
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C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
Corrective action 2.58 The corrective actions taken, or proposed to be taken, to
prevent a recurrence of the event or to correct the situation,
including, for an event that involves inappropriate human
actions, those actions that result from a root-cause analysis
2.59 Comments and/or recommendations of the plant management,
including comments on the appropriateness of the actions taken
by operating staff
Completeness of report 2.60 A statement addressing whether the Detailed Event Report is
complete, or whether an Additional Report will be made and, if
so, the Additional Report number that has been assigned
2.61 Signature of the designated representative of the licensee
CONTENT: ADDITIONAL EVENT REPORTS
Updated information 2.62 The required information that is missing from the Detailed
Event Report
2.63 Any necessary correction of information in the Detailed Event
Report
Completeness of report 2.64 A statement as to whether or not the Additional Report is
complete and all necessary follow-up actions have been taken
2.65 Signature of the designated representative of the licensee
3. Operation Quarterly Reports
Every licensee who operates a nuclear power plant shall make, for each
quarter of a calendar year, a written or electronic Operation Quarterly
Report. The reports shall be submitted to the Project Officer or to a
person authorized by the Commission to receive the reports.
TIMING
Quarterly submission 3.1 Each quarterly report shall be submitted within ninety days of
the end of the period covered by the report, except the fourth
quarterly report for the calendar year, which shall be submitted
by March 1 of the next calendar year.
CONTENT
Staffing update 3.2 A summary of changes in station personnel organization and
staffing, procedure, equipment or fuel design, which could
impact upon the Safety Report (5.0) or other licensing
documents
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Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
3.3 A list of the names of all persons occupying a certified position
and - for each person occupying a certified shift position -
any temporary assignment to another position, the position
filled during this assignment, and the assignment
commencement date and duration
Summary of reports 3.4 A list of the events requiring report under 2.1 to 2.30 of this
submitted document, which occurred during the reporting period,
including a brief description, with report titles and numbers
except for any security event referred to in 2.21 to 2.23
3.5 A list and/or brief description of the Additional Event Reports
described in 2.20 for events other than security-related events,
which were submitted during the quarter. Such list or
description shall contain titles and numbers of the Detailed
Event Reports that were submitted, and those intended for
submission
Systems important to 3.6 Identification of events that occurred when systems important
safety to safety were unable to meet their defined specifications. The
fourth quarterly report for each calendar year shall also include
the reliability performance indices for each system important to
safety (refer to details in Regulatory Guide G-98, 2.30)
Monitoring 3.7 The results of monitoring of routine radioactive effluent and
other hazardous substances specified in the licence, including,
for each month of the quarter, the total activity released and the
cooling water flow volume
3.8 The results of non-routine off-site monitoring that was triggered
as a result of any unplanned release of radioactive nuclear
substances or other hazardous substances
Surveys 3.9 A summary of the results of:
(a) all routine surveys of the radiation field,
(b) all routine surveys of the surface contamination,
(c) the concentration of airborne radioactive materials that
were taken in various locations within the plant
(d) any assessment to detect increases of radiation hazard
over time
Exposure 3.10 The dose, received by any person, that resulted from any event
described in paragraphs 2.5 and 2.6, the collective dose of all
workers and dose statistics for the various groups of workers.
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C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
Exercises 3.11 A summary of emergency exercises and drills that were carried
out, and a description of any change made to non-security-
related emergency procedures. Once per year, one quarterly
report shall also include the results of the annual review,
conducted by the licensee, of the off-site emergency procedures
and of the arrangements with off-site authorities involved in
these procedures;
Prescribed substances 3.12 The acquisition and transfer of prescribed substances, including
any revisions to the inventory to account for radioactive decay.
The fourth quarterly report for each calendar year shall also
include the inventory as of year end.
Fires 3.13 The number of fires that occurred at the plant, with an
evaluation of their safety significance.
Performance indicators 3.14 A report of the safety-related station Performance Indicator
data in accordance with the Specification Sheets in Appendix A
and Data Sheets in Appendix B
4. Security Quarterly Reports
Every licensee who operates a nuclear power plant shall make, for each
quarter of the calendar year, a written or electronic Security Quarterly
Report. Reports shall be submitted to the Project Officer or to a person
authorized by the Commission to receive the reports.
TIMING
4.1 The Security Quarterly Report shall be submitted within ninety
days of the end of the period covered by the report, except the
fourth Security Quarterly Report for the calendar year, which
shall be submitted by March 1 of the next calendar year.
CONTENT
Changes 4.2 A summary of changes in station security organization and
staffing, procedure or equipment that could impact the licensing
Security Report;
Events 4.3 A list and /or a brief description of the security events, including
report titles and numbers, of the reportable events specified
under 2.2 to 2.25 that occurred during the reporting period.
Update reports 4.4 A list and /or brief description of Additional Reports required
under 2.36, which were submitted during the quarter for
security-related events and of those that remain to be submitted
Other information for security-related events, with the report titles and numbers;
4.5 The number of escorted visitors in the protected area (visitor
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Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
days)
4.6 The total number of entries and exits of vehicles to and from the
protected area at each point of entry
4.7 A summary of security exercises and drills that were carried out
during the quarter
System performance 4.8 A history of failure and impairment of the security system
equipment and a summary of faults, combinations of faults, or
events that prevented the security systems from meeting their
defined specifications
5. Safety Reports
Every licensee who operates a nuclear power plant shall review the
Safety Report, and submit an update, in writing or electronically to the
Project Officer or to a person authorized by the Commission to receive
the report.
TIMING
5.1 The description of the nuclear power plant in the Safety Report
shall be submitted every three years from the last update.
5.2 Each analysis of the Safety Report for the nuclear power plant
shall be reviewed every three years and revised where necessary
to ensure its continued validity. Any revised analysis shall be
submitted to the Commission within three years from the last
update, unless otherwise approved in writing by the Commission
or by a person authorized by the Commission.
CONTENT
5.3 The Safety Report update shall include, where necessary, a revised
description of the changes made to the site, structures and systems
of the nuclear power plant.
5.4 The update of the analyses of the Safety Report shall take into
consideration necessary changes resulting from any event or
occurrence, that was reported pursuant to subsection 2.2 to 2.30,
which brings the validity of the results of the analyses into
question. In such cases, the analyses shall be repeated using
current methods and information, and the new results forwarded
to the Commission for incorporation into the Safety Report.
12
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
6. Environmental Monitoring Reports
Every licensee who operates a nuclear power plant shall make, in writing,
an annual report of the results of the environmental monitoring program.
The report shall be submitted to the Project Officer or to a person
authorized by the Commission to receive the report.
TIMING
6.1 The report shall be submitted by May 1 of the next calendar year.
CONTENT
6.2 An analysis of the results of the environmental monitoring
program
6.3 The individual doses that were calculated as doses to the critical
group
6.4 A description of the dosimetric models used
6.5 A review of the environmental monitoring quality assurance
program
6.6 Any unusual findings during the calendar year.
7. Research and Development Progress Reports
Every licensee who operates a nuclear power plant shall make a written
annual Research and Development (R&D) Progress Report. Reports
shall be submitted to the Project Officer or to a person authorized by
the Commission to receive the reports.
TIMING
7.1 The R&D Progress Report shall be submitted by July 1 of the
next calendar year.
CONTENT
7.2 The progress report shall describe research and development
programs that are planned or are being carried out during the
calendar year, or that are planned for future years, to resolve
safety issues. The report shall describe schedules, milestones
and results of the programs.
13
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
8. Problem Reports: Analysis or Research and
Development
Every licensee who operates a nuclear power plant shall make a written
Analysis or Research and Development (R&D) Problem Report for the
discovery of a safety problem, arising from research findings or
improved methods for safety analysis, that reveals a hazard to
radiological health or nuclear safety; or that is different in nature,
greater in probability, or greater in magnitude than was previously
represented to the CNSC in the licensing documents.
Reports shall be submitted to the Project Officer or to a person
authorized by the Commission to receive the reports.
REPORTABLE PROBLEMS
Invalid or inadequate 8.1 The discovery that assumptions, inputs, analytical methods or
information results of the safety analyses may be invalid
8.2 Information revealing that the limits in the Operating Policies
and Principles document, or in the appendices to the document,
are inadequate
8.3 Information revealing that the analyses from which the limits in
the Operating Policies and Principles document were derived
may be invalid or uncertain, such that the minimum margins of
safety are less than predicted
8.4 Information revealing that the defined specifications of a special
safety system or of a safety-related system are invalid
8.5 The discovery of a mistake in a licensing document that, if
relied upon or acted upon, would increase the risk to the
environment and to the health and safety of persons
8.6 Information revealing that the measures for protecting the
environment are inadequate
TIMING
8.7 For all problems referred to in 8.1 through 8.6 above, a licensee
shall submit to the CNSC a written Problem Report within 21
calendar days after the licensee has been notified of the problem
CONTENT
8.8 Where relevant, identification of the plant and reactor unit(s)
affected
8.9 Where relevant, identification of the systems, components or
14
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
functions affected
8.10 A statement of the safety significance of the problem
8.11 Where relevant, an evaluation of the degree of impairment of
any special safety system or safety-related system
8.12 Corrective actions taken or proposed to be taken to correct the
situation
8.13 Signature of the designated representative of the licensee
9. Periodic Inspection Program Reports
Following a scheduled maintenance outage, every licensee who
operates a nuclear power plant shall make a Periodic Inspection
Program Report, in writing, to the Project Officer or to a person
authorized by the Commission to receive the reports.
TIMING
9.1 The reports shall be submitted within 90 days of the completion
of a scheduled maintenance outage.
CONTENT
9.2 The Periodic Inspection Program Reports shall describe the
results of any inspection carried out in accordance with the
Periodic Inspection Program requirements of CSA Standards
N285.4 and N285.5
Note: Requirements to 10. Reliability Reports
report reliability on an Every licensee who operates a nuclear power plant shall submit an
annual basis do not annual Reliability Report in writing to the Project Officer or to a person
relieve the licensee of
obligations (1) to detect authorized by the Commission to receive the report. The Reliability
any unacceptable Report shall contain an evaluation, for the calendar year being reported,
decline in reliability, of the reliability of each system important to safety, as detailed in
and (2) to respond by Regulatory Guide G-98
taking appropriate
actions. TIMING
10.1 The annual Reliability Report shall be submitted by April 1 of
the year that follows the reporting period.
CONTENT
10.2 A list of incidents during which systems important to safety
were unable to meet their defined specifications
10.3 A listing and description of unsafe component faults of systems
15
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
important to safety
10.4 Reliability Performance Indices of systems important to safety
10.5 The predicted reliability of each system important to safety
using the latest failure data
10.6 A comparison between the predicted reliability, the reliability
performance indices and the reliability targets for each special
safety system
10.7 An impairment report for each system important to safety
10.8 A description of changes made to the system design, operating
or maintenance practices, which affected the predicted reliability
of the systems important to safety
10.9 A description of changes made to the reliability model of
systems important to safety
10.10 A listing of surveillance activities that have been credited in the
reliability assessments and identification of missed activities
10.11 The component failure rates used in the reliability assessments
and the site-specific failure data
10.12 The human interaction data used in the reliability assessments
10.13 The common cause failure data used in the reliability
assessments.
11. Fissionable and Fertile Substances Reports
Every licensee who operates a nuclear power plant shall make, to the
Commission, reports on the inventory and transfer of fissionable and
fertile substances, in accordance with AECB-1049, Reporting
Requirements for Fissionable and Fertile Substances.
12. Certified Personnel Notification Reports
REPORTABLE INFORMATION
Every licensee who operates a nuclear power plant shall make, in
writing to the Project Officer or to a person authorized by the
Commission to receive the reports, a Certified Personnel Notification
report, as stated below.
12.1 Upon becoming aware that a certified person meets any of the
conditions of the Commission Standard for the Certification of
Staff at Nuclear Power Plants, s.9 (Decertification)
12.2 Upon removing a person from a certified shift position for any
reason, unless a Certified Personnel Notification report has
16
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
previously been submitted in accordance with 12.1
TIMING
12.3 The Certified Personnel Notification report shall be submitted
within 7 calendar days of the licensee becoming aware that a
certified person meets any of the conditions of S.9
(Decertification), of the Standard for the Certification of Staff
at Nuclear Power Plants, or within 7 calendar days of removing
a person from the certified position, unless otherwise approved
in writing by the Commission or a person authorized by the
Commission.
CONTENT
12.4 The name and position of the person, and the condition that
would justify decertification or the reason for removal of the
person from the certified position
17
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
Glossary
Act Nuclear Safety and Control Act (NSC Act)
AECB Atomic Energy Control Board
ASME American Society of Mechanical
Engineering
certified certified by the Commission under paragraph
21(1)(I) of the Act or by a designated officer
authorized under paragraph 39(2)(b) of the Act
indicating that a person is certified
certified position a position, referred to in the nuclear plant
operating licence, requiring written permission
of the CNSC to allow an individual to perform
the duties and exercise the responsibilities of
that position
certified shift position the certified position of shift supervisor, reactor
control room operator and, where applicable,
unit 0 control room operator
CNSC Canadian Nuclear Safety Commission, also
called "the Commission," established by section
8 of the Nuclear Safety and Control Act
common cause failure a dependant failure in which two or more
component fault states exist simultaneously or
within a short time interval and are a direct result
of a shared cause
designated officer a person designated as a designated officer under
section 37 of the Nuclear Safety and Control Act
defined specifications the criteria set out in the licensing documents for
a special safety system or a safety-related system
that designate the minimum functional capability
and performance levels required for effectiveness
18
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
discovery of a problem the earliest time when the licensee uncovers a
situation revealing a health, safety, security or
environmental problem or decides that specific
resources should be allocated to ascertain
whether or not such a problem exists
hazardous substance or a substance or waste, other than a nuclear
hazardous waste substance, that is used to connection with or
produced in the course of carrying on a licensed
activity and that may pose a risk to the health
ans safety of persons or to the environment
IAEA International Atomic Energy Agency
impairment report a report of the impairment history of the system
for a given time, and includes, for each
impairment, its duration and an assessment of
the ability of the system to perform with respect
to the reliability measures in the licensing
documents
initiating event an event (such as loss of instrument air, loss of
class IV power, loss of main feedwater etc.)
which initiates a sequence that could lead to a
serious process failure in the absence of action
by any other system important to safety e.g.,
setback stepback, auxiliary feedwater etc.
needed to prevent a serious process failure
inspector a person designated as an inspector under
section 29 of the Nuclear Safety and Control
Act
licence a licence issued under section 24 of the
NSC Act
licence condition a requirement of a power reactor operating
licence (PROL)
licensed activity an activity described in paragraph 26(e) of the
Act that a licence authorizes the licensee to carry
on in relation to a Class I nuclear facility
19
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
licensee a person who is licensed to carry on an activity
described in paragraph 26(e) of the NSC Act in
relation to a Class I nuclear facility
NSC Act Nuclear Safety and Control Act
nuclear energy any form of energy released in the course of
nuclear fission or nuclear fusion or of any other
nuclear transmutation
nuclear energy worker a person who is required, in the course of the
person's business or occupation in connection
with a nuclear substance or nuclear facility, to
perform duties in such circumstances that there
is a reasonable probability that the person may
receive a dose of radiation that is greater than
the prescribed limit for the general public
nuclear facility any of the following facilities:
(a) a nuclear fission or fusion reactor or
subcritical nuclear assembly,
(b) a particle accelerator,
(c) a uranium or thorium mine or mill,
(d) a plant for the processing, reprocessing
or separation of an isotope of uranium,
thorium or plutonium,
(e) a plant for the manufacture of a product
from uranium, thorium or plutonium,
(f) a plant for the processing or use, in a
quantity greater than 10 Bq per
15
calendar year, of nuclear substances
other than uranium, thorium of
plutonium,
(g) a facility for the disposal of a nuclear
substance generated at another nuclear
facility,
(h) a vehicle that is equipped with a nuclear
reactor, and power
nuclear power plant a facility that employs one or more power
20
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
reactors to generate electric or thermal energy
on a commercial scale
nuclear substance (a) deuterium, thorium, uranium or an
element with an atomic number greater
than 92,
(b) a derivative or compound of deuterium,
thorium, uranium or of an element with
an atomic number greater than 92,
(c) a radioactive nuclide,
(d) a substance that is prescribed as being
capable of releasing nuclear energy or as
being required for the production or use
of nuclear energy
(e) a radioactive by-product of the
development, production or use of
nuclear energy, and
(f) a radioactive substance or radioactive
thing that was used for the development
or production, or in connection with the
use, of nuclear energy
Operating Policies a document identified as the Operating Policies
and Principles and Principles in the licensing documents, that
sets out the authorities and responsibilities of
managerial and operating staff, and the principles
and guidelines to be followed for safe operation
of the plant systems
potential serious an event that could have become a serious
process failure process failure, but did not, due to fortuitous
circumstances rather than design provisions or
approved procedures
predicted reliability the reliability of a system in its nominal state
during some future period and/or, for poised
systems, at some future time
prescribed prescribed by regulation of the CNSC
prescribed equipment the equipment prescribed by section 19 of the
General Nuclear Safety and Control
21
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
Regulations
prescribed information the information prescribed by section 20 of the
General Nuclear Safety and Control
Regulations
pressure boundary any pressure-retaining vessel or system
component that is subject to registration or that
is registered under the applicable boiler and
pressure vessel legislation, whether a
conventional system or a nuclear system
prompt event report information transmitted in a verbal or
other form acceptable to the CNSC
radiation the emission by a nuclear substance, the
production using a nuclear substance, or the
production at a nuclear facility of, an atomic or
subatomic particle or electromagnetic wave with
sufficient energy for ionization
reliability the probability that a system in a given state will
be able to perform a stated mission under stated
conditions according to its defined specifications
for a stated mission time and/or, for poised
systems, when required to do so
reliability performance the actual experienced reliability measure for a
indices system over a specific period of time, expressed
in terms of failure rate (time, demand based),
availability or reliability
safeguards a verification system that is established in
accordance with a safeguards agreement
22
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
safety-related system those structures, systems and
components that either are identified as
safety-related in the licensing documents,
or whose malfunction or failure could
lead directly to radiation exposure of site
personnel or the public, or could directly
increase the severity of accidental
releases of radioactive material from the
plant
serious process failure a failure of a process system, component,
structure, or an inappropriate procedure or
human action:
(a) that led to a systematic fuel failure or to
a significant release from the plant, or
(b) that could have led to a systematic fuel
failure or a significant release in the
absence of action by any special safety
system
significant release a release of radioactive material that arises from
an event and that results in a whole body or
committed effective dose in excess of 0.0005 Sv
(50 mrem) or a committed or received thyroid
dose of 0.005 Sv (500 mrem) to the most
exposed member of the public at or beyond the
exclusion boundary
special safety system the shutdown system no.1, the shutdown system
no. 2, the containment system, or the emergency
core cooling system
surveillance activities the activities of monitoring, checking, testing,
calibration or inspecting systems or components
related to assuring that the systems important to
safety remain capable of meeting their defined
specifications
systematic fuel failure fuel that has no defect prior to an event, fails or
exceeds the fuel integrity criteria defined in the
licensing documents as a result of the event
system important to safety a system associated with the initiation, detection,
prevention or mitigation of any failure sequence
which could lead to damage to the fuel and the
23
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
release of radionuclides
unacceptable decline a safety-related system, subsystem or
in reliability component that:
(a) does not meet the predicted reliability
targets that are set out in the licensing
documents or
(b) shows a continued trend of reduced
reliability such that those targets will not
be met
unsafe component fault the condition of a component, whether due to a
failure, maintenance outage or any other reason
for removing the component from service, that
increases the probability of the system being
unable to meet its defined specifications
written report information transmitted in a written or electronic
form acceptable to the CNSC
24
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
Appendices
A: Performance Indicator Specification Sheets
B: Performance Indicator Data Sheets
Consultation Notice
Appendices A and B to this standard have been in trial use at nuclear power
plants during the past year, and are attached for completeness only. No
additional comment is expected from licensees. Following the consultation
period, the standard and the appendices will be edited as a complete document.
25
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
APPENDIX A
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: ACCIDENT SEVERITY RATE (1)
2. Purpose: To indicate the accident severity rate at Canadian nuclear generating stations.
To monitor performance in meeting nuclear industry standards in the area of
worker safety.
To compare performance between Canadian CANDU nuclear generating stations
and internationally.
3. Definition: The accident severity rate is the total number of days lost or charged for all
disabling injuries per 200 000 person hours worked at a station.
4. Calculation: Accident Severity Rate = No. days / 200 000 person hrs
5. Attribute: Reactive
6. Coverage: Global
7. Performance Areas: Worker Safety
8. Primary Area: Worker Safety
9. Data: See attached Data Sheet
10. Data Applicability: Station
11. Trending Frequency: Quarterly
12. Comparability: World
13. Means by which AECB will present data (no licensee input required):
A.1
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
14. Notes:
14.1 The Canadian federal reporting requirement for severity includes shifts not worked. For example, a
person is hurt on the last regularly scheduled shift and then is away for two days that were regularly
scheduled off. If the person would not have been able to work those two days, but was able to return to
work on the first regularly scheduled day, those two days would be counted as lost days.
14.2 Recurrent injuries are attributed back to the originating accident. For example, if an injury from an
accident that resulted in a lost time injury occurred in 1994, recurred in 1996 (with no new accident) the
lost days would not appear in 1996 totals. These days are attributed back to 1994.
14.3 Factors which can influence the length of recovery, such as compensation programs, should be
considered when making comparisons
A.2
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: CHEMISTRY INDEX (2)
2. Purpose: To indicate long-term unit control of important chemical parameters.
To monitor performance in meeting licensee's requirements in chemistry.
To compare performance between Canadian CANDU units.
3. Definition: The average percent success in the monitoring/sampling (for validity, frequency
and range) of selected chemical parameters, for the quarter.
4. Calculation: Chemistry Index (%) = (3 a
m /A
1 1 , a
/A
2 ,... a
2 /A
m ) /m
m
a :1 number of successful samples for parameter 1 during the quarter;
A :
1 number of required samples for parameter 1 for the quarter. The sampling period is based on the
licensee program;
a /A
1 :
1 ratio of success for parameter 1;
m: number of parameters monitored during the period, usually the 12 parameters on the list below.
(3 a
m /A
1 1 , a
/A
2 ,... a
2 /A
m ):
m sum of individual success ratios for each parameter monitored in the
index.
All data is dimensionless. Chemistry Index (CI) results will range between 0 % and 100 %.
Parameters Monitored:
PHT: Annulus Gas [O ]2 Steam Generators:
- pHa - dissolved [Cl]
- dissolved D2 - [SO
]4
- [Na]
Feedwater:
- dissolved [O ]2
- suspended iron
- suspended copper (both copper and iron measured on a 0.45 micron filter paper)
- hydrazine
Condensate Extraction Pump
- dissolved [O ]2
- pHa
5. Attribute: Predictive/Reactive
A.3
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
6. Coverage: Program
7. Performance Areas: Operations
8. Primary Area: Operations
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Quarterly
12. Comparability: CANDU
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 "Success": measurement of a parameter is taken according to the specified minimum frequency, is valid
and within range; "valid": a given sample is representative, acceptable to the licensee chemistry
department; "within range": the result is within the acceptability range defined by the licensee, where no
further action is required of the licensee.
14.2 A missed scheduled minimum measurement cannot be credited later, past the allowable period.
When the safety of chemical technicians or employees could be adversely affected by new hazards, during
normal execution of their tasks, or when the status of the plant is such that the chemical measure is
useless, or unrepresentative, the representative period will be adjusted without penalty. Such measures
will be qualified "void". The data should be auditable.
14.3 The minimum sampling/monitoring frequency is determined by the licensee's current requirements. It is
not permissible to increase sampling frequency in order to reduce the effect of missed or failed previous
or future anticipated failures in sampling.
14.4 Sampling or monitoring equipment failure or unavailability of personnel are not valid reasons for voiding
results. Such results are "failures".
14.5 Where on-line monitoring equipment is available, the success ratio will be calculated as the ratio of time
where the monitoring is on-line and valid data is available and within range over total time. When
monitoring equipment fails, it is permissible to replace the monitoring with manual sampling techniques at
a reasonable frequency.
14.6 The failure period starts from the last valid sample, until the sample is a success.
14.7 Parameters making up the list of the index, and the rules of success, failures, validity and voiding are
reviewed by the AECB.
A.4
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: COMPLIANCE CHEMISTRY INDEX (3)
2. Purpose: To indicate unit control of safety-related chemical and radiochemical parameters.
To monitor performance in meeting regulatory and licensee requirements in
chemistry.
To compare performance between Canadian CANDU units.
3. Definition: The average percent success in the monitoring/sampling (for validity, frequency
and range) of selected chemical parameters, for the quarter.
The list is found in section 4. below.
A.5
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
4. Calculation:
Sampling frequencies and specifications should be defined in the station's operating documentation.
Rules for calculation of the chemistry index in section 3.1 apply also to this compliance chemistry index.
The parameters are selected as compliance parameters in accordance to most OP&P requirements and on
the basis of safety.
List of Parameters Monitored:
Gadolinium isotopic analysis (or boron) (for depletion of neutron absorbing isotopes) for all batch lots
prior to their use in reactor systems;
poison injection tanks [Gd] (unit not in GSS);
poison injection tanks pH during GSS;
a
Moderator D O isotopic (unit not in GSS);
2
[H ] (unit not in GSS);
3
excess reactivity (unit not in GSS);
dissolved [D ] (unit not in GSS);
2
cover gas [D ] (unit not in GSS);
2
[Gd] during GSS;
D O conductivity during GSS (except for G-2);
2
pH during GSS;
a
PHT D O isotopic (unit not in GSS);
2
[H ] (unit not in GSS);
3
[I131] concentration (unit not in GSS);
D2O storage tank cover gas [D
2];
[Cl];
Moderator to heat transport coolant D O isotopic purity difference (unit not in GSS);
2
Annulus gas dewpoint (unit not in GSS);
End shield cooling water pH ;a
cover gas [H ] (for Point Lepreau, G-2 and Pickering B);
2
Emergency coolant injection high pressure tanks water pH;a
hydrazine concentration;
Liquid zone control cover gas [H ]2
5. Attribute: Predictive/Reactive
6. Coverage: Program
7. Performance Areas: Operations, Compliance
8. Primary Area: Compliance
9. Data: See attached Data Sheet
10. Data Applicability: Unit
A.6
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
11. Trending Frequency: Quarterly
12. Comparability: Canadian
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 "Success": measurement of a parameter is taken according to the specified minimum frequency, is valid
and within range; "valid": a given sample is representative, acceptable to the licensee chemistry
department; "within range": the result is within the acceptability range defined by the licensee, where no
further action is required of the licensee.
14.2 A missed scheduled minimum measurement cannot be credited later, past the allowable period.
When the safety of chemical technicians or employees could be adversely affected by new hazards, during
normal execution of their tasks, or when the status of the plant is such that the chemical measure is
useless, or unrepresentative, the representative period will be adjusted without penalty. Such measures
will be qualified "void". The data should be auditable.
14.3 The minimum sampling/monitoring frequency is determined by the licensee's current requirements. It is
not permissible to increase sampling frequency in order to reduce the effect of missed or failed previous
or future anticipated failures in sampling.
14.4 Sampling or monitoring equipment failure or unavailability of personnel are not valid reasons for voiding
results. Such results are "failures".
14.5 Where on-line monitoring equipment is available, the success ratio will be calculated as the ratio of time
where the monitoring is on-line and valid data is available and within range over total time. When
monitoring equipment fails, it is permissible to replace the monitoring with manual sampling techniques at
a reasonable frequency.
14.6 The failure period starts from the last valid sample, until the sample is a success.
14.7 Parameters making up the list of the index, and the rules of success, failures, validity and voiding are
reviewed by the AECB.
A.7
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: CHANGE CONTROL INDICES (4)
2. Purpose: To indicate the control over changes in equipment and procedures for the
safety-related systems at Canadian nuclear generating stations.
To monitor performance in the management of change for the safety-related
systems.
3. Definition: Change control indices are defined as factors important for maintaining control
over equipment and procedural changes.
4. Calculation:
4.1 Number of temporary procedural changes (number of pages) (see note 14.1).
4.2 Number of temporary equipment changes.
4.3 Number of incomplete permanent equipment changes (see note 14.2).
4.4 Number of temporary equipment and procedural changes over six months old (listed separately).
4.5 Number of instances where changes in equipment and/or procedures contributed to an event as recorded
by the licensee.
5. Attribute: Predictive
6. Coverage: Global
7. Performance Areas: Operations, Maintenance
8. Primary Area: Operations
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Data collected monthly and submitted to the AECB each quarter.
12. Comparability: Station
13. Means by which AECB will present data (no licensee input required):
A.8
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
14. Notes:
14.1 All equipment or procedural changes for the indicator shall include the safety-related systems identified
by the station staff according to S-99.
14.2 Permanent changes are considered incomplete until all testing, design, installation and operating
documentation have been amended according to the changes in the field.
14.3 All data gathered shall be presented on a quarterly basis, as described in the data sheet.
A.9
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: EMERGENCY PREPAREDNESS DRILL COMPLETION RATIO (5)
2. Purpose: To indicate adherence to licensee's requirements on emergency preparedness
drills.
To compare performance between Canadian nuclear generating stations.
To be used as a component in the measurement of emergency preparedness.
3. Definition:
3.1 A drill is a means by which a licensee tests and evaluates through simulation the capability of the station's
planning, procedures, resources, and staff training to respond to an emergency.
3.2 Each drill evaluates the response capability of a single crew, team, or unit, and uses the actual facility and
staff.
3.3 Examples of the type of emergency for which drills are conducted include but are not limited to radiation,
fire, search and rescue, first-aid, contaminated casualty, and hazardous materials.
3.4 Station-wide exercises and practices (e.g., rehearsals, walk-through, table-top training) are not included
in this definition.
4. Calculation:
Emergency Preparedness Drill Completion Ratio = number of drills completed by the end of each
quarter/number of drills scheduled that year
3 3
i c
j ij
=
3 3
i s
j ij
where c = number of completed drills at the end of each
quarter
s = number of scheduled drills for the year
i = crew
j = drill type
c max
ij =
s , if c
ij
ij ó s
ij ú i,j
5. Attribute: Predictive
6. Coverage: Program
7. Performance Areas: Public Safety, Worker Safety
8. Primary Area: Public Safety
A.10
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
9. Data: See attached Data Sheet
10. Data Applicability: Station
11. Trending Frequency: Quarterly
12. Comparability: Canadian
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 A drill schedule that shows the annual commitment of different drill types for each crew is to be
produced by each licensee, and be brought to the attention of the AECB prior to the beginning of the
year.
14.2 Completed drills are reported to the AECB in the quarterly technical reports, pursuant to sections 3.11
and 4.7 of S-99.
14.3 Reporting must depict clearly the number of drills completed, and the crew or team that completed
them.
14.4 Drills testing the interfaces between crews and teams may take place concurrently.
14.5 Overachievement by one crew or team will not be credited unless all crews and teams have completed
the scheduled drills.
A.11
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: EMERGENCY RESPONSE EQUIPMENT CALL-UP COMPLETION RATIO (6)
2. Purpose: To indicate adherence to a licensee's requirements for completing scheduled
maintenance, inventory checks and testing of dedicated emergency response
equipment as set out in the document referenced in A.A.4 of the operating
licence.
To compare performance between Canadian nuclear generating stations.
To be used as a component in the measurement of emergency preparedness.
3. Definition: The ratio is the number of completed maintenance, inventory checks and tests for
all equipment dedicated to emergency response to that which was scheduled in a
quarter.
4. Calculation: Emergency Response Equipment Call-Up Completion Ratio
= number of emergency equipment maintenance, inventory checks and
testing call-ups completed in a quarter/number scheduled that quarter
c
=
s
where: c = number of completed call-ups; and
s = number of scheduled call-ups.
5. Attribute: Predictive
6. Coverage: Program
7. Performance Areas: Public Safety, Worker Safety
8. Primary Area: Worker Safety
9. Data: See attached Data Sheet
10. Data Applicability: Station
11. Trending Frequency: Quarterly
12. Comparability: Canadian
13. Means by which AECB will present data (no licensee input required):
A.12
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
14. Notes:
14.1 The types of emergency to be covered extend beyond those related only to radiation, but also to any
contingency for which a licensee may have a response plan in place, such as chemical, fire, and toxic gas
emergencies.
14.2 Only dedicated equipment is to be included in the definition. Equipment that is used in normal operation
but required to implement a plan is not included here.
14.3 Equipment includes fixed systems, facilities, portable instruments, communications and other equipment
called up in the licensee's plan as needing to be available and in a state of readiness. Licensees are
expected to produce a list as part of their plans.
14.4 Examples of dedicated equipment are listed in elements 9 and 10 of the document entitled
"Recommended Criteria for the Evaluation of On-Site Nuclear Emergency Plans", which was provided
to all licensees in 1994. Any equipment in the lists that is used in normal operation would not be
applicable.
14.5 Although the equipment covered is not in regular use, the expectation is that it will always be in a state
of readiness and in its proper location. Programmes should be established to check regularly if
equipment is present and ready.
14.6 The aspect being measured is that tests and inventory checks are done according to a fixed schedule, not
whether the test failed or equipment was missing. It is expected that programmes exist to remedy
malfunctions and replace that which is missing. Other means are in place to assess these latter aspects of
readiness.
14.7 The frequencies chosen for the checks in the call-up schedule should be defendable.
14.8 Pre-determined dates for the call-ups will be used to measure schedule compliance. There will be
provision for a grace period only when a licensee justifies it to AECB staff satisfaction.
A.13
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: NUMBER OF MINIMUM SHIFT COMPLEMENT NON-COMPLIANCES (7)
2. Purpose: To indicate compliance with the minimum shift complement requirement as identified in
the reactor operating licence.
To monitor performance in meeting regulatory requirements to ensure adequate staffing
for normal operations and emergencies.
To compare performance between Canadian CANDU units.
To be used as a component in the measurement of an emergency preparedness index, if
necessary.
3. Definition: The number of shifts, during the quarter, where the minimum shift complement is not met
for the unit.
4. Calculation: PI = # of shifts without minimum shift complement, during the quarter.
5. Attribute:Predictive/Reactive
6. Coverage: Program
7. Performance Areas: Public Safety, Worker Safety
8. Primary Area: Public Safety
9. Data: See attached Data Sheet
10. Data Applicability: Station
11. Trending Frequency: Quarterly
12. Comparability: Canadian
13. Means by which AECB will present data (no licensee input required):
A.14
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
14. Notes:
14.1 The "minimum shift complement" is defined in terms of staff number with proper authorizations and
qualifications and is referenced in the reactor operating licence or supporting licensing submissions for
the purpose of normal and abnormal operations, radiation emergencies, fires, special procedures, first
aid, search and rescue.
14.2 Records should be able to demonstrate shift staff qualifications and minimum requirements.
14.3 There is no grace period. Failure to satisfy the minimum requirement for the full shift is a failure for that
shift.
A.15
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: NON-COMPLIANCE INDEX (8)
2. Purpose: To indicate the number of occurrences where the licensee's operation of a unit
failed to comply with its Operating Policies and Principles, Radiation Protection
Regulations, other Operating Licence conditions, or with other Regulations.
To compare performance between Canadian CANDU units.
3. Definition:
3.1 A non-compliance (or violation) is a failure to comply with the Physical Security Regulations, the
Transport Packaging of Radioactive Materials Regulations, the Atomic Energy Control Regulations or
any condition of the Reactor Operating Licence, including documents referenced therein, such as the
licensee's OP&Ps or its Radiation Protection Regulations.
3.2 The index is a composite total in each quarter of all types of non-compliances for a unit in which
simultaneous non-compliances of the same safety principle are counted only once.
A.16
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
4. Calculation:
The number may be represented by an expression that removes any intersections
that represent multiple accounting.
a b
Number of Non-Compliances =
a + b + c + d c d
- (a 1 b) - (b 1 c) - (c 1 d) - (d 1 a)
- (a 1 c) - (b 1 d)
+ (a 1 b 1 c) + (b 1 c 1 d) + (c 1 d 1 a) + (d 1 a 1 b)
- (a 1 b 1 c 1 d)
where: a = number of OP&P non-compliances for a reactor unit plus, for multi-unit
stations, those for unit 0 (station-wide);
b = number of RPR non-compliances for a reactor unit plus, for multi-unit
stations, those for unit 0 (station-wide);
c = number of other licence non-compliances for a reactor unit plus, for multi-
unit stations, those for unit 0 (station-wide);
d = number of non-compliances for a reactor unit of the Physical Security
Regulations, the Transport Packaging of Radioactive Materials
Regulations, or the Atomic Energy Control Regulations plus, for multi-
unit stations, those for unit 0 (station-wide).
5. Attribute: Reactive
6. Coverage: Global
7. Performance Areas: Operations, Maintenance, Public Safety, Worker Safety, Compliance
8. Primary Area: Compliance
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Quarterly
12. Comparability: Canadian
13. Means by which AECB will present data (no licensee input required):
A.17
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
14. Notes:
14.1 Under sections 2.1, 2.2 and 2.3 of S-99, licensees are required to report all non-compliances defined by
3.1 above.
14.2 Written reports for 1.1(a)-related events are to be submitted no later than 15 calendar days from the
discovery of the event.
14.3 The values to be used for the variables a, b, c, and d in the calculation will be based on the date of
discovery of the event.
14.4 In instances where AECB and licensee staff disagree on a non-compliance classification, the opinion of
the AECB shall prevail.
14.5 For each event reported as non-compliance, the data in the report must indicate which of the four
categories (a, b, c, d) apply.
14.6 In multi-unit stations, a non-compliance that affects more than one unit should be recorded as a
non-compliance for each unit.
14.7 The presentation of this measurable is capable of showing the composite index as well as the total
number of non-compliances in a quarter for each of the four categories.
14.8 Full comparison of this measurable between stations may not be appropriate because the OP&Ps and
RPRs vary from station to station.
A.18
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: NUMBER OF PRESSURE BOUNDARY DEGRADATIONS (9)
2. Purpose: To indicate the number of pressure boundary degradations and failures which can
be attributed or related to the operation and management of each of the CANDU
stations.
To monitor the performance in meeting nuclear industry codes and standards.
To compare performance between Canadian CANDU units.
3. Definition:
Pressure boundary: Any pressure-retaining vessel or system component that is subject to registration or
that is registered under the applicable boiler and pressure vessel legislation, whether a conventional
system or a nuclear system. (Note: For older stations such as Pickering A this should be "that under the
current N285.0 would require registration.")
Pressure boundary degradation: A degradation of the pressure boundary that exceeds a limit that is
specified in the design analysis or in the applicable boiler and pressure vessel code, standard or act under
which the pressure boundary was registered and includes the following elements currently listed in
AECB regulatory document S-99 section 2.18:
(a) a pressure boundary excessive deformation, a crack, a hole or a rupture, or a leakage in excess
of a limit that is specified in the Operating Policies and Principles;
(b) the occurrence of an abnormal loading transient that exceeds:
A) a pressure boundary design condition, or
B) a Service Level B condition, for any nuclear component that is designed in accordance with
rules of ASME III subsection NB;
(c) a change to the size, rating or material property of the pressure boundary beyond that allowed
for in the design;
(d) a repair or modification that changes the strength of a component of the pressure boundary that
did not receive the prior authorization required by CSA Standard N285.0;
(e) a reduction of the wall thickness beyond that allowed in the design by the applicable pressure
vessel code, standard or act under which the pressure boundary was registered; and
(f) degradation of the overpressure protection equipment for the pressure boundary that violates a
limit of the overpressure protection report or any other licensing document.
A.19
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
4. Calculation:
Number of pressure boundary degradations:
N = a + a
n1 + a
n2 + a
n3 + a
n4 - (b + c)
c
where: an1 = number of pressure boundary degradations in Class 1 nuclear systems
reported under S-99 in a given quarter;
an2 = number of pressure boundary degradations in Class 2 nuclear systems
reported under S-99 in a given quarter;
an3 = number of pressure boundary degradations in Class 3 nuclear systems
reported under S-99 in a given quarter;
an4 = number of pressure boundary degradations in Class 4 nuclear systems
reported under S-99 in a given quarter;
ac = number of pressure boundary degradations in conventional systems
reported under S-99 in a given quarter;
b = number of pressure boundary degradations reported under S-99 in a given
quarter on station conventional systems; and
c = number of pressure boundary degradations reported under S-99 in a given
time period (quarter, year etc...) which were not attributed or related to
the operation of the station (i.e. manufacturing defects, poor original
design, etc...) Failures or degradations due to deficiencies in the recent
(within 5 years) design of modifications and repairs, as well as failures due
to site manufacturing and installation are not included in c.
Note that in the above categorization, classes 1, 2, 3 and 4 refer to what the code class of the failed or
degraded component would be under the requirements of the current N285.0.
5. Attribute: Predictive/Reactive
6. Coverage: Program
7. Performance Areas: Operations, Maintenance
8. Primary Area: Maintenance
A.20
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Quarterly
12. Comparability: Canadian
13. Means by which AECB will present data (no licensee input required):
14. Notes:
A.21
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: PREVENTIVE MAINTENANCE COMPLETION RATIO (10)
2. Purpose: To indicate the fraction of preventive maintenance (PM) jobs to total
maintenance jobs completed.
To monitor performance in meeting nuclear industry expectations in the area of
PM.
To compare performance between Canadian CANDU units.
3. Definition: Preventive maintenance completion ratio is the ratio of preventive maintenance
jobs completed divided by the preventive maintenance plus corrective
maintenance jobs completed.
3.1 Preventative maintenance jobs
The PM jobs are jobs performed on equipment in the field, but is in working order when PM job
commences. The PM jobs completed shall include those that are frequency or condition based.
3.2 Corrective maintenance jobs
The corrective maintenance (CM) jobs completed shall include data on all maintenance performed as a
result of a reported deficiency or failure of equipment in the safety-related system as defined in notes
14.2. It shall not include design modifications.
3.3 Deficiency Reports (DRs)
DRs that are written to correct an identified deficiency in safety-related systems.
4. Calculation:
(PM jobs per quarter)
--------------------------------------- x 100 = % PM jobs completed per quarter
(PM jobs + CM jobs per quarter)
The data for this indicator shall only include safety related systems as defined by the CSA standards.
5. Attribute: Predictive
6. Coverage: Program
7. Performance Areas: Operations, Maintenance, Public Safety
A.22
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
8. Primary Area: Maintenance
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Quarterly
12. Comparability: Station
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 The indicator requires the total number of PM and CM jobs scheduled to be completed in the quarter.
14.2 PM, CM and DRs to be included in the data shall be from systems identified by the station staff as being
safety related according to S-99.
14.3 The term "open" refers to DRs which have been identified but remain to be checked before introducing
them into the system for corrective action.
14.4 The term "close" or complete refers to jobs which are completed in the field and marked as complete in
the database.
A.23
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: NON-COMPLIANCES WITH NUMBER OF SPECIFIC RADIATION
PROTECTION PROCEDURES (11)
2. Purpose: To indicate compliance with station specific radiation protection rules and procedures.
3. Definition: Indicator is intended to measure the staff adherence to radiation protection.
(a) Radiation signposting out of date.
(b) Evidence of food and cigarette consumption in Zone 2/3.
(c) Improper setup of rubber areas.
(d) Whole body monitor contamination incidents not recorded in radiation log.
4. Calculation: None
5. Attribute: Predictive/Reactive
6. Coverage: Program
7. Performance Areas: Worker Safety
8. Primary Area: Worker Safety
9. Data: See attached Data Sheet
10. Data Applicability: Station
11. Trending Frequency: Quarterly
12. Comparability: Station
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 The data shall be presented as individual totals for (a) to (d).
14.2 The data provided to the AECB should be auditable.
14.3 In case of Pickering NGS, Pickering NGS "A" and Pickering NGS "B" data are combined as one
measure.
A.24
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
A.25
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: RADIATION OCCURRENCE INDEX (12)
2. Purpose: To indicate the number and equivalent severity of radiation occurrences which
have taken place at each CANDU station.
To monitor the performance in meeting the AECB's expectations in the area of
worker radiation protection safety.
To compare performance between Canadian nuclear generating stations.
3. Definition: A radiation occurrence is an occurrence where one or more of the following
consequences has occurred:
- fixed body contamination exceeding 50 kBq/square metre (1.35 uCi/m ) has been detected,
2
- an unplanned acute whole body dose (resulting from an external exposure) exceeding 5 mSv
(500 mrem) has been received,
- an acute intake of radioactive material resulting in an effective dose greater than 2 mSv (200
mrem) has taken place,
- an acute or committed dose in excess of any of the limits specified in the Radiation Protection
Regulations Part 1 (Ontario Hydro), the Directives de Santé et Normes de Radioprotection
(Hydro-Quebec), or the Radiation Protection Regulations (New Brunswick Power) has been
received.
4. Calculation:
Radiation occurrence index = a + 5b + 5c + 50d.
where: a = number of occurrences in a quarter where fixed body contamination in
excess of 50 kBq/square metre (1.35 uCi/m ) was measured;
2
b = number of occurrences in a quarter where an unplanned acute whole body
dose (resulting from an external exposure) exceeding 5 mSv (500 mrem)
was received;
c = number of occurrences in a quarter where an intake of radioactive
material resulted in an effective dose greater than 2 mSv (200 mrem),
normalized to 2 mSv;
d = number of occurrences in a quarter where an acute or committed dose in
excess of any of the limits specified in the Radiation Protection
Regulations Part 1 (Ontario Hydro), the Directives de Santé et Normes de
Radioprotection (Hydro-Québec), or the Radiation Protection
Regulations (New Brunswick Power) was received.
A.26
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
5. Attribute: Predictive/Reactive
6. Coverage: Program
7. Performance Areas: Worker Safety, Compliance
8. Primary Area: Worker Safety
9. Data: See attached Data Sheet
10. Data Applicability: Station
11. Trending Frequency: Quarterly
12. Comparability: Canadian
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 The values of the weighting factors (1, 5, 5, 50) required to differentiate between the least safety
significant radiation occurrences and the most serious ones are not risk-based. They are meant to
indicate the degree of radiation protection practices breakdown required for each type of event to occur.
14.2 The total for 'c' above is normalized to 2 mSv. This means that the actual dose received as a result of
the occurrence(s) is divided by 2 mSv. For example: two exposures as a result of separate intake
incidents, one of 3 mSv and one of 4 mSv, would result in a value for 'c' = 3 mSv/2 mSv + 4 mSv/2
mSv = 1.5 + 2 = 3.5. This value 'c' would then be multiplied by the weighting factor 5.
14.3 Only the number of occurrences with a higher level of consequence will be calculated for a single
occurrence to prevent double counting. For example: If a single occurrence results in an unplanned
acute whole body dose exceeding 5 mSv (500 mrem), and in the affected workers whole body dose
exceeding the quarterly legal limit, the occurrence would be counted as a "d" type of occurrence (the
most serious type) and not as one "b" type and one "d" type.
A.27
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: UNAVAILABILITY OF SAFETY-RELATED SYSTEMS (13)
2. Purpose: To indicate the actual past unavailability and to predict the future unavailability of
safety-related systems for a reactor unit in a quarter (see note 14.1 below).
To monitor the performance of safety-related systems in meeting AECB and
licensee requirements.
To compare performance between Canadian CANDU units in a given station.
3. Definition:
3.1 Actual Past Unavailability: The fraction of a time period during which a system was required to be
available but did not meet the minimum performance standards as identified in the Safety Report or other
design/operating documents to perform its function if it was called upon to operate.
3.2 Predicted Future Unavailability: The fraction of a time period during which a system is required to be
available but is not expected to meet the minimum performance standards as identified in the Safety
report or other design/operating documents to perform its function if it is called upon to operate.
3.3 Safety-Related Systems: For this indicator at present, the systems include SDS1, SDS2, ECCS,
Containment, Emergency (Service) Water System, Boiler Emergency Cooling System, Standby
Generators and Emergency Power System (see note 14.2 below ).
4. Calculation:
Unavailability of a system in a reactor unit = a + b
where: a = unavailability value of the components specific to a unit within a quarter;
and
b = unavailability value of the components common to a number of unit(s)
within a quarter. This is only applicable to a station with multiple units.
For single unit stations, the value of b is zero.
5. Attribute: Predictive/Reactive
6. Coverage: Global
7. Performance Areas: Maintenance, Public Safety, Compliance
8. Primary Area: Public Safety
A.28
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Quarterly
12. Comparability: Station
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 Licensees are currently required to report the actual past unavailability and predicted future unavailability
of SDS1, SDS2, ECCS and Containment. In the future, licensees will be required to report the reliability
values for the safety-related systems according to AECB regulatory requirements (currently given in C-
98, revision 1, Reliability Requirements for Safety-Related Systems of Nuclear Reactor Facilities).
14.2 At present, the Safety-Related Systems only include those systems identified in section 3.3. In the future
these systems may be expanded to include those as required by the AECB regulatory requirements
(currently given in C-98, revision 1, Reliability Requirements for Safety-Related Systems of Nuclear
Reactor Facilities).
14.3 Currently our licensees do not report System Reliability/Unreliability value according to the regulatory
document S-99 because the required mission time of some systems is not yet defined, e.g. ECCS and
Containment System.
14.4 At present, the models used in the unavailability calculation vary from station to station. Consequently, it
is not appropriate to compare the unavailability values between stations. For meaningful comparison, it
is important to have acceptable common models for the unavailability calculations.
14.5 The Actual Past Unavailability should include the unavailability values due to design deficiencies and
operational faults found in the specified time period.
14.6 The methodology and the processes of analysis should be auditable.
A.29
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: STATION WHOLE BODY DOSE (14)
2. Purpose: To indicate the total dose due to ionizing radiation received by all individuals
working at the nuclear station and its related facilities, required to operate the
nuclear station.
To monitor the performance in meeting a station whole body dose ALARA.
To compare performance internationally.
3. Definition: The sum of the whole body doses due to ionizing radiation received by
individuals (including permanent, part-time and temporary staff, external
contractors, consultants etc.) working at the nuclear station and its related
facilities, over the year.
4. Calculation:
PI = a + b + 3 N (c
i + d
i ) ... in units of millisievert; decimals are not required.
i
a = total whole body dose to station personnel from external irradiation during power
operation and forced outages, at the nuclear station;
b = total whole body dose (actual or committed) to station personnel from internal uptakes
during power operation and forced outages, at the nuclear station;
ci = total whole body dose to station personnel from external irradiation during individual unit
planned outage i, as applicable; and
di = total whole body dose (actual or committed) to station personnel from internal uptakes
during individual unit planned outage i, as applicable.
N = Number of planned outages at the nuclear station during the year.
5. Attribute: Predictive
6. Coverage: Global
7. Performance Areas: Worker Safety
8. Primary Area: Worker Safety
9. Data: See attached Data Sheet
A.30
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
10. Data Applicability: Station
11. Trending Frequency: Annually
12. Comparability: World
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 A unit outage is considered "planned" when preparations and the decision to shutdown the unit have
been made at least 6 weeks in advance. For the purposes of convenience to the utility, the dose related
to work during unplanned power outages, forced shutdowns or power manoeuvres, or short outages
such as poison prevent operation, should be recorded under "station operation" doses (parameters a and
b in the equation).
14.2 Doses from neutron exposures should normally be accounted for, in this indicator.
14.3 The internal and external dosimetry programs should satisfy the requirements of the AECB.
14.4 Doses attributed to an outage (measurables c and d
i ) include the doses directly related to the
i
preparation of the outage, the start-up and related power manoeuvres.
14.5 The results of dose reduction processes can be credited for individual dose commitments, provided that
an auditable process is in place.
14.6 For the purposes of the AECB performance indicator program, whole body doses can be considered
equivalent to deep doses at Canadian nuclear stations.
14.7 For the purposes of this indicator, doses resulting from activities on a unit which is decommissioned,
mothballed or in lay-up is not considered. However, licensees should notify the AECB where significant
doses resulting from activities on one of these units could affect station dose.
14.8 Dose commitments to staff resulting from common station services such as Unit zero, heavy water
upgrading, or from activities in facilities directly related to the operation of the unit such as laundry,
decontamination, fuelling, waste handling, can be accounted under "station operation".
14.9 Doses to visitors, or from initiatives related to public information programs, are considered negligible.
However, these doses can be excluded from the total dose, as seen appropriate by the licensee
A.31
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: NUMBER OF MISSED MANDATORY SAFETY SYSTEM TESTS (15)
2. Purpose: To indicate successful completion of tests required by licence condition, including
those referenced in documents submitted in support of a licence application.
To monitor performance in meeting regulatory and licensee availability
requirements.
3. Definition: The number of missed safety system tests is the sum of those tests that are not
completed for each of the three groups of safety-related system (i.e., the Special
Safety Systems, the Standby Safety Systems, and other Safety Related Systems)
in a quarter.
4. Calculation: Number of Missed Tests
= a + b + c
where: a = number of missed tests reported under 10.10 of S-99 for the SSSs of a
unit plus, for multi-unit stations, those for unit 0 (station-wide);
b = number of missed tests reported under 10.10 of S-99 for the SbSS of a
unit plus, for multi-unit stations, those for unit 0 (station-wide);
c = number of missed tests reported under 10.10 of S-99 for other SrS of a
unit plus, for multi-unit stations, those for unit 0 (station-wide)
Missed tests means tests not completed within the allowable period.
5. Attribut: Predictive/Reactive
6. Coverage: Program
7. Performance Areas: Operations, Public Safety
8. Primary Area: Operations
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Quarterly
A.32
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
12. Comparability: Canadian
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 Licence condition A.A.12 requires that tests requested by the Board be completed expeditiously. Unless
otherwise directed in writing by the Board, A.A.13 requires licensees to test all systems at a frequency
sufficient in the opinion of the Board to substantiate the reliability claimed or implied in the Safety
Report or in the documents listed in the application. There is also a specific condition requiring a leak
rate test of containment.
14.2 Section 10 of S-99 requires that each licensee make a reliability report annually. Included as sub-section
(a) is the requirement for a report on the completion of all tests that were required to be carried out
during the reporting period by a licence condition or that were required by a routine test program that
was referred to in the licensing documents.
14.3 To comply with A.A.13, licensees have documentation in place that establishes a test program for safety
systems, outlining a frequency/time interval for each test.
14.4 The number of missed tests reported under 10.10 of S-99 therefore will provide the data for the
calculation.
14.5 For the purpose of this definition, the following shall apply:
Special Safety Systems (SSS) - SDS1, SDS2 (SDSE for PNGS-A), ECC, and Containment;
Standby Safety Systems (SbSS) - Boiler Emergency Cooling, Emergency Power Supply, Standby
Generators, Emergency Filtered Air Discharge, Emergency
Water, Inter-Unit Feedwater Tie;
Safety Related Process Systems (SrS) - Reactor Regulating, Heat Transport, Moderator, Class I, II,
and III, Auxiliary Boiler Feed, Service Water.
This list may be expanded in future.
14.6 All licensees currently report missed tests in the Quarterly Technical Report for the fourth quarter or in
their annual Reliability Report.
14.7 Missed tests refer to those not completed, as opposed to those that fail. It is expected that causes of
failures are corrected promptly.
14.8 Tests conducted beyond the maximum allowable time interval permitted by the reliability calculation or
by an applicable engineering code will count as a miss unless approval has been obtained from the
AECB to extend the test interval.
14.9 If provision is made in approved procedures to miss a particular test, it will not be used in the
calculation.
A.33
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: NUMBER OF UNPLANNED TRANSIENTS (16)
2. Purpose: To indicate the number of reactor power transients which have exceeded or have
a potential to exceed the allowable safe operating envelope as defined in the
Safety Report, design documents or operating manuals due to equipment failures
or operator errros while the reactor is not in a guaranteed shutdown state.
To compare operation of Canadian CANDU units.
3. Definition:
The unplanned transients are the events that result in a change of reactor operating states due to:
(a) Unplanned reactor setbacks and stepbacks, both automatic and manual, which occur while the
reactor is not in a guaranteed shutdown state. These reactor setbacks and stepbacks are the
events resulting from the corrective actions taken by operator or the internal plant equiment
failure, spurious signal, human error or external events such as severe weather, earthquake,
airplane strike, grid instability, railway explosion, etc.
(b) Unplanned reactor trips, both automatic and manual, which occur while the reactor is not in a
guaranteed shutdown state. These reactor trips are the result of the events arising from
corrective actions taken by operator or internal plant equipment failure, spurious signal, human
error, or external events such as severe weather, earthquake, airplane strike, grid instability,
railway explosion, etc.
4. Calculation: The total number of unplanned transients in a quarter for a particular reactor
unit = number of (a) + number of (b) as defined above.
5. Attribute: Predictive/Reactive
6. Coverage: Global
7. Performance Areas: Operations, Public Safety
8. Primary Area: Public Safety
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Quarterly
12. Comparability: Canadian
A.34
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
13. Means by which AECB will present data (no licensee input required):
14. Notes:
14.1 The manual reactor trips, setbacks or stepbacks which are required by planned (as opposed to forced) outage
maintenance or routine testing should not be included in (a) and (b) above. The manual reactor trip on SDS2
once every two years is one of the examples. The use of manual reactor setback to deal with flux tilt during
refuelling is not appropriate and therefore should be included in (a) calculation. (Note: The proper way of
operation is to reduce power first then start refuelling to prevent reactor setback, stepback or trip)
14.2 The main function of setback and/or stepback, as installed in the regulating system, is to reduce the number of
shutdown system actuations. This is a CANDU design feature. That is why the setback and stepback are
included in the calculation of transient numbers which have exceeded or have potential to exceed the
allowable safe operating envelope.
14.3 If an event results in a reactor setback, stepback and trip in sequence, then the total transient number will be
counted in (b) as 1.0. If an event results in a reactor setback and stepback in sequence, again the total number
of transients will be counted in (a) as 1.0.
14.4 If an event results in a reactor trip on both shutdown systems, the number of reactor trips should only be
counted as one, even though two shutdown systems are actuated on different trip parameters. This is to take
into account that CANDU plants are installed with two shutdown systems, if the number of reactor trips is to
be compared internationally.
14.5 After a reset of reactor setback, stepback and/or trip by operator and the reactor power is allowed to increase,
if (a) or (b) occurs again due to failure to correct causes of the initial transient, then the subsequent reactor
setback, stepback and/or trip should be included in the calculation of the number of unplanned transients.
A.35
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
PERFORMANCE INDICATOR
SPECIFICATION SHEET
Rev. date: 1998.06.30
1. Title: UNPLANNED CAPABILITY LOSS FACTOR (17)
2. Purpose: To indicate how a unit is managed, operated, maintained in order to avoid
unplanned outages.
To compare performance internationally.
3. Definition: This measurable is from WANO, Implementing Guideline 19.1 (1993), addendum
September 1996.
Some of its elements are found under the indicator Unit Capability Factor, found
in the same document.
4. Calculation: The accounting period considered will be the quarter.
5. Attribute: Reactive
6. Coverage: Global
7. Performance Areas: Operations, Maintenance
8. Primary Area: Maintenance
9. Data: See attached Data Sheet
10. Data Applicability: Unit
11. Trending Frequency: Quarterly
12. Comparability: World
13. Means by which AECB will present data (no licensee input required):
14. Notes:
The same comments and remarks as in the referenced WANO document also apply (attached), WANO
IG 19.1 ANR1.OR, September 1996.
A.36
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
APPENDIX B
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: ACCIDENT SEVERITY RATE (1)
STATION: QUARTER, YEAR:
DATA:
a = number of days lost during the quarter at the station
=
b = number of person-hours worked during the quarter at the station
=
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO
AECB BY: DATE:
FOR AECB USE:
accident severity rate = (a / b) x 200000
=
AECB PREPARED BY: AECB VERIFIED BY:
B.1
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: CHEMISTRY INDEX (2)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per unit)
DATA:
a = number of successful samples during the quarter for parameter i.
i
A = number of required samples during the quarter for parameter i.
i
PHT pHa a1 =
A1 =
PHT Dissolved [D2] a2 =
A2 =
Annulus Gas [O2] a3 =
A3 =
Steam Generators a4 =
Dissolved [Cl] A4 =
Steam Generators a5 =
[SO4] A5 =
Steam Generators a6 =
[Na] A6 =
Feedwater Dissolved a7 =
[O2] A7 =
Feedwater Suspended a8 =
Iron (could be total A8 =
combined with copper)
Feedwater Suspended a9 =
Copper A9 =
Feedwater (or Steam a10 =
Generator) A10 =
Hydrazine
AECB PERFORMANCE INDICATOR
B.2
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
DATA SHEET
Rev. Date: 1998.06.30
TITLE: CHEMISTRY INDEX (CONT'D)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per unit)
DATA:
Condensate a =
11
Dissolved [O2] A11 =
(at extraction pump
or HP feedheater
outlet)
Condensate pHa a =
12
(extraction pump) A =
12
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO
AECB BY: DATE:
B.3
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: CHEMISTRY INDEX (CONT'D)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per
unit)
FOR AECB USE:
a x 100 = %
1 a x 100 = %
2 a x 100 = %
3 a x 100 = %
4
A1 A2 A3 A4
a x 100 = %
5 a x 100 = %
6 a x 100 = %
7 a x 100 = %
8
A5 A6 A7 A8
a x 100 = %
9 a x 100 = %
10 a x 100 = %
11 a x 100 = %
12
A9 A10 A11 A12
Chemistry index = Average of above percentages
= %
AECB PREPARED BY: AECB VERIFIED BY:
B.4
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: COMPLIANCE CHEMISTRY INDEX (3)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per unit)
DATA:
a = number of successful samples during the quarter for parameter i.
i
A = number of required samples during the quarter for parameter i.
i
Gadolinium (or boron) batch a1 =
isotopic analysis A1 =
LISS poison injection tanks [Gd] a2 =
A2 =
Moderator D2O isotopic
a3 =
A3 =
Moderator [H13] a4 =
A4 =
Moderator excess reactivity a5 =
A5 =
Moderator dissolved [D2 ] a6 =
A6 =
Moderator cover gas [D2] a7 =
A7 =
PHT D2O isotopic
a8 =
A8 =
PHT [H3] a9 =
A9 =
PHT [I131] a10 =
A10 =
B.5
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: COMPLIANCE CHEMISTRY INDEX (CONT'D)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per unit)
DATA:
PHT D O storage tank cover gas
2 a =
11
[D2] A =
11
PHT [Cl] a =
12
A =
12
Moderator to PHT D 2O isotopic
a =
13
purity difference check A =
13
Annulus gas dewpoint a =
14
A =
14
End shield cooling water pHa a =
15
A =
15
For selected stations:
End shield cooling cover gas [H ]2 a =
16
A =
16
ECI HP water tank(s) pHa a =
17
A =
17
ECI HP water tank(s) hydrazine a =
18
concentration A =
18
Liquid zone control system cover a =
19
gas [H2] A19 =
B.6
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: COMPLIANCE CHEMISTRY INDEX (CONT'D)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per unit)
DATA:
For all units in guaranteed shutdown state (GSS) during the quarter, or part of the quarter:
b =
i number of successful samples (during GSS) for parameter i.
B =
i number of required samples (during GSS) for parameter i.
LISS poison injection tanks b =
1
pH when SDS2 available
a B =
1
[Gd] in moderator b =
2
B =
2
Moderator D O conductivity
2 b =
3
(except for G-2) B =
3
Moderator D O pH
2 a b4 =
B =
4
Supplementary parameter(s) b =
5
sampled B =
5
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO
AECB BY: DATE:
B.7
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: COMPLIANCE CHEMISTRY INDEX (CONT'D)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per
unit)
FOR AECB USE:
a x 100 = %
1 a x 100 = %
2 a x 100 = %
3 a x 100 = %
4
A1 A2 A3 A4
a x 100 = %
5 a x 100 = %
6 a x 100 = %
7 a x 100 = %
8
A5 A6 A7 A8
a x 100 = %
9 a x 100 = %
10 a x 100 = %
11 a x 100 = %
12
A9 A10 A11 A12
a x 100 = %
13 a x 100 = %
14 a x 100 = %
15 a x 100 = %
16
A13 A14 A15 A16
a x 100 = %
17 a x 100 = %
18 a x 100 = %
19 a x 100 = %
20
A17 A18 A19 A20
(if required)
b x 100 = %
1 b x 100 = %
2 b x 100 = %
3 b x 100 = %
4
B1 B2 B3 B4
b x 100 = %
5 b x 100 = %
6 b x 100 = %
7 b x 100 = %
8
B5 B6 B7 B8
(if required) (if required)
B.8
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: COMPLIANCE CHEMISTRY INDEX (CONT'D)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per
unit)
Compliance chemistry index = Average of above percentages
= %
AECB PREPARED BY: AECB VERIFIED BY:
B.9
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: CHANGE CONTROL INDICES (4)
STATION: QUARTER, YEAR:
DATA:
QUARTER:
Parameter Unit 0 Unit 1 Unit 2 Unit 3 Unit 4
1.0 Number of temporary procedural
changes*
2.0 Number of temporary equipment
changes*
3.0 Number of incomplete permanent
equipment changes* (see spec.
sheet note number 14.2)
4.0 Number of equipment or
procedural changes over six
months old*
5.0 Number of instances where
procedural or equipment change
contributed to an event, as
recorded by the licensee
Explanation of unusually high numbers in 4.1, 4.2, 4.3, 4.4 or 4.5 is required (attach supplementary pages if
necessary):
*Note: For safety-related (as defined by the utility) systems only (see specification sheet).
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO
AECB BY: DATE:
*See specification sheet note number 14.1
B.10
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: CHANGE CONTROL INDICES (CONT'D)
STATION: QUARTER, YEAR:
FOR AECB USE:
(Develop a line chart showing items 4.1, 4.2, 4.3, 4.4 and 4.5 on a quarterly basis).
No further calculation or manipulation of data required.
AECB PREPARED BY: AECB VERIFIED BY:
B.11
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: EMERGENCY PREPAREDNESS DRILL COMPLETION RATIO (5)
STATION: QUARTER, YEAR:
DATA: Table "S": Number of drills scheduled for the year by crew and by drill type.
QTR: DRILL DRILL DRILL DRILL DRILL DRILL
TYPE 1 TYPE 2 TYPE 3 TYPE 4 TYPE 5 TYPE 6
CREW 1
CREW 2
CREW 3
CREW 4
CREW 5
Table "C": Number of drills completed quarter-end by crew and drill type.
QTR: DRILL DRILL DRILL DRILL DRILL DRILL
TYPE 1 TYPE 2 TYPE 3 TYPE 4 TYPE 5 TYPE 6
CREW 1
CREW 2
CREW 3
CREW 4
CREW 5
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE DATE:
TO AECB BY:
B.12
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: EMERGENCY PREPAREDNESS DRILL COMPLETION RATIO (CONT'D)
STATION: QUARTER, YEAR:
FOR AECB USE:
c = sum of Table "C"entries, replacing any entry by the corresponding Table "S" value
whenever the latter is smaller, unless all are smaller, without exception
=
s = sum of Table "S"entries
=
emergency preparedness drill completion ratio = c / s
=
AECB PREPARED BY: AECB VERIFIED BY:
B.13
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: EMERGENCY RESPONSE EQUIPMENT CALL-UP COMPLETION
RATIO (6)
STATION: QUARTER, YEAR:
DATA:
a = number of completed call-ups in a quarter
=
b = number of scheduled call-ups in the same quarter
=
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO
AECB BY: DATE:
FOR AECB USE:
emergency response equipment call-up completion ratio = a / b
=
AECB PREPARED BY: AECB VERIFIED BY:
B.14
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: NUMBER OF MINIMUM SHIFT COMPLEMENT NON-COMPLIANCES (7)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data sheet per unit)
DATA:
PI = Number of shifts without "minimum shift complement" during the quarter.
=
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR
RELEASE TO AECB BY: DATE:
FOR AECB USE:
AECB PREPARED BY: AECB VERIFIED BY:
B.15
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: NON-COMPLIANCE INDEX (8)
STATION: UNIT NO.: QUARTER, YEAR:
DATA: In the table below, list each event where a non-compliance has taken place of an
OP&P, RPR, other licence condition, Physical Security Regulation, Transport Packaging of
Radioactive Materials Regulations, or the Atomic Energy Control Regulations. For each event,
note which of the following non-compliance types are applicable by putting an "X" in the
appropriate column. Ensure that a separate line entry is made for any non-simultaneous
occurrences that may have transpired during the event.
a / an OP&P non-compliance at the Unit including, for multi-unit stations, any for Unit 0
(station-wide)
b / an RPR non-compliance at the Unit including, for multi-unit stations, any for Unit 0
(station-wide)
c / any other licence non-compliance at the Unit including, for multi-unit stations, any
for Unit 0 (station-wide)
d / a non-compliance at the Unit of the Physical Security Regulations, the Transport
Packaging of Radioactive Materials Regulations, or the Atomic Energy Control
Regulations including, for multi-unit stations, any for Unit 0 (station-wide)
Event a b c d
Explanation of data, if required (attach supplementary pages if necessary):
AECB PERFORMANCE INDICATOR
B.16
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
DATA SHEET
Rev. Date: 1998.06.30
TITLE: NON-COMPLIANCE INDEX (CONT'D)
STATION: UNIT NO.: QUARTER,YEAR:
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO DATE:
AECB BY:
FOR AECB USE:
non-compliance index = total number of lines in the above table where there is at least one
"X"
=
AECB PREPARED BY: AECB VERIFIED BY:
B.17
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: NUMBER OF PRESSURE BOUNDARY DEGRADATIONS (9)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per unit)
DATA:
Note: In this data sheet, PBD stands for pressure boundary degradation(s)
a = number of PBD in Class 1 nuclear systems, reported under S-99 =
N1
a = number of PBD in Class 2 nuclear systems, reported under S-99 =
N2
a = number of PBD in Class 3 nuclear systems, reported under S-99 =
N3
a = number of PBD in Class 4 nuclear systems, reported under R=99 =
N4
b = number of PBD reported under S-99 on unit conventional systems =
c = number of PBD reported under S-99 in the quarter, which were not
attributed to the operation of the station (manufacturing defects,
poor original design, etc.). However, failures or degradations due
to deficiencies in the recent (within five years) design of
modifications and repairs, as well as failures due to site
manufacturing and installation are not included in c =
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR
RELEASE TO AECB BY: DATE:
FOR AECB USE:
N = a + a
N1 + a
N2 + a
N3 + b - (b + c) =
N4
AECB PREPARED BY: AECB VERIFIED BY:
B.18
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: PREVENTIVE MAINTENANCE COMPLETION RATIO (10)
STATION: QUARTER, YEAR:
DATA:
Parameter Unit 0 Unit 1 Unit 2 Unit 3 Unit 4
(i) Total number of preventive
maintenance jobs completed by quarter.
(ii) Total number of corrective
maintenance jobs completed per
quarter.
(iii) Total number of preventive
maintenance jobs planned to be
completed in the quarter.
(iv) Total number of corrective
maintenance jobs planned to be
completed in the quarter.
(v) Total number of preventive and
corrective maintenance jobs scheduled
to be done per calendar year.
(vi) Total number of deficiency reports per
quarter.
(vii) Total number of open (unverified)
deficiency reports per quarter which are
defined as design changes.
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO
AECB BY: DATE:
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
B.19
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
TITLE: PREVENTIVE MAINTENANCE COMPLETION RATIO (CONT'D)
STATION: QUARTER, YEAR:
FOR AECB USE:
(i) x 100 = % PM jobs done per unit per quarter
(ii) + (i)
AECB PREPARED BY: AECB VERIFIED BY:
B.20
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: NON-COMPLIANCES WITH NUMBER OF SPECIFIC RADIATION PROTECTION
PROCEDURES (11)
STATION: QUARTER, YEAR:
DATA:
Provide auditable data relating to the number of instances (per quarter) for the following:
(Please include references for each event for AECB follow-up)
Location and Date
Radiation Protection Number of of Non-Compliance Comments
Procedure Non-Compliances in Each Case
(a) Radiation signposting out of date
(b) Evidence of food, drink and
cigarette consumption in zones 2
and 3
(c) Improper setup of rubber areas
(d) Whole body monitor
contamination incidents not
recorded in radiation log
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO
AECB BY: DATE:
B.21
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: NON-COMPLIANCES WITH NUMBER OF SPECIFIC RADIATION PROTECTION
PROCEDURES (CONT'D)
STATION: QUARTER, YEAR:
FOR AECB USE:
AECB PREPARED BY: AECB VERIFIED BY:
B.22
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: RADIATION OCCURRENCE INDEX (12)
STATION: UNIT: QUARTER, YEAR:
DATA:
a Event Title Dose Received Event Date Licensee's Event No.
b Event Title Dose Received Event Date Licensee's Event No.
c Event Title Dose Received Event Date Licensee's Event No.
d Event Title Dose Received Event Date Licensee's Event No.
B.23
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: RADIATION OCCURRENCE INDEX (CONT'D)
STATION: UNIT: QUARTER, YEAR:
Note: a, b, c and d are defined in section 4.
Occurrence Index = a + 5b + 5c + 50d
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO AECB DATE:
BY:
FOR AECB USE:
AECB PREPARED BY: AECB VERIFIED BY:
B.24
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: UNAVAILABILITY OF SAFETY-RELATED SYSTEMS (13)
STATION: UNIT: QUARTER, YEAR:
DATA:
Data of SDS1:
Predicted
Failed Affected Trip Actual Past Future Licensee's
Component Parameter Unavailability Unavailability* Event Date Event No.
Data of SDS2:
Predicted
Failed Affected Trip Actual Past Future Licensee's
Component Parameter Unavailability Unavailability* Event Date Event No.
*Data are required only when the actual past unavailability values are greater than design requirements.
B.25
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: UNAVAILABILITY OF SAFETY-RELATED SYSTEMS (CONT'D)
STATION: UNIT: QUARTER, YEAR:
DATA:
Data of ECCS:
Failed (or Predicted
affected) Affected Trip Actual Past Future Licensee's
Component Parameter Unavailability Unavailability* Event Date Event No.
Data of Containment:
Failed (or Predicted
affected) Affected Trip Actual Past Future Licensee's
Component Parameter Unavailability Unavailability* Event Date Event No.
*Data are required only when the actual past unavailability values are greater than design requirements.
B.26
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: UNAVAILABILITY OF SAFETY-RELATED SYSTEMS (CONT'D)
STATION: UNIT: QUARTER, YEAR:
DATA:
Data of Emergency Water System:
Failed (or Predicted
affected) Affected Trip Actual Past Future Licensee's
Component Parameter Unavailability Unavailability* Event Date Event No.
Data of Steam Generator Emergency Cooling System:
Failed (or Predicted
affected) Affected Trip Actual Past Future Licensee's
Component Parameter Unavailability Unavailability* Event Date Event No.
*Data are required only when the actual past unavailability values are greater than design requirements.
B.27
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: UNAVAILABILITY OF SAFETY-RELATED SYSTEMS (CONT'D)
STATION: UNIT: QUARTER, YEAR:
DATA:
Data of Standby Generators:
Failed (or Predicted
affected) Affected Trip Actual Past Future Licensee's
Component Parameter Unavailability Unavailability* Event Date Event No.
Data of Emergency Power System:
Failed (or Predicted
affected) Affected Trip Actual Past Future Licensee's
Component Parameter Unavailability Unavailability* Event Date Event No.
Explanation of data, if required (attach supplementary pages if necessary):
*Data are required only when the actual past unavailability values are greater than design requirements.
B.28
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: UNAVAILABILITY OF SAFETY-RELATED SYSTEMS (CONT'D)
STATION: UNIT: QUARTER, YEAR:
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO AECB DATE:
BY:
FOR AECB USE:
AECB PREPARED BY: AECB VERIFIED BY:
B.29
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: STATION WHOLE BODY DOSE (14)
STATION: YEAR:
DATA: In units of millisievert, no decimals required
WHOLE BODY
WHOLE BODY DOSES FROM PLANNED UNIT DOSES FROM
OUTAGES STATION
OPERATION STATION WHOLE
INCLUDING BODY DOSE
Unit # Unit # Unit # Unit # SHORT FORCED (EXTERNAL,
Outage Outage Outage Outage OUTAGES, INTERNAL, AND
# # # # EXCLUDING TOTAL)
PLANNED UNIT
OUTAGE OUTAGES
DURATION
(DAYS)
EXTERNAL a =
(c )i
INTERNAL b =
(d )i
SUM OF
c AND d
i i
ABOVE
Explanation of data, if required:
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO DATE:
AECB BY:
B.30
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: STATION WHOLE BODY DOSE (CONT'D)
STATION: YEAR:
FOR AECB USE:
AECB PREPARED BY: AECB VERIFIED BY:
B.31
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: NUMBER OF MISSED MANDATORY SAFETY SYSTEM TESTS (15)
STATION: UNIT NO.: QUARTER, YEAR:
DATA:
a = number of missed tests reported under 10.10 of S-99 for the SSSs of a unit plus, for
multi-unit stations, those for Unit 0 (station-wide)
=
b = number of missed tests reported under 10.10 of S-99 for the SbSSs of a unit plus,
for multi-unit stations, those for Unit 0 (station-wide)
=
c = number of missed tests reported under 10.10 of S-99 for the SrSs of a unit plus, for
multi-unit stations, those for Unit 0 (station-wide)
=
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO DATE:
AECB BY:
B.32
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: NUMBER OF MISSED MANDATORY SAFETY SYSTEM TESTS (CONT'D)
STATION: UNIT NO.: QUARTER,YEAR:
FOR AECB USE:
number of missed mandatory safety system tests = a + b + c
=
AECB PREPARED BY: AECB VERIFIED BY:
B.33
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: NUMBER OF UNPLANNED TRANSIENTS (16)
STATION: UNIT: QUARTER, YEAR:
DATA:
Total hours during which the reactor is in guaranteed shutdown state:
Data of Reactor Trips:
SDS1/SDS2
Auto or Failed Affected Trip Licensee's Event
Manual Component Parameters % Full Power Event Date No.
Data of Reactor Stepbacks:
Auto or Failed Affected Trip Licensee's Event
Manual Component Parameters % Full Power Event Date No.
Data of Reactor Setbacks:
Auto or Failed Affected Trip Licensee's Event
Manual Component Parameters % Full Power Event Date No.
Explanation of data, if required (attach supplementary pages if necessary):
B.34
C-099 (Rev. 1) (E) Reporting Requirements for Operating Nuclear Power Plants
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. date: 1998.06.30
TITLE: NUMBER OF UNPLANNED TRANSIENTS (CONT'D)
STATION: UNIT: QUARTER, YEAR:
PREPARED BY: DATE:
AUTHORIZATION FOR RELEASE TO AECB DATE:
BY:
FOR AECB USE:
AECB PREPARED BY: AECB VERIFIED BY:
B.35
Reporting Requirements for Operating Nuclear Power Plants C-099 (Rev. 1) (E)
AECB PERFORMANCE INDICATOR
DATA SHEET
Rev. Date: 1998.06.30
TITLE: UNPLANNED CAPABILITY LOSS FACTOR (17)
STATION: UNIT NO.: QUARTER, YEAR:
(Use one data set per unit)
DATA:
Licensee to submit a unit power history for the quarter, preferably as a graph, a brief description of all
energy losses (power reduction, duration in hours, reason for the reduction from the reference capacity)
and the classification of the energy loss(es) as either unplanned losses, planned losses or losses due to
"external" effects.
a = Reference Capacity = MW
b = Reference Period (Quarter) = Hours
Explanation of data, if required (attach supplementary pages if necessary):
PREPARED BY: DATE:
AUTHORIZATION FOR
RELEASE TO AECB BY: DATE:
FOR AECB USE:
REG = Reference Energy Generation = a x b = MW hours
UEL = Total Unplanned Energy Losses = MW hours
UCLF = UEL x 100 = %
REG
AECB PREPARED BY: AECB VERIFIED BY:
B.36
DRAFT
REGULATORY
GUIDE
Radiobioassay Protocols
for Responding
to Abnormal Intakes
of Radionuclides
C-147
Issued for public comments by the
Canadian Nuclear Safety Commission
August 2001
DRAFT REGULATORY GUIDE
Radiobioassay Protocols for Responding to Abnormal
Intakes of Radionuclides
C-147
Issued for public comments by the
Canadian Nuclear Safety Commission
August 2001
CNSC REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY GUIDE
Radiobioassay Protocols for Responding to Abnormal
Intakes of Radionuclides
C-147
August 2001
About this document
This draft regulatory guide describes two radiobioassay protocols that may be used by CNSC
licensees to respond to situations where persons who perform duties in connection with activities
authorized by the Nuclear Safety and Control Act and regulations may have experienced an
abnormal intake of radioactive material. The document also provides advice on how to collect and
handle radiobioassay samples.
Comments
The CNSC invites interested persons to assist in the further development of this draft regulatory
document by commenting in writing on the document's content and potential usefulness. Please
respond by November 30, 2001. Direct your written comments to the postal or e-mail address
below, referencing file 1-8-8-147.
The CNSC will take the comments received on this draft into account when developing it further.
These comments will be subject to the provisions of the federal Access to Information Act.
Document availability
This document can be viewed on the CNSC Internet site at (www.nuclearsafety.gc.ca). To order
a printed copy of the document in English or French, please contact:
Operations Assistant
Regulatory Documents Group
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 996-9505
Facsimile: (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
i
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
ACKNOWLEDGEMENTS
The Canadian Nuclear Safety Commission acknowledges the contributions of the Working Group
on Internal Dosimetry to the development of the radiobioassay protocols that are described in this
guide. The members of the Working Group are:
Dr. M. Billinghurst Winnipeg Health Sciences Centre
Mr. K. Bundy Canadian Nuclear Safety Commission
Dr. G. Kramer Health Canada
Dr. M. Limson-Zamora Health Canada
Dr. R. Richardson Atomic Energy of Canada Limited
Mr. B. Thériault Canadian Nuclear Safety Commission
Dr. K. Thind Consultant
Dr. C. Webber Hamilton Health Sciences Corporation
Dr. D. Whillans Ontario Power Generation
ii
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
CONTENTS
About this document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
ACKNOWLEDGEMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory framework . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Relevant legislation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.3 Radiobioassay methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1.4 "Routine" and "non-routine" radiobioassays . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1.5 Selecting and applying radiobioassay methods . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2 Response Protocols . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
2.1 Application . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
2.2 Alignment with ICRP recommendations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
2.3 Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
2.3.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
2.3.2 A response protocol triggered by a routine radiobioassay . . . . . . . . . . . . . . . . . . 7
2.3.3 A response protocol triggered by an abnormal incident. . . . . . . . . . . . . . . . . . . . 9
3 Collecting and Handling Radiobioassay Samples. . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3.1 General rules . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3.2 Labelling samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3.3 Treating and storing urine samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3.4 Collecting and storing faecal samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.5 Packaging and transporting radiobioassay samples . . . . . . . . . . . . . . . . . . . . . 12
References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
iii
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
Appendices . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
Table 1: Conducting Radiobioassays When the Preliminary Estimate of Intake is
Equal To or Greater Than an Action Level . . . . . . . . . . . . . . . . . . . . . A.1
Figure 1: A Typical Schedule for Sampling Faecal Excretions that May Contain
Type S Compounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.2
Figure 2: A Typical Schedule for Sampling Urine Excretions that May Contain
Type F Compounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.3
Figure 3: Events Leading to Non-routine Radiobioassays in Response to the Results
of Routine Radiobioassays . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.4
Figure 4: Events Leading to Non-routine Radiobioassays in Response to an
Abnormal Incident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.5
iv
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
Purpose
This Regulatory Guide is intended to help licensees of the Canadian Nuclear Safety Commission
(CNSC, Commission) ascertain and control radiation exposures and doses to workers in
accordance with regulatory requirements, including the Radiation Protection Regulations and any
relevant licence conditions.
Scope
This guide:
* describes two radiobioassay protocols that may be used by CNSC licensees to respond
to situations where persons who perform duties in connection with activities authorized
by the Nuclear Safety and Control Act and regulations may have experienced an
abnormal intake of radioactive material; and
* provides advice on how to collect and handle radiobioassay samples.
1 Background
1.1 Regulatory framework
The CNSC is the federal agency that regulates the use of nuclear energy and materials to
protect health, safety, security and the environment, and to respect Canada's international
commitments on the peaceful use of nuclear energy.
The NSC Act requires persons or organizations to be licensed by the CNSC for carrying out
the activities referred to in section 26 of the Act, unless otherwise exempted. The associated
regulations stipulate prerequisites for CNSC licensing, and the obligations of licensees and
workers.
1.2 Relevant legislation
The General Nuclear Safety and Control Regulations and the Radiation Protection
Regulations contain provisions that are relevant to this guide.
In particular:
* Paragraph 3(1)(e) of the General Nuclear Safety and Control Regulations stipulates
that an application for a CNSC licence shall contain the proposed measures to ensure
compliance with the Radiation Protection Regulations.
* Section 5 of the Radiation Protection Regulations stipulates:
- "For the purpose of keeping a record of doses of radiation in accordance with section
27 of the Act, every licensee shall ascertain and record the amount of exposure to
radon progeny of each person referred to in that section, as well as the effective dose
and equivalent dose received by and committed to that person."[Subsection 5(1)]
1
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
- "A licensee shall ascertain the amount of exposure to radon progeny and the effective
dose and equivalent dose (a) by direct measurement as a result of monitoring; or (b)
if the time and resources required for direct measurement as a result of monitoring
outweigh the usefulness of ascertaining the amount and doses using that method, by
estimating them."[Subsection 5(2)]
* Section 6 of the Radiation Protection Regulations stipulates:
- "In this section, `action level' means a specific dose of radiation or other parameter
that, if reached, may indicate a loss of control of part of a licensee's radiation
protection program and triggers a requirement for specific action to be
taken."[Subsection 6(1)]
- "When a licensee becomes aware that an action level referred to in the licence for the
purpose of this subsection has been reached, the licensee shall
(a) conduct an investigation to establish the cause for reaching the action level;
(b) identify and take action to restore the effectiveness of the radiation protection
program implemented in accordance with section 4; and
(c) notify the Commission within the period specified in the licence".
[Subsection 6(2)]
* Paragraph 16(c) of the Radiation Protection Regulations stipulates:
- "When a licensee becomes aware that a dose of radiation received by and committed
to a person or an organ or tissue may have exceeded an applicable dose limit
prescribed by section 13, 14 or 15, the licensee shall conduct an investigation to
determine the magnitude of the dose and to establish the causes of the exposure."
The Radiation Protection Regulations do not stipulate how CNSC licensees are to ascertain
exposures and doses to persons by "direct measurement as a result of monitoring". In the
absence of such direction, section 2 of this guide describes two response protocols that
could be used by CNSC licensees to ascertain radiation doses to individuals by non-routine
radiobioassays when abnormal intakes of radionuclides are known or suspected to have
occurred.
1.3 Radiobioassay methods
The radiation protection programs that the Radiation Protection Regulations require of
CNSC licensees will typically include provisions for radiobioassays. These radiobioassays
may be "direct" or "indirect".
A "direct" (or " in vivo") radiobioassay is a measurement on the human body for the
purpose of determining the amount of radioactive material in the body, utilizing
instrumentation that detects the radiation emitted from the radioactive material.
2
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
An "indirect" (or "in vitro") radiobioassay consists of the collection and analysis of a
sample of human hair, tissue, nasal fluid, urine or faeces for the purpose of determining the
amount of radioactive material that might have been taken into the body.
1.4 "Routine" and "non-routine" radiobioassays
Direct and indirect radiobioassays to ascertain radiation doses may be further characterized
as "routine" or "non-routine", as follows:
* a "routine" radiobioassay is any radiobioassay that involves collecting and analysing
samples or taking measurements on the body at scheduled intervals, or at predetermined
times, during normal operations.
* a "non-routine" radiobioassay is any radiobioassay that is implemented as part of an ad
hoc response to a particular circumstance, such as a known or suspected intake of
radioactive material due to an abnormal incident in the workplace. "Non-routine"
radiobioassays are often termed "ad hoc" or "special" radiobioassays.
By definition, a dose monitoring program that includes routine radiobioassays is pro-active
and precautionary in nature. Typically, such a program is intended to provide routine and
timely detection, measurement and confirmation of any radioactive intakes that occur on an
on-going basis during normal operations. An example of a routine radiobioassay is the
submission of a biweekly (every 14 days) urine sample for analyses for the presence of
tritiated water.
A monitoring program that consists only of non-routine radiobioassays is typically reactive
and ad hoc in nature. Such a program is usually custom-designed for the purpose of
obtaining key parameters that are necessary in order to conduct a specific dose assessment
in response to a specific, identified need. To avoid prejudicing the results, a non-routine
radiobioassay is typically performed with the subject individual removed from further
contact with, or exposure to, radioactive substances.
Both routine and non-routine radiobioassays may involve one or more of the radiobioassay
methods described in publications 54 and 78 of the International Commission on
Radiological Protection (ICRP), or as otherwise determined appropriate to address case-
specific needs.
1.5 Selecting and applying radiobioassay methods
In situations where ad hoc response protocols are implemented, the associated program for
radiobioassays and analyses will typically depend upon case-specific factors, including:
* the time of intake of the radioactive contaminant(s);
* the mode of intake of the radioactive contaminant(s);
3
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
* the preliminary assessment of the radioactive intake and resulting dose, using the
precipitating radiobioassay result and default parameters;
* whether the radiation is due to a single radionuclide or a mixture of radionuclides;
* the chemical and physical forms (e.g., particle size) of the radioactive contaminant(s);
* the types and magnitudes of the radiation emitted by the contaminant(s);
* the rate of decay of the radioactive contaminant(s);
* the metabolic characteristics and behaviour of each suspected radioactive contaminant
(e.g., retention time within the body, lung-solubility class, gut-transfer factor, rate of
excretion);
* when the radiobioassay results must be available;
* the number of radiobioassay results required; and
* the convenience, sensitivity, quality and suitability of the available radiobioassay
equipment and facilities.
To assess radiation doses from internal sources, the commonly used radiobioassay methods
are in vivo counting, and the analysis of collected samples of excreta, such as urine and
faeces. In some situations, such as those involving radionuclides with no gamma-ray
emissions or only low energy photon emissions, radiobioassays of excreta may be the only
reasonable option.
Where a person may have been internally exposed to a mixture of radionuclides that emits
penetrating gamma photons, a combination of in-vivo counting, and the collection and
analysis of excreta may be useful. With appropriate adjustments, metabolic data - such as
organ or whole-body retention times, and urine and faecal excretion rates - can often be
used along with the results of radiobioassays to ascertain the radiation exposures and doses
from radioactive intakes.
For individuals that inhale or ingest a mixture of radionuclides, standard radiobioassay
methods may not be able to detect all radionuclides in the mixture. However, if certain
radionuclides are detected by the standard methods, other radionuclides that are normally
associated with the detected radionuclides will also be present. For example, 144Ce is
typically associated with the transuranic radionuclides formed in irradiated uranium fuel.
Accordingly, when the detection of a surrogate radionuclide, such as 144Ce, by a standard
radiobioassay method (e.g., in vivo counting or gamma spectroscopy on faecal samples,
using germanium detectors) indicates that a more difficult-to-detect radionuclide may also
be present, supplementary in vivo counting techniques or special analyses of radiobioassay
samples (e.g., fission track analysis for 239Pu in urine) may be necessary.
4
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
2 Response Protocols
2.1 Application
The response protocols that are described in this section are intended to be used in non-
routine situations, such as when it is necessary to ascertain the committed effective dose to
an individual following a suspected or actual intake of radioactive substances due to the
occurrence of an abnormal incident in the workplace. Such scenarios include:
* a breach or failure of a sealed source that could result in the intake of a radioactive
substance;
* abnormal intakes of radioactive materials when handling unsealed radioactive sources;
* the failure of personnel protection (e.g., respiratory equipment) during the maintenance
or servicing of contaminated equipment or systems; and
* accidents involving air-borne contamination (e.g., fires and explosions).
Where an incident that occurs at a CNSC-licensed facility or during a CNSC-regulated
activity involves a possible or actual internal radiation exposure, the incident could trigger a
requirement for non-routine radiobioassays in accordance with regulations, a CNSC licence
or a licensee's radiation protection program.
For example, when a licensee becomes aware that a dose to a person or an organ or a tissue
may have exceeded an applicable dose limit prescribed by sections 13, 14 or 15 of the
Radiation Protection Regulations, the licensee must conduct an investigation in order to
determine the magnitude of the dose and to establish the causes of the exposure (Paragraph
16(c) of the Radiation Protection Regulations).
Similarly, section 6 of the CNSC Radiation Protection Regulations requires an
investigation when an "action level" in a licence is reached, and requires the licensee to take
any necessary corrective or remedial measures. Although section 6 does not refer directly to
monitoring, an investigation into the cause or magnitude of a radiation exposure incident
could involve some form of supporting monitoring and assessment. Accordingly, a
licensee's radiation protection program should provide for appropriate responses, including
routine and non-routine radiobioassays, to both normal and abnormal situations involving
radiation exposures. If an "action level" in a licence is reached, and the licensee suspects
that someone may have experienced a significant intake of radionuclides, the licensee should
typically investigate to determine the magnitude of the dose and related circumstances.
If the results of non-routine radiobioassays are to be credible, care should be paid to details
such as the choice and application of assay methods, the timing and number of in vivo
counts, or the timing of the collection of excreta samples relative to the time of intake of
radiation. Accordingly, persons who are responsible for designing and implementing
response protocols must exercise competent judgement on key matters. For example, they
must decide whether to collect and retain samples for confirmatory analyses, they must
select appropriate times for truncation of sampling, and they must weigh and balance the
5
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
associated advantages and disadvantages.
When selecting the preferred radiobioassay methods and identifying any complementary
requirements for additional biological monitoring, the responsible persons should take into
account the factors discussed in section 1.5 above.
The response protocols that are described in subsection 2.3 of this guide can be used by
licensees to ascertain the committed effective dose resulting from an intake of radionuclides.
However, users should exercise sound judgement. They should adjust and refine the
recommended protocols to suit their specific needs and individual circumstances. These
needs and circumstances will typically depend upon case-specific factors, including radiation
hazards in the workplaces and the circumstances associated with the internal exposures.
2.2 Alignment with ICRP recommendations
The response protocols described in this guide are similar to those recommended in relevant
annals of the ICRP. For example, "ICRP Publication 54" and "ICRP Publication 78"
recommend, for a worker that intakes a significant quantity of radionuclides, that the
workers's retention or excretion patterns and other relevant parameters be estimated, using
individual-specific data. The ICRP prefers the use of individual-specific data over the
application of standardized bio-kinetic models in such evaluations, because observed rates
are typically more realistic than default values, and, thus, more likely to result in more
realistic estimates of the associated radiation doses.
ICRP recommendations and the radiobioassay protocols described in this guide recognize
that an individual's rate of retention or excretion cannot be adequately constructed on the
basis of two or three randomly collected measurements or samples. Accordingly, these
recommendations and protocols are designed to systematically yield individual-specific
radiobioassay data that are sufficient to generate scientifically defensible dose assessments.
2.3 Description
2.3.1 Overview
The response protocols that are described below in subsections 2.3.3 and 2.3.2, and
illustrated graphically in Figures 3 and 4 on pages A.4 and A.5 of this guide, address
two specific situations. These situations are:
* when a routine radiobioassay program yields a result that is abnormal, and that
indicates that a person may have been exposed to abnormal levels of radiation;
and
6
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
* when the occurrence of an abnormal incident (e.g., fire, explosion, failure of
ventilation systems) that has a recognized potential to give rise to significant
intakes of radionuclides by an affected person is known or suspected.
Figures 1 and 2 on pages A.2 and A.3 of this guide illustrate excretion-sampling
patterns that could be appropriate in specific situations. These figures take into
account the anatomical and physiological characteristics defined by the ICRP for
reference individuals in ICRP Publication 23.
Figure 1 shows a typical schedule for sampling faecal excretions that may contain a
"Type S compound". A "Type S compound" is a compound that is relatively insoluble
in the human respiratory tract, and thus only absorbed into the blood of the tract
slowly.
Figure 2 shows a typical schedule for sampling urine excretions that may contain a
"Type F compound". A "Type F compound" is a compound that is relatively soluble
in the human respiratory tract, and thus readily absorbed into the blood of the tract.
2.3.2 A response protocol triggered by a routine radiobioassay
Removing workers, confirming radiobioassay results and retaining samples
As the first phase of this response protocol:
* Remove the exposed individual from any possibility of further intake.
* Confirm the precipitating radiobioassay result as soon as practical, using the
laboratory that performed the analysis.
* If the precipitating result is unusually high (i.e., much greater than a relevant
"action level", as defined in subsection 6.(1) of the CNSC Radiation Protection
Regulations consider confirming it with a laboratory that is independent of the
laboratory that obtained the precipitating result.
* Where possible, retain the precipitating sample until all investigations associated
with the incident are complete.
Depending on circumstances, it may or may not be possible or practical to retain
samples or to repeat sample analyses in accordance with the above recommendations.
It may be most practical to retain samples for repeat analyses when only a small
portion of the total collection is required for each analysis. For example, each analysis
of tritium in tritiated water in urine may require only 5 mL of sample per analysis,
whereas the total volume of urine collected may range from 0.1 to 1 L.
7
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
When routine radiobioassays yield elevated results, it may be prudent, where possible,
to save the unused portion of the collected urine or faecal sample for further analyses.
For example, when non-destructive gamma spectroscopy of a faeces sample indicates
significantly elevated levels of fission and activation products, the sample can be
ashed, and a portion of the resulting ash saved for confirmatory or additional analyses.
In some situations, it may be possible and worthwhile to retain unused samples for
extended periods of time in case improved analytical technique and equipment
become available over the storage period, enabling more sensitive or more accurate
determinations of results.
Section 3 of this guide provides guidance on how to handle and store radiobioassay
samples.
Determining and implementing requirements for further radiobioassays
When determining the need for further radiobioassays in accordance with this
protocol, the responsible persons should consider relevant factors, such as those
discussed in section 1.5 of this guide.
Table 1 on page A.1 of this guide recommends sampling frequencies for three
contiguous periods (1-10 days, 10-100 days, and over 100 days) following a
preliminary estimate of a radioactive intake that equals or exceeds a relevant "action
level", as defined in subsection 6.(1) of the CNSC Radiation Protection Regulations.
The recommended sampling protocol applies over the time period during which the
worker is removed from further work involving radiation or radioactive substances in
the workplace.
For radionuclides that decay quickly (i.e., half-life < 3 days) the effective available
sampling period may be relatively short, and consequently the sampling regime
recommended in Table 1 may not be entirely appropriate. Similar considerations apply
for long-lived radionuclides incorporated into compounds having short biological
half-lives (< 3 days). For these situations, sample daily during the first 10 days after
intake to accurately define the shape of the individual's retention and/or excretion
curves.
Less frequent sampling than recommended in Table 1 could result in insufficient data
for reliable assessments of the associated radiation doses. Accordingly, to provide
better estimates of radioactive intake over periods of less-frequent sampling, urine
samples should be collected, as well as samples of faeces where such are likely to be
useful for dose-assessment purposes (e.g., after intakes of insoluble radioactive
compounds).
8
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
Following a preliminary estimate of a radioactive intake that equals or exceeds a
relevant "action level", as defined in subsection 6.(1) of the CNSC Radiation
Protection Regulations, the exposed individual(s) should be instructed to collect and
submit the recommended samples, and to be available for in vivo counting as per the
frequency recommended for this period.
Where a preliminary dose assessment (i.e., on the basis of the initial few sample
results) indicates that the associated intake may be much greater than a relevant
"action level", as defined in subsection 6.(1) of the CNSC Radiation Protection
Regulations, arrange for supplementary biological sampling (e.g., blood, saliva,
breath), and for sampling at the frequencies recommended in Table 1. These
frequencies are: daily for the first 10 days after intake, every 14 days from 10 to 100
days after intake, and every 30 days thereafter until such time as the worker is
authorized to resume work involving radiation or nuclear substances in the
workplace.
2.3.3 A response protocol triggered by an abnormal incident
Application of this response protocol
This protocol is intended to be implemented in response to abnormal incidents in the
workplace, such as accidents involving fire, explosions, or failure of ventilation
systems. Abnormal incidents typically increase the levels of airborne radioactivity, and
can result in increased intakes of radioactive materials by affected workers.
Assessing whether and when an intake may have occurred
Because incidents such as fires, explosions or ventilation failures may be self-evident
or typically trigger protective alarms or monitors, their times of occurrence are
usually well-known. When this is so, the time of any associated intake of radioactive
contaminants by workers can be established with similar accuracy.
Following an abnormal incident at a nuclear facility, an increased intake by workers
may be suspected from indirect evidence. For example, the detection of facial or nasal
contamination by portal monitors or hand-held detectors, or the presence of surface
cuts or sores that are radioactively contaminated, may indicate that the individuals has
been subjected to airborne contamination. Such evidence can be sufficient reason to
immediately initiate a non-routine radiobioassay, instead of awaiting the results of
routine radiobioassay monitoring.
Where an intake of radioactive contaminants is suspected but not confirmed, the
timely collection of non-radiobioassay and radiobioassay samples may help establish
whether or not such an incident has occurred.
9
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
Non-radiobioassay samples include swabs of nasal fluid, and surface wipes of
protective clothing or workplace surfaces. Either the presence of radionuclides in
such media, or the lack thereof, can serve as a reasonable indicator of whether an
inhalation incident has occurred.
Collecting samples and confirming results
* Arrange for timely collection of radiobioassay samples and in vivo counting from
the exposed individuals, since the initial results of radiobioassay or in vivo
monitoring will influence decisions about further sampling.
* If widespread contamination is present, take particular care to obtain
uncontaminated radiobioassay samples. (See section 3 below).
* While awaiting the results of the initial radiobioassay sampling and in vivo
monitoring, continue to sample at the frequencies recommended in Table 1 on
page A.1.
* Review the results of the first in vitro radiobioassay or in vivo count results, and
compare them with the results of available non-radiobioassay sample, such as
nasal swabs or surface contamination wipe samples.
* If the radionuclides detected in the nasal swab and workplace wipe samples are
the same as those reported in the radiobioassay results, consider the agreement to
be a confirmation that a corresponding radioactive intake has occurred.
* Perform preliminary assessments of intake and dose using the results of the initial
radiobioassays results, taking factors such as those presented in section 1.5 of this
guide into account.
Possible responses to the preliminary assessments of intake and dose
* If the estimated intake or dose is less than a relevant "action level", as defined in
subsection 6.(1) of the CNSC Radiation Protection Regulations, adjust the
protocols recommended in Table 1 accordingly. Since the intake is relatively low,
consider ending sampling much sooner.
* If the estimated intake or dose is equal to or greater than a relevant "action level",
as defined in subsection 6.(1) of the CNSC Radiation Protection Regulations,
follow the protocols recommended in Table 1. If appropriate, arrange for
additional specialized analyses and biological monitoring, as discussed in sections
1.5 and 2.1, respectively.
10
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
3 Collecting and Handling Radiobioassay Samples
3.1 General rules
* Ensure that all persons who handle radiobioassay samples are properly instructed
in the safe handling of biological and radioactive specimens.
* Use the services of a qualified medical agency or practitioner to collect blood
samples.
* When collecting samples in restricted zones, work in areas where the chances of
contamination are lowest.
* Collect samples in sterile disposable containers.
* When faecal samples are required from subjects, the subjects should be instructed
not to contaminate the samples with urine.
* After collecting samples in a workplace, shower or wash your hands carefully
before removing the samples to an unrestricted area.
3.2 Labelling samples
* After collecting a radiobioassay sample from a person, label the sample container
with the name or identification number of the person, and the date and time of
sample collection.
3.3 Treating and storing urine samples
* After collecting urine samples, immediately place them in cold storage, such as in
a refrigerator or cooler.
* Urine samples that are to be stored for more than a day should be acidified or
otherwise treated (taking into account the characteristics and chemical forms of
the radionuclides present in the excretion) in order to prevent or minimize losses
to container walls. Samples containing fission and activation products may
especially warrant such precautions
* If necessary to prevent bacterial growth and precipitation, treat urine samples with
an appropriate preservative such as thymol. Urine samples that contain tritium or
carbon-14 may require such treatment.
* Where appropriate, urine samples may be preserved by freezing. Freezing may be
particularly appropriate and convenient for urine samples containing organically-
bound tritium, particularly those which may require a repeat or confirmatory
analysis for organically-bound tritium. Where possible, retain part of the collected
samples for repeat or multiple analyses.
11
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
3.4 Collecting and storing faecal samples
* To collect faecal samples, use specialty kits (e.g., a "commode specimen
collection system.") that are designed for the purpose, and readily available from
commercial medical equipment suppliers.
* Upon receiving faecal samples, freeze them immediately.
3.5 Packaging and transporting radiobioassay samples
* To prevent discharge, emission or loss of radiobioassay samples during transport,
package them securely in accordance with paragraph 2.3.3(b) of the
Transportation of Dangerous Goods Regulations. Pay particular attention to the
packaging of liquids and fluid samples.
* If liquid radiobioassay samples are to be more than two hours in transport,
package them in a cooler containing dry ice.
* Maintain faecal samples in a frozen state during transport.
12
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
REFERENCES
General Guidelines for Bioassay Programs, Bioassay Guideline 1, Health Canada Report
81-EHD-56.
Individual Monitoring for Intakes of Radionuclides by Workers: Design and Interpretation,
ICRP Publication 54, 1988.
Individual Monitoring for Internal Exposure of Workers, ICRP Publication 78, 1997.
Report of the Task Group on Reference Man, ICRP Publication 23, 1975.
13
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
APPENDICES
Table 1: Conducting Radiobioassays When the Preliminary Estimate of Intake is Equal To
or Greater Than an Action Level
Figure 1: A Typical Schedule for Sampling Faecal Excretions that May contain Type S
Compounds
Figure 2: A Typical Schedule for Sampling Urine Excretions that May Contain Type F
Compounds
Figure 3: Events Leading to Non-routine Radiobioassays in Response to the Results of
Routine Radiobioassays
Figure 4: Events Leading to Non-routine Radiobioassays in Response to an Abnormal
Incident
14
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
Table 1: Conducting Radiobioassays When the Preliminary Estimate of Intake is
Equal To or Greater Than an Action Level
Period After Urine Sampling Faecal Sampling In Vivo Count Comments
Radioactive Frequency Frequency Frequency
Intake
1-10 days Collect a 24-hour Collect a 24-hour Perform in vivo End sampling and/or in
urine sample each faecal sample each day. counting each day. vivo counting when
day. results fall below
detection limits or reach
chronic baseline values.
10-100 days Collect a 24-hour Collect 24-hour faecal Perform in vivo End sampling or in vivo
urine sample samples on 3 counting every 14 counting when results fall
every 14 days. consecutive days. days. below detection limits or
Repeat the collection reach chronic baseline
program every 14 days. values.
More than Collect a 24-hour Collect 24-hour faecal Perform in vivo End sampling and/or in
100 days urine sample samples on 3 counting every 30 vivo counting when
every 30 days. consecutive days. days. results fall below
Repeat collection detection limits or reach
program every 30 days. chronic baseline values.
Notes:
1. The radiobioassay schedule recommended in Table 1 above should typically be followed until
the subject returns to work. However, users may need to modify the protocol to take into
account individual circumstances. For example, the three discrete sampling periods shown in
this table may not be appropriate for radionuclides with half-lives < 3 days, for long-lived
radionuclides in chemical form that have biological half-lives < 3 days, and when the intake is
above a detection limit but below an action level.
2. In Table 1 above, a "24-hour" sample is a sample integrated over 24 consecutive hours.
A.1
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
Figure 1: A Typical Schedule for Sampling Faecal Excretions that May Contain Type S
Compounds
10 1
Early
10 0
10 -1
10 -2
10 -3 Intermediate
10 -4 Late
Daily Biweekly
10 -5 sampling sampling Monthly
sampling
10 -610 -1 10 0 10 1 10 2 10 3 10 4
Time (days)
A.2
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
Figure 2: A Typical Schedule for Sampling Urine Excretions that May Contain Type F
Compounds
10 0
Early
10 -2
Late
10 -4
10 -6
Daily Biweekly
sampling sampling
10 -8
10 0 10 1 10 2 10 3
Time (days)
A.3
Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides C-147
Figure 3: Events Leading to Non-routine Radiobioassays in Response to the Results of
Routine Radiobioassays
Intake may have occured
in this interval
Routine bioassay results
may trigger non-routine
bioassays
Sampling intervals for
routine bioassays Time
Lapsed time may be 24 to
48 hours due to the time
required for analyzing and
reporting routine bioassays,
and for locating and Start of non-routine
contacting individuals bioassays
A.4
C-147 Radiobioassay Protocols for Responding to Abnormal Intakes of Radionuclides
Figure 4: Events Leading to Non-routine Radiobioassays in Response to an Abnormal
Incident
Start of non - routine
bioassays
Time of abnormal
incident
Time
Lapsed time may be 2 to 24 hours
due to time required for preliminary
evaluation of incident and any
decontamination or mitigation
needed
A.5
PROPOSED
REGULATORY
GUIDE
C-200 (E)
RADIATION SAFETY TRAINING
FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND
TRANSPORTATION WORKERS
Consultative Document
Published by the
Atomic Energy Control Board
(October 15, 1998)
Atomic Energy Commission de contrôle
Control Board de l'énergie atomique
PROPOSED
REGULATORY GUIDE
C-200 (E)
RADIATION SAFETY TRAINING
FOR RADIOISOTOPE, MEDICAL ACCELERATOR
AND TRANSPORTATION WORKERS
Consultative Document
Published by the
Atomic Energy Control Board
(October 15, 1998)
Preface
The Atomic Energy Control Board (AECB) has approved the release of C-200 (E), a proposed
regulatory guide, for trial use over a period of one year as part of its public consultation program.
The consultation trial use period of one year is beneficial to the development of working
experience by both the AECB and the licensees. The approach allows for a more generous
consultation period.
The NSC Act and pursuant regulations will come into effect during the trial use period. The
AECB will become Canadian Nuclear Safety Commission (CNSC). The language used in C-
200 (E) reflects the philosophy of the NSC Act and the proposed new regulations.
A letter will be sent at the end of the trial use period during November 1999 to all licensees and
public participants asking to provide comments. CNSC staff will provide input based on their
experience.
You are invited to provide comments and suggestions during or immediately following the trial
use period. Please direct comments to:
Operations Assistant
Corporate Documents Section
Atomic Energy Control Board
P.O. Box 1046, Station B
Ottawa ON K1P 5S9
or
e-mail: reg@atomcon.gc.ca
Notice
On March 20, 1997, Bill C-23, the Nuclear Safety and Control Act, received Royal Assent.
When this Guide was published it was not yet in force. When the NSC Act and regulations come
into force, this document will be revised. Proposed Regulatory Guide C-200 (E) references the
existing Atomic Energy Control Act and its Regulations, which remain in force until further
notice.
Table of Contents
Page
About this Regulatory Guide i
How to use this guide ii
The AECB, Nuclear Legislation and Regulatory Process ii
Background iii
Abbreviations used in this Guide v
Section 1: Introduction
Introduction 1.1
Application 1.1
Training is a regulatory requirement 1.2
Developing a safety culture 1.2
Program design 1.3
Section 2: Content and format
Program content 2.1
Learning objectives 2.1
Topics to be covered: 2.1
Worker classification and required training 2.2
Formats to provide the information 2.2
Statements on performance, conditions and standards 2.3
Section 3: Workplace Hazardous Materials Information System [WHMIS]
About WHMIS 3.1
Guidance on basic WHMIS-equivalent training 3.2
Section 4: Induction training, basic training and retraining
Induction training 4.1
Basic training 4.2
Retraining 4.3
Training of Supervisors 4.3
Section 5: Administering training programs
Keeping records 5.1
Training program records 5.1
Certification 5.1
Scheduling 5.2
Section 6: Selecting and training instructors
Instructor qualifications 6.1
Instructor training 6.1
Learning objectives 6.2
Section 7: Program verification
Evaluation and validation 7.1
Objectives and criteria for evaluating training programs 7.1
Preparation of the radiation training program 7.2
Implementing the training program 7.5
Evaluating the training program 7.7
Table of Contents (continued)
Page
Section 8: AECB assessment of radiation safety training programs
AECB assessment of radiation safety training programs 8.1
Section 9: Radioisotopes
Radioisotope licensees 9.1
Radiation safety training modules for those dealing with
radioisotopes (reference) 9.1
Section 10: Particle accelerators
Particle accelerator licensees 10.1
Radiation safety training modules for particle accelerator
licensees (reference) 10.2
Section 11: Transportation of radioactive nuclear substances
Transportation of radioactive materials 11.1
Radiation safety training modules for transportation of
radioactive nuclear materials (reference) 11.2
Section 12: Radiation safety training modules
Modules 1 to 13 12.1
Section 13: Recommended matrix tables for worker training programs:
Table 13-1: Radioisotope worker training matrix
Table 13-2: Medical accelerator training matrix
Table 13-3: Transportation training matrix
Section 14: Review objectives
Review objectives 14.1
Appendices
Appendix 1 Systems Approach to Training (SAT) Diagram A1.1
Appendix 2 Training Evaluation Form (example) A2.1
Training Delivery Evaluation Form A2.4
Appendix 3 Validation of Radiation Training Form (examples) A3.1
Appendix 4 Examples of Schedules for Training A4.1
Appendix 5 Radiation Protection Training Program Validation: Summary of
recommended topics A5.1
Appendix 6 Radiation Safety Information Sheets A6.1
Appendix 7 Training record A7.1
Appendix 8 Radiation Safety Data Sheet A8.1
Appendix 9 References A9.1
Radiation Safety Training
About this Regulatory Guide
The AECB provides instruction, assistance and information in Regulatory Guidance
Documents (RGDs). These documents are intended to guide the conduct of anyone
subject to regulatory requirements, such as AECB licensees, and others involved with
the AECB's regulatory process.
This Regulatory Guide helps licensees develop and implement radiation safety training
programs for workers and supervisors that are effective and are acceptable to the
AECB.
The information in this Guide is intended for current AECB licensees, licence
1
applicants and radiation safety instructors. In addition to sections describing the
general requirements for radiation safety training, specialized sections are provided
for:
C radioisotope licensees using open source radioactive material, sealed sources and
devices containing radioactive prescribed substances
C Medical accelerator licensees
C transportation and packaging operators
C radiation safety instructors
If you fit one of the above categories, you should read the appropriate sections of the
Guide, and retain the Guide for reference. The following chart will assist you in
determining which sections you should read.
The training for a qualified operator (QO) or any other person who is helping with industrial radiography is
1
not addressed in this guide. There is a separate AECB program for industrial radiography safety training.
i
Radiation Safety Training
How to use this Guide
If you are a... you should read Sections... and Appendixes...
Radioisotope licensee or 1 to 9 1 to 9
applicant
Radioisotope safety training 1 to 9, 12, 13, 14 1, 2, 3, 4, 6, 7, 8, 9
instructor
Medical accelerator licensee or 1 to 8 plus 10 1, 2, 3, 4, 5, 7, 8,
applicant
Medical accelerator radiation 1, 2, 4, 5, 6, 8, 10, 12, 1, 2, 3, 4, 7, 8, 9
safety instructor 13, 14
Transportation and packaging 1 to 5, plus 8, 11, 12, 1, 2, 3, 4, 5, 7, 8, 9
personnel 13, 14
The AECB, Nuclear Legislation and Regulatory Process
The AECB is a federal regulatory agency with responsibilities for radiation safety. It
ensures that nuclear facilities and the use of nuclear materials do not pose undue risk to
health, safety, security, and the environment.
At present, the AECB operates under the authority of the Atomic Energy Control (AEC)
Act and related regulations. A new Act, the Nuclear Safety and Control (NSC) Act,
received Royal Assent on March 20, 1997, and will be implemented on a date to be fixed
by order of the Governor in Council. The accompanying regulations are being developed
in accordance with federal policies and process, which include consultations with industry
and the public.
The NSC Act will replace the AEC Act. However, until the new Act and regulations are
proclaimed, the existing legislation continues to apply. Accordingly, this Consultative
Document pertains to activities conducted under the authority of the AEC Act and
regulations.
The AECB will consider the comments received in response to this document for purposes
of finalizing or revising this draft Guide.
ii
Radiation Safety Training
Background
All radioisotope licences issued by the AECB include a condition requiring licensees to
ensure that only persons properly trained and informed of the hazards are allowed to
handle radioactive materials. This licence condition has been a requirement for many years.
Moreover, the process of applying for many types of radioisotope licence requires that the
applicant submit a description of the proposed radiation safety training program. AECB
radioisotope licensing staff review these training program descriptions, provide feedback
to licensees about their programs, and require that training be improved where necessary.
Similarly, particle accelerator operating licences require that licensees train every person
who works in the facility in radiation safety.
AECB radioisotope licensing staff prepared a brief description of the standards for
radiation safety training in Consultative Document C-121 "Requirements for a Radiation
Safety Program for Consolidated Radioisotope Licensees" (August, 1992). C-121 was
directed at institutions holding consolidated radioisotope licences, mainly universities. The
document was later revised and the training standards were removed because they
belonged in a stand-alone document which should apply to the radiation safety training of
workers, including supervisors, for all activities licensed or regulated through MRD by the
AECB. This Guide fulfills that need for a stand-alone document.
Further impetus for the development of a regulatory guide describing radiation safety
training was provided by an article in the April, 1995, issue of the Canadian Radiation
Protection Association (CRPA) Bulletin. The article described the results of a survey of
radiation safety training programs at institutions holding consolidated radioisotope
licences. It concluded that radiation safety training is varied and further highlighted the
need for a comprehensive regulatory guide.
Seeking input from the regulated community, AECB staff consulted more than 70
radiation safety officers (RSOs) about specific radiation safety training requirements for
workers. RSOs were consulted at a workshop held during the CRPA meeting in Victoria
in May of 1997. Feedback from workshop participants has been incorporated into the
training matrices in this Guide.
The guidelines are not mandatory. However, they do clearly indicate what the AECB
expects to see in training programs required to fulfill licensing requirements, as well as the
methods for administering and maintaining those programs. Conformance with the
guidelines will expedite the licence assessment process.
iii
Radiation Safety Training
In preparing a set of general guidelines it is simply not practical to prepare a separate
guideline that addresses all the training necessary for each specific job. Accordingly, these
guidelines are general in nature and should not be assumed to be comprehensive. The
intent is to provide a foundation for developing a training program that is acceptable to the
AECB.
Licensees should not regard the guidelines as a replacement for the proper identification of
the specific training needs of their staff and the development of training programs in
radiation safety that meet those needs. It remains the responsibility of the licensee to
ensure that their training programs meet the particular needs of both the job and the
individual worker.
iv
Radiation Safety Training
Abbreviations used in this Guide
µSv microsievert
AEC Act Atomic Energy Control Act
AEC Regulations Atomic Energy Control Regulations
AECB Atomic Energy Control Board
ALARA as low as reasonably achievable
CNSC Canadian Nuclear Safety Commission
CGSB Canadian General Standards Board
HVL half-value layer
IAEA International Atomic Energy Agency
LAO Licence Assessment Officer
LET linear energy transfer
MBq megabecquerel
MRD Materials Regulation Division (of the AECB)
MSDS Material Safety Data Sheets (associated with WHMIS)
mSv millisievert
NSC Act Nuclear Safety and Control Act
PMT photomultiplier tube
QA quality assurance
QC quality control
QO qualified operator
RSDS Radiation Safety Data Sheets (associated with WHMIS)
RSO radiation safety officer
rem roentgen equivalent man [unit of dose equivalent. SI: sievert (Sv)]
SAT Systems (also Systematic) Approach to Training
TLD badge thermoluminescent dosimeter badge
WHMIS Workplace Hazardous Materials Information System
WLM working level month
v
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 1: Introduction
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION
SAFETY INSTRUCTORS
Radiation Safety Training
Introduction
1.1 This Regulatory Guide advocates a systems approach to radiation safety
training and includes suggestions for integrating radiation safety into work
practices.
1.2 The systems approach includes the development, implementation and
evaluation of initial and continuing training programs as well as the
development of training as a solution to operational problems associated with
radiation safety. The focus is on the actual job requirements. This will result in
improved radiation safety and protection, minimize future regulatory problems,
and may have a significant positive impact on the licensee's business.
1.3 Site-specific factors that may have to be considered when developing a training
program for a particular licensee should be discussed with staff of the AECB.
Application
1.4 This Guide applies to the radiation safety training of radioisotope, medical
accelerator and transportation workers.
[A note to industrial radiography licensees: the training for a qualified operator
(QO) or any other person who is helping with industrial radiography is not
addressed in this guide. There is a separate AECB program to become a QO
which includes the Study Guide for the Qualified Operator Examination, a
written examination and registration. The persons responsible for industrial
radiography radiation safety are the licensee and the operator (who must be
either a QO, or a supervised trainee.)]
1.5 References made to licensees in this document also apply to non-licensed
persons regulated by the AECB in the packaging and transportation of nuclear
substances.
1.6 For the transportation of nuclear substances there are two types of workers:
C licensee staff, such as persons working at nuclear facilities
C non-licensed persons, regulated by the AECB, such as air carrier
personnel and couriers.
Training is a regulatory requirement
Page 1.1
Radiation Safety Training
1.7 The Act authorizes the making of regulations. The regulations authorize the
issue of licences containing conditions. Standard licence condition 565 for
radioisotope licences issued by the AECB, for example, requires the licensee to
ensure that only persons properly trained in work with radioactive prescribed
substances, and informed of the hazards involved, are allowed to handle such
substances.
Failure to comply with the regulations is an offence under the Act.
1.8 Licences issued by the AECB may contain training conditions covering matters
such as instructions to be given to atomic radiation workers respecting the
hazards of ionizing radiation and the procedures to be followed to limit
exposure to ionizing radiation;
the qualifications, training and experience of any person who is to use or
supervise the use of the prescribed substance or any device or equipment to
which the licence applies.
1.9 The AECB expects that applications for a radioisotope licence will include a
description of the training and experience required of all personnel, including
trainees, who work with, will be handling or are in the vicinity of radioactive
material.
1.10 The scope and depth of radiation safety training will vary significantly with the
job requirements and the responsibilities of the individual. These requirements
and responsibilities are determined by the licensee and assessed by AECB
licensing staff by reviewing the task analysis, training course outlines and
training manual contents submitted as part of the licence application. If the
training program is suitable, AECB staff recommends its acceptance to the
designated officer of the AECB.
Developing a safety culture
1.11 Under the regulations and licences, the development and implementation of
radiation safety training programs is the responsibility of the licensee.
Licensees are encouraged to discuss their training programs and the
recommendations contained in this Guide with AECB staff.
1.12 The ultimate goal of radiation safety training is the development of a safety
culture. While the overall responsibility for the development of such a culture
rests with management, the ultimate responsibility for the practice of safety
rests with each individual.
Page 1.2
Radiation Safety Training
1.13 AECB staff have observed that poor compliance with radiation safety practices
is usually a direct result of workers not receiving appropriate information. Staff
have also noted that time spent discussing safety issues with the licensee's staff
during inspection visits has a positive impact on future compliance.
1.14 All licensees should allocate resources in proportion to the magnitude and
scope of their work so that the basic recommendations of this Guide are
addressed.
1.15 Licensees should meet the standards of the radiation safety training program
described in Section 2 of this Guide, or through an equivalent program
developed by the licensee and accepted by the AECB.
Program design
1.16 To be acceptable to the AECB, a radiation safety training program must follow
the Systems Approach to Training (SAT). SAT is now internationally
recognized, and is used by agencies and departments such as the U.S. Nuclear
Regulatory Commission (USNRC), the International Atomic Energy Agency
(IAEA) and the Public Service Commission of Canada. SAT is the standard by
which the quality of a radiation safety training program will be judged and the
procedures approved by the AECB.
1.17 SAT places emphasis on the provision and management of training for all
employees by the licensee. It provides consistency, efficiency, management
control and accountability in the training and qualification of workers. SAT can
be integrated easily into any licensee's existing management system. In small
operations with no full-time instructors, other functional specialists, with
suitable training, can operate SAT in the area of specialization on a limited
scale.
1.18 The model for SAT, which is illustrated in Appendix 1, has five phases:
training analysis, evaluation design, training design, conduct and evaluation,
and validation. The detailed standards for these phases are described in Section
7. Examples of evaluation and validation questionnaires are provided in
Appendixes 2 and 3. More detailed descriptions of SAT can be found in
documents such as the IAEA's Technical Report Series No.380, Nuclear
Power Plant Personnel Training and its Evaluation - A Guidebook
(International Atomic Energy Agency, Vienna), and the Systems Approach to
Page 1.3
Radiation Safety Training
Training, revised edition, published by the Public Service Commission of
Canada.
1.19 The first steps in designing a training program are to establish performance
standards and to select the most appropriate training environment-classroom,
laboratory, on-the-job, or self study, or a combination of these.
1.20 Training programs should include a series of activities intended to develop
learning objectives, tests and other indicators of trainee performance. The
design is based on job classification and task analysis, as well as feedback from
on-the-job experiences. The design phase should also include the development
of a training plan that guides the development phase of the training program.
1.21 The development phase starts with the learning objectives produced during the
design phase and identifies the content that meets the trainees' needs. See
Section 2 for information about how to determine the content of training
programs, also Sections 9 through 13 for details of training modules for
different types of licensed facilities.
Page 1.4
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 2: Content and Format
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION
SAFETY INSTRUCTORS
Radiation Safety Training
Program content
2.1 A radiation safety training program acceptable to the AECB includes the
following elements:
C learning objectives
C topics to be covered, including tasks to be performed
C format to provide the information
C statements on performance, conditions and standards
Each of these elements is discussed in this section.
Licensee organizations should also state their policies and philosophy on
training in their radiation safety programs.
Learning objectives
2.2 Licensees should develop learning objectives for the tasks selected for training.
This ensures that training is linked directly to job performance and that the
content of the training program is defined. Learning objectives describe what is
to be learned in terms of measurable trainee performance. A learning objective
consists of:
C a statement of the action that the trainee should perform
C the conditions under which the action should take place
C the standards of acceptable performance of the action.
See 2.10, Statements on performance, conditions and standards, for
additional information.
Topics to be covered
2.3 The topics to be covered in radiation safety training can be grouped under
three broad headings:
C Definitions and terms, including essential terminology, radiation protection
and the effects of radiation
C Concepts, ranging from good work practices to the worker's responsibility
for radiation safety
C Regulations and licensing, including current and proposed regulations
governing training requirements, and the license issued for that particular
facility
Page 2.1
Radiation Safety Training
2.4 Appendix 5 contains a summary of recommended topics. Depending on the
type of program, and the level of knowledge of the trainees, it may not be
necessary to include all the recommended topics. Determine which are "nice to
know" and which are "must know" topics. Keep in mind that in job-related
training, topics such as radiation protection equipment are essential.
2.5 Prepare a timetable showing how much time will be allocated to each topic.
Ensure that there is enough time to address the "must know" topics in detail.
2.6 On completing the training program workers should be able to understand why
specific work practices and procedures are required, to perform these work
practices and to know when and how to respond to upset conditions and
emergency situations.
Worker classification and required training
2.7 Section 13 of this Guide includes matrix tables designed to aid in determining
workers' job classification. In each table, the column headed Recommended
Training Modules specifies the subject areas in which training is required.
Appendix 5 provides, in summary, specific suggestions for the content of each
module.
Formats to provide the information
2.8 If the training is to be delivered in a traditional classroom or lecture setting, the
use of audio-visuals such as overheads, films, slides, and videos, will improve
comprehension. Group discussions and practical demonstrations are also
effective.
2.9 Computer-based training can also be used for individual or group training, or
for self-paced study. However, competent persons knowledgeable in radiation
and related safety matters should always be available to assist and answer
questions that may arise. Practical workshops and demonstrations are also
recommended to complement computerized and self-paced training modules.
Page 2.2
Statements on performance, conditions and standards
2.10 Before a training session begins the licensee should have a clear definition of
what the trainee is expected to know by the end of that session. To this end,
the training program should contain statements on performance, conditions and
standards:
Performance statement: this should clearly state what is expected of the
trainee on completion of the training session, and what results the instructor,
supervisor and inspector can initially observe and evaluate. Specifically:
What the participant will be able to do. For example, to identify the
radiological risks when cleaning up spills.
What the participant will be able to demonstrate. For example, to select the
appropriate equipment for monitoring dose rates.
Conditions statement: describes the performance level expected of the
trainee at the end of the training period and on commencement of work
without supervision or consultation. This is the context of the action.
Standards statement: provides a definition of the minimum acceptable
standard of performance and the recognition of success. Indicates how well the
action must be performed.
Page 2.3
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 3: Workplace Hazardous Materials Information
System (WHMIS)
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR ANY LICENCE PREPARED BY THE
AECB
Radiation Safety Training
About WHMIS
3.1 WHMIS (pronounced whim-iss) is a nationwide communication system that
provides information to workers and employers on hazardous materials used in
the workplace. WHMIS applies to all Canadian workplaces. It requires that all
workers who work with or near a hazardous substance, as defined in the
Controlled Products Regulations, are informed about potential hazards and
recommended safe work practices.
3.2 As a result of the definition in the AEC Act for "prescribed substance" which
includes any substances containing radioisotopes, hazardous materials containing
radioisotopes are currently excluded from WHMIS. This will change when the
NSC Act comes into force replacing the definition for "prescribed substances" by
"nuclear substances".
3.3 Under the NSC Act a "nuclear substance" does not include the substances being
used that carry radionuclides. This means that the non-radioactive carrier material,
if a controlled substance within the meaning of the Hazardous Products Act, will
be subject to the rules of WHMIS and will, therefore, require WHMIS labelling.
There will be quantity exemptions for the non-radioactive component of mixtures
based on a combination of volume or weight, and hazard.
3.4 WHMIS requires that information be provided in three ways:
C All controlled products used in the workplace must have a WHMIS label
on the container.
C Material Safety Data Sheets (MSDS) and hazard information must be
readily available in the workplace. An MSDS summarizes the health and
safety information about the product.
C Workers must receive training to be able to recognize and work safely
with the controlled products.
3.5 Trainees should be informed that the HAZARD SYMBOL is an important part of
the WHMIS label.
3.6 The AECB is developing a program of "WHMIS-equivalence", where all
workers who may be exposed to radioactive materials in the workplace will,
regardless of risk, be afforded basic safety information and training. This basic
program will be based on the requirements that will be specified in the new
CNSC Worker Safety Information Regulations. The program will likely be based
on the following being in place:
Page 3.1
Radiation Safety Training
a. criteria for worker training, covering all workers in the industries
regulated by the AECB who may come into contact with radioactive
material or who may be exposed to radiation
b. a product labelling system equivalent to the WHMIS labelling system.
(AECB's current labelling requirements are sufficient)
c. Radiation Safety Data Sheets (RSDS) similar to the WHMIS MSDS.
[Unlike the MSDS, which are produced by the manufacturers of
conventionally hazardous products, the RSDS may be produced and
supplied by the AECB. Appendix 8 contains a draft sample RSDS]
d. the right of a worker representative or OSH Committee to participate in
the development and review of WHMIS-equivalent training programs.
The AECB will advise its licensees when the program will come into effect. WHMIS-
equivalence training will be necessary.
Guidance on basic WHMIS-equivalent training
3.7 Every worker encountering radioactive materials at the workplace should receive
a level of training that is consistent with the risks associated with their proximity
to and/or handling of the radiological hazard. Even workers subjected to only
very low risk should receive at least a basic level of training.
3.8 Training programs should include instruction in:
a. the types, quantities and forms of radioactive material that may be
encountered by the worker
b. the significance of labelling and signage for radioactive material that may
be encountered by the worker
c. the contents of radioisotope safety data sheets relevant to the workplace
and the significance of the information contained therein
d. procedures for those licensed activities that are relevant to the worker's
duties or understanding the safety implications of the radioactive material
e. procedures to be followed when unplanned releases of radiation or
radioactive material are present or suspected
f. procedures to be followed in case of emergency involving radioactive
material.
Page 3.2
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 4: Induction Training, Basic Training
and Retraining
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION
SAFETY INSTRUCTORS
Radiation Safety Training
Induction training
4.1 The induction or initial training program is important because it provides the
basis of all training and work that follows. It presents to the new worker the
licensee's safety philosophy and standards that should be stated in supporting
documents to the licence application. It also makes the new worker aware of
the regulatory requirements for training.
4.2 Induction training can be provided as a single training session. Alternatively,
the instructor may prefer to divide the training into shorter periods or modules
to make it easier for the trainees to absorb and retain the new knowledge.
4.3 Develop timetables to indicate how much time is to be spent on each topic.
This helps to determine how much detail can be included. For suggestions on
the timing for each topic, refer to the guidelines in Sections 9 through 12.
4.4 The major goals of induction training are to:
a. provide a brief introduction to radiation safety and the licensee's
existing radiation safety program that will enable trainees to recognize
and deal with situations that require precautions and explain to them
their responsibility in reducing their own radiation exposure
b. provide new workers with the knowledge and skills to understand
radiation safety and procedures to perform initial work assignments
safely
c. confirm that the worker has the ability to understand radiation safety
and procedures and is motivated to perform assignments in an
acceptable manner
d. show that operations are monitored by licensees and regulatory
agencies.
4.5 The AECB recommends that induction training include:
a. emphasis on the licensee's commitment to radiation safety by reviewing
the licensee's existing policies, programs, radiation health and safety
procedures, and awareness of the applicable legal provisions
b. explanation of the methods used and problems encountered in the
licensee's operation, as well as the control mechanisms in place, such as
radiation protection equipment, codes of practice, hygiene, ventilation
and dosimetry
c. identifying the location of the posted local radiation safety rules and
procedures
Page 4.1
Radiation Safety Training
d. informing the new worker about the health effects of radiation
exposure
e. providing the names of persons to be contacted on matters of health
and safety emergencies
f. training on how to handle upset conditions, instruction for emergency
procedures and evacuation plans.
4.6 Key topics for both the induction and subsequent training modules and
programs must relate to practical on-the-job situations and conditions. Even at
the induction level, these should include a basic understanding of:
C radiation theory
C radiation protection
C radiation units
C types of radiation exposure and doses
C risks associated with radiation exposure and doses
C understanding personal dose and exposure records.
4.7 Trainers should encourage participants to ask questions, but it is important to limit
your answers to essentials. Too much detail can be confusing. Participants should be
tested after completing their induction training. Supervisors should follow-up and
make recommendations concerning additional instruction that may be required.
Basic training
4.8 Detailed basic training for site-specific working areas and conditions is provided
only after successful completion of induction training. This type of training can
take place both in a classroom and on the job.
4.9 Basic training should:
a. instruct workers in the procedures submitted by the licensee and accepted
by the AECB to ensure regulatory compliance
b. inform workers of their responsibility to report promptly to the licensee
any condition(s) that might violate the AEC Regulations or licence
conditions
c. instruct personnel who do not work directly with prescribed radioactive
substances but who may work in the vicinity of radioactive
materials-such as janitorial, security, maintenance, nursing, and clerical
staff-in how to recognize warning symbols, actions expected in case of
radiation hazard, and who to contact about radiation safety issues
d. educate trainees about the radiation characteristics, radiation risks and
hazard levels of the materials with which they will be working
Page 4.2
Radiation Safety Training
e. describe the regulatory process and the measures taken by the licensee to
protect workers
f. instruct supervisors about the persons who are to be constantly
supervised while they remain in areas where a potential for radiation
exposure exists, for example, persons less than 18 years of age
g. inform female workers that subsection 19.(4) of the AEC Regulations
requires them to advise the licensee if they are pregnant, and that special
reduced limits apply during the term of the pregnancy (also see
subsection 11.(1) of the proposed Radiation Protection Regulations).
Retraining
4.10 The AECB recommends retraining programs both to enhance and maintain
employee competence. Workers should be retrained after receiving new
assignments and when procedures change. The retraining must reflect the new
working conditions and environment to which the worker will be exposed.
4.11 As a general rule the licensee should provide retraining annually or bi-annually.
More detailed radiation retraining programs should be provided every three-to-
five years. In addition, radiation safety topics should be discussed at informal
safety meetings held for workers at least once a month.
4.12 The degree of retraining required depends on each worker's knowledge and
experience, and will be reflected on their training record (see Keeping records in
Section 5). Validation (refer to Section 7 dealing with program verification)
provides a means of assessing the overall level of knowledge retained by
workers, and identifying the areas where retraining is essential.
Training of supervisors
4.13 The radiation safety training program for supervisors and workers should be the
same, with the exception that the supervisor is expected to provide on-the-job
guidance to the workers and include instruction on the interpretation of regulations
and licence conditions, and the necessity of compliance. In addition to radiation
safety training, the AECB recommends instruction for effective supervision.
Management should evaluate the effectiveness of the supervisor's training.
4.14 It is important that supervisors are trained to set an example to the workers by
following the rules and regulations, using protective equipment correctly, wearing
their assigned dosimeters, and supporting the licensee's radiation safety programs.
Page 4.3
Radiation Safety Training
4.15 Supervisors should be trained to observe the attitudes and performance of workers
who report to them, note any deficiencies and make recommendations concerning
requirements for any additional training.
Page 4.4
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 5: Administering Training Programs
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION
SAFETY INSTRUCTORS
Radiation Safety Training
Keeping records
5.1 It is the licensee's responsibility to ensure that a complete record is maintained
of the qualifications and training received by every worker (refer to Section 1
for a more detailed description of the licensee's responsibilities for training).
These records should indicate if and when previous experience and training
have been taken into consideration.
5.2 If there are site-specific factors that may have to be considered in developing,
implementing and maintaining training records you can discuss them with
AECB staff.
Training program records
5.3 The training program records maintained by the licensee should include:
C training plans
C training group procedures and lesson plans
C training materials
C test/examination blanks
C training schedules
C trainee participation (names and occupations of those attending each
lecture)
C results of training examinations and course evaluations
C analysis results (job and task)
C changes to the program and when implemented (note that significant
changes must be approved by the AECB before implementation.)
C data gathered to assess the program's effectiveness
C certification records
Certification
5.4 The licensee should provide every worker with a certificate indicating that the
worker has successfully completed the radiation safety training program.
5.5 The certificate should reference the AECB requirement for certification, and
should include the trainee's name, title of the course, modules completed, date
of issue, and the names of the licensee and the person responsible for issuing
the certificate.
Page 5.1
Radiation Safety Training
Scheduling
5.6 Schedules for both new worker training and retraining for established workers
are important. In preparing and maintaining schedules take into consideration
the needs of the trainees and their supervisors, and the best conditions for
learning.
5.7 For example, the program can be concentrated into a solid week of training, or
spread out over a period of time. To minimize disruptions, set the schedule and
communicate it well in advance to the participants and their supervisors.
5.8 In addition to dates, the schedules should indicate the location, type of training,
the instructor's name, and the names and work classifications of the trainees.
5.9 Some sample schedules are shown in Appendix 4.
Page 5.2
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 6: Selecting and Training Instructors
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION
SAFETY INSTRUCTORS
Radiation Safety Training
Instructor qualifications
6.1 The AECB expects that instruction will be provided either by qualified staff of
the licensee or by an outside training institution or consultant, using a training
program acceptable to the AECB.
6.2 While the format and materials in a training program are approved by the
AECB, the effectiveness of the training depends to a large extent on how well
it is delivered. It is important, therefore, that the instructor selection profile
include:
C subject matter expertise in the relevant tasks or environmental factors, gained
through experience and training
C effective oral communication skills
C the successful completion of an instructional techniques course that includes
rigorous evaluation.
See also Objective 2 in Section 7 of this Guide.
6.3 In addition, instructors selected for radiation safety training should have:
C adequate knowledge of the applicable regulations and licence conditions
C successfully completed studies in radiation safety
C adequate support staff
C all necessary materials, equipment and tools
C a reliable system of document control and records management
Instructor training
6.4 The licensee should establish, for AECB acceptance, the technical and
instructional qualifications criteria, procedures and programs for selecting,
training and authorizing qualified instructors.
6.5 Most instructors achieve the required instructional qualifications by completing a
specific training program. Instructors are expected to continuously upgrade their
training in order to maintain their technical knowledge and instructional skills.
Page 6.1
Radiation Safety Training
Learning objectives
6.6 Learning objectives for training instructors include:
C role of the instructor
C understanding how adults learn
C using appropriate training techniques
C using lesson plans and other instructional materials and media
C conducting lectures, discussions and practical demonstrations
C establishing job performance measures to evaluate on-the-job training
C assisting trainees in solving learning problems
C assessing trainees
C maintaining and using individual trainee records and program records.
Page 6.2
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 7: Program Verification
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION
SAFETY INSTRUCTORS
Radiation Safety Training
Evaluation and validation
7.1 Evaluation and validation of training programs are essential in order to
determine their effectiveness. Licensees should establish quality controls for
radiation safety training programs and their delivery based on ongoing
evaluation and validation. For continuing evaluation of their radiation safety
training programs, licensees should review the five essential phases of SAT
(see paragraph 1.18 of this Guide) to verify compliance. This section outlines
the objectives for such a review
7.2 Evaluations of radiation safety training programs should be comprehensive and
consistent over time. Acceptable measurable objectives and criteria for
regulatory evaluations are included in this section. Appendix 2 contains sample
evaluation forms.
7.3 Validation closes the loop, through linkage of all previous components or
phases of a training program. It proves the soundness of the training program,
and indicates if it is defensible and efficient. It is the process that identifies
whether the abilities learned and evaluated are required on the job and are
being used to produce acceptable performance.
7.4 The licensee should validate the training program within three months
following the completion of the worker's training. In addition, AECB
inspectors will periodically evaluate and validate training programs, including
training materials and methods of instruction. Schedules and records will also
be verified through inspections. (See Appendixes 4 and 7 in this Guide).
Objectives and criteria for evaluating training programs
7.5 The following are objectives and criteria for regulatory evaluations of radiation
safety programs, where
C an objective specifies an end result of an effective, well-managed
training program
C supporting criteria specify conditions or actions that satisfy the intent
of the objectives.
Page 7.1
Radiation Safety Training
Preparation of the radiation safety training program
7.6 This section covers the first three phases of SAT: analysis, evaluation design,
and training design.
Objective 1 Training programs are effectively organized, directed and
supported.
Criterion 1.1 The goals of the training program are clearly defined and
understood.
Criterion 1.2 The specific responsibilities of personnel who manage,
supervise, develop, conduct and evaluate training are clearly
stated and understood.
Criterion 1.3 Personnel in a licensed facility whose duties could have an
impact on the environment, the public or on the safe and reliable
operation of the facility, have clearly defined and documented
qualification needs, and well documented initial and continuing
training programs.
Criterion 1.4 A framework for establishing, conducting and maintaining
training programs in accordance with the principles of SAT is
documented in written policies and procedures.
Criterion 1.5 Personnel are aware of the training required for their respective
positions and of the performance expected of them during and
after training.
Criterion 1.6 Management ensures that all personnel receive the required
training and that the records of the training process are
maintained that are easily understood and readily accessible for
review and inspection purposes.
Criterion 1.7 Instructional facilities such as workshops, classrooms and
laboratories are adequately equipped to meet the specific
objectives of the training program.
Criterion 1.8 Adequate funding and personnel are available to support
effectively all the required training programs and activities.
Page 7.2
Radiation Safety Training
Criterion 1.9 Training developed or implemented by contractors or outside
organizations meets the same specifications as the AECB-
accepted training provided by the licensee's training department.
Objective 2 Training personnel have the subject matter expertise,
experience and instructional skills needed to discharge their
assigned duties.
Criterion 2.1 Subject matter expertise and instructional skills of training
personnel meet documented training and qualification standards.
Criterion 2.2 Training personnel receive additional training periodically to
maintain and improve their technical knowledge and
instructional skills.
Criterion 2.3 Current revisions of radioactive materials handling procedures,
technical references and radiation protection procedures are
readily available to training personnel.
Criterion 2.4 Personnel who conduct on-the-job training and assessment
receive adequate training on the policies, practices, methods
and standards for delivering this training and for assessing
trainee performance.
Criterion 2.5 Training personnel monitor training delivered by contractors,
beginner instructors and personnel from outside the training
department, and obtain feedback on their performance from the
course participants.
Criterion 2.6 The performance of all training personnel is periodically
monitored and assessed. The results are fed back to the
instructors and used to improve their performance.
Objective 3 The job performance requirements are determined by an
analysis of the duties, where such duties could have an
impact on the environment, the workers, the public or on the
safe and reliable operation of the licensed facility. This
analysis serves as the basis for the development of learning
objectives, training materials and job performance measures.
Page 7.3
Radiation Safety Training
Criterion 3.1 Qualified personnel occupying positions for which training is
being developed participate in the analysis of the jobs that they
perform to provide subject matter expertise.
Criterion 3.2 The analysis identifies the tasks requiring initial training and
those requiring continuing training, with due consideration to
tasks performed infrequently during non-routine and upset
conditions.
Criterion 3.3 The analysis identifies the skills and knowledge needed for the
tasks selected for training and is sufficiently detailed to enable
the development of learning objectives, test items and job
performance measures.
Criterion 3.4 Learning objectives are developed and kept current to establish
the training content for task-related knowledge and skills,
taking into consideration the initial knowledge, skills and
experience of the trainees.
Criterion 3.5 Learning objectives state clearly the satisfactory standards of
trainee performance and are linked to job performance
requirements.
Criterion 3.6 Learning objectives are sequenced, grouped and organized
appropriately according to the required progression of learning.
Criterion 3.7 Job performance measures and test items for written and oral
examinations are developed to measure job-related knowledge
and performance of trainees, including higher cognitive abilities
such as analytical and diagnostic skills.
Objective 4 Training materials encompass the knowledge and skills
required to meet the learning objectives.
Criterion 4.1 Necessary training materials, including lesson plans, laboratory
guides, individualized study guides and on-the-job training
guides, are developed to promote effective and consistent
delivery of training.
Page 7.4
Radiation Safety Training
Criterion 4.2 Training materials are accurate, support the learning objectives
and test items, and are maintained so as to reflect current
knowledge.
Implementing the training program
Objective 5 Training delivery employs principles of good instructional
presentation and conveys accurate information consistently
and clearly.
Criterion 5.1 Classroom, on-the-job and skills training are implemented as
outlined in the training materials approved by the licensee are
conducted by individuals qualified to perform the job who
possess adequate instructional and assessment skills.
Criterion 5.2 Lesson plans or equivalent training guides approved by the
licensee are used in all instructional settings to ensure consistent
training delivery directed towards specific learning objectives.
Criterion 5.3 Instructors are prepared adequately to deliver training
effectively and consistently with access to the necessary
instructional aids and equipment.
Criterion 5.4 Instructors use techniques appropriate to the learning
objectives, lesson content and instructional setting.
Criterion 5.5 Individualized or self-paced instruction, if used, gives the
trainees sufficient guidance and supporting material to master
the learning objectives.
Criterion 5.6 When the task is simulated because it cannot be performed in
the actual job setting, the differences between the simulated task
and the actual must be reviewed with the trainees prior to
training.
Criterion 5.7 The procedures used during training are those used during
licensed activities.
Page 7.5
Radiation Safety Training
Criterion 5.8 Training reinforces the expectations of management for the
conduct of licensee activities by their personnel. The diagnostic
and teamwork skills required for the operations, according to
the expectations of the licensee's management, are developed.
Objective 6 Trainees are evaluated on their mastery of the learning
objectives and receive prompt feedback on their performance.
Criterion 6.1 There is a clear link between the test items or job performance
measures and the initial analysis of the job performance
requirements.
Criterion 6.2 Trainees are evaluated regularly and receive prompt feedback
on their strengths and weaknesses, and on the need for any
remedial training.
Criterion 6.3 Trainees who fail to meet minimum performance standards
receive remedial training and reassessment.
Criterion 6.4 The licensee's personnel successfully complete all training
requirements for their position before being assigned to work
independently. Exemptions from training are based on
appropriate test results or other objective evidence that the
subject training is not required.
Criterion 6.5 There is a formal procedure for verifying that contractors,
temporary personnel, or other non-plant personnel have been
appropriately trained for the work to which they are assigned.
Criterion 6.6 Tests and assessments are prepared and administered in a
consistent manner. All answers expected during written or oral
tests are specified in advance in appropriate marking guides that
are consistent with accepted working practices.
Criterion 6.7 Adequate precautions are taken to ensure confidentiality of
examinations, tests and making guides.
Page 7.6
Radiation Safety Training
Evaluating the training program
Objective 7 Training programs are systematically evaluated and, if
necessary, revised so that on-the-job competence is attained
and maintained.
Criterion 7.1 Training programs are evaluated on a basis determined by the
AECB with all the individual evaluation activities integrated
into a comprehensive evaluation system.
Criterion 7.2 Training program improvements and changes are proposed,
initiated, tracked and incorporated into the program in a timely
manner.
Criterion 7.3 Feedback on training activities is solicited regularly from both
trainees and instructors during training program evaluations.
Criterion 7.4 For validation purposes, former trainees are asked for feedback
on the strengths and weaknesses of the training program and its
individual courses, approximately three months after the end of
the program.
Criterion 7.5 Supervisors and managers are interviewed regularly to obtain
feedback on job performance problems, training priorities, and
the effectiveness of recent training in enabling the licensee's
personnel to do their jobs.
Criterion 7.6 Training programs are revised, where necessary, to reflect
procedural and equipment changes, changes in job descriptions,
industry-wide significant events, as well as other aspects of
operating experience.
Criterion 7.7 Inspections and evaluations of training programs performed by
groups external to the training department, for quality
assurance, are factored into the training program evaluation
process.
Criterion 7.8 Job analyses, learning objectives, test items, job performance
measures, lesson plans and training material are revised to
reflect the findings of training program evaluations.
Page 7.7
Radiation Safety Training
Criterion 7.9 Training delivery in the workshop, classroom, laboratory field
and on-the-job is evaluated regularly and evaluation results are
communicated to instructors.
Criterion 7.10 Test results and other indicators of trainee performance are
analysed to identify areas where the training program can be
improved.
Criterion 7.11 Training provided by contractors or outside organizations is
evaluated to ensure that it meets the job needs and that its
quality is consistent with accepted training standards.
Criterion 7.12 Records are maintained of evaluations, and decisions resulting
from evaluations, that are easily understood and readily
accessible for review and inspection.
Page 7.8
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 8: AECB assessment of radiation safety training
programs
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION
SAFETY INSTRUCTORS
Radiation Safety Training
AECB assessment of radiation safety training programs
8.1 The AECB requires applicants to include a description of their
radiation safety training programs as part of the submission for a
licence.
8.2 Radiation safety training program descriptions are assessed by AECB
staff using the standards and criteria in this Guide.
8.3 Radiation safety training programs that conform to this Guide expedite
the licence assessment process.
8.4 Other radiation safety training programs should conform with the intent
of this Guide or be revised to conform. AECB licensing staff inform
applicants in writing of deficiencies in their radiation safety training
programs.
8.5 Following acceptance, the applicant's radiation safety training program
description is referenced in a condition of the licence. Any substantial
change to the radiation safety training program requires the prior
written acceptance of the Board.
8.6 AECB staff monitor compliance with accepted and referenced licensee
radiation safety training programs during site visits.
Page 8.1
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 9: Radioisotopes
THIS SECTION APPLIES TO RADIOISOTOPE LICENSEES AND APPLICANTS FOR ANY RADIOISOTOPE
LICENCE PREPARED BY THE AECB, AND TO RADIATION SAFETY INSTRUCTORS
Radiation Safety Training
Radioisotope licensees
9.1 All radioisotope licences issued by the AECB pursuant to section 7 of the AEC
Regulations include a licence condition requiring the licensee to ensure that
only persons properly trained in work with radioactive prescribed substances
and informed of the hazards involved are allowed to handle such substances.
As a licensee it is your responsibility to ensure that every person working in the
facility:
a. is competent to work safely
b. is adequately trained in radiation safety
c. has actual notice of and complies with the conditions of the licence
The philosophy behind these regulatory requirements has not changed for the
proposed Nuclear Substances and Devices Regulations to be implemented
under the NSC Act.
9.2 You also have a responsibility to provide training to their workers before they
start to work with radioisotopes, except when on-the-job training is carried out
under the direct supervision of a person with a level of training and expertise
that is acceptable to the AECB. The scope and depth of the radiation safety
training for radioisotope workers should reflect the actual radioisotope
operations and radiation risks associated with their work.
9.3 Some groups of workers, such as nuclear medicine technologists and medical
physicists, whose work routinely requires the manipulation of radioisotopes
and exposure to radiation, receive extensive radiation safety training as part of
their specialized education or certification. These workers should also comply
with the radiation safety training recommended in this Guide, but may
demonstrate the required level of training through their academic background
and professional certification.
Safety training modules for those dealing with radioisotopes
9.4 Refer to section 12 for a description of the suggested module content and also
to section 13, table 13-1: Radioisotope worker training matrix.
Page 9.1
Radiation Safety Training
for Radioisotope, Industrial Radiography, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 10: Medical particle accelerators
THIS SECTION APPLIES TO MEDICAL ACCELERATOR LICENSEES, APPLICANTS FOR ANY PARTICLE
ACCELERATOR LICENCE PREPARED BY THE AECB AND TO RADIATION SAFETY INSTRUCTORS
Radiation Safety Training
Particle accelerator licensees
10.1 All particle accelerator licences issued pursuant to sections 7 and 9 of the AEC
Regulations include a licence condition requiring the licensee to ensure that
every person working in the facility:
a. is competent to work safely
b. is adequately trained in radiation safety
c. has actual notice of and complies with the conditions of the licence
10.2 You also have a responsibility to provide training to your workers before they
start to work with the accelerator, except when on-the-job training is carried
out under the direct supervision of a person with a level of training and
expertise that is acceptable to the AECB. The scope and depth of the radiation
safety training for particle accelerator workers should reflect the actual
operations and radiation risks associated with their work. Modules should
include:
C accelerator operation
C radiation production (direct radiation and radiation from activation)
C shielding (see 10.4)
C personnel safety interlocks and warning displays (see 10.5)
C personnel dosimetry special to the facility (neutrons, for example)
C operating practices to avoid personnel exposure to radiation
10.3 Some groups of accelerator employees have received extensive radiation safety
training as part of their specialized education or certification. Such employees
are expected to comply with the radiation safety training recommended in this
Guide, but may demonstrate the required degree of competence through their
academic background and/or professional certification.
10.4 Time, distance and shielding are all important factors in controlling the risks
associated with radiation. However, the training program should also be sure
to remind participants that accelerators have a fourth control: they can simply
be turned off.
Page 10.1
Radiation Safety Training
10.5 Paragraph 4.(5)(b) of this Guide states that worker protection, including
personal protective equipment and clothing, and the importance of personal
hygiene in the prevention of spreading and ingesting contaminants, should be
included in radiation safety training. Note that the interlock system is also part
of radiation safety in accelerators.
Safety training modules for particle accelerator licensees
10.6 The topics in the modules should reflect what is required to be delivered
according to the analysis phase of SAT (see Appendix 1).The instructor can
vary the length and content of each module depending on the understanding,
qualifications and experience of the participants. Refer to section 12 of this
guide for a description of the suggested module content and also to section 13,
Table 13-2: Medical Accelerator Training Matrix.
Page 10.2
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 11: Transportation of radioactive nuclear
substances
THIS SECTION APPLIES TO THOSE DEALING WITH THE PACKAGING AND TRANSPORTATION OF
RADIOACTIVE NUCLEAR SUBSTANCES/MATERIALS AND TO RADIATION SAFETY INSTRUCTORS
Radiation Safety Training
Transportation of radioactive materials
11.1 Section 3.(1) of the Transport Packaging of Radioactive Materials
Regulations [TPRM Regulations] references the Transportation of Dangerous
Goods Regulations [TDG Regulations]. As a result the requirements set out in
the TDG Regulations are mandatory.
11.2 Basic training should include:
a. fundamental requirements of the regulations mentioned in 11.1 above,
governing the transport of radioactive material
b. preparing packages
i. classification, packaging, contamination levels and qualification of
workers
ii. consignor responsibilities, checks before transportation, including
measuring surface dose rates, and determining the transport index
[meaning the number for a package or transport container derived in
accordance with the procedures described in Schedule XI of the TPRM
Regulations]
c. labelling of transportation containers
d. describing nuclear substances (radioactive materials) on shipping papers
e. emergency response.
11.3 Module content should include the procedures necessary for paragraphs a. to
e., plus:
f. reporting requirements
g. receipt of nuclear substances (radioactive materials), including survey
requirements and receipt records
h. shipment of nuclear substances (radioactive materials), including preparing
the nuclear substances for shipment, completing shipping documentation
and surveying the packages
i. transportation of nuclear substances (radioactive materials), including the
packaging of nuclear substances, the labelling of transport containers and
completing the shipping documents
j. placarding of a transport vehicle.
Page 11.1
Radiation Safety Training
Safety training modules for those dealing with packaging and
transportation of radioactive nuclear substances/materials
11.4 Refer to section 12 for a description of the suggested module content and also
section 13, Table 13-3:Transportation Training Matrix.
Page 11.2
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 12: Radiation Safety Training Modules
THIS SECTION APPLIES TO RADIOISOTOPE AND MEDICAL ACCELERATOR LICENSEES AND
APPLICANTS FOR ANY RADIOISOTOPE OR PARTICLE ACCELERATOR LICENCE PREPARED BY THE
AECB, TO THOSE DEALING WITH THE PACKAGING AND TRANSPORTATION OF RADIOACTIVE
NUCLEAR SUBSTANCES/MATERIALS, AND TO RADIATION SAFETY INSTRUCTORS.
Radiation Safety Training
The topics in the modules should reflect what is required to be delivered
according to the analysis phase of SAT (see Appendix 1). The instructor can
vary the length and content of each module depending on the understanding,
qualifications and experience of the participants.
Module 1: Radiation Orientation Lecture (Radioisotopes, Accelerators and
Transportation)
OBJECTIVE:
1. To provide a brief introduction to the radiation safety training program and
radiation safety practices.
MODULE CONTENT:
1. General (non-technical) summary of radiation and radiation effects. Technical
terms should be carefully selected and worded to avoid misunderstanding. The
aim is to give a general idea of the whole problem of radiological protection or
of some particular field of it.
2.(a) (Radioisotopes) Basic training to enable employees and others to recognize
and deal with situations that require precautions. These trainees may include
technical staff, personnel in shipping/receiving, janitorial or housekeeping,
secretarial or clerical, management, contractors, maintenance staff, security,
emergency services, students and visitors.
2.(b) (Accelerators and transportation) Basic training to enable employees and
others to recognize and deal with situations that require precautions.
3.(a) (Radioisotopes) Problem recognition and superficial risk analysis, including
whom to contact in the event of problems and what simple steps to take to
protect themselves and others until radiation safety personnel arrive. For
example, a member of the cleaning staff should be able to recognize a radiation
warning symbol and be familiar with the organization's radioactive waste
handling procedures. A shipping/receiving clerk should know what steps to
take if a radioactive package is damaged and whom to call for assistance.
3.(b) (Accelerators and transportation) Problem recognition and superficial risk
analysis, including whom to contact in the event of problems and what simple
steps to take to protect themselves and others until radiation safety personnel
arrive.
Page 12.1
Radiation Safety Training
Module 2: Regulatory Requirements (Radioisotopes, Accelerators and Transportation)
OBJECTIVES:
1. To inform trainees of the regulatory requirements, including the responsibilities
of licensees to provide the training prescribed in standard licence conditions.
2. To acquaint the trainees with precautionary practices and procedures and with
records, reports and notification procedures.
3. To inform trainees of the qualifications and skills their job requires.
4. To inform trainees on worker responsibilities under the regulations, including
the procedure for reporting unusual occurrences.
MODULE CONTENT:
A. Atomic Energy Control Act and Regulations/Nuclear Safety and Control Act
and highlights of the proposed new regulations.
B. AECB licence conditions
C. Notices, regulatory documents and AECB instructions
D. Other federal, provincial or local regulatory agency requirements that affect
radioisotope workers.
Module 3: Operating and Emergency Procedures (Radioisotopes, Accelerators and
Transportation)
Note: This module should be tailored to complement the licensee's own needs
and procedures.
OBJECTIVES:
1. To acquaint the trainee with the operating and emergency procedures.
2. To provide information about the organization of operating and emergency
procedures.
3. To show the trainee an example of procedure implementation.
Page 12.2
Radiation Safety Training
MODULE CONTENT:
A Operating procedures
1. Radiation safety program personnel (duties and responsibilities)
a. Radiation safety officer
b. Radiation safety instructor
c. Area supervisor
d. Assistant supervisory personnel
2. Facility requirements
a. Use areas
b. Storage areas
3. Radiation Safety Program
a. Personnel monitoring
- film/TLD badges
- pocket dosimeters
- ring badges (for example, finger dosimeters) (radioisotopes)
- internal dosimetry (urine analysis, whole body counting)
(radioisotopes)
b. Radiation detection instruments
- preparation for use
- daily check
- use of
c. Use of security
- posting warning signs
- calculating boundary of restricted area
d. Surveys
- area surveys
- vehicle survey (if applicable)
- wipe survey
- other as applicable
e. Transportation of radioactive materials
f. Receipt and disposal of radioactive material
g. Leak testing of radioactive sources
h. Inventory, inspection and maintenance of equipment
- requirements
- records
i. Records management
B Emergency procedures
1. Contamination
2. Other
Page 12.3
Radiation Safety Training
3. Emergency response training
C Hands-on exercises (where applicable)
1. Procedures for receiving and opening packages of radioactive material
2. Locking and securing sources
3. Use of open source radioactive material
a. Handling equipment
b. Protective clothing
c. Prevention of contamination
d. Decontamination procedures
e. Ventilation
4. Use of sealed radioactive sources
Module 4: Structure of Matter (Radioisotopes, Accelerators and Transportation)
OBJECTIVES:
1. To acquaint the trainee with the basic concepts on the structure of matter.
2. To provide information about the relationship of the particles to matter as they
relate to radioactive isotopes specific to the task.
MODULE CONTENT:
Start the module with matter on a macroscopic scale, relate to molecules and then to
atoms.
A. Structure of the atom
1. The nucleus
a. Proton - characteristics
b. Neutron - characteristics
2. Electrons
a. Characteristics
b. Orbits
c. Orbit jumping and escape
B. Atomic number - determination
C. Atomic weight - determination
D. Isotopes
1. Definition
Page 12.4
Radiation Safety Training
2. Radioactive vs stable
3. Types used by the trainee, or those isotopes with which the trainee has a
possibility of coming in contact.
Module 5: Radiation and Radioactivity (Radioisotopes, Accelerators and
Transportation)
OBJECTIVES:
1. To acquaint the trainee with the basic concepts of radiation.
2. To provide information about the types of radioactive decay.
3. To introduce the units of radioactive decay including the International Standard
Units (SI)
4. To introduce the concepts of radiation interaction with matter.
5. To provide information about the hazards of various types of radiation.
MODULE CONTENT:
Start the module with a definition of radiation.
A. Discussion of the types of radiation
1. Alpha and beta -- type, hazard and shielding
2. Gamma and X-ray -- type, hazard and shielding
3. Neutrons --- type, hazards and shielding
4. Electron capture and positron emission (Optional)
B. Electron Volt - unit of radiation energy
1. Definition
2. Energy spectrum
C. Mechanism of decay -- decay to stable form
D. Activation of matter
E. Units of radioactivity
1. Becquerel/Curie
2. Half-life concept
a. Decay charts
b. Calculation of half-life
F. Interaction of radiation with matter
Page 12.5
Radiation Safety Training
1. Ionization - effect on electrical stability
A basic definition of ionization and a simple description of the production of
electron - ion pairs in matter should suffice.
a. Particle ionization
b. Electromagnetic ionization
- photoelectric effect
- Compton effect
- pair production
2. Effects: energy lost by radiation to the material through which it is passing
a. Range in tissue (linkage with biological effects in Module 8)
b. Range in air
c. Linear energy transfer (LET)
Module 6: Radiation Units (Radioisotopes, Accelerators and Transportation)
OBJECTIVES:
1. To acquaint the trainee with the units used to measure radiation.
2. To provide information concerning the correct usage of the units as they apply to
different types of radiation.
3. To provide information concerning the conversion of the old units to the new units
(SI).
MODULE CONTENT:
A. The coulomb (Roentgen) - X-ray and gamma ray exposure to air. May mention
that the coulomb is the historical unit for measuring exposure to air by X- and
gamma rays
B. The Gray (Gy) [RAD (R)]- Absorbed dose
1. Definition and discussion
C. The sievert (Sv) [REM (rem) - Unit of dose equivalent
ICRP 60 introduced new units: "equivalent dose" and "effective dose". (The latter
is a combined radiation and biological unit.)
1. Definition
2. Qualify factor (QF) (for the new units QF has been replaced with "w -
R
radiation weighting factor")
3. Conversion from and to Si units
D. Submultiples, for example, changing mSv to Sv [mR to R]
Page 12.6
Radiation Safety Training
E. Relationship between units
1. Difference between units
2. Changing from one unit to next
Module 7: Radiation Detection and Measurement (Radioisotopes, Accelerators and
Transportation)
OBJECTIVES:
1. To develop a basic understanding of radiation detection and measurement.
2. To learn the basic concepts of radiation measuring instruments.
3. To acquaint trainees with the more commonly used radiation measuring
instruments required for the operation/environment in which they will be
employed.
4. To learn how to differentiate between a survey meter (dose rate meter) and a
contamination meter.
5. To provide information concerning interpretation of radiation protection
instrument readings.
MODULE CONTENT:
A. Dose vs. dose rate devices
1. Dosimeters - total dose device
2. Survey instruments - dose rate devices
B. Survey instruments
1. Ion chamber instruments:
a). Typical examples of commercially available instruments
b). Specifications of radiation protection importance
c). Use, including the interpretation of radiation instruments measurements,
recognizing problem areas and recording the results of instrument readings
d). Advantages and disadvantages
2. Geiger Müller (GM) instruments:
- as for B. 1.
3. Proportional counter instruments:
- as for B. 1.
Page 12.7
Radiation Safety Training
4 Solid state instruments:
- as for B. 1.
C. Contamination instruments:
- as for B. 1.
D. 1. Personal dosimeters
a) Typical examples of commercially available dosimeters
b) Specifications of radiation protection importance
c) Energy response
d) Procedures use, wearing and storing personal pocket dosimeters
e) Advantages and disadvantages
f) Interpretation of results
2. Thermoluminescent dosimeter (TLD) badges
a). Film badge design, basic operation and use
b). Use of control badges
c). Procedures for wearing badges
d). TLD badge vs. film badge
3. Pocket dosimeters
a) Typical examples of commercially available dosimeters
b) Specifications of radiation protection importance
c) Energy response
d) Procedures use, wearing and storing personal pocket dosimeters
e) Advantages and disadvantages
f) Interpretation of results
4. Electronic dosimeters
a) Typical examples of commercially available dosimeters
b) Specifications of radiation protection importance
c) Energy response
d) Procedures use, wearing and storing personal pocket dosimeters
e) Advantages and disadvantages
f) Interpretation of results
E. Solid state detectors
Page 12.8
Radiation Safety Training
1. Energy bonds, holes, electron traps
2. Scintillation detectors
a) Basic operation
b) Components - crystal, PMT tube
3. Advantages and disadvantages
F. Maintenance of equipment and the calibration of radiation measurement devices
1. Regulatory requirements
2. Inspection frequency for equipment
3. Specify the following:
a. Practical instrument appearance check
b. Battery check
c. Calibration certificate check
d. Check source response check
Module 8: Biological Effects (Radioisotopes, Accelerators and Transportation)
OBJECTIVES:
1. To inform trainees of the relative sensitivity of various cells of the body to
radiation.
2. To develop an understanding of the types of biological effects of radiation on the
various organs and tissues of the body.
3. To acquaint the trainee with the stochastic and deterministic effects of radiation on
living matter.
4. To acquaint trainees with the genetic effects of radiation.
MODULE CONTENT:
A. Types of effects
1. Stochastic and non-stochastic
2. Somatic
a) short term vs long term
b) early radiation effects
c) late radiation effects
3. Genetic
B. Radiosensitivity
Page 12.9
Radiation Safety Training
C. Dose-effect relationship
1. Classification of doses
2. Effects of acute radiation
3. Chronic doses and late effects
D. Sources of radiation exposure
1. Hazards from external radiation
2. Hazards from internal radiation
E. Clinical effects on Humans
1. Factors that determine what effect a given dose will have
a) Part of body exposed
b) Rate of exposure
c) Extent of body part that receives exposure
d) Age of the individual
e) Biological variations among individuals
F. Radiation hazard in proper perspective
1. The philosophy of radiation benefits and risks
2. Personal exposure
a) man-made sources
b) background
3. Radiation risks to trainees from their specific job task(s)
4. Maximum permissible doses for workers
Module 9: Effects of Radiation on the Foetus (Radioisotopes, Accelerators and
Transportation)
OBJECTIVES:
1. To provide workers with knowledge of radiation effects on an unborn child.
2. To enable workers to make better judgements regarding radiation risks while
pregnant.
3. To explain the declaration of pregnancy procedure (see also Module 2, Regulatory
Requirements).
MODULE CONTENT:
A. Genetic effects:
Page 12.10
Radiation Safety Training
1. Definition: abnormality observed in offspring due to previous irradiation of a
parent's reproductive system
2. Increase in risk with any exposure
B. Teratogenic Effects:
1. Abnormality observed in offspring due to irradiation in utero
2. Increase in risk above threshold value
C. Explanation of the risks to the foetus from internal and external doses:
1. Review of specific risks associated with specific tasks
D. Protective measures
E. Specify activities that are not acceptable/personal choice.
F. Declaration of pregnancy [AEC Regulations 19(4)]
G. Dose limits while pregnant and the relative risks involved.
Module 10: Controlling Radiation Exposure (Radioisotopes, Accelerators and
Transportation)
OBJECTIVES:
1. To relate time, distance and shielding as methods of reducing radiation exposure.
2. To impress on the trainees the importance of the ALARA principle - keeping
exposures as low as reasonably achievable, taking into account economic and
social factors.
3. To develop an understanding of the hazard from contamination when handling
loose radioactive material, and contamination control practices.
MODULE CONTENT:
A. Control of external radiation exposure
1. Time - Dose = Dose Rate x Time
2. Distance
a. Definition of inverse square law
- explain that not all fields diminish at a rate proportional to the square of
the distance from the source since many decrease at much slower rates,
depending on the geometry of the conditions of exposure
b. Example problems
3. Shielding
a. Definition of half-value layer (HVL)
Page 12.11
Radiation Safety Training
b. HVL for various shielding materials
c. Example problems
4. Use of time, distance and shielding in actual radioisotope work
5. Contamination control
B. Control of internal radiation exposure
1. Modes of entry into body
a. Inhalation
b. Ingestion
c. Absorption through skin
2. Leak testing of sealed sources
a. Requirement
b. Procedures
3. Contamination control
a. Monitoring procedures
- instrument survey
- wipe survey
- regulatory limits
b. Contamination prevention
- handling equipment
- personnel protective equipment
c. Decontamination techniques
4. Bioassay
a. Requirements
b. Method and frequency
c. Body counting vs urinalysis
d. Contamination control
Module 11: Transportation Requirements (Radioisotopes, Accelerators and
Transportation)
OBJECTIVES:
1. To acquaint the trainee with the basic requirements of regulations governing the
transport of radioactive material.
2. To provide information necessary to properly label transportation containers.
3. To provide information necessary to properly describe radioactive material on
shipping documents.
Page 12.12
Radiation Safety Training
MODULE CONTENT:
A. Receipt of radioactive material
1. Survey requirement
2. Receipt record
B. Shipment of radioactive material
1. Preparing material for shipment
2. Completing shipping document
3. Surveying package
C. Transportation of radioactive material
1. Packaging radioactive material
2. Package certificate/special form material certificate/special arrangement
certificate requirements
3. User licensing requirements
4. Transport security
5. In-transit transport licence
6. Labelling transport container
7. Completing shipping document
8. Quality Assurance program
D. Placarding of transport vehicle
Module 12: Practical Exercises [dependent on job tasks] (Radioisotopes, Accelerators
and Transportation)
OBJECTIVES:
1. To provide the trainee with the practical examples of the application of time,
distance and shielding.
2. To demonstrate correct handling practices and precautions for nuclear substances.
3. To demonstrate the method for performing a wipe and survey and how to evaluate
the wipe.
4. To demonstrate the correct method of leak testing sealed sources. (radioisotopes)
Page 12.13
Radiation Safety Training
MODULE CONTENT:
Recommended practical exercises
A. Time, distance and shielding calculations
B. Radiation detection instrument verification (instrument's accuracy, precision and
reproducibility under conditions of intended use)
C. Radiation scattering
D. Contamination surveying and analysis
E. Leak testing of sealed sources
F. Surface decontamination methods
G. Contamination control
H. Personnel contamination and decontamination:
1. radioactive spills
2. damage to sealed sources
3. device malfunctions
I. Radiation emergency procedures
J. Hands-on exercises (simulations, role playing)
Module 13: Accelerator (accelerators)
OBJECTIVES:
1. To acquaint the trainee with the sources of radiation at the accelerator, including
activation of components.
2. To acquaint the trainee with the safety interlock system at the accelerator facility.
3. To inform the trainee of the safety procedures at the accelerator facility.
MODULE CONTENT:
A. Sources and potential sources of radiation from the accelerator.
B. Shielding principles used at the accelerator.
C. Description of the interlock system and of the radiation warning system.
D. Procedures to use the interlock system.
E. Other procedures including entry to the accelerator areas and handling of activated
parts.
F. Use of radiation instrumentation.
Page 12.14
Radiation Safety Training
The topics in the modules should reflect what is required to be delivered according to the analysis
phase of SAT (see Appendix 1). The instructor can vary the length and content of each module
depending on the understanding, qualifications and experience of the participants.
Page 12.15
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 13: Recommended matrix tables for worker
training programs:
Table 13-1: Radioisotope Worker Training Matrix
Table 13-2: Medical Accelerator Training Matrix
Table 13-3: Transportation Training Matrix
THIS SECTION APPLIES TO RADIOISOTOPE, MEDICAL ACCELERATOR AND TRANSPORTATION
LICENSEES AND APPLICANTS FOR ANY RADIOISOTOPE, MEDICAL ACCELERATOR OR
TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION SAFETY INSTRUCTORS
Table 13-1: Radioisotope Worker Training Matrix
Module 1: Radiation Orientation Lecture; Module 2: Regulatory Requirements;
Module 3: Operating and Emergency Procedures; Module 4: Structure of Matter;
Module 5: Radiation and Radioactivity; Module 6: Radiation Units;
Module 7: Radiation detection and Measurement; Module 8: Biological Effects;
Module 9: Effects of Radiation on the Foetus Module 10: Control of Radiation Exposure;
Module 11: Transportation Requirements; Module 12: Practical Exercises
Recommended Training Modules
1
Job Task/Use Description 1 2 3 4 5 6 7 8 9 10 11 12
O R P M R U D B C T P
R *
E R A A N E I O R R
D - means detailed course I G O T D I T O F N A A
E U C T I T E O T N C
I - means introductory course N L E E A S C E E T S T
P - means professional qualification T A D R T T F T R P I
A T U I I F U O O C
* S
- indicates female worker and T O R O O E L R A
I R E N N C T L
supervisor of female workers O Y S T
N S
Sealed Sources
Sealed source manufacturers D D D D D D D D I D I D
Radioactive device maintenance, installation. P D D I D D D I I D D D
Radioactive device service, dismantling P D D I D D D I I D D D
Radioactive device calibration P D D I D D D I I D D D
Handles check source <50 MBq I D D I I I I
Handles check source >50 MBq D D D I D D D I *I D I D
Electron capture detector/ Gas Chromatography /X- I I D I I * I
ray Fluorescence
Bone analyser D I D I I I I D I
Table 13-1 continued
Irradiators/calibrators D D D I P P D P I D I
Self-shielded irradiators (e.g. Gammacell) D D D I P P P I * D I
Brachytherapy D D D P P P P D D D I D
Teletherapy D D D P P P P D D D D
Static eliminators I D D I I I
External radiation fields < 25 µSv/h I I I I I I I * I I
1Refer to pages 12.1 to 12.15 for a description of training module content
Job Task/Use Description 1 2 3 4 5 6 7 8 9 10 11 12
O R P M R U D B C T P
R *
E R A A N E I O R R
D - means detailed course I G O T D I T O F N A A
E U C T I T E O T N C
I - means introductory course N L E E A S C E E T S T
P - means professional qualification T A D R T T F T R P I
A T U I I F U O O C
* S
- indicates female worker and T O R O O E L R A
I R E N N C T L
supervisor of female workers O Y S T
N S
External radiation fields > 25 µSv/h D D D I D D D D D D I D
Tower scanning (licensed as portable gauge) D I D I I D I I I I D I
Unsealed Sources
Laboratory / benchwork < 10 MBq I I D I I D I * I D
Laboratory / benchwork > 10 MBq D D D I D D D D D D D
Nuclear medicine pharmacy D D D D D D D D D D D D
Radioisotope processor/supplier < 10 GBq P D D P P P D D I D D D
Radioisotope processor/supplier > 10 GBq P D D P P P D D D D D D
Diagnostic human studies P D D P P P P D D D I D
Therapeutic human studies P D D P P P P D D D I D
Caring of radioactive patients D I D I I I I D I D
Table 13-1 continued
Veterinary studies D D D D D I D I D D D D
Tracer field studies D D D P D P D D D D D D
Subsurface tracer studies D D D I D I D D D D D D
Industrial or Academic
Instrument technician I I D I I I I I I I
Isotope technician D D D D D D D I D I I
Laboratory technician D I D I D D D D D D D
Research students I I D D D D D D D D I
Scientists/Engineers [field] P D I I * I
Scientists/Engineers [lab] P D I I I I * I
Job Task/Use Description 1 2 3 4 5 6 7 8 9 10 11 12
O R P M R U D B C T P
R *
E R A A N E I O R R
D - means detailed course I G O T D I T O F N A A
E U C T I T E O T N C
I - means introductory course N L E E A S C E E T S T
P - means professional qualification T A D R T T F T R P I
A T U I I F U O O C
* S
- indicates female worker and T O R O O E L R A
I R E N N C T L
supervisor of female workers O Y S T
N S
Well logger D D D I I D D D I D D D
Gauge user [fixed] D D D D D D I I I I D
Gauge user [portable] I D D I I * I D I
Radiation Safety Officer (portable gauges) I D D D I I I I I D I
Animal caretakers I I D I I I I I D I I
Tracer technicians [unsealed sources] D D D P D P D D D D D D
Table 13-1 continued
Irradiator operator D D D I I D D D D D D D
Calibration technician D D D I I D P D D D I I
Radiation Safety Officer (Industrial, general) P D D P P P D D D D D D
Medical
Nuclear medicine physician [therapeutic] P P D P P P P P P P I
Nuclear medicine physician [diagnostic] P P D P P P P P P P D
Medical technologist [diagnostic nuclear medicine] P P D P P P P P P P D
Medical technologist [therapeutic nuclear medicine] P P D P P P P P P P D
Medical technologist [radiotherapy nuclear medicine] P P D P P P P P P P D
Laboratory technician D D D I D D D I D D D
Medical physicist P P D P P P P P P P I D
Radio pharmacist P P D P P P P P P P D D
Radio pharmacy technician D D D D D D D D D D D D
Radiation Safety Officer D D D D D D D D D D D D
Nurse [diagnostic] D D D I I I I I I
Nurse [therapeutic] D D D D D D D D D I
Nurse [radiotherapy] D D D I D D D D D I
Job Task/Use Description 1 2 3 4 5 6 7 8 9 10 11 12
O R P M R U D B C T P
R *
E R A A N E I O R R
D - means detailed course I G O T D I T O F N A A
E U C T I T E O T N C
I - means introductory course N L E E A S C E E T S T
P - means professional qualification T A D R T T F T R P I
A T U I I F U O O C
* S
- indicates female worker and T O R O O E L R A
I R E N N C T L
supervisor of female workers O Y S T
N S
Table 13-1 continued
Ward aid/orderly I D I I I
Support staff
Security personnel I D I I I D
Housekeeping/janitorial I D I I
Shipping, receiving and distributing I D I I I I D D I
Waste disposal I D I D D I D D D D
Service/maintenance I D I I
Administrative
Administrators I I I I I
Reception staff I I I I I
Safety officer D I I D D I
Other
Note: If the job task for the radioisotope worker does not appear in Table 13-1, contact the
AECB Licence Assessment Officer (LAO) responsible for the licensee. In such cases a
determination of job classification is made on the basis of the worker's task list supplied to
the LAO.
Table 13-2: Medical Accelerator Training Matrix
Module 1: Radiation Orientation Lecture; Module 2: Regulatory Requirements;
Module 3: Operating and Emergency Procedures; Module 4: Structure of Matter;
Module 5: Radiation and Radioactivity; Module 6: Radiation Units;
Module 7: Radiation detection and Measurement; Module 8: Biological Effects;
Module 9: Effects of Radiation on the Foetus Module 10: Control of Radiation Exposure;
Module 11: Transportation Requirements; Module 12: Practical Exercises
Module 13: Medical Accelerator module(s)
Recommended Training Modules
2
Job Task/Use Description 1 2 3 4 5 6 7 8 9 10 11 12 13
D - means detailed course O R P M R U D B * C T P A
I - means introductory course R E R A A N E I F O R R C
I G O T D I T O O N A A C
P - means that this training should be part of the E U C T I T E E T N C E
person's professional qualification N L E E A S C E T R S T L
T A D R T T F U O P I E
* - required for female workers and all supervisors A T U I I F S L O C R
T O R O O E R A A
I R E N N C T L T
O Y S T O
N S R
Medical particle accelerator
Radiation-oncologists P I P P P P P P P P P
Medical physicists D P D P P P P P P P I P P
D D
Radiotherapy technologists D I P P P P P P P P P I
Support technologists (dosimetry, electronics) D I D D D D D D I D D D
Administration/office workers D I I
Table 13-2 continued
Service/maintenance (building workers) D I
RSO D D D D D D D D D D D D D
Head of the safety group D D D D D D D D D D D D D
Non-technical support staff D I
Note: If the job task for the accelerator worker does not appear in Table 13-2, contact the
AECB Licence Assessment Officer (LAO) responsible for the licensee. In such cases a
determination of job classification is made on the basis of the worker's task list supplied to
the LAO.
Refer to pages 12.1 to 12.15 for a description of the training module content
2
Table 13-3: Transportation Training Matrix
Module 1: Radiation Orientation Lecture; Module 2: Regulatory Requirements;
Module 3: Operating and Emergency Procedures; Module 4: Structure of Matter;
Module 5: Radiation and Radioactivity; Module 6: Radiation Units;
Module 7: Radiation detection and Measurement; Module 8: Biological Effects;
Module 9: Effects of Radiation on the Foetus Module 10: Control of Radiation Exposure;
Module 11: Transportation Requirements; Module 12: Practical Exercises
Recommended Training Modules
3
Job Task/Use Description 1 2 3 4 5 6 7 8 9 10 11 12
D - means detailed course O R P M R U D B C T P
*
I - means introductory course R E R A A N E I O R R
I G O T D I T O F N A A
P - means that this training should be E U C T I T E O T N C
part of the person's professional N L E E A S C E E R S T
T A D R T T F T O P I
qualification A T U I I F U L O C
* S
- required for female workers and all T O R O O E R A
I R E N N C T L
supervisors O Y S T
N S
Transport packaging of prescribed
nuclear substances (and licences)
RSO D D D P P P P P P P D D
Technical staff D D D P P P P P P P D D
D D D D D D D
Conveyance operator I I D I I I I I I I D D
Table 13-3 continued
Shippers/Receivers D D D I D D D D D D D D
I
Stores personnel I D D I I I I I I I D D
Package design approval certificate;
and, endorsement of a foreign design
approval certificate
Designer I D D D D D D D D D
Fabrication personnel I D D
Quality Assurance Inspector I D D
Refer to pages 12.1 to 12.15 for a description of the training module content
3
Job Task/Use Description 1 2 3 4 5 6 7 8 9 10 11 12
D - means detailed course O R P M R U D B C T P
*
I - means introductory course R E R A A N E I O R R
I G O T D I T O F N A A
P - means that this training should be E U C T I T E O T N C
part of the person's professional N L E E A S C E E R S T
T A D R T T F T O P I
qualification A T U I I F U L O C
* S
- required for female workers and all T O R O O E R A
I R E N N C T L
supervisors O Y S T
N S
Maintenance and inspection personnel I D D I D D D D D D D D
Special form (radioactive material)
nuclear substances certificate
Designer D D D D D D D D I D
Fabrication technicians D D D D D D D D D D I D
Quality control technicians D D D D I D D I I D I D
Authorization to transport nuclear
substances in non-conforming
packages
Consignor D D D D D D D D D D D D
Conveyance operator D I D I I I I I I I D D
Table 13-3 continued
Receiver D D D D D D D D D D D D
Surveyor or escort D D D P P P P P D D D D
Emergency response team D D D D D D D D D D D D
Documentation
Shipping and receiving I I I I I I I I
Surveyor I I I I I P I I
Labelling and placarding
RSO D D D P P P D P I P D D
D
Shipping and receiving clerk I D D I I I D I D I D I
I
Surveyor P D D I I P P D D D D I
I
Conveyance operator I I D I I I I I I I I
Job Task/Use Description 1 2 3 4 5 6 7 8 9 10 11 12
D - means detailed course O R P M R U D B C T P
*
I - means introductory course R E R A A N E I O R R
I G O T D I T O F N A A
P - means that this training should be E U C T I T E O T N C
part of the person's professional N L E E A S C E E R S T
T A D R T T F T O P I
qualification A T U I I F U L O C
* S
- required for female workers and all T O R O O E R A
I R E N N C T L
supervisors O Y S T
N S
Stores personnel I D D I I I I I I I I
Surveying and determining transport
index
RSO D D D P P P D P D P D D
Radiation surveyor D D D D P P P P D D D D
I
Note: If the job task for the transportation worker does not appear in Table 13-3, contact the
AECB. In such cases a determination of job classification is made on the basis of the
worker's task list supplied to the Licence Assessment Officer.
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Section 14: Review objectives
THIS SECTION APPLIES TO LICENSEES AND APPLICANTS FOR RADIOISOTOPE, MEDICAL
ACCELERATOR AND TRANSPORTATION LICENCE PREPARED BY THE AECB, AND TO RADIATION
SAFETY INSTRUCTORS
Radiation Safety Training
Review objectives
This section provides an opportunity to review the objectives for your radiation safety
program outlined in this Guide. You can use this section like a check list to verify that your
training program meets the recommendations of the AECB. If you find areas where your
training program does not meet a specific objective, review the appropriate sections of the
Guide or consult with AECB staff for assistance.
1. Objectives for Systematic Analysis of Jobs to be Performed
Verify that:
1.1 A systematic method is used for identifying and selecting tasks for training to prepare individuals to do
their jobs.
1.2 Tasks for continuing and initial training are differentiated.
1.3 The analysis is adequate for development of learning objectives.
1.4 The analysis is kept current as job performance requirements change.
2. Objectives for the Development of Learning Objectives
Verify that:
2.1 There are learning objectives related to knowledge, skills and abilities for each task.
2.2 Learning objectives contain actions, conditions and standards needed for job performance.
2.3 There are procedures to modify learning objectives as job performance requirements change.
3. Objectives for Design Implementation
Verify that:
3.1 The goals, objectives, responsibilities and authority of the training organization and staff are clearly stated.
3.2 Qualifications and training requirements for the training staff address both the acceptable subject matter
and instructional skills.
3.3 Training is acceptably organized and sequenced and instructional setting are appropriate to tasks.
3.4 Lesson plans provide for consistent training delivery.
3.5 Existing instructional materials have been evaluated based on training requirements.
3.6 Training is conducted in an acceptable manner and records are maintained.
Page 14.1
Radiation Safety Training
4. Objectives for Training Evaluation
Verify that:
4.1 Exemptions from training are objectively determined.
4.2 Trainee performance is regularly evaluated using job performance measures and objectives.
4.3 Trainees who perform below minimum standards during initial and re-qualification training
receive remedial training and are retested.
4.4 Precautions are in place to prevent test compromise.
5. Objectives for Program Evaluation
Verify that:
5.1 Methods are in place to systematically evaluate the effectiveness of training programs and
that training programs are updated.
5.2 Feedback from trainee tests, on-the-job experiences, and supervisors are used in program
evaluation.
5.3 Instructor and trainee critiques are used in program evaluation.
5.4 Both internal and external program inspections are used to evaluate the program.
5.5 Training staff are routinely and objectively evaluated.
Page 14.2
Radiation Safety Training
for Radioisotope, Medical Accelerator
and Transportation Workers
Proposed Regulatory Guide
Appendixes
Radiation Safety Training
Appendix 1
Systems Approach to Training (SAT)
A chart illustrating the phases of SAT is shown on the next page. The schematic
outlines the process used in developing an effective training program. The main
principles of SAT are summarized as follows:
C Training meets actual job needs: the job performance needs are determined by
analysis of the duties where such duties could have an impact on the environment,
the public or on the safe and reliable operation by the licensee. Such an analysis
also serves as a basis for the development of learning objectives, training materials
and job performance measures.
C The training program is systematic: training programs are effectively organized,
directed and supported in a systematic way.
C The training is documented: training material defines the knowledge and skills
required to meet the learning objectives. Training delivery employs the principles
of good instructional presentation and conveys accurate information consistently
and clearly.
C Training is conducted by qualified staff: training staff possess the necessary
subject matter expertise, experience and instructional skills to discharge their
assigned duties.
C Training is evaluated for effectiveness: the training program is evaluated, and
revised as necessary, so that on-the-job competence is attained and maintained.
Trainees are evaluated on their mastery of the learning objectives and receive
prompt feedback on their performance.
Page A1.1
Radiation Safety Training
Analysis Evaluatio Training Conduct Validatio
research n Design Design implementation n
activity for for monitors
effectiveness development effectiveness
and efficiency and initial after 90 days
implementati
on
Conduct needs Establish Preliminary Implement Serves a *
and task performance research, training plan percentage of
analyses standards Prepare course F
training plan, Conduct training graduates
Identify Determine Develop through
training needs trainee entry training Evaluate trainee validation
level package, performance surveys and E
Develop during and observation of
Develop: Develop learning admin following workers'
Performance objectives package training performance
objectives E
Training Assess Document the Assesses
objectives Organize
effectiveness by: training accuracy of
Supporting learning
Predicting on- original analysis
specifications objectives
the-job and D
performance effectiveness of
Specify
Diagnosing the design and
learning
participant conduct phases
objectives
shortfalls B
Identifying Determines
Produce
incidents of effectiveness
training
over-training and removes
package impediments to A
Efficiency: the application
Select and
maximum of learning
train
number of instructors
trainees meeting C
the training
objectives in the
shortest time K
* Denotes that feedback applies to all phases and components of SAT
Page A1.2
Radiation Safety Training
Appendix 2
Training Evaluation
READ APPENDIXES 2 AND 3 IF YOU ARE A RADIOISOTOPE LICENSEE, A MEDICAL ACCELERATOR
LICENSEE, AN INSTRUCTOR OR A TRAINING EVALUATOR
The following is an example of one style of reporting form for the evaluation of
AECB licensees' radiation safety training programs
File:
Date:
Licensee:
Instructor(s) name(s)
Title(s)
Evaluator(s):
For each of the statements below, circle the number that comes closest to expressing the inspector's opinion.
Scoring guidance:
C 1 is unacceptable
C 2 is conditionally acceptable
C 3 is acceptable
A score of
C 21 to 41 indicates an unacceptable training program. Complete review required.
C 42 to 62 indicates a conditionally acceptable training program. Improvement needed.
C 63 indicates an acceptable training program. No action is required.
Instructor Remarks
1) Organization and preparation
1 2 3
Unacceptable Acceptable
Page A2.1
Radiation Safety Training
Other questions in this section evaluating the instructor could include:
C Knowledge and mastery of subject
C Organization and preparedness
C Confidence and voice projection
C Enthusiasm
C Clear and understandable presentation of course content
C Ability to involve trainees
C Ability to respond to questions
C Overall assessment
Additional sections of the evaluation report form may include, but are not necessarily
limited to:
Preparation
C Was the trainer present before the lesson?
C Were the course notes distributed in advance?
C Were the classroom and equipment ready?
Classroom presentation
C Did the class start on time?
C Did the instructor encourage trainees to participate?
C Were trainees' responses handled well?
C Were AV materials used effectively?
C Was there a good working atmosphere?
C Did the instructional methods maintain interest?
C How suitable was the amount of lecturing?
C How suitable was the amount of discussion?
C How useful were the discussions?
C How was the speed of presentation?
C Did participants have enough practice with new skills?
Training Content
C Were the lesson objectives clearly stated at the outset?
C Were the stated training objectives met?
C Was the course content as outlined?
C Was the course content relevant to the workplace?
C Was the course content appropriate for the level of trainees?
C Was the length of time taken for this course appropriate?
C Was all the course material covered in the time available?
C Was there time for a question-and-answer period?
Page A2.2
Radiation Safety Training
C How thoroughly were the subjects covered?
C What was the overall value of this course?
C Were the results achieved on this course acceptable?
Quality of training facilities
C Was the classroom appropriate for the size of class and type of course?
C Was the room well lit and ventilated?
C Was the room properly equipped?
The evaluation form should also provide space for:
Notes on any changes and new materials used in the program
C Was the program presented as accepted by the AECB?
- List any new materials used:
- List any changes made to the accepted program:
Comments and suggestions
C Information of most value was:
C Information of least value was:
C Suggestions to improve the course:
C Specific concerns the inspector would like to see addressed:
C Additional comments:
Another style of evaluation form acceptable to the AECB is shown on the following
page.
Page A2.3
Radiation Safety Training
Training Delivery Evaluation Form
Trainer ................. Course ................... Level ..................
AECB Evaluator ................. Class Size ................... Date ..................
Training Centre ................. Supervisor ................... Time ..................
Code: S: satisfactory I: some improvement needed U: unsatisfactory
# ITEM S I U COMMENTS
PREPARATION
Trainer present before lesson
Course notes distributed before teaching
started
Audio/visual aids ready
Class room ready
FACILITIES
Appropriate for teaching-well lit,
ventilated, appropriate temperature
Quiet, free of distractions
Room layout appropriate for:
Class size
Type of course
Use of AVs
Demonstrations
TRAINING MATERIAL CONTENT
Relevant to the workplace
Relevant to trainees' jobs
Current and technically correct
"Nice to Know" material is clearly identified
Page A2.4
Radiation Safety Training
Appendix 3
Validation of Radiation Training
Validation (refer to Section 7 dealing with program verification) provides a means of
assessing the overall level of knowledge retained by workers It identifies the areas
where retraining is essential. This appendix contains sample forms and topics for
trainee testing designed to assess the effectiveness of radiation training after 90 days.
Licensee: File No
Date:
1. Employee's Background
1.1 Department ___________________________________________________________
1.2 Position _____________________________________________________________
1.3 How long employed by this licensee?
_________________________________________
1.4 When last did you receive radiation training or re-training by this company?
________________________________________________________________
1.5 Have you received any previous training in radiation before joining this company?
(If yes, give details)
________________________________________________________________
1.6 When? ________________________________________________________________
1.7 Where? _______________________________________________________________
(Note: responses to this page do not count towards the total marks)
Page A3.1
Radiation Safety Training
Questionnaire
The test should take the form of a questionnaire. A number of approaches can be used
to determine the effectiveness of the training.
For example, multiple choice questions, where the participants check or circle the
correct answer:
The types of radiation usually associated with licensee's activities are mainly:
a) ionizing radiation
b) non-ionizing radiation.
Are you more likely to get lung cancer:
a) from working with radioactive materials
b) from smoking
Multiple choice questions can also ask participants to explain their choices:
Are the radiation exposure levels and doses received by workers in this type of licensed
facility today, as compared to those received by workers 20 years ago, on the average,
likely to be:
a) lower
b) higher
c) the same
Briefly explain your choice
_________________________________________________________
_________________________________________________________
Page A3.2
Radiation Safety Training
Alternatively, participants can be required to complete statements that demonstrate
their knowledge of the subject:
Identify, and write in the space below, the type of radiation being described.
a) - Relatively heavy particles, each consists of two protons and two neutrons
- Very low penetrating power. A sheet of paper will easily stop them
- Most important for their possible effects in the lungs.
This form of ionizing radiation is known as ___________ radiation.
b) - Consists of electromagnetic radiation
- Will penetrate most materials including wood, concrete, and steel to various depths
- Generally affects all organs of the body from an external source.
This form of ionizing radiation is called _____________ radiation.
Another approach is to use the true/false or yes/no testing technique:
TRUE FALSE
Time, distance and shielding are important factors in
radiation protection.
a TIME: To reduce exposure, workers may have to be
rotated by working for shorter periods in high-
radiation areas
b DISTANCE: You can reduce your exposure by
keeping as far from the source of radiation as
practicable.
c SHIELDING: You can reduce exposure from density
of level gauges by placing heavy materials, such as
lead, between the radiation source and the work area.
Page A3.3
Radiation Safety Training
Participants can also be asked simply to respond to direct questions about radiation:
a) To reduce exposure, spills of radioactive material must be cleaned up quickly. What
methods are used for cleaning up spills in your workplace?
______________________________________________________________
______________________________________________________________
Testing topics
Whichever method, or combination of methods is used, the following topics should be
covered in sufficient detail to reflect the material covered in the training course that is
being validated:
C Terms, concepts and types of radiation
C Radiation
C Protective measures
C Regulations
C Dosimetry
C Worker responsibility
Refer to Appendix 5 for a detailed list of the topics covered under these headings.
Page A3.4
Radiation Safety Training
Appendix 4
Training Schedules
The following is an example of a training schedule. It formalizes a licensee's
program schedule and also helps AECB Inspectors plan for evaluation and validation
exercises.
1999 SCHEDULE: RADIATION SAFETY TRAINING FOR WORKERS
Initial___ Basic Retraining___
Module:
Date ID Trainee Classification Location Instructor
(YMD) Number Name Init.
99-1-20 12345 Doe J.H. Operator Training Centre Thomas D.
Alternatively time posters can be used to display monthly, quarterly or yearly
training schedules.
Page A4.1
Radiation Safety Training
Appendix 5
Radiation Safety Training Program Validation:
Summary of recommended topics (referred to in Appendix 3).
Note: the following list does not replace the information shown in the modules described in
section 12, but outlines topics that may be considered for developing validation exercises,
depending on specific needs:
1 Definitions, terms and concepts
1.1 Definitions and terms:
alpha ("), beta ($) and gamma (()
Atomic Radiation Worker (ARW)
atoms
decay chain
electrical charge
electrons, protons and neutrons
elements
half-life
ionization, and ionizing radiation
isotopes
maximum permissible concentration
maximum permissible exposure and dose
radioactivity
radon progeny
units [sievert (Sv), gray (Gy), becquerel (Bq), WL, WLh, WLM, joules]
uranium
1.2 Radiation safety:
function and use of radiation safety manual(s)
detection of radiation
radiation in the licensed facility (alpha, beta and gamma)
absorption of radiation
kinds of radiation exposure
control of external radiation exposure
control of internal radiation exposure
units for measuring radiation exposure
contamination and decontamination
restricted areas
warning signs
Page A5.1
Radiation Safety Training
protective devices and clothing
dosimetry - types (e.g. TLD badge and how to comply)
occupational dose
whole body exposure
bioassay
radiological surveys
direct and indirect measurement of radioactive contamination
work place monitoring (air quality and quantity, dose rates, dust control, wipe
samples)
controls for spills and unusual occurrences and procedures for clean-up
operating methods
managing waste materials
actions to be taken for personal contamination, including if a medical injury is
involved
actions to be taken in case of fire and other emergencies
conditions which require bioassay
prenatal exposure limits and duties of a pregnant employee
1.3 Effects of radiation:
how radiation affects the human body
genetic and somatic effects of radiation
background and low level radiation
evidence of radiation effects
risk from radiation exposure
pregnant workers
2 Concepts:
2.1 Explain the following concepts:
Radiation safety in the work place
Good work practices
As low as reasonably achievable (ALARA principle)
Codes of practice
Time-Distance-Shielding, as it relates to dose reduction
No eating, drinking, smoking, chewing or application of cosmetics in a "radioactive
" work area
Worker responsibilities in radiation safety (to self, co-workers and public)
Ventilation controls
The licensee's operations cycle
Waste disposal methods
Page A5.2
Radiation Safety Training
Actions required for a lost, late return or damaged personal dosimeter
Importance of informing supervision of any radiological incident
Common problems found during radiation safety reviews and inspections
3 Regulations and licensing
3.1 Regulations
The Atomic Energy Control Act (AEC Act)
Atomic Energy Control Regulations (AEC Regulations)
Transport Packaging of Radioactive Materials Regulations (TPRMR)
NSC Act and pursuant regulations
WHMIS
3.2 Licensing
The type of licence authorized by the AECB for this licensee.
Supporting documents, such as procedures manuals.
Page A5.3
Radiation Safety Training
Appendix 6
Radiation Safety Information Sheets
Devices, licensed activities, types and quantities of radioisotopes and source sizes for
which Radiation Safety Information Sheets are used, include:
C Electron capture/gas chromatography detectors
C Liquid scintillation counters
C Dewpointers
C Smoke detectors
C Surge voltage detectors
C Static detectors
C Static eliminators
C Bacterial growth incubators such as Bactec
C Tritium exit signs (aircraft and buildings)
C Tritium watches and gun sights
C All radioactive prescribed substances, sources and devices for which no licence is
required
C Small radioisotope mini-generators
C Sealed sources <50 MBq scheduled quantities
Page A6.1
Radiation Safety Training
Appendix 7
Training record
The following is a sample training record form that is acceptable to the AECB. The
detail of the topics will vary. Test scores should be recorded. What was taught will be
monitored by AECB staff.
Employee name:
Job classification:
Licensee:
Topic Training Coordinator's Name Pass (p) Fail (f) Date
Module 1: Orientation lecture
Module 2: Structure of matter
Module 3: Radiation and radioactivity
Module 4: Radiation units
Module 5: radiation detection and
measurement
Module 6: Control of radiation exposure
Module 7: Biological effects
Module 8: Regulatory requirements
Module 9: Operating and emergency
procedures
Module 10: Transportation requirements
Module 11: Laboratory exercises (dependent
on job tasks)
Other training (specify)
Page A7.1
Radiation Safety Training
Appendix 8
Radiation Safety Data Sheet (RSDS)
Page A8.1
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B Tel: (613) 995-5894 Fax: (613) 995-5086
Ottawa, Canada 24 Hour Emergency Hotline: (613) 995-0479
K1P 5S9
Radiation Safety Data Sheet
This data sheet presents information on radioisotopes only.
For information on chemical compounds incorporating this radionuclide, see the relevant Material Safety Data Sheet.
Detailed information is available on the CNSC Radioisotope Fact Sheet
Part 1 - Radioactive Material Identification
Chemical Symbol___________________ Common Names____________________
Atomic Weight ___________________ Atomic Number____________________
Part 2 - Radiation Characteristics
Physical Half-Life_____________________________
Principle Max
E eff
E Dose Rate at Distance Shielding Required
Emissions (MeV) (MeV) (mSv/h.GBq)
Neutrons (@ 1m) TVL Paraffin___________cm
Gamma (() / X- (@ 1m) TVL Lead________________
Rays
Beta* ($) (@ 1m) Range in Plexiglass______cm
Alpha (") (not applicable) (not applicable)
* Where Beta radiation is present, Bremsstrahlung radiation will be produced. Shielding may be required.
Progeny_______________________________________________________________________
Part 3 - Detection and Measurement
Methods of detection (in order of preference)
1.
2.
3.
4
Dosimetry
Whole Body 9 Skin 9 Extremity 9 Neutron 9
Internal: Critical Organ(s):
Part 4 - Preventive Measures
Always use the principles of time, distance and shielding to minimize dose
Engineering Controls:
Personal Protective Equipment (for normal handling, unsealed sources only. Always use gloves and glass and other
personal protective equipment and clothing as appropriate to the material handled)
Special Storage Requirements
Part 5 - Control Levels
F (fast) M (moderate) S (slow)
Ingestion Inhalation Ingestion Inhalation Ingestion Inhalation
Maximum Release Concentration
Exemption Quantity (EQ)
Part 6 - Non-Radiological Hazards
Elemental Toxicity (LD )
50
Emergency Procedures
The following is a guide for first responders. The following actions, including remediation, should be carried out by qualified
individuals. In cases where life threatening injury has resulted, first treat the injury, second deal with personal
decontamination.
Personal Decontamination Techniques
C Wash well with soap and water and monitor skin
C Do Not abrade skin, only blot dry
C decontamination of clothing and surfaces are covered under operating and emergency
procedures
Spill and Leak Control
C Alert everyone in the area
C Confine the problem or emergency (includes the use of absorbent material)
C Clear area
C Summon Aid
Emergency Protective Equipment, Minimum Requirements
C Gloves
C Footwear Covers
C Safety Glasses
C Outer layer or easily removed protective clothing
C Suitable respirator selected
Radiation Safety Training
Appendix 9
References
The AEC Regulations
Nuclear Safety and Control Act
Proposed Regulations
AECB Compliance Policy
AECB Radiation Protection Training Policy
Public Service Commission of Canada, Systems Approach to Training, revised
edition, April 1984.
Treasury Board Manual: Training Guide
OCD Objectives and Criteria for Regulatory Evaluations of NGS Training Programs,
as revised May, 1995
Ontario Hydro, Nuclear Generation Division, Objectives and Criteria for Effective
Training Programs, dated October 1989
IAEA-TecDoc-525, Guidebook on Training, etc. dated 1989.
US Nuclear Regulatory Commission, NUREG-1220, Training Review Criteria and
Procedures
Evaluating Training Programs, The Four Levels, by Donald L. Kirkpatrick, 1994
Guideline for the Training of Workers at Uranium and Thorium Mining Facilities,
First draft, Prepared by SDS, July 26, 1996.
Page A9.1
DRAFT
REGULATORY
STANDARD
Certification of
Persons Working at
Nuclear Power Plants
C-204
Issued for public consultation by the
Canadian Nuclear Safety Commission
May 2001
DRAFT REGULATORY STANDARD
Certification of Persons Working at
Nuclear Power Plants
C-204
Issued for public consultation by the
Canadian Nuclear Safety Commission
May 2001
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in regulatory programs.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY STANDARD
CERTIFICATION OF PERSONS WORKING AT
NUCLEAR POWER PLANTS
C-204
About this Document
This standard describes the qualifications, training, the examinations and certifications that may be
required of nuclear power plant (NPP) personnel for positions referred to in a term or condition
of their operating licence.
The requirements described in the document become mandatory if this standard is incorporated as
a term or condition of the NPP's operating licence. Otherwise, the information contained in the
standard constitutes guidance material only.
This standard does not address refusal to certify, decertification and any rights to an opportunity
to be heard by the CNSC in either of these two situations. These matters are covered in
sections 11-13 of the Class I Nuclear Facilities Regulations.
Comments
The CNSC invites interested persons to assist in the further development of this draft regulatory
document by commenting in writing on the document's content and potential usefulness. Please
respond by August 31, 2001. Direct your comments to the postal or e-mail address below,
referencing file 1-8-8-204.
The CNSC will take the comments received on this draft into account when developing it further.
These comments will be subject to the provisions of the federal Access to Information Act.
Document availability
This document can be viewed on the CNSC Internet site at www.nuclearsafety.gc.ca. To order a
printed copy of the document in English or French, please contact:
Operations Assistant
Corporate Documents Section
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B, 280 Slater Street
Ottawa, Ontario K1P 5S9 CANADA
Telephone (613) 996-9505 Facsimile (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
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Certification of Persons Working at Nuclear Power Plants C-204
ii
C-204 Certification of Persons Working at Nuclear Power Plants
CONTENTS
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory Framework - licensing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 CNSC licensing process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1.3 Relevant legislation and justification for qualification, training, examination
and certification of workers and other persons . . . . . . . . . . . . . . . . . . . . . . . . . . 2
PART I - OBLIGATIONS OF PERSONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
SUBPART A - SENIOR HEALTH PHYSICISTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
2 Qualification Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3 Initial Training Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4 CNSC Examinations for Initial Certification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
5 Senior Health Physicist Transferring to Another Plant . . . . . . . . . . . . . . . . . . . . . . . 6
6 Continuing Training Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
7 Requalification Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
8 Certification Following Decertification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
SUBPART B - REACTOR OPERATORS, UNIT 0 OPERATORS, . . . . . . . . . . . . . . . . . . . . . . . . 9
SHIFT OPERATING SUPERVISORS AND SHIFT SUPERVISORS . . . . . . . . . . . . . . . . . . . . . . . 9
9 Qualification Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
Reactor Operators and Unit 0 Operators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
Shift Operating Supervisors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
Shift Supervisors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
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Certification of Persons Working at Nuclear Power Plants C-204
10 Initial Training Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
Reactor Operators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
Unit 0 Operators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
Shift Operating Supervisors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
Shift Supervisors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
11 CNSC Examinations for Initial Certification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
Reactor Operators . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
Shift Supervisors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
12 Extending the Period of Validity of an Examination . . . . . . . . . . . . . . . . . . . . . . . . 20
13 Advancement from Reactor Operator or Shift Operating Supervisor to Shift
Supervisor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
Reactor Operators to Shift Supervisors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
Shift Operating Supervisors to Shift Supervisors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
14 Continuing Training Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
15 Requalification Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
16 Certification Following Decertification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
PART II - OBLIGATIONS OF THE LICENSEE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
17 Fitness-for-Duty Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
18 Selection and Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
19 Simulator Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
20 Licensee's Examinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
21 Removal of a Person from a Position Requiring Certification . . . . . . . . . . . . . . . . . 29
22 Reinstatement of a Person to a Position Requiring Certification . . . . . . . . . . . . . . 30
23 Temporary Assignment to Other Positions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30
24 Retention of Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
GLOSSARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
iv
C-204 Certification of Persons Working at Nuclear Power Plants
Purpose
The purposes of this Canadian Nuclear Safety Commission (CNSC) regulatory standard are:
* To help applicants for CNSC licences and CNSC licensees understand their obligations
for the training, the examination and certification of persons for positions referred to in
a term or condition of the operating licence for nuclear power plants (NPPs or plants);
* To help applicants for CNSC licences and CNSC licensees understand their obligations
to ensure that only competent persons operate or supervise the operation of an NPP.
* To help persons seeking certification for positions at NPPs to understand their
obligations to complete the required training and examinations.
Scope
This standard describes the qualifications, the training and the examinations that may be required
of persons seeking certification for positions referred to in a term or condition of the licensee's
operating licence. In addition, this standard describes the licensee's obligations concerning the
certification of persons and also describes their obligations to ensure that only competent persons
operate or supervise the operation of an NPP.
The requirements described in this document become mandatory if this standard is incorporated as
a term or condition of the NPP's operating licence. Otherwise, the information contained in this
standard constitutes guidance material only.
This standard does not address refusal to certify, decertification and any rights to an opportunity
to be heard by the CNSC in either of these two situations. These matters are covered in
sections 11 to 13 of the Class I Nuclear Facilities Regulations.
Definitions
A glossary of terms defined specifically for the purpose of this Standard can be found on page 34.
1 Introduction
1.1 Regulatory Framework - licensing
The CNSC is the federal agency that regulates the use of nuclear energy and materials to
protect health and safety of persons, security and the environment, and to respect Canada's
international commitments on the peaceful use of nuclear energy.
The Nuclear Safety and Control Act ("the Act") requires persons or organizations to be
licensed by the CNSC for carrying out the activities referred to in section 26 of the Act
unless otherwise exempted. The associated regulations stipulate prerequisites for CNSC
licensing, and the obligations of licensees, workers and other persons.
1
Certification of Persons Working at Nuclear Power Plants C-204
1.2 CNSC licensing process
The CNSC typically applies a phased process to its licensing of nuclear facilities and
activities. For major facilities, this process begins with a consideration of the environmental
impacts of the proposed project, and proceeds progressively through site preparation,
construction, operation, decommissioning and abandonment phases.
The Act and regulations require licence applicants to provide certain information at each
licensing stage. The type and level of detail of this information will vary to accommodate
the licensing stage and specific circumstances.
At all licensing stages, applications may incorporate (directly or by reference) new or
previously submitted information, in accordance with legislated requirements and the best
judgement of the applicant. An application that is submitted at one licensing stage can
become a building block for the next stage.
Upon receipt of an application that is complete, the CNSC reviews it to determine whether
the applicant is qualified to carry on the proposed activity, and has made adequate provision
for the protection of the environment, the health and safety of persons, and the maintenance
of national security and measures required to implement international obligations to which
Canada has agreed. If satisfied, the CNSC may issue, renew, amend or replace a licence that
contains relevant conditions. Typically, this licence will incorporate the applicant's
undertakings and will contain other conditions that the CNSC considers necessary,
including those that reference or incorporate a CNSC regulatory standard.
1.3 Relevant legislation and justification for qualification, training,
examination and certification of workers and other persons
Paragraph 21(1)(i) of the Act empowers the CNSC to certify and decertify persons referred
to in paragraph 44(1)(k) of the Act. Paragraph 44(1)(k) of the Act empowers the CNSC to
make regulations and subsection 24(5) of the Act empowers the CNSC to impose licence
conditions, both respecting the qualifications for, and training and examination of nuclear
energy workers and other persons employed in a nuclear facility. Paragraph 37(2)(b) of the
Act states that the Commission may authorize a Designated Officer to certify and decertify
persons referred to in paragraph 44(1)(k).
Paragraph 12(1)(a) and (b) of the General Nuclear Safety and Control Regulations require
every licensee to ensure the presence of a sufficient number of qualified workers to carry on
the licensed activity safely and to train the workers to carry on the licensed activity in
accordance with the Act, the regulations made under the Act and their licence.
Subsection 9(2) of the Class I Nuclear Facilities Regulations stipulates that the CNSC or a
Designated Officer authorized under paragraph 37(2)(b) of the Act, may certify a person for
a position at a Class IA facility, such as an NPP, when that position is referred to in the
facility operating licence. NPP operating licences require incumbents of the positions of
2
C-204 Certification of Persons Working at Nuclear Power Plants
senior health physicist, reactor operator, unit 0 operator, shift operating supervisor and shift
supervisor to obtain a certification by the CNSC before they can assume their
responsibilities. This is because, in the opinion of the CNSC, these positions may impact
directly on the safety of an NPP and on the health and safety of workers, the public and the
environment.
Certification may be granted to a person, pursuant to subsection 9(2) of the Class I Nuclear
Facilities Regulations, upon receipt of a satisfactory application from the facility licensee.
The application must state that the person meets the qualification requirements and has
successfully completed the training program and the examinations specified in the facility
licence, and that the person, in the opinion of the licensee, is capable of performing the
duties of the position. Subsection 10(1) of the Class I Nuclear Facilities Regulations
addresses the requirement in a licence for a person to successfully complete examinations
administered by the CNSC in order to be certified for a position. The purpose of these
examinations is to give the CNSC additional assurance that the person has the knowledge
and skills required by the position prior to certification.
Pursuant to subsection 9(4) of the Class I Nuclear Facilities Regulations, certifications
issued by the CNSC expire five years after the date of their issuance or renewal. Renewal of
certification may be granted to a person, pursuant to subsection 9(3) of the Class I Nuclear
Facilities Regulations, upon receipt of a satisfactory application from the facility licensee.
The application must state that the person has performed the duties of the position safely
and competently, has participated in the continuing training for the position and has
successfully completed the requalification tests referred to in the licence, and that the
person, in the opinion of the licensee, is capable of performing the duties of the position.
The CNSC may, pursuant to subsection 11(1) of the Class I Nuclear Facilities Regulations,
refuse to certify a person. The CNSC may refuse certification where information available
to the CNSC indicates that such a refusal is warranted to protect the health and safety of
persons, to protect the environment, to maintain national security or to fulfil Canada's
international obligations.
The CNSC may, pursuant to subsection 12(1) of the Class I Nuclear Facilities Regulations,
decertify a person. The CNSC may initiate the process for the decertification of a person
when the CNSC has reasonable grounds to believe that the person is no longer qualified to
perform the duties of the position in a safe manner, or where the person's conduct has been
such as to give reasonable concern that the person is not acting responsibly while
performing the duties of the position.
Where the CNSC proposes not to certify or to decertify a person, the CNSC is bound,
pursuant to subsection 11(1) or 12(1) of the Class I Nuclear Facilities Regulations, to
inform both the facility licensee and the person of the proposed decision and of its reasons
at least 30 days before the decision may come into effect. During this period, section 13 of
the Class I Nuclear Facilities Regulations provides the licensee and the person with the
opportunity to be heard on that matter, either orally or in writing. If a hearing is held, the
3
Certification of Persons Working at Nuclear Power Plants C-204
licensee and person will be notified of the decision and the reasons for it.
4
C-204 Certification of Persons Working at Nuclear Power Plants
PART I - OBLIGATIONS OF PERSONS
SUBPART A - SENIOR HEALTH PHYSICISTS
2 Qualification Requirements
2.1 At the time of certification, a senior health physicist shall meet the requirements
specified in paragraphs 2.1.1 to 2.1.3.
2.1.1 EDUCATION: Baccalaureate degree in engineering or science from a
recognized university, with additional specialized courses in the field of
radiation protection.
2.1.2 EXPERIENCE: A minimum of four years related experience in a nuclear
facility. At least two years of this experience must be at an NPP, of which at
least six months must be at the plant where the position is to be filled. At least
one year of the total four years related experience must be in a supervisory
position.
2.1.3 TRAINING: As specified in section 3.
3 Initial Training Requirements
3.1 A senior health physicist shall meet the requirements specified in paragraphs 3.1.1 and
3.1.2.
3.1.1 Have successfully completed training appropriate to the knowledge
requirements of the position, covering:
* the relevant provisions of the Act
* the regulations made pursuant to the Act and, specifically:
- General Nuclear Safety and Control Regulations
- Radiation Protection Regulations
- Class I Nuclear Facilities Regulations
- Nuclear Substances and Radiation Devices Regulations
- Packaging and Transport of Nuclear Substances Regulations
* safety culture and its implementation
* responsibilities and authority of the senior health physicist
* responsibilities and authority of persons who interact with the senior
health physicist
* the plant's operating licence, including documents referenced in the licence
* the licensee's and plant's policies, standards and procedures
* plant design, operation and maintenance
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Certification of Persons Working at Nuclear Power Plants C-204
3.1.2 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the senior
health physicist.
4 CNSC Examinations for Initial Certification
4.1 At the time of certification, a senior health physicist shall have successfully completed
an interview by CNSC staff that samples the topics specified in paragraph 3.1.1 and
current radiation protection principles, methods and practices related to the operation
of an NPP.
5 Senior Health Physicist Transferring to Another Plant
5.1 A senior health physicist at a plant, to obtain a certification as such at another plant,
shall meet the requirements specified in paragraphs 5.1.1 to 5.1.3.
5.1.1 Have successfully completed training appropriate to the knowledge
requirements of the position at that plant, covering:
* responsibilities and authority of the position
* responsibilities and authority of persons who interact with the senior
health physicist
* the plant's operating licence, including documents referenced in the licence
* the licensee's and plant's policies, standards and procedures
* plant design, operation and maintenance
5.1.2 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the senior
health physicist at that plant.
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C-204 Certification of Persons Working at Nuclear Power Plants
5.1.3 Have successfully completed an interview by CNSC staff that samples the
topics specified in paragraph 5.1.1, and current radiation protection principles,
methods and practices related to the operation of that plant.
6 Continuing Training Requirements
6.1 A senior health physicist shall successfully complete training appropriate to the
knowledge requirements of the position, covering:
* changes to plant systems and equipment
* changes to licensee's and plant's policies, standards and procedures
* changes to regulatory requirements
* changes to the plant's operating licence or to documents referenced in the licence
* plant and industry experience and operating events
7 Requalification Tests
7.1 A senior health physicist shall successfully complete an interview by CNSC staff that
samples the topics specified in paragraph 3.1.1, and current radiation protection
principles, methods and practices related to the operation of an NPP.
This interview must be successfully completed within the six month period prior to
the expiry date of the person's certification.
8 Certification Following Decertification
8.1 A senior health physicist who has been decertified as such may, within the three year
period following the decertification, be certified again when the reasons that led to the
decertification are no longer a concern and the person meets the requirement specified
in paragraph 8.1.1.
8.1.1 Have successfully completed training appropriate to the knowledge
requirements of the position, covering changes or events that have occurred
during the absence of the person from the position, including:
* changes to plant systems and equipment
* changes to licensee's and plant's policies, standards and procedures
* changes to regulatory requirements
* changes to the plant's operating licence or to documents referenced in the
licence
* plant and industry experience and operating events
8.2 Where circumstances give rise to reasonable concerns respecting the person's
continuing ability to perform the duties of the position, in addition to meeting the
requirement specified in paragraph 8.1.1, the person shall successfully complete an
interview by CNSC staff, covering the topics specified in paragraph 3.1.1 and current
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Certification of Persons Working at Nuclear Power Plants C-204
radiation protection principles, methods and practices related to the operation of an
NPP.
8.3 A senior health physicist who has been decertified as such may, after a three year
period following the decertification, be certified again when the reasons that led to the
decertification are no longer a concern and the person meets the requirements for
initial certification specified in sections 3 and 4.
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C-204 Certification of Persons Working at Nuclear Power Plants
SUBPART B - REACTOR OPERATORS, UNIT 0 OPERATORS,
SHIFT OPERATING SUPERVISORS AND SHIFT SUPERVISORS
9 Qualification Requirements
Reactor Operators and Unit 0 Operators
9.1 At the time of certification, a reactor operator or unit 0 operator shall meet the
requirements specified in paragraphs 9.1.1 to 9.1.3.
9.1.1 EDUCATION: High school diploma that includes credits in science and
mathematics.
9.1.2 EXPERIENCE: A minimum of five years of plant experience at a Canadian
NPP, with at least two years of this experience at the plant where the position
is to be filled. At least one year of the experience at the plant must be
immediately prior to the selection for training for the position.
9.1.3 TRAINING: As specified in section 10.
Shift Operating Supervisors
9.2 At the time of certification, a shift operating supervisor shall meet the requirements
specified in paragraphs 9.2.1 to 9.2.3.
9.2.1 EDUCATION: High school diploma that includes credits in science and
mathematics.
9.2.2 EXPERIENCE: A minimum of four years of plant experience as a reactor
operator at the plant where the position is to be filled. At least one year of the
experience as a reactor operator must be immediately prior to the selection for
training for the position.
9.2.3 TRAINING: As specified in section 10.
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Certification of Persons Working at Nuclear Power Plants C-204
Shift Supervisors
9.3 At the time of certification, a shift supervisor shall meet the requirements specified in
paragraphs 9.3.1 to 9.3.3.
9.3.1 EDUCATION: Baccalaureate degree in engineering or science from a
recognized university. A certificate of qualification as stationary engineer first
class, or a certification as reactor operator at a Canadian NPP are acceptable
alternatives to a university degree.
9.3.2 EXPERIENCE: A minimum of six years of plant experience at a Canadian NPP,
with at least two years of this experience at the plant where the position is to
be filled. At least one year of the experience at the plant must be immediately
prior to the selection for training for the position.
9.3.3 TRAINING: As specified in section 10.
10 Initial Training Requirements
Reactor Operators
10.1 A reactor operator shall meet the requirements specified in paragraphs 10.1.1 to
10.1.8.
10.1.1 Have successfully completed training appropriate to the knowledge
requirements of the position, covering:
* science fundamentals relevant to the operation of the plant
* principles of operation of plant equipment
This training shall be followed by a comprehensive written examination set by the
licensee. This examination must be successfully completed before the person may take
the CNSC examination specified in paragraph 11.1.1.
10.1.2 Have successfully completed training appropriate to the knowledge
requirements of the position, covering:
* radiation fundamentals
* radiation hazards
* radiation protection theory and practices
* radiation protection procedures used during normal, abnormal and
emergency operation of the plant
This training shall be followed by a comprehensive written examination set by the
licensee. Subject to the provisions of section 12, this examination must be successfully
completed within the three year period prior to certification and before the CNSC
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C-204 Certification of Persons Working at Nuclear Power Plants
examination specified in paragraph 11.1.2 may be taken.
10.1.3 Have successfully completed training appropriate to the knowledge
requirements of the position, covering:
* design and operation of plant systems
* plant systems integrated operation including, where applicable, interaction
between unit systems and those of the other units and of unit 0
* administrative procedures related to plant operation and maintenance
* principles of reactor fuelling, fuel handling and storage, irradiated fuel
cooling, fuelling limitations
* principles of nuclear safety
* responsibilities of the reactor operator
This training shall be followed by a comprehensive written examination set by the
licensee. This examination must be successfully completed before the person may take
the CNSC examination specified in paragraph 11.1.2.
10.1.4 Have successfully completed training, on the plant full scope replica simulator,
appropriate to the knowledge and skill requirements of the position, covering:
* operation of unit systems and equipment under normal, abnormal and
emergency conditions including, where applicable, the effects that unit
operation may have on the other units and unit 0
* interaction with other members of the shift crew
This training shall be followed by the successful completion of a comprehensive
simulator-based examination set by the licensee.
10.1.5 Have successfully completed on-the-job training appropriate to the knowledge
and skill requirements of the position, covering:
* standard control room operating practices
* maintenance and repair of unit systems and equipment
* operations in the control equipment room
* operation of unit systems from the emergency control room
This training shall include job performance measures to confirm that the person has
the required knowledge and skills.
10.1.6 Have satisfactorily performed the duties of the position under the supervision
of a certified reactor operator for a minimum of 480 hours on shift after
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Certification of Persons Working at Nuclear Power Plants C-204
having met the requirements specified in paragraphs 10.1.1 to 10.1.5.
10.1.7 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the reactor
operator.
10.1.8 Have met the requirements specified in paragraphs 10.1.1 to 10.1.7 before
taking the examination specified paragraph 11.1.3.
Unit 0 Operators
10.2 A unit 0 operator shall meet the requirements specified in paragraphs 10.2.1 to
10.2.7.
10.2.1 Have successfully completed training appropriate to the knowledge
requirements of the position, covering:
* science fundamentals relevant to the operation of the plant
* principles of operation of unit 0 equipment
This training shall be followed by a comprehensive written examination set by the
licensee. Subject to the provisions of section 12, this examination must be successfully
completed within the three year period prior to certification.
10.2.2 Have successfully completed training appropriate to the knowledge
requirements of the position, covering:
* radiation fundamentals
* radiation hazards
* radiation protection theory and practices
* radiation protection procedures used during normal, abnormal and
emergency operation of the plant
This training shall be followed by a comprehensive written examination set by the
licensee. Subject to the provisions of section 12, this examination must be successfully
completed within the three year period prior to certification.
10.2.3 Have successfully completed training appropriate to the knowledge
requirements of the position covering:
* design and operation of plant systems
* plant system integrated operation including interaction between unit 0
systems and those of other units
* administrative procedures related to plant operation and maintenance
* fuel storage and irradiated fuel cooling
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C-204 Certification of Persons Working at Nuclear Power Plants
* principles of nuclear safety
* responsibilities of the unit 0 operator
This training shall be followed by a comprehensive written examination set by the
licensee. Subject to the provisions of section 12, this examination must be successfully
completed within the two year period prior to certification.
10.2.4 Have successfully completed training, on the plant full scope replica simulator,
appropriate to the knowledge and skill requirements of the position, covering:
* operation of unit 0 systems and equipment under normal, abnormal and
emergency conditions and the effects that unit 0 operation may have on
the other units
* interaction with other members of the shift crew
This training shall be followed by a comprehensive simulator-based examination set
by the licensee. Subject to the provisions of section 12, this examination must be
successfully completed within the one year period prior to certification.
10.2.5 Have successfully completed on-the-job training appropriate to the knowledge
and skill requirements of the position, covering:
* standard control room operating practices
* maintenance and repair of unit 0 systems and equipment
* operations in the control equipment room
* operation of unit 0 systems from the emergency control room
This training shall include job performance measures to confirm that the person has
acquired the necessary knowledge and skills.
10.2.6 Have satisfactorily performed the duties of the position under the supervision
of a certified unit 0 operator for a minimum of 480 hours on shift after having
met the requirements specified in paragraphs 10.2.1 to 10.2.5.
10.2.7 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the unit 0
operator.
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Certification of Persons Working at Nuclear Power Plants C-204
Shift Operating Supervisors
10.3 A shift operating supervisor shall meet the requirements specified in paragraphs
10.3.1 to 10.3.5.
10.3.1 Have successfully completed training appropriate to the knowledge
requirements of the position that are in addition to those of a reactor operator,
covering:
* fuelling strategies, properties of irradiated fuel and physics of fuel failures
* primary and back-up heat sinks
* conventional and radiation hazards to plant personnel and to the public,
including radiological hazards from postulated accident conditions
* handling of conventional and radiation emergencies
* expected response of systems and units to equipment failures and accident
conditions
* plant operating strategies
* configuration of systems and equipment isolation required for maintenance
activities
* design and operation of plant systems for which the reactor operators do
not have direct operational control
* the plant's operating licence and documents referenced in the licence
* the licensee's policies, standards and procedures
* situations that may result in the violation of conditions in the Operating
Policies and Principles
* requirements pertaining to NPP operation in federal and provincial acts
and regulations, and in relevant standards and codes
* responsibilities and authority of a shift operating supervisor and of other
plant personnel who report to or interface with the shift operating
supervisor
* qualification requirements of plant personnel who report to the shift
operating supervisor
This training shall be followed by a comprehensive written examination set by the
licensee. Subject to the provisions of section 12, this examination must be successfully
completed within the one year period prior to certification.
10.3.2 Have successfully completed training, on the plant full scope replica simulator,
appropriate to knowledge and skill requirements of the position, covering:
* operation and monitoring of plant systems and equipment for which
reactor operators do not have direct operational control, under normal,
abnormal and emergency conditions
* independent monitoring of plant systems and equipment under normal,
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C-204 Certification of Persons Working at Nuclear Power Plants
abnormal and emergency conditions
* independent diagnosis and decision making
* supervision and direction of plant operations under normal, abnormal and
emergency conditions
* interaction with other members of the shift crew
This training shall be followed by a comprehensive simulator-based examination set
by the licensee. Subject to the provisions of section 12, this examination must be
successfully completed within the one year period prior to certification.
10.3.3 Have successfully completed on-the-job training appropriate to knowledge
and skill requirements of the position, covering:
* where applicable, standard unit 0 control room operating practices
* operation and monitoring of plant systems and equipment performed by
the fuel handling operators under normal, abnormal and emergency
conditions
* supervision and direction of plant operations in the control room, control
equipment room, emergency control room, in the field and, where
applicable, at the Tritium Removal Facility under normal, abnormal and
emergency conditions
* maintenance and repair of plant systems and equipment
This training shall include job performance measures to confirm that the person has
acquired the necessary knowledge and skills.
10.3.4 Have satisfactorily performed the duties of the position under the supervision
of a certified shift operating supervisor for a minimum of 480 hours on shift
after having met the requirements specified in paragraphs 10.3.1 to 10.3.3.
10.3.5 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the shift
operating supervisor.
Shift Supervisors
10.4 A shift supervisor shall meet the requirements specified in paragraphs 10.4.1 to
10.4.7.
10.4.1 Have successfully completed the training and the examinations for reactor
operators specified in paragraphs 10.1.1 to 10.1.3.
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Certification of Persons Working at Nuclear Power Plants C-204
10.4.2 Have successfully completed training appropriate to the knowledge
requirements of the position that are in addition to those of a reactor operator,
covering:
* reactor physics, principles of reactor operation and fuelling strategies
* phenomena that may significantly affect core reactivity and flux shape
* properties of irradiated fuel, principles of fuel cooling and physics of fuel
failures
* operating constraints and limits associated with reactor fuelling and
irradiated fuel cooling
* reactor safety, heat transfer, thermodynamics and fluid mechanics
* primary and back-up heat sinks
* conventional and radiation hazards to plant personnel and to the public,
including radiological hazards from postulated accident conditions
* handling of conventional and radiation emergencies
* design requirements of safety related equipment and systems
* design features and limitations of equipment and systems
* chemical control of systems
* diagnosis of equipment failures and assessment of abnormal plant
conditions
* expected response of systems and units to equipment failures and accident
conditions
* plant operating strategies
* major assumptions in the NPP accident analyses and technical bases for
emergency operating procedures
* the configuration of systems and equipment isolation required for
maintenance activities
* design and operation of plant systems for which the reactor operators do
not have direct operational control
* the plant's operating licence and documents referenced in the licence
* the licensee's policies, standards and procedures
* situations that may result in the violation of conditions of the plant's
operating licence and of the Operating Policies and Principles
* requirements pertaining to NPP operation in federal and provincial acts
and regulations, and in relevant standards and codes
* responsibilities and authority of a shift supervisor and of other plant
personnel who report to or interface with the shift supervisor
* qualification requirements of plant personnel who report to the shift
supervisor
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C-204 Certification of Persons Working at Nuclear Power Plants
This training shall be followed by a comprehensive written examination set by the
licensee. This examination must be successfully completed before the person may take
the CNSC examination specified to in paragraph 11.2.3.
10.4.3 Have successfully completed training, on the plant full scope replica simulator,
appropriate to the knowledge and skill requirements of the position, covering:
* operation and monitoring of plant systems and equipment by the operators
under normal, abnormal and emergency conditions
* independent monitoring of the plant systems and equipment under normal,
abnormal and emergency conditions
* independent diagnosis and decision making
* supervision and direction of plant operations under normal, abnormal and
emergency conditions
* at single unit plants, operation of plant equipment and systems when
replacing the reactor operator
* interaction with other members of the shift crew
This training shall be followed by a comprehensive simulator-based examination set
by the licensee.
10.4.4 Have successfully completed on-the-job training appropriate to the knowledge
and skill requirements of the position, covering:
* standard control room operating practices and, where applicable, standard
unit 0 operating practices
* operation and monitoring of plant systems and equipment performed by
the fuel handling operators under normal, abnormal and emergency
conditions
* where applicable, operation and monitoring of systems and equipment of
the Tritium Removal Facility performed by the operators of the facility
under normal, abnormal and emergency conditions
* supervision and direction of plant operations in the control room, in the
control equipment rooms, in the emergency control room, in the field and,
where applicable, at the Tritium Removal Facility, under normal, abnormal
and emergency conditions
* maintenance and repair of plant systems and equipment
This training shall include job performance measures to confirm that the person has
acquired the necessary knowledge and skills.
10.4.5 Have satisfactorily performed the duties of the position under the supervision
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Certification of Persons Working at Nuclear Power Plants C-204
of a certified shift supervisor for a minimum of 480 hours on shift after having
met the requirements specified in paragraphs 10.4.1 to 10.4.4.
10.4.6 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the shift
supervisor.
10.4.7 Have met the requirements specified in paragraphs 10.4.1 to 10.4.6 before
taking the CNSC examination specified to in paragraph 11.2.4.
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C-204 Certification of Persons Working at Nuclear Power Plants
11 CNSC Examinations for Initial Certification
Reactor Operators
11.1 At the time of certification a reactor operator shall meet the requirements specified in
paragraphs 11.1.1 to 11.1.3.
11.1.1 Have successfully completed the CNSC examination for reactor operators,
which samples topics covered in the training specified in paragraph 10.1.1.
Subject to the provisions of section 12, this examination must be successfully
completed within the three year period prior to certification and before the CNSC
examination specified in paragraph 11.1.2 may be taken.
11.1.2 Have successfully completed the CNSC examination for reactor operators,
which samples topics covered in the training specified in paragraph 10.1.3 and
covering those areas of unit operation, that may result in the discharge of
radioactivity to the environment, or which could affect the safety of station
personnel or of members of the public.
Subject to the provisions of section 12, this examination must be successfully
completed within the two year period prior to certification and before the CNSC
examination specified in paragraph 11.1.3 may be taken.
11.1.3 Have successfully completed the CNSC simulator-based examination covering
unit operations under abnormal and emergency conditions.
Subject to the provisions of section 12, this examination must be successfully
completed within the six month period prior to certification.
Shift Supervisors
11.2 At the time of certification, a shift supervisor shall meet the requirements specified in
paragraphs 11.2.1 to 11.2.4.
11.2.1 Have successfully completed the CNSC examination for reactor operators
specified in paragraph 11.1.1.
Subject to the provisions of section 12, this examination must be successfully
completed within the three year period prior to certification and before the CNSC
examination specified in paragraph 11.2.2 may be taken.
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Certification of Persons Working at Nuclear Power Plants C-204
11.2.2 Have successfully completed the CNSC examination for reactor operators
specified in paragraph 11.1.2.
Subject to the provisions of section 12, this examination must be successfully
completed within the two year period prior to certification and before the CNSC
examination specified in paragraph 11.2.3 may be taken.
11.2.3 Have successfully completed the CNSC examination which samples topics
covered in the training specified in paragraph 10.4.2.
Subject to the provisions of section 12, this examination must be successfully
completed within the two year period prior to certification and before the CNSC
examination specified in paragraph 11.2.4 may be taken.
11.2.4 Have successfully completed of the CNSC simulator-based examination
covering:
* monitoring of plant systems and equipment under abnormal and
emergency conditions
* diagnosis and decision making
* supervision and direction of control room operations under abnormal and
emergency conditions
* at single unit plants, operation of plant equipment and systems when
replacing the reactor operator
Subject to the provisions of section 12, this examination must be successfully
completed within the six month period prior to certification.
12 Extending the Period of Validity of an Examination
12.1 Where any provision of sections 10 or 11 calls for successful completion of an
examination within a period of time prior to certification, that period may, on
application from a licensee, be extended for a further period not exceeding one year,
under the following conditions:
* the person needs to perform the duties of the position under the supervision of a
certified person for longer than the period specified in section 10, prior to taking
the CNSC simulator-based examination
* the person needs to retake a single examination and an extension is required for
that purpose
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C-204 Certification of Persons Working at Nuclear Power Plants
* the person's training has been delayed due to sickness, injury or family related
responsibilities
12.2 When applying for an extension pursuant to subsection 12.1, the licensee shall submit
information confirming:
* that one or more of the conditions specified in subsection 12.1 apply
* the measures taken to ensure that the person has maintained the knowledge and
skills required to work competently in the position
13 Advancement from Reactor Operator or Shift Operating
Supervisor to Shift Supervisor
Reactor Operators to Shift Supervisors
13.1 At the time of certification as shift supervisor, a reactor operator advancing to that
position at the plant shall meet the requirements specified in paragraphs 13.1.1 to
13.1.8.
13.1.1 Have a minimum of four years experience as reactor operator at the plant.
13.1.2 Have successfully completed the shift supervisor training specified in
paragraph 10.4.2.
This training shall be followed by a comprehensive written examination set by the
licensee. This examination must be successfully completed before the person may take
the CNSC examination referred to in paragraph 11.2.3.
13.1.3 Have successfully completed, on the plant full scope replica simulator, the
components of the shift supervisor training specified in paragraph 10.4.3,
covering:
* operation and monitoring of the plant's systems and equipment for which
the reactor operators do not have direct operational control, under normal,
abnormal and emergency conditions
* independent monitoring of plant systems and equipment under normal,
abnormal and emergency conditions
* independent diagnosis and decision making
* supervision and direction of plant operations under normal, abnormal and
emergency conditions
* interaction with other members of the shift crew
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Certification of Persons Working at Nuclear Power Plants C-204
This training shall be followed by a comprehensive simulator-based examination set
by the licensee.
13.1.4 Have successfully completed the shift supervisor on-the-job training specified
in paragraph 10.4.4.
13.1.5 Have satisfactorily performed the duties of the position under the supervision
of a certified shift supervisor for a minimum of 480 hours on shift after having
met the requirements specified in paragraphs 13.1.2 to 13.1.4.
13.1.6 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the shift
supervisor.
13.1.7 Have met the requirements specified in paragraphs 13.1.2 to 13.1.6 before
taking the CNSC examination specified in paragraph 11.2.4.
13.1.8 Have successfully completed the CNSC examinations for shift supervisors
specified in paragraphs 11.2.3 and 11.2.4.
13.1.9 Have met the requirements specified in paragraphs 13.1.2 to 13.1.6 and in
paragraph 13.1.8 within the two year period prior to certification as a shift
supervisor.
Shift Operating Supervisors to Shift Supervisors
13.2 At the time of certification as shift supervisor, a shift operating supervisor advancing
to that position at the plant, shall meet the requirements specified in paragraphs
13.2.1 to 13.2.7.
13.2.1 Have successfully completed the components of the shift supervisor training
specified in paragraph 10.4.2 that were not covered in the shift operating
supervisor training specified in paragraph 10.3.1.
This training shall be followed by a comprehensive written examination set by the
licensee. This examination must be successfully completed before the person may take
the CNSC examination specified in paragraph 11.2.3.
13.2.2 Have successfully completed, on the plant full scope replica simulator, the
components of the shift supervisor training specified in paragraph 10.4.3,
covering:
* supervision and direction of plant operations under normal, abnormal and
emergency conditions
* interaction with other members of the shift crew
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C-204 Certification of Persons Working at Nuclear Power Plants
This training shall be followed by a comprehensive simulator-based examination set
by the licensee.
13.2.3 Have successfully completed the shift supervisor on-the-job training specified
in paragraph 10.4.4.
13.2.4 Have satisfactorily performed the duties of the position under the supervision
of a certified shift supervisor for a minimum of 480 hours on shift after having
met the requirements specified in paragraphs 13.2.1 to 13.2.3.
13.2.5 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the shift
supervisor.
13.2.6 Have met the requirements specified in paragraphs 13.2.1 to 13.2.5 before
taking the CNSC examination specified in paragraph 11.2.4.
13.2.7 Have successfully completed the CNSC examinations for shift supervisors
specified in paragraphs 11.2.3 and 11.2.4.
13.2.8 Have met the requirements specified in paragraphs 13.2.1 to 13.2.5 and in
paragraph 13.2.7 within the two year period prior to certification as a shift
supervisor.
14 Continuing Training Requirements
14.1 Reactor operators, unit 0 operators, shift operating supervisors and shift supervisors,
during the period of their certification, shall meet the requirements specified in
paragraphs 14.1.1 to 14.1.3.
14.1.1 Participate in continuing training appropriate to the knowledge and skill
requirements of their position, covering:
* a review of the knowledge gained in their initial training, specified in
section 10, that is required to work competently in their position but for
which competency is not maintained during the day-to-day operation of
the plant
* simulator-based exercises covering infrequently performed plant
manoeuvres
* simulator-based exercises covering a sufficiently varied number of
situations that challenge their diagnostic and problem-solving abilities and
ensure that they are, at all times, proficient in selecting and using abnormal
and emergency operating procedures
* in-plant exercises and drills conducted on a regular basis to practice their
response to accidents and emergencies
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Certification of Persons Working at Nuclear Power Plants C-204
14.1.2 Participate in training appropriate to the knowledge and skill requirements of
the position, covering:
* changes to plant systems and equipment
* changes to licensee's and plant's policies, standards and procedures
* changes to regulatory requirements
* changes to the plant's operating licence or to documents referenced in the
licence
* plant or industry experience and operating events
14.1.3 Have successfully completed written and simulator-based tests set by the
licensee to confirm that the person possesses the knowledge of the material
and the skills covered in each training session.
15 Requalification Tests
15.1 Reactor operators, unit 0 operators, shift operating supervisors and shift supervisors
shall successfully complete written and simulator-based requalification tests within the
five-year period of certification, as specified in the document Common Standard for
Requalification Testing of Certified Staff at Canadian CANDU Nuclear Generating
Stations, as amended from time to time.
16 Certification Following Decertification
16.1 A reactor operator, unit 0 operator, shift operating supervisor or shift supervisor,
who has been decertified as such may, within the two year period following the
decertification, be certified again when the reasons that led to the decertification are
no longer a concern and the person meets the requirements specified in paragraphs
16.1.1 to 16.1.3.
16.1.1 Have successfully completed training appropriate to the knowledge and skill
requirements of the position covering topics identified as a result of changes
or events that have occurred during the absence of the person from the
position, including:
* changes to plant systems and equipment
* changes to licensee's and plant's policies, standards and procedures
* changes to regulatory requirements
* changes to the plant's operating licence or to documents referenced in the
licence
* plant or industry experience and operating events
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C-204 Certification of Persons Working at Nuclear Power Plants
16.1.2 Have successfully completed training covering selected topics from the initial
training to ensure that the person has the knowledge required to work
competently in the position.
16.1.3 Have successfully completed simulator-based exercises that cover a
sufficiently varied number of situations that challenge the diagnostic and
problem solving abilities of the person, and ensure that the person has the
skills required to work competently in the position.
16.2 Where circumstances give rise to reasonable concerns respecting the person's
continuing ability to perform the duties of the position, in addition to meeting the
requirements specified in paragraphs 16.1.1 to 16.1.3, the person shall meet the
requirements specified in paragraphs 16.2.1 to 16.2.3.
16.2.1 Have successfully completed written and simulator-based requalification tests
set by the licensee to confirm that the person has the knowledge and skills
required to work competently in the position.
16.2.2 Have performed the duties of the position under the supervision of a person
currently certified in the position for a number of shifts sufficient to confirm
that the person can perform those duties competently and safely.
16.2.3 Have successfully completed an interview by plant management that confirms
and documents the person's competence to perform the duties of the position.
16.2.4 The requirements specified in paragraphs 16.1.1 to 16.2.3 must be met within
one year of the licensee's application for the certification of the person.
16.3 A reactor operator, unit 0 operator, shift operating supervisor or shift supervisor who
has been decertified as such may, after a two year period following the decertification,
be certified again when the reasons that led to the decertification are no longer a
concern and the person meets the requirements for initial certification specified in
sections 10 and 11.
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Certification of Persons Working at Nuclear Power Plants C-204
PART II - OBLIGATIONS OF THE LICENSEE
17 Fitness-for-Duty Program
17.1 The licensee shall have in place a fitness-for-duty program that addresses the physical
and mental conditions required of a person seeking certification, renewal of
certification or holding a certification.
18 Selection and Training
18.1 The licensee shall establish and document policies, standards and procedures for
selecting, training and qualifying persons seeking certification or renewal of
certification as:
* senior health physicist
* reactor operator
* unit 0 operator
* shift operating supervisor
* shift supervisor
18.2 The licensee shall establish and document initial training programs, specific to each
applicable position referred to in subsection 18.1, to address the training requirements
specified to in sections 3 and 10.
18.3 The licensee shall establish and document continuing training programs specific to
each applicable position specified in subsection 18.1, to address the training
requirements specified in sections 6 and 14.
18.3.1 The continuing training for the senior health physicist position shall be
delivered as specified in subsection 6.1.
18.3.2 The continuing training for the reactor operator, unit 0 operator, shift
operating supervisor and shift supervisor positions shall be delivered on a
regular basis over a cycle not exceeding three years.
18.4 The licensee shall ensure that the initial and continuing training programs specified in
subsections 18.2 and 18.3 are developed using the principles of a systematic approach
to training.
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C-204 Certification of Persons Working at Nuclear Power Plants
18.4.1 Existing training programs that have been developed without using the
principles specified in subsection 18.4, but adequately address the training
requirements specified in this Standard, may be continued for a period not
exceeding five years from the day this Standard comes into effect.
19 Simulator Requirements
19.1 Each plant shall have in service a full scope replica simulator facility for training and
examining persons seeking certification or renewal of certification as reactor operator,
unit 0 operator, shift operating supervisor or shift supervisor.
19.1.1 The simulator shall be capable of simulating, realistically and in real time, all
significant plant manoeuvres and transients, including:
* plant start-ups and shutdowns
* all significant failures of systems and their equipment and the
consequences of such failures
* major plant upsets and accident conditions
19.1.2 The simulator shall be provided with all functional devices which replicate
those of the plant's main control room, including:
* a telephone system
* radiation alarms
* fire alarms
19.2 In addition, for the simulator-based examinations specified in section 11 and for those
specified in paragraphs 10.2.4 and 10.3.2, the simulator shall be equipped with the
data recording devices specified in paragraphs 19.2.1 to 19.2.4. These devices must
be capable of being synchronized to within two seconds of each other.
19.2.1 The simulator shall be equipped with a trainee action monitor capable of
printing in chronological order, with their respective time of occurrence, all
malfunctions initiated by the simulator operator and all the actions performed
on the control panels during a test scenario.
19.2.2 The simulator shall have provisions for either:
* tracing, with adequate precision, graphics of any selection of 48 system
parameters versus time for up to two hours and for printing those
graphics, or
27
Certification of Persons Working at Nuclear Power Plants C-204
* storing and printing the values versus time of any selection of 48 system
parameters sampled at an adequate frequency during a period of up to two
hours
19.2.3 The simulator shall be equipped with a video system capable of recording all
actions performed during the examination by the person being examined. The
system must possess sufficient resolution to permit the examiners to identify,
with the aid of the corresponding control panel photographs, the controls and
instruments used by the person being examined and must also have provision
for displaying time on the recordings.
19.2.4 The video system shall have provisions for recording clearly all verbal
communications and telephone conversations, during the examination,
between the person being examined and members of the operating team. This
audio recording must have sufficient fidelity to allow easy identification of the
voice of the person being examined.
19.3 The simulator operating facility shall be located such that the person being examined
cannot become aware of data recorded or of the simulator inputs being made by the
person operating the simulator.
20 Licensee's Examinations
20.1 The written examinations, specified in paragraphs 10.1.1 to 10.1.3, 10.2.1 to 10.2.3,
10.3.1 and 10.4.2 and the simulator-based examinations, specified in paragraphs
10.1.4, 10.3.2 and 10.4.3, shall be administered by the licensee in accordance with
procedures that cover the items listed below:
* the guidelines and standards applicable to:
- the preparation, conduct and marking of each comprehensive written
examination
- the design, development, conduct and assessment of comprehensive
simulator-based examination
* the requirements and procedures for ensuring examination confidentiality
* the qualification requirements of the persons responsible for:
- the preparation and marking of each comprehensive written examination
- the design and development of each comprehensive simulator-based
examination
- the assessment of the performance of candidates taking each comprehensive
simulator-based examination
* the provisions for ensuring a sufficient level of independence between:
- the delivery of training and the examinations
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C-204 Certification of Persons Working at Nuclear Power Plants
- the examiners and the candidates taking a given examination
* the interval between successive attempts at a given examination
* the provisions to ensure the reliability and validity of the results of each
examination
20.2 The questions on plant's systems and procedures of the comprehensive written
examination specified in paragraph 10.2.3 shall be within the envelope defined by the
CNSC document Generic Station System Knowledge Objectives for Control Room
Operators, dated June 1998, as amended from time to time, or by station specific
knowledge objectives derived from this document and endorsed by the CNSC.
20.3 The comprehensive simulator-based examination, specified in paragraph 10.2.4 shall
be designed, developed, conducted and assessed based upon the guidelines and
criteria applicable to the position and specified in the CNSC Operational Procedure
PQAD ST-6(RO), Rev. 2, Simulator Based Examinations for Reactor Operators of
Nuclear Power Plants, dated February 2001, as amended from time to time.
20.4 The written tests and the simulator-based tests of the continuing training program
specified in paragraph 14.1.3 shall be such as to confirm that the person possesses the
knowledge and skills covered in each session and that are required for the person to
work competently in the position.
20.5 The requalification tests, specified in subsection 15.1, shall be designed, developed,
conducted and assessed in accordance with the requirements of the Common
Standard for Requalification Testing of Certified Staff at Canadian CANDU Nuclear
Generating Stations, as amended from time to time.
21 Removal of a Person from a Position Requiring Certification
21.1 The licensee shall immediately remove a person from the position of reactor operator,
unit 0 operator, shift operating supervisor or shift supervisor under the conditions
specified in paragraphs 21.1.1 to 21.1.3.
21.1.1 The person has failed any written or simulator-based requalification test.
21.1.2 The person has shown a lack of competence in performing the duties of the
position.
21.1.3 The person and the licensee have been informed that the CNSC has initiated
procedures for the decertification of the person.
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Certification of Persons Working at Nuclear Power Plants C-204
22 Reinstatement of a Person to a Position Requiring Certification
22.1 The licensee may reinstate a person removed from the position of reactor operator,
unit 0 operator, shift operating supervisor or shift supervisor pursuant to section 21 if
the person meets the requirements specified in paragraphs 22.1.1 or 22.1.2 as
applicable.
22.1.1 For a person removed from the position under paragraph 21.1.1, successful
completion of remedial training and testing appropriate to the knowledge and
skills requirements of the position. The testing shall consist of a requalification
test equivalent to, but different from the one the person has failed.
22.1.2 For a person removed from the position under paragraph 21.1.2, successful
completion of remedial training and testing to ensure that the person can
competently perform the duties of the position.
22.2 Where the remedial training and testing specified in paragraphs 22.1.1 and 22.1.2
indicates that the person's deficiencies cannot be rectified, the licensee shall notify
the CNSC in writing.
23 Temporary Assignment to Other Positions
23.1 The licensee shall ensure that reactor operators, unit 0 operators, shift operating
supervisors and shift supervisors, who are temporarily assigned to other positions at
the plant, maintain the competence required to perform the duties of the position for
which they hold a certification by requiring those persons to meet the requirements
specified in paragraphs 23.1.1 to 23.1.4 during the temporary assignment.
23.1.1 Perform the duties of the position for which they hold a certification for a
minimum of five complete shifts per calendar quarter.
23.1.2 Participate in the continuing training applicable to the position for which they
hold a certification and successfully complete the associated tests.
23.1.3 Successfully complete all the requalification tests, applicable to the position
for which they hold a certification.
23.1.4 Perform the duties of the position for which they hold a certification on a full
time basis for a minimum of six months in any three year period.
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C-204 Certification of Persons Working at Nuclear Power Plants
24 Retention of Records
24.1 The licensee shall retain the records specified in paragraphs 24.1.1 to 24.1.12 for the
period specified in subsection 14(4) of the Class I Nuclear Facilities Regulations.
24.1.1 The policies, standards and procedures for selecting, training and qualifying
personnel for positions that require certification.
24.1.2 The responsibilities of the line organization with respect to training and
qualification of personnel for positions that require certification.
24.1.3 The structure of the training organization, including the responsibilities of
personnel in the organization.
24.1.4 The procedures for the retention of a person's training records.
24.1.5 The procedures for administering written and simulator-based examinations
and tests, oral examinations or interviews to persons seeking or holding a
certification.
24.1.6 The performance capabilities of the full scope replica simulator and the
performance test procedures and test results which indicate that the simulator
meets regulatory requirements.
24.1.7 The outcome of the analysis performed to identify training needs for positions
that require certification, including:
* a description of the process followed to conduct the analysis
* the names and qualifications of the persons who participated in the
analysis
* the task list obtained from the analysis
* the criteria used in selecting tasks for training
* the list of the knowledge and skills required to perform the selected tasks
24.1.8 The outcome of the design of training programs for positions that require
certification including:
* the training objectives
* the training settings
* the test items for written and simulator-based examinations and tests, and
for field checkouts
* the marker's assessment guides
24.1.9 The outcome of the development of training programs for positions that
require certification including:
* the lesson plans, simulator training guides and other similar guides
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Certification of Persons Working at Nuclear Power Plants C-204
* the training manuals and any other training material used by the trainers
and the trainees
24.1.10 The outcome of the implementation of training programs for positions that
require certification including:
* the names of the trainees
* the names of the instructors
* the dates when the training was delivered
* the content of the training delivered
* the examinations, tests and field checkouts, including the results
* the names and positions of the persons involved in the preparation and
grading of examinations, tests and field checkouts
24.1.11 The outcome of the evaluation of training programs for positions that
require certification including:
* evaluations conducted by the licensee or by external organizations
* feedback from supervisors, job incumbents, trainers and trainees
* disposition of feedback, including date of receipt of the feedback, name of
its author, date of disposition
* changes made to training programs due to operational experience
feedback
* changes made to the training program due to plant modifications
* reviews of the effectiveness of training programs
* analyses of the results of examinations, tests and field checkouts
24.1.12 The procedures for reviewing and approving the outcome of each phase of
the systematic approach to training.
24.2 The licensee shall retain the records specified in paragraphs 24.2.1 to 24.2.11 for each
person seeking or holding a certification. These records shall be kept for the time
specified in subsection 14(5) of the Class I Nuclear Facilities Regulations.
24.2.1 The name and address of the educational establishments where the certificates
or degrees required by the position were obtained and certified copies of those
degrees or certificates.
24.2.2 The experience gained and qualifications obtained prior to employment with
the licensee, indicating the number of years of experience, where the
experience was gained and where the qualifications were obtained.
24.2.3 The qualifications obtained during employment by the licensee.
24.2.4 The training for initial certification, continuing training and any remedial
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C-204 Certification of Persons Working at Nuclear Power Plants
training.
24.2.5 The dates and the results of all licensee's tests, field checkouts, comprehensive
written and simulator-based examinations, job performance measures and
interviews required for initial certification or renewal of certification.
24.2.6 The licensee's periodic evaluations of performance on the job.
24.2.7 Any temporary assignments away from the position for which the person
holds a certification, including the duration and the nature of the assignments.
24.2.8 When temporarily assigned to another position, the time spent performing the
duties of the position for which the person holds a certification, with the
continuing training and the requalification tests taken during the assignment.
24.2.9 Any temporary removal from the position by the licensee, the reasons for the
removal and actions taken to reinstate the person in the position.
24.2.10 The physical and mental fitness of a person requiring a certification.
24.2.11 Absences of more than six months from the position with the reasons for
the absence.
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Certification of Persons Working at Nuclear Power Plants C-204
GLOSSARY
The following terms are defined specifically for the purpose of this Standard.
Certification (accréditation)
A written attestation from the Commission, or from a Designated Officer authorized by the
Commission, that a person meets the qualification requirements for a position as required by a
regulation or by a nuclear power plant operating licence.
CNSC (also Commission) (Commission)
The Canadian Nuclear Safety Commission established by section 8 of the NSC Act.
Job Performance Measure (évaluation pratique)
A test used to evaluate the proficiency of a person in performing a specific job task.
Licence (also Operating Licence) (permis ou permis d'exploitation)
A licence issued by the CNSC to operate a nuclear power plant.
Licensee (titulaire de permis)
The holder of a licence issued by the Commission to operate a nuclear power plant.
NSC Act (Loi)
Nuclear Safety and Control Act.
Nuclear Facility (installation nucléaire)
A facility included in the definition "Nuclear Facility" set out in section 2 of the Act.
Nuclear Power Plant (also NPP or plant) (centrale nucléaire ou centrale)
Any electricity generating plant powered by a nuclear reactor. Where a licence is issued for the
operation of multiple reactors, plant means all the reactors identified in the licence.
On-the-job Training (formation sur les lieux de travail)
The training undertaken in the actual work environment to obtain required job related knowledge
and skills.
Plant Experience (expérience en centrale)
The experience gained in an NPP during commissioning, start-up testing or operation which is
relevant to the position being sought by a person.
34
C-204 Certification of Persons Working at Nuclear Power Plants
Qualification (qualifications)
The combination of education, training and experience required to meet specific job performance
criteria.
Reactor Operator (opérateur de réacteur)
The person in a nuclear power plant who is responsible for operating the reactor and for
manipulating the supporting plant controls from the plant's control room in compliance with the
plant's operating licence, policies and procedures.
Recognized University (université reconnue)
A Canadian university having a federal or provincial charter, or a foreign university whose
degrees are recognized in Canada.
Related Experience (expérience pertinente)
The experience gained in performing duties related to those of the position for which a person
seeks certification.
Systematic Approach to Training (approche systématique à la formation)
A phased approach to training consisting of :
* an analysis phase which is the identification of the competencies in terms of knowledge
and skills required by a position
* a design phase which is the conversion of competency requirements into training
objective and the production of a training plan
* a development phase which is the preparation of the training material to meet the
training objectives
* an implementation phase which is conducting the training using the material developed,
and
* an evaluation phase which is the collection and collation of data obtained during each of
the phases of the systematic approach to training to be followed by appropriate actions
to improve training effectiveness
Senior Health Physicist (chef radioprotection)
The person in a nuclear power plant who is responsible for interpreting the regulations, policies
and procedures applicable to radiation protection, and for providing procedure related approvals
where required.
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Certification of Persons Working at Nuclear Power Plants C-204
Shift Operating Supervisor (superviseur de l'exploitation)
The person in a multi-unit nuclear power plant who is responsible to the shift supervisor to ensure
that the main control room staff function safely within their authority limits, and that the conduct
of operations within the main control room is performed to approved procedures in a manner
which meets the standards set by the utility.
Shift Supervisor (chef de quart)
The person in a nuclear power plant who is responsible for the direct supervision of the operation
of a nuclear power plant and for ensuring that operations and maintenance are conducted in
accordance with the plant's operating licence, policies and procedures, and other applicable
documents. The shift supervisor is the senior management's representative on shift.
Unit 0 Operator (opérateur de la tranche 0)
The person in a multi-unit nuclear power plant who is responsible for operating a group of safety
and process systems common to all reactor units from the plant control room unit 0 panels in
accordance with plant's procedures.
36
DRAFT
REGULATORY
GUIDE
Access Control
for Protected and
Inner Areas of
Nuclear Facilities
C-205
Issued for public comments by the
Canadian Nuclear Safety Commission
October 2000
DRAFT REGULATORY GUIDE
Access Control for Protected and Inner Areas
of Nuclear Facilities
C-205(E)
Issued for public comments by the
Canadian Nuclear Safety Commission
October 2000
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY GUIDE
Access Control for Protected and Inner Areas
of Nuclear Facilities
C-205(E)
About this Document
The draft regulatory guide Access Control for Protected and Inner Areas of Nuclear Facilities
provides information on how to meet the regulatory requirements of the Nuclear Security
Regulations regarding physical protection measures at nuclear facilities, specifically access control
for protected and inner areas. This guide will be of particular interest to nuclear facilities'
licensees, their workers, heads of security and others involved in providing physical protection
measures to nuclear facilities.
This document identifies the regulatory requirements relevant to access control for protected and
inner areas, outlines the Canadian Nuclear Safety Commission (CNSC, the Commission)
expectations of licensees with respect to controlling access to those areas, and describes
recommended practices for providing effective physical protection measures for the protected and
inner areas of nuclear facilities.
Comments
The CNSC invites interested persons to assist in the further development of this document by
commenting in writing on its content and potential usefulness. Please respond by December 15,
2000, referencing file 1-8-8-205. Direct your comments and any related enquiries to the addresses
below.
The CNSC will study the above comments to determine how best to improve and finalize this
draft.
Unless responders request otherwise, the CNSC will place a copy of their comments in its library,
in Ottawa.
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Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
Document availability
This document, C-205, can be viewed on the CNSC Internet website at www.nuclearsafety.gc.ca,
or an English or French printed copy may be ordered using the following information:
Operations Assistant
Corporate Documents Section
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9 CANADA
Telephone (613) 996-9505
Facsimile (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
ii
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
CONTENTS
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory framework and relevant legislation . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Guide overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1.3 Responsibilities of licensee for security . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1.4 Responsibilities of workers for security . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2 Access to Protected and Inner Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2.1 Protected areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2.2 Inner areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.3 Authorization for access to inner and protected areas . . . . . . . . . . . . . . . . . . . . 4
2.4 Unescorted access to a protected area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
2.5 Escorted access to protected areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
2.6 Unescorted access to an inner area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
2.7 Escorted access to inner areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
2.8 Revoking an authorization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
2.9 Keeping records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3 Security Issues Related to Access . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.1 Nuclear security guards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.2 Worker training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.3 Monitoring entry and exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.4 Vehicle access . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
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Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
3.5 Unauthorized entry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
iv
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
Purpose
This guide provides information on how to meet the regulatory requirements of the Nuclear
Security Regulations regarding physical protection measures at nuclear facilities, specifically,
access control for protected and inner areas.
Scope
This guide identifies the regulatory requirements relevant to access control for protected and inner
areas, outlines the CNSC's expectations of licensees with respect to controlling access to those
areas, and describes recommended practices for providing effective physical protection measures
for the protected and inner areas of nuclear facilities.
This guide applies to all licensees of nuclear facilities, their workers, heads of security and others
involved in providing physical protection measures to nuclear facilities.
1 Introduction
1.1 Regulatory framework and relevant legislation
The Canadian Nuclear Safety Commission is the federal agency that regulates nuclear
facilities and materials to prevent undue risk to health, safety, security and the environment.
The CNSC operates under the authority of the Nuclear Safety and Control Act (NSC Act)
and associated regulations.
Regulatory requirements specific to access control for protected and inner areas are
primarily contained in the Nuclear Security Regulations.
The Nuclear Security Regulations section 2 stipulates that "these regulations apply in
respect of (a) Category I nuclear material, Category II nuclear material and Category III
nuclear material; and (b) a nuclear facility consisting of a nuclear reactor that may exceed
10 MW thermal power during normal operation."
Additionally, Nuclear Security Regulations section 7 requires that Category I nuclear
material be processed, used and stored in an inner area and Category II nuclear material be
processed, used and stored in a protected area.
Category III nuclear material may, however, be processed, used and stored in either a
protected area, a place that is under the direct visual surveillance of the licensee or a secure
place to which access is controlled by the licensee. Nuclear Security Regulations sections
17 to 29 further require that access to protected and inner areas of nuclear facilities be
strictly controlled.
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Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
1.2 Guide overview
To assist licensees in meeting their regulatory requirements with regard to access to
protected and inner areas, this guide :
* addresses the responsibilities of licensees and workers in maintaining security at
nuclear facilities
* discusses protected and inner areas
* outlines the CNSC's expectations of licensees in controlling access to protected and
inner areas
* explains how authorization to enter these areas is granted or revoked
* discusses the search requirements for persons and items that enter or exit protected
or inner areas
* notes requirements and expectations with regard to the record keeping practices of
licensees
* identifies the factors that impact the physical protection measures of inner and
protected areas, including those relevant to nuclear security guards and workers,
individual and vehicle access.
1.3 Responsibilities of licensee for security
The licensee is ultimately responsible for the security of the nuclear facility and has an
obligation to act immediately to counter any threat to the facility.
With regard to access, the licensee has the authority to grant or deny access to the
protected area and, within limits, to grant or deny escorted access to the inner area. Section
2 of this guide provides licensees with specific legal requirements and recommended
practices for access to protected and inner areas.
As a general guide, all workers involved in authorizing persons to enter inner and protected
areas should refuse access to these areas whenever there is any reservation as to a person's
eligibility to enter. Assuring the safety and security of the nuclear facility is of greater
importance than the possibility of offending someone by refusing them entry.
The CNSC expects the licensee to implement a security awareness program to keep
workers informed of the facility's security program and their obligations under that
program. The program should also explain the workers' obligations under CNSC
regulations.
2
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
As an additional physical protection measure, Nuclear Security Regulations section 36
requires the licensee to conduct security drills at least once every six months to test the
operation of the facility's security equipment, systems and procedures at the nuclear facility.
1.4 Responsibilities of workers for security
CNSC General Nuclear Safety and Control Regulations paragraph 17(b) states that every
worker at a nuclear facility is expected to "comply with the measures established by the
licensee to...maintain security...."
The General Nuclear Safety and Control Regulations paragraph 17(c) requires workers to
notify their supervisor or the licensee promptly if they believe there may be "a threat to the
maintenance of security or an incident with respect to security" or "an act of sabotage, theft,
loss or illegal use or possession of a nuclear substance, prescribed equipment or prescribed
information...."
Nuclear Security Regulations subsection 24(2) requires anyone who sees an unauthorized
person within a protected or inner area to report the sighting to the nearest nuclear security
guard. Additional information on the nuclear security guards' responsibilities is presented in
section 3 of this guide.
2 Access to Protected and Inner Areas
The Nuclear Security Regulations subsection 17(1) requires that all persons entering a protected
area must have "physical proof of the recorded authorization of the licensee." In addition, prior to
being allowed to enter any inner area, all unescorted persons must have "the recorded
authorization of the Commission" and "the recorded authorization of the licensee to enter the
protected area" [subsection 18(1)].
2.1 Protected areas
The protected area is defined in the Nuclear Security Regulations section 1 as "an area that
meets the requirements of sections 9, 10 and 11" of the Regulations.
The protected area is established by the licensee and approved by the Commission via the
operating license. Nuclear Security Regulations sections 9 to 11 specify the requirements
concerning a protected area.
* Section 9 states that every protected area must be enclosed by a barrier that is
located at its perimeter and elaborates the physical requirements concerning this
barrier.
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Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
* Section 10 requires the protected area be surrounded by an unobstructed area. This
section also specifies the unobstructed area's minimum extension from the barrier
and the need for its continuous illumination.
* Section 11 completes the requirements of a protected area by characterizing the
necessary intrusion detection devices to be mounted in a protected area.
2.2 Inner areas
The inner area is defined in the Nuclear Security Regulations section 1 as "an area that
meets the requirements of sections 12, 13 and 14" of the Regulations.
Nuclear Security Regulations sections 12 to 14 state the requirements for an inner area.
* Section 12 identifies the inner area as an area inside a protected area where, in
accordance with Nuclear Security Regulations subsection 7(1), all Category I nuclear
material must be processed, used and stored.
* Section 13 specifies that the inner area must be enclosed in a structure or barrier that
is capable of resisting attempts to gain entry using hand tools, firearms or explosives.
All open access points of an inner area, such as doors and windows, must be under
the direct visual surveillance of a nuclear security guard dedicated to watching the
area. Access points to the inner area must be kept locked at all times by devices
which can only be unlocked at the same time by two authorized persons, one of
whom should be a nuclear security guard.
* Section 14 goes on to specify the requirements for intrusion detection devices with
which the inner area must be equipped.
2.3 Authorization for access to inner and protected areas
It is the responsibility of the licensee to ensure that persons being granted authorized access
to inner and protected areas are bona fide and pose no threat to the facility. Nuclear
Security Regulations sections 17 through 29 set out specific requirements for controlling
access to these areas.
The licensee is also responsible for issuing all authorizations for access to protected areas.
Authorization for unescorted access to inner areas may only be granted by the Commission
at the licensee's request. The process for authorizing unescorted or escorted access to the
protected and inner areas of nuclear facilities is detailed further in sections 2.4 to 2.7 of this
document.
Note: In the case of CNSC inspectors who are designated under subsection 29(1) of the
NSC Act to inspect protected or inner areas of nuclear facilities, the required authorization
4
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
for access to protected areas is issued by the Commission only.(Nuclear Security
Regulations, section 29)
Recommended To aid in the identification of authorized persons, the use of colour-coded
Practice: badges indicating authorization status is recommended. All persons who enter
a protected or inner area should be issued a badge that both identifies them and
indicates their authorized category. They should also be required to wear their
assigned badge so that it is clearly visible during the entire duration of their
stay in all protected and inner areas. Different coloured badges could indicate
that the wearer has either:
* authorized access to all areas,
* authorized access to the protected area only, or
* escorted access only.
Such badges should be difficult to counterfeit, and their distribution should be
carefully controlled and recorded.
2.4 Unescorted access to a protected area
Nuclear Security Regulations subsection 17(1) specifies that no person may enter a
protected area unescorted without physical proof of the licensee's recorded authorization.
Access to the protected area should be limited to persons who have a business requirement
to be in that area. The licensee is responsible for granting authorization for unescorted
access to workers and to other individuals, such as contractors, who routinely have business
inside the protected area.
Prior to issuing authorization for a person to have unescorted access to a protected area,
the licensee must prepare an identification report on the individual being granted access.
Nuclear Security Regulations subsection 17(2) requires that this identification report
include the following:
(a) the person's name, date of birth, and place of birth
The person's name should be written in full; initials are not sufficient. For example, write,
John Michael Smith, not John M. Smith on the application; the person's place of birth
should include the city or town, province and country. This information should be verified
using original documents such as a passport or other government issued photo
identification.
(b) documentary proof of the person's lawful presence in Canada
5
Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
Examples of documentation include:
* a Canadian birth certificate, citizenship certificate or valid passport
* proof of landed immigrant status
* a Quebec baptismal certificate
* if the person is in Canada temporarily, a valid passport bearing the appropriate entry
stamp or visa
(c) the address of the person's principal residence
This is the person's home address. The address should be complete, including street address
and apartment number where applicable. Rural addresses should include identifiers such as
lot number, concession and rural route number. This should be verified using documentation
such as a valid driver's licence. If the person is an external worker or a visitor not residing
in the area, also obtain and verify that person's local address (hotel, for example).
(d) a photograph depicting a frontal view of the person's face
Photographs must be clear and well-defined, showing a full frontal view of head and
shoulders without any form of casual head covering.
(e) the person's occupation
It is important to verify the trustworthiness of each person who will have unescorted access
to the protected area. This can be done through checks with former employers, police or
other authorities to verify both the identity and the reliability of the individual.
In a case where the access is for a short period of time, and it is not possible (or practical)
to verify the trustworthiness of the individual, only escorted access shall be provided
[Nuclear Security Regulations paragraph 17(3)(b) and subsection 17 (4)].
Identification reports are not required for designated CNSC inspectors who are authorized
to enter the protected or inner area of the nuclear facility. (Refer to Nuclear Security
Regulations section 29.) The inspector's certificate, however, should always be verified
prior to granting access to the facility, and details of the certificate noted in the
organization's records.
6
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
2.5 Escorted access to protected areas
The licensee may authorize access to a protected area for visitors and infrequent workers,
such as contractors or external workers, without preparing an identification report only
when providing an escort for whom an identification report has been prepared [Nuclear
Security Regulations paragraph 17(3)(b)].
In accordance with Nuclear Security Regulations subsection 17(3)(a), persons who will be
escorted during their stay inside the protected area must provide the licensee with
documentary proof of their name and address.
Recommended For identification purposes, this documentary proof may be a valid driver's
Practice: licence with photograph, or similar form of official photo identification. The
local address (hotel, for example) of any person whose principal residence is
not in the surrounding area should be obtained and verified using the licensee's
established security procedures when warranted.
If there is any reason for doubt regarding identification of an individual,
precautions should be taken. For example, in the case of an external worker,
the person's employer could be contacted to verify that this person is in fact
the individual assigned to work at this site.
It is good practice to require advance notification from all persons who seek
access to a nuclear facility. This advance notification would provide the
licensee with the time to complete any necessary verifications.
It is also good practice to place limits on the authorization. For example,
access could be limited to specific hours or to weekdays only; the access
authorization could include an expiry date after which access is denied unless
the authorization is renewed.
To establish adequate control of visiting groups, visitor/escort ratios should be
clearly defined. The maximum permissible visitor/escort ratio should be set at
10 to 1. Further reductions may be necessary depending upon the persons
being escorted and on the sensitivity of areas to be visited.
Workers assigned to escort duty are expected to be able to maintain control of
the persons being escorted and must accompany them at all times while in the
protected area. This includes mealtimes and other essential breaks. The
escort's proximity is necessary both for the safety of the visitor and the security
of the facility.
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Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
2.6 Unescorted access to an inner area
Nuclear Security Regulations subsection 18(1) requires that any person entering an inner
area have:
* recorded authorization from the licensee to enter the protected area, and
* recorded authorization from the CNSC to enter the inner area.
Recommended In general, the inner area of a nuclear facility should be accessible to as few
Practice: people as possible. Few individuals need to access the inner area, and the
licensee should limit this access.
It is recommended practice for licensees to limit unescorted access to workers
whose duties require regular access to this area. Licensees should also apply
the "two-person rule," which means that, whenever accessed, at least two
persons be present in the inner areas at any time.
Authorization for a person to have unescorted access to the inner area can only be granted
by the CNSC. Obtaining this authorization requires the licensee to submit to the
Commission an application signed by the licensee and the person for whom the
authorization is sought. This application must also include an explanation why this person
needs unescorted access to the inner area. The Nuclear Security Regulations paragraphs
18(2)(a) to (f) specify that the "application must contain the following information and
documents:
(a) a copy of the identification report referred to in Nuclear Security Regulations
subsection 17(2);
(b) Personnel Security Clearance Questionnaire TBS/SCT 330-60, as amended from
time to time, completed and signed by the person for whom the authorization is
sought;
(c) Personnel Screening Request and Authorization form TBS/SCT 330-23, as amended
from time to time, completed and signed by the person for whom the authorization
is sought;
(d) a description of the purpose for which entry into the inner area is required;
(e) a record emanating from the Canadian Police Information Centre, showing the
results of the Centre's criminal record name check on the person for whom the
authorization is sought; and
8
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
(f) at the request of the Commission, any other information that the Commission
requires for the purpose of subsection (3)."
In accordance with Nuclear Security Regulations subsection 18(5), if requested, the
licensee must provide the applicant with copies of any documents or information used in the
application for authorization.
If the application is approved, the Commission will issue an authorization allowing the
person unescorted access to the inner area. This authorization is valid only for the period of
time specified by the Commission, which may be up to a maximum of five years.
Recommended Original documents should be submitted to the Commission with this
Practice: application.
For renewals, the licensee should submit applications at least six months prior
to the expiry of the current authorization.
2.7 Escorted access to inner areas
The licensee may grant temporary escorted access for persons not authorized by the
Commission who must have access to the inner area to perform authorized work. While in
the inner area, these persons must be escorted at all times by someone who is authorized by
the Commission to access the inner area.
Nuclear Security Regulations subsection 20(2) requires the licensee to obtain the following
information for individuals prior to granting written authorization for escorted access to
inner areas:
(a) the name of the person who is authorized,
(b) the address of the person's principal residence, and
(c) the name and business address of the person's employer.
Recommended For identification purposes, a valid driver's licence with photograph, or similar
Practice: official photo identification is adequate. If the person seeking authorization is
not an area resident, the licensee should obtain and verify the person's local
address (hotel, for example). If there is any reason to doubt the identification
of the person seeking access to the inner area, precautions should be taken. For
example, the employer of an external worker could be contacted to verify that
this is in fact the person who was assigned to do work at this site.
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Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
The identity of persons should be verified in advance. Just as for access to
protected areas, it is recommended that all persons who seek access to the
inner area be required to provide advance notification. This advance
notification would provide the licensee with the time to complete any necessary
verifications.
It is also good practice to place access restrictions on the authorization.
Consider, for example, restricting access to specific hours or to weekdays only
and including an expiry date after which access is denied until the authorization
is renewed.
It is expected that workers assigned to escort duty will be able to maintain
control of the persons being escorted, and will accompany them at all times
while in the inner area. This includes mealtimes and other essential breaks. The
escort's proximity is necessary both for the safety of the visitor and the security
of the facility.
2.8 Revoking an authorization
Pursuant to Nuclear Security Regulations section 21, a licensee has the authority at any
time to revoke an individual's authorization to enter a protected area issued under
section 17 or an authorization to enter an inner area issued under section 20.
The Nuclear Security Regulations subsection 22(1), gives the Commission authority to
revoke any authorization to enter a protected area issued under section 17, or an
authorization to enter an area issued under section 18 or 20 if there are reasonable grounds
to believe the entry of the person into the protected or inner area will pose a security risk.
This means that only the Commission may revoke an authorization for unescorted access to
an inner area.
Although licensees may not revoke an authorization for unescorted access to an inner area,
the licensee can, for security reasons, deny access to the inner area by revoking access to
the protected area (Nuclear Security Regulations section 21).
Recommended In the event that the licensee utilizes this method to prevent an authorized
Practice: person from entering the inner area, the Commission should be immediately
notified.
If the Commission revokes an authorization, it will, in accordance with the Nuclear Security
Regulations subsection 22(2), immediately notify both the licensee and the person
concerned and give reasons for its action.
10
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
2.9 Keeping records
Nuclear Security Regulations section 37 requires that the licensee record the name of each
person who has been authorized to enter either a protected area or an inner area. These
records must be kept for at least one year from the time the authorization expires or is
revoked. A current list of all persons authorized to enter the protected and inner areas must
be available to the nuclear security guards who control access to the areas. The list may be
on paper or in electronic form, such as a computer database, and must be kept up to date.
Recommended It is good practice to maintain daily records of all persons who enter and exit
Practice: the protected area or the inner area, noting the times of each entry and exit. In
the case of persons requiring escort, the name of the escort should also be
recorded.
3 Security Issues Related to Access
The CNSC General Nuclear Safety and Control Regulations and Nuclear Security Regulations
also regulate other factors which impact the security of protected and inner areas. Additional
factors which need to be considered in order to provide optimum security for the protected and
inner areas of a nuclear facility include the use of nuclear security guards, licensee responsibility
for ongoing security training for workers, the control of persons, materials and vehicles entering
and exiting protected and inner areas, and procedures for dealing with unauthorized entry.
3.1 Nuclear security guards
A primary component in the provision of security for nuclear facilities is the use of nuclear
security guards. Refer to Nuclear Security Regulations sections 30 to 34 for the licensee's
regulatory requirements with regard to nuclear security guards.
Nuclear security guards are employed by the licensee (with the written consent of the
Commission) to control the movement of people, material and vehicles in and out of the
protected and the inner area. These guards are responsible for ensuring that no weapons or
explosives are brought into these areas and that no unauthorized nuclear material is
removed from these areas. Nuclear security guards are also responsible for conducting the
necessary security patrols, apprehending and detaining unarmed intruders, and operating
security equipment and systems.
11
Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
3.2 Worker training
General Nuclear Safety and Control Regulations paragraph 12(1)(j) requires the licensee to
implement an ongoing security awareness program to ensure that all workers, including
contract workers and students, are instructed on the facility's security program and their
obligations under that program. Security-related issues should also be discussed regularly at
staff meetings.
3.3 Monitoring entry and exit
The purpose of monitoring is to intercept weapons and explosives upon entry and to
prevent removal of nuclear material without authorization. Nuclear Security Regulations
paragraphs 27(1)(a) and (b) require that all escorted persons and everything in their
possession be searched by a nuclear security guard, both upon entering and when leaving a
protected or inner area. Where there is reasonable suspicion, all other persons authorized
for unescorted access to enter protected or inner areas and everything in their possession,
must be searched upon entering and leaving a protected or inner area. This includes a search
for all firearms and any tool or object not specifically required for work in the area that
could be used as a weapon or to cause damage. The search is to include all carried items
such as briefcases, bags and coats.
Recommended At many facilities, searches can be performed by means of a hand-held scanner.
Practice: For large facilities, the use of additional equipment such as portal metal
detectors and x-ray scanners similar to those used at airports is recommended.
Where detection equipment is used, the licensee is responsible for making sure
that all security staff operating the equipment are properly trained in its use,
and are aware of the appropriate procedure to follow when the equipment
alarm is activated. For example, if a metal detector appears to have been
triggered by a set of keys in someone's pocket, the nuclear security guard
should have the person remove the keys and go through the detector again to
verify that the keys were not being used as a decoy for a weapon.
All detection equipment should be tested regularly and properly maintained by
trained technicians. The procedures established for testing, calibration and
maintenance should be in accordance with the manufacturer's
recommendations.
In cases where the results of the detection device are inconclusive, the nuclear security
guard must carry out a hands-on or "frisk" search. Nuclear Security Regulations
12
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
paragraph 27(5)(b) requires that any hands-on search be performed by a person of the same
sex as the person being searched.
Recommended The objective of searching people as they leave the inner or protected area is to
Practice: ensure that no unauthorized nuclear material is removed. For this purpose,
radiation detection devices should be used as part of the search. The nuclear
security guards performing the search must also check everything in the
person's possession that might possibly contain shielding materials. During a
search for nuclear material, the nuclear security guard should always consider
the smallest component that could be removed.
Any person may refuse to be searched, but pursuant to Nuclear Security Regulations
subsection 28(1), that person must be denied entry to or exit from the protected or inner
area. Refusal to be searched when leaving the protected or inner area would immediately
give the nuclear security guards reasonable grounds to detain that person and call the local
law enforcement agency.
3.4 Vehicle access
Pursuant to Nuclear Security Regulations paragraphs 27(1)(a) and (b), if a vehicle must be
brought into the protected or the inner area, it must be thoroughly searched for weapons
and explosives prior to entry, and again searched to prevent the unauthorized removal of
nuclear material prior to exit from protected or inner areas. These searches must be
repeated each time any vehicle enters or leaves either area.
Recommended Vehicle access into the protected or inner areas of nuclear facilities should only
Practice: be allowed when absolutely necessary. This includes all private vehicles, such
as contractors' vehicles and delivery vans, as well as the licensee's own
vehicles.
If a licensee's vehicle requires frequent access to the protected area, it may be
preferable to keep the vehicle inside the area to be used as a "shuttle." The
shuttle vehicle would then remain within the designated protected area at all
times.
The Commission expects that a vehicle search will include a thorough
inspection of the passenger compartment, the engine compartment, the cargo
compartment and any cargo, as well as the top and the underside of the
vehicle. Security staff should enter the cargo compartment to verify the cargo.
13
Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
Parking for both workers and visitors, regardless of seniority or executive
privilege, should never be provided inside the protected area. In the case of
persons with disabilities, the licensee should ensure that the workplace is
accessible without compromising security requirements.
3.5 Unauthorized entry
Nuclear Security Regulations subsection 24(1) states that unauthorized persons must not be
permitted to "enter or remain in a protected or inner area."
If it is suspected that any individual who has access to either the protected area or the inner
area may pose a threat to the nuclear facility in any way, immediate steps should be taken to
mitigate that threat. Security is the first priority, and whatever action is necessary should
not be delayed.
The Commission expects the licensee to deal firmly with intruders. The licensee should
instruct all workers to immediately report the presence of any unauthorized person in the
inner or protected area to the nearest nuclear security guard [Nuclear Security Regulations
subsection 24(2)]. Pursuant to Nuclear Security Regulations paragraph 30(e), nuclear
security guards are then required to "apprehend and detain unarmed intruders" until they
can be turned over to the response force for subsequent action as deemed appropriate.
Recommended If armed intruders succeed in gaining access to the protected or the inner area,
Practice: engagement should be avoided. Armed intruders should be kept under
observation, and their movements should be continuously reported to the
security monitoring room until the response force arrives and takes control of
the situation.
14
C-205(E) Access Control for Protected and Inner Areas of Nuclear Facilities
GLOSSARY
escort an individual (normally a staff member or a nuclear security guard) who is
authorized by the Commission and licencee to have access to the protected
or inner area of a nuclear facility, and who has been assigned to accompany
persons granted escorted access to the area by the licensee. The escort is
expected to maintain control over the activity of the person(s) under escort
at all times.
external worker a person, employed by a firm or organization other than the licensee who
performs work that is referred to in a licence
inner area an area that is inside a protected area, where Category I nuclear material is
used or stored and that meets the requirements of sections 12, 13 and 14 of
the Nuclear Security Regulations.
nuclear
security guard a person who is authorized by a licensee, in accordance with section 31 of
the Nuclear Security Regulations, to act as a nuclear security guard at a
nuclear facility referred to in paragraph 2(b) of the Nuclear Security
Regulations.
response force a local, provincial or federal police force detachment, a Canadian Armed
Forces unit, or any other force trained in the use of firearms, that is
authorized under any Act or regulation to carry firearms and is qualified to
use them.
protected area an area under constant surveillance by a nuclear security guard or by
electronic means, surrounded by a physical barrier and having a limited
number of controlled access points that meets the requirements of sections
8, 10 and 11 of the Nuclear Security Regulations.
worker a person who performs work that is referred to in a licence including
anyone who is a student, attached staff, etc.
15
Access Control for Protected and Inner Areas of Nuclear Facilities C-205(E)
REFERENCES
IAEA. The Convention on the Physical Protection of Nuclear Material, INFCIRC/274/Rev. 1.
Vienna: May 1980.
IAEA. Guidance and Considerations for Implementation of INFCIRC/225/Rev. 3, (The Physical
Protection of Nuclear Material), IAEA Technical Document-967. Vienna: September 1997.
IAEA. The Physical Protection of Nuclear Material and Nuclear Facilities,
INFCIRC/225/Rev. 4, Vienna: March 1999.
Canada. CNSC. General Nuclear Safety and Control Regulations. Ottawa: 2000.
Canada. CNSC, Nuclear Security Regulations. Ottawa: 2000
16
DRAFT
REGULATORY
GUIDE
Transport Security for
Category I, II and III
Nuclear Material
C-208
Issued for public comments by the
Canadian Nuclear Safety Commission
December 2000
DRAFT REGULATORY GUIDE
Transport Security for Category I, II and III
Nuclear Material
C-208(E)
Issued for public comments by the
Canadian Nuclear Safety Commission
December 2000
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in regulatory programs.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY GUIDE
Transport Security for Category I, II and III
Nuclear Material
C-208(E)
About this Document
This guide is intended to apprise license applicants of the information they must submit to the
Canadian Nuclear Safety Commission (CNSC, the Commission) when preparing a transportation
security plan pursuant to the Nuclear Security Regulations. This plan must be submitted when
applying for any licence to transport Category I, II or III nuclear material outside of the
authorized area for that nuclear material.
Comments
The CNSC invites interested persons to assist in the further development of this document by
commenting in writing on its content and potential usefulness. Please respond by January 31,
2001, referencing file 1-8-8-208. Direct your comments and any related enquiries to the addresses
below.
The CNSC will study the comments received to determine how best to improve and finalize this
draft. Unless responders request otherwise, the CNSC will place a copy of their comments in its
library in Ottawa.
Document availability
This document can be viewed on the CNSC Internet website at www.nuclearsafety.gc.ca. A copy
of C-208 in English or French may be ordered using the contact information below:
Operations Assistant
Corporate Documents Section
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9 CANADA
Telephone (613) 996-9505
Facsimile (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
i
Transport Security for Category I, II and III Nuclear Material C-208(E)
Contents
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory framework . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Relevant legislation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2 The Transportation Security Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.2 General guidelines for the preparation of a transportation security plan . . . . . 2
2.3 Content of the transportation security plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2.4 Description of the nuclear material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.5 Threat assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.6 Mode of transportation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
2.7 Security measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
2.8 Communication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
2.9 Route planning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
2.10Response force security arrangements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
2.11Import/export transportation of Category I, II and III nuclear material . . . . . 8
2.12In-transit transport of Category I, II and III nuclear material through Canada
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
2.13Transport of Category I, II and III nuclear material under special arrangements
9
3 Guidelines for the Protection Nuclear Material During Transportation. . . . . . . . . . 9
3.1 Category I nuclear material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
3.1.1 Locks and seals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
3.1.2 Guards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
3.1.3 Security of transportation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
3.1.4 Transportation by road . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3.1.5 Transportation by rail . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
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C-208(E) Transport Security for Category I, II and III Nuclear Material
3.1.6 Transportation by ship . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3.1.7 Transportation by air . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3.1.8 Communication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3.2 Category II nuclear material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.2.1 Locks and seals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.2.2 Security of transportation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.2.3 Transportation by road . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
3.2.4 Transportation by rail . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
3.2.5 Transportation by ship . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
3.2.6 Transportation by air . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
3.2.7 Communication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
3.3 Category III nuclear material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
3.3.1 Locks and seals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
3.3.2 Security of transportation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
3.3.3 Transportation by road . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
3.3.4 Transportation by rail . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
3.3.5 Transportation by ship . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
3.3.6 Transportation by air . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
3.3.7 Communication . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
Appendix A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
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C-208(E) Transport Security for Category I, II and III Nuclear Material
Purpose
This guide is intended to apprise license applicants of the information they must submit to the
Canadian Nuclear Safety Commission (CNSC, the Commission) when preparing a transportation
security plan pursuant to the Nuclear Security Regulations. This plan must be submitted when
applying for any licence to transport Category I, II or III nuclear material outside of the
authorized area for that nuclear material.
Scope
This guide will be helpful for persons who are or will be involved with the transport and security
of Category I, II or III nuclear material. The guide explains the information requirements of the
transportation security plan as set out in the Nuclear Security Regulations and describes what the
CNSC considers to be effective ways to comply with these transportation security requirements.
This guide is concerned with transportation security issues only. Information regarding packaging,
placarding or other issues related to the transportation of nuclear material can be found in the
CNSC Packaging and Transport of Nuclear Substances Regulations and the Transport Canada
Transportation of Dangerous Goods Regulations.
1 Introduction
1.1 Regulatory framework
The Canadian Nuclear Safety Commission is the federal agency that regulates nuclear
facilities and nuclear material to prevent undue risk to health, safety, security and the
environment.
The CNSC operates under the authority of the Nuclear Safety and Control (NSC) Act and
associated regulations. The NSC Act and regulations prohibit persons or organizations from
transporting Category I, II and III nuclear material without a licence from the CNSC. The
regulations also stipulate prerequisites for licensing, including the information that is to be,
or may be, included in applications for specific types of licences.
1.2 Relevant legislation
NSC Act paragraph 26 requires that "Subject to the regulations, no person shall, except in
accordance with a licence, (a) possess, transfer, import, export, use or abandon a nuclear
substance, prescribed equipment or prescribed information; (b) mine, produce, refine,
convert, enrich, process, reprocess, package, transport, manage, store or dispose of a
nuclear substance; ..."
Information requirements for the CNSC licences to transport nuclear material are contained
in sections 3, 4 and 5 of the Packaging and Transport of Nuclear Substances Regulations.
Section 5 of the Nuclear Security Regulations further requires that any application to
transport Category I, II and III nuclear material include the submission of a written
transportation security plan. This guide explains the recommended content of such a
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transportation security plan regarding the movement, transport, transfer and security of
these nuclear materials. A listing of nuclear material contained in Category I, II and III is
provided in Appendix A of this document.
2 The Transportation Security Plan
2.1 Background
Nuclear material is vulnerable to theft or acts of sabotage when it is being transported.
Licensees must, therefore, ensure that the protection of nuclear material being transported
equals or exceeds the protection provided during use or storage.
In the transportation security plan, licence applicants need to identify the security measures
that will be applied to the transportation of Category I, II and III nuclear material in order
to minimize the risks and threats associated with the transportation of the nuclear material.
This transportation security plan is an integral part of the application for a licence to
transport nuclear material. This plan must be submitted and approved by the CNSC prior to
any licence to transport nuclear material being issued.
The envelope containing the plan should be:
* addressed to "NSSD, Security Advisor,"
* sealed and clearly marked "Confidential" or "Protected - Security",
* labelled "To be opened by the addressee only" and
* placed in a second, sealed envelope addressed to:
Canadian Nuclear Safety Commission
P.O. Box 1046, Station "B"
Ottawa, ON K1P 5S9
2.2 General guidelines for the preparation of a transportation security
plan
The transportation security plan details how physical protection of the nuclear material will
be provided during transportation. The content of the security plan will vary depending on
the category of nuclear material that is to be transported. In general, security requirements
for transport of Category I nuclear material are more stringent than for the transport of
Category II nuclear material; similarly, requirements for the transport of Category II nuclear
material are more stringent than for the transport of Category III nuclear material.
Regardless of the category of nuclear material being transported, applying the following
general guidelines should assist in maximizing the physical protection of nuclear material
while in transport:
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C-208(E) Transport Security for Category I, II and III Nuclear Material
* The total time during which the nuclear material remains in transport should be
minimized.
* The number and duration of nuclear material transfers, i.e., transfers from one
conveyance to another, transfers to and from temporary storage, temporary storage
while awaiting the arrival of a vehicle, etc. should be minimized.
* Regular transport schedules for the movement of nuclear material should be avoided.
* Different routes should be used to transport nuclear material whenever possible.
* Information about the movement of nuclear material should be restricted to those
who need to know.
* Prior arrangements for the planned shipment should be made with the receiver and
details, such as the mode of transport, handover point and arrival time, should be
confirmed.
* The trustworthiness of everyone involved in planning and executing the transport of
the nuclear material should be verified, using the licensee's established procedures,
prior to the transport.
* A transport control centre should be established to coordinate the operation and to
assure that secure and reliable communications are in place at all times during the
transport of Category I, II and III nuclear material.
2.3 Content of the transportation security plan
This document describes recommendations for specific information that should be included
in the transportation security plans for each category of nuclear material. The remainder of
section 2 of this guide discusses each of the information requirements listed in section 5 of
the Nuclear Security Regulations:
* description of the nuclear material (paragraph 5(a))
* threat assessment (paragraph 5(b))
* mode of transportation (paragraph 5(c))
* security measures (paragraph 5(d))
* communication arrangements (paragraph 5(e))
* route planning (paragraphs 5(g) and (h))
* response force security arrangements (paragraph 5(f))
CNSC information requirements for the licensing of international shipments of nuclear
material including in-transit shipments are also discussed.
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2.4 Description of the nuclear material
As the first item in a transportation security plan, a detailed description of the nuclear
material to be transported should include:
* the name of the nuclear material (chemical and/or industrial)
* the quantity of nuclear material to be transported (gross mass, net mass and mass of
nuclear material)
* the chemical and physical characteristics of the nuclear material
* the isotopic content of the nuclear material to indicate the degree of enrichment or
dilution (specifically uranium-235, uranium-233 and plutonium)
* the radiation level of the overall shipment as well as its discrete parts
In addition to the above, the transportation security plan should also include a description of
how the nuclear material will be packaged for shipment. This should include such
information as the type of container, its size and weight, as well as the method of securing
the container on the transport conveyance.
2.5 Threat assessment
As per paragraph 5(b) of the Nuclear Security Regulations, applicants must prepare a threat
assessment for the transport of Category I, II and III nuclear material. This threat
assessment is part of the transportation security plan, and identifies possible threats to the
security of the shipment. Assessment for Category I and II shipments should be
considerably more thorough than assessments for Category III shipments. These
assessments should (a) outline the nature and likelihood of any threat related to the
shipment and (b) describe the probable consequences of any such threats.
The CNSC receives assessments from the responsible federal security agencies identifying
known criminal, extremist or terrorist threats related to the movement of nuclear material.
The applicant should verify the extent of the threat assessment required with the CNSC
prior to preparing their application. The CNSC expects licensees to include in this
assessment all local public concerns with respect to any proposed transfer of nuclear
material. For example, news media coverage that might give rise to unwarranted fears
should be identified, and all planned or likely demonstrations by protest groups should be
noted. In addition, information on all activities that might be relevant to the transportation
security plan, i.e., road closures, construction or emergency detours, etc., should be
obtained from local and provincial response forces.
2.6 Mode of transportation
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C-208(E) Transport Security for Category I, II and III Nuclear Material
The next step in the preparation of a transportation security plan is to provide a complete
and detailed description of all conveyances involved in the transport, from the time it leaves
its originating location until it reaches its planned destination.
The journey may involve more than one transportation mode - for example, by road to a
rail terminal, then by rail for the main portion of the journey, and finally by road from
another rail terminal to the planned destination site. Details should be provided for each
segment of the journey, including a description of how the nuclear material will be
transferred from one conveyance to another. For example, when nuclear material is
transferred from a freight car to a truck, the transportation plan should specify the time and
location of the transfer and the name of the person responsible for the transfer and for
verifying that the shipment is intact.
Where interim storage of the nuclear material in transport is required, levels of physical
protection during storage should be equivalent to the physical protection measures required
for storage of the same category of nuclear material at a licenced nuclear facility. Details of
these storage security measures should be provided in the transportation security plan.
Consideration will be given to alternate physical protection arrangements as may be
applicable to the location of the interim storage and prevailing circumstances, for example
the nuclear material's appeal to thieves or terrorist groups.
The plan should provide evidence that all vehicles used for transporting the nuclear material
have been properly maintained and provide adequate protection for the shipment. The
transportation plan should also include provisions for a detailed search of these vehicles
prior to shipment to detect any possibility of sabotage. As mentioned earlier, the CNSC
expects that this search will be considerably more thorough for Category I and II nuclear
material than for Category III nuclear material. The intensity of the search should also be
determined by the results of the threat assessment.
2.7 Security measures
The security measures' section of a transportation security plan provides details of the
specific security measures proposed for this transport of nuclear material. A number of
factors will influence the proposed security measures. These include the category of nuclear
material to be transported, the size and type of the shipment, the distance and type of terrain
to be covered, the mode of transportation and the threat assessment.
Some of the issues to be addressed include whether the shipment will be sealed or unsealed;
vehicle staffing; requirements for guards, armed escort and escort vehicles; arrangements
for a response force support and notification of local response forces along the route, as
addressed in section 2.10 of this guide.
The plan should include contingencies for events such as a mechanical breakdown of the
vehicle carrying the nuclear material or its escort vehicle. Action to be taken by the
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Transport Security for Category I, II and III Nuclear Material C-208(E)
responsible licensee if the shipment fails to arrive at its destination at the expected time
should also be indicated.
For a detailed review of the security measures recommended for each category of nuclear
material, refer to section 3 of this guide.
2.8 Communication
Communication arrangements that will be in place throughout the transport of the nuclear
material should be detailed. This may include communication among:
* the transport control centre for the operation (if one exists)
* the operator(s) of the vehicle(s) transporting the nuclear material
* the licensee responsible for the shipment
* the licensee receiving the shipment
* all response forces and other law enforcement authorities involved along the route in
all jurisdictions through which the shipment will pass
Although cellular telephones generally provide adequate communication, it is important for
all those involved in the transport and security of nuclear material to be aware that
communication by cellular phone is not secure. Unencrypted cellular phones are therefore
not recommended for use during the transport of Category I nuclear material. Encrypted
radio systems provide a more secure means of communication. Note that if the
transportation route traverses remote regions, there may be gaps in cellular or radio
coverage. Whether using a radio or a cellular phone, it is important to assure that coverage
is adequate along the entire route. If it is not possible to avoid "blackout" areas, additional
arrangements should be made to establish alternate communications. For example, the
driver or escort could call in from public phones at pre-arranged locations. Consideration
should be given to using newer, more reliable and secure communication technologies as
they become available.
If using cellular phones for communications during the transport of Category II or III
nuclear material, use should be limited and messages should be encoded where possible.
Whatever communication method is used, the plan should incorporate backups. For
example, when using cellular phones, the vehicle should be equipped with more than one
phone and supplementary power sources.
Facilities making regular shipments of nuclear material may find it more efficient to establish
a transport control centre. This centre would be operational whenever a shipment of nuclear
material is being transported. Staff would be trained in techniques used to monitor the
movement of shipments during transportation, and should be familiar with the
communication protocols established with emergency response units and other authorities.
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C-208(E) Transport Security for Category I, II and III Nuclear Material
The communications section of the plan should also indicate the action to be taken if
communication with the vehicle carrying the nuclear material is lost. Applicants may wish to
consider the use of electronic tracking devices, such as transponders, which can be
concealed on the vehicle and/or in the load. These devices would be helpful in establishing
the location of the vehicle and the nuclear material, particularly if communications are
interrupted for any reason.
2.9 Route planning
As per Nuclear Security Regulations paragraphs 5(g) and (h), the transportation security
plan must include a detailed route plan, which incorporates an alternate route in the event
that some unexpected incident makes the primary route unsafe.
When planning the route for the transport of nuclear material, avoid urban areas wherever
possible. If the vehicle passes through a city or town, the plan should specify the precise
route to be taken through this area. The journey should be timed so that the shipment will
not pass through urban areas during peak traffic hours. Check local conditions along the
route, and identify any potential natural hazards, such as rock slides, floods or forest fires at
relevant times of the year.
In choosing an alternative route, consider that it may be necessary to switch from the
primary route to the alternative route during the journey. Where possible, there should be
connecting points between the two routes, and these should be identified in the plan. The
plan should also indicate what action will be taken in the event of unexpected delays caused
by natural or other hazards. Under such circumstances, the importance of reliable
communications is paramount, as addressed in 2.8 above.
If the journey is expected to take more than one day, overnight stays should be at a
pre-arranged location where the vehicle can be immobilized and kept in a physically secure
monitored area. The cargo should be secured on the vehicle to avoid easy removal. The
appropriate response force should be notified of the exact overnight location of the vehicle.
In general, the level of physical protection provided during an overnight stop should be
equivalent to the physical protection measures required for storage of the same category of
nuclear material at a licenced nuclear facility. Every effort should be made to ensure that all
practicable requirements of the physical protection measures required at the nuclear facility
are applied to each overnight stop.
2.10 Response force security arrangements
Prior to the transport, the response force in each jurisdiction involved should be notified of
the movement of nuclear material, and clear channels of communication established. The
local law enforcement agency will usually provide an escort and serve as an armed response
force where required. All arrangements made with the response force should be clearly
7
Transport Security for Category I, II and III Nuclear Material C-208(E)
indicated in the transportation security plan. If for any reason the local law enforcement
agency is unable to provide the necessary security, alternate arrangements, such as
employing a private security firm, should be made.
In cases where the established transport route involves response forces provided by more
than one law enforcement agency, the plan should detail all arrangements for the transfer of
responsibility from one response force to another. For example, transport might involve
crossing a provincial border, with transfer from the jurisdiction of the Ontario Provincial
Police to the Sûreté du Québec or Royal Canadian Mounted Police. All communication
changes resulting from such a transfer - such as radio frequency or encryption method -
should also be clearly documented in the transportation security plan.
2.11 Import/export transportation of Category I, II and III nuclear
material
The Convention on the Physical Protection of Nuclear Material (INFCIRC/274/Rev. 1) calls
on states to cooperate in providing protection for the international transport of nuclear
material. In general, it is important to ensure that the appropriate regulatory agency in the
country (or countries) involved are aware of the transport and are in agreement as to who
will be responsible for the shipment at any stage. In addition, the agreement between the
shipper and the receiver should clearly state at what point responsibility for physical
protection of the shipment is transferred to the receiver. When the nuclear material is within
Canadian jurisdiction, it must be the responsibility of a CNSC licensee (NSC Act, section
26).
For example, in the case of shipments between Canada and the United States, responsibility
for physical protection begins and ends at the border between the two countries. This means
that a shipment entering Canada from the United States comes under the jurisdiction of the
CNSC licensee as soon as it enters Canada.
As outlined in The Physical Protection of Nuclear Material and Nuclear Facilities, IAEA
INFCIRC /225/Rev. 4, subsection 8.1.5, before nuclear material is transported
internationally, the shipper should ensure that the arrangements are in accordance with the
physical protection regulations of the receiving state and of other states which are transited.
Prior to a nuclear material's entry into Canada, the receiver of the nuclear material is
responsible for obtaining the required import and transportation licences from the CNSC.
For nuclear material being exported from Canada, the shipper is responsible for obtaining
the required export and transportation licences from the CNSC.
An international shipment might also pass through the territory of one or more other
countries, including territorial waters and air space. In this case, the sending and receiving
countries should include those "transit" countries in their arrangements in order to enlist
their cooperation in providing physical protection for the shipment
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C-208(E) Transport Security for Category I, II and III Nuclear Material
2.12 In-transit transport of Category I, II and III nuclear material
through Canada
In-transit transport of Category I, II and III of nuclear material (i.e., nuclear material whose
point of origin and destination are outside of Canada) being transported through Canada
requires a licence to transport a nuclear substance. Applicants will find details of the
information requirements for licence applications for in-transit shipment of nuclear material
in section 4 of the Packaging and Transport of Nuclear Substances Regulations.
2.13 Transport of Category I, II and III nuclear material under special
arrangements
Transport of Category I, II and III nuclear material under special arrangements requires a
licence to transport a nuclear substance. Applicants will find details of the information
requirements for licence applications for special arrangement shipment of nuclear material in
section 5 of the packaging and Transport of Nuclear Substances Regulations.
3 Guidelines for the Protection Nuclear Material During
Transportation
This section provides a more detailed description of the specific measures which should be taken
to provide an acceptable level of protection for transporting Category I, II and III nuclear
material. These guidelines are based on recommendations contained in the International Atomic
Energy Agency (IAEA) circulars on The Physical Protection of Nuclear Material and Nuclear
Facilities (INFCIRC/225/Rev. 4) and The Convention on the Physical Protection of Nuclear
Material (INFCIRC/274/Rev. 1), as well as the requirements under the Nuclear Security
Regulations.
Note: The packaging and placarding of nuclear material for transportation are covered in the
CNSC Packaging and Transport of Nuclear Substances Regulations and the Transport Canada
Transportation of Dangerous Goods Regulations.
3.1 Category I nuclear material
The following guidelines apply specifically to Category I nuclear material during
transportation. Note that the extent of the security arrangements (such as the size and
preparedness of the armed escort) will depend on the threat assessment. Consequently, an
increased threat level to the shipment would require that security measures be increased
accordingly.
3.1.1 Locks and seals
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Packages containing Category I nuclear material should be transported in closed,
locked or sealed vehicles, compartments or freight containers. Locked or sealed
packages weighing more than 2000 kg may, if necessary, be transported on open
vehicles with the package secured to the vehicle or freight container. Locks and seals
should be checked before departure, during the journey, and on arrival at the final
destination to verify that they have not been tampered with.
3.1.2 Guards
Shipments of Category I nuclear material should be accompanied by armed guards (or
other armed escort) and an escort vehicle to maintain constant surveillance and to
protect the shipment against the potential threats identified in the threat assessment
and/or other unauthorized actions. The guards will maintain regular communication
with the shipper, the receiver and local response forces throughout the transportation
of the Category I nuclear material, up to and including the handover point of the
shipment.
3.1.3 Security of transportation
For Category I nuclear material, regardless of the mode of transportation, the
conveyance used should be dedicated to the shipment.
The licensee is responsible for ensuring that the carrier is aware of, and can comply
with, the required physical security measures. When dealing with third-party carriers,
the licensee should emphasize the need for confidentiality and the trustworthiness of
everyone involved in the transport of nuclear material.
Before loading, a detailed search of the conveyance should be carried out. The search
should be performed by qualified personnel to ensure that there has been no attempt
to sabotage the conveyance. Following the search, the conveyance should be kept in a
secured area until loaded.
3.1.4 Transportation by road
If the Category I nuclear material is to be transported by road, the vehicle used to
carry the nuclear material should be designed to resist attack. The cargo should be
firmly secured to the vehicle. The vehicle should carry a crew of two (a driver and a
guard). An escort vehicle should be provided with one or more guards to maintain
constant surveillance of the shipment.
3.1.5 Transportation by rail
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If all or part of the journey is by rail, the Category I nuclear material should be
transported by freight train in a dedicated freight car. The car should be locked and/or
sealed. Two or more guards should accompany the shipment, and they should travel
in the rail car nearest the nuclear material to maintain constant surveillance. The
guards should be directed to check the locks and seals at regular intervals.
3.1.6 Transportation by ship
If all or part of the journey is by ship, the Category I nuclear material should be
transported in a sealed and/or locked freight containers on a dedicated cargo vessel.
Two or more guards should accompany the shipment and maintain constant
surveillance. The guards should be directed to check the locks and seals on the
containers at regular intervals.
3.1.7 Transportation by air
If all or part of the journey is by air, the Category I nuclear material should be
transported by dedicated cargo charter aircraft in a sealed and/or locked freight
containers. Two or more guards should accompany the shipment and maintain
constant surveillance. The guards should be directed to check locks and seals on the
containers at regular intervals.
3.1.8 Communication
The shipper is responsible for notifying the receiver in advance of the planned
shipment, providing details of the nuclear material, the mode of transportation, and
the estimated time of arrival. Prior to commencement of the transport of Category I
nuclear material, the receiver should confirm readiness to accept delivery at the
expected time and place. Immediately upon the arrival of the shipment, the receiver is
responsible for notifying the shipper of its arrival. It is important for the shipper and
the receiver to agree on a reasonable delay interval, after which the receiver will
notify the shipper in the event that the shipment does not arrive.
During the transport of Category I nuclear material, continuously available
communication using encrypted two-way radio should be used. Individuals
accompanying the shipment should remain in regular contact with the shipper, the
receiver and local authorities along the route. In the case of shipment by road, the
load vehicle should remain in constant contact with the escort vehicle(s). It is
important to establish, in advance, a plan of action in the event that communication is
lost for any reason. The establishment of a transport control centre, as described in
section 2.8 of this guide, should be considered.
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Transport Security for Category I, II and III Nuclear Material C-208(E)
If for any reason the shipment is lost, stolen or diverted, the CNSC must be notified
immediately (General Nuclear Safety and Control Regulations, paragraph 29 (1)(a),
and NSC Act, paragraph 27(b)).
3.2 Category II nuclear material
The following guidelines apply specifically to Category II nuclear material during
transportation.
3.2.1 Locks and seals
Packages containing Category II nuclear material should be transported in closed,
locked or sealed vehicles, compartments or freight containers. Locked or sealed
packages weighing more than 2000 kg may, if necessary, may be transported on open
vehicles with the package secured to the vehicle or freight container. Locks and seals
should be checked before departure, during the journey, and on arrival at the final
destination to verify that they have not been tampered with.
3.2.2 Security of transportation
For Category II nuclear material, regardless of the mode of transportation, the
Commission recommends that one or more escorts be provided to maintain constant
surveillance. The number of cargo transfers and the length of time the shipment is
actually being transported should be minimized. The licensee is responsible for
ensuring that the carrier is aware of and can comply with the required physical
security measures. When dealing with third-party carriers, the licensee should
emphasize the need for confidentiality and the trustworthiness of everyone involved in
the transport of nuclear material.
3.2.3 Transportation by road
If the Category II nuclear material is to be transported by road, arrange for a detailed
search of the vehicle before loading. The search should be performed by qualified
personnel to ensure that there has been no attempt to sabotage the vehicle. Following
the search, the vehicle should be kept in a secured area until it is loaded and ready to
move. Once loaded, the vehicle should be kept locked when not on the move. While
transporting Category II nuclear material, the vehicle should never be left unattended.
If escorts are used, they may accompany the driver in the vehicle transporting the
nuclear material.
3.2.4 Transportation by rail
12
C-208(E) Transport Security for Category I, II and III Nuclear Material
If all or part of the journey is by rail, the Category II nuclear material should be
transported by freight train or in a separate freight car attached to a passenger train.
The car should be locked and/or sealed.
3.2.5 Transportation by ship
If all or part of the journey is by ship, the Category II nuclear material should be
transported in a sealed and/or locked freight container.
3.2.6 Transportation by air
If all or part of the journey is by air, the Category II nuclear material should be
transported by designated cargo charter aircraft or scheduled cargo aircraft in a sealed
and/or locked freight container.
3.2.7 Communication
The shipper is responsible for notifying the receiver in advance of the planned
shipment, providing details of the nuclear material, the mode of transportation, and
the estimated time of arrival. Prior to commencement of the transport of Category II
nuclear material, the receiver should confirm readiness to accept delivery at the
expected time and place. Immediately upon the arrival of the shipment, the receiver is
responsible for notifying the shipper of its arrival. It is important for the shipper
13
Transport Security for Category I, II and III Nuclear Material C-208(E)
and the receiver to agree on a reasonable delay interval, after which the receiver will
notify the shipper in the event that the shipment does not arrive.
During the transport of Category II nuclear material, frequent communication using
two-way radios or cellular telephones should be used. Individuals accompanying the
shipment should remain in regular contact with the shipper, the receiver and local
authorities along the route. It is important to establish, in advance, a plan of action in
the event that communication is lost for any reason. The establishment of a transport
control centre, as described in section 2.8 of this guide, should be considered.
If for any reason the shipment is lost, stolen or diverted, the CNSC must be notified
immediately (General Nuclear Safety and Control Regulations, paragraph 29 (1)(a),
and NSC Act, paragraph 27(b)).
3.3 Category III nuclear material
The following guidelines apply specifically to Category III nuclear material during
transportation.
3.3.1 Locks and seals
Packages containing Category III nuclear material should be transported in closed,
locked or sealed vehicles or containers whenever feasible.
3.3.2 Security of transportation
For Category III nuclear material, regardless of the mode of transportation, the
number of cargo transfers and the length of time the shipment is actually being
transported should be minimized. The licensee is responsible for ensuring that the
carrier is aware of, and can comply with, the required physical security measures.
When dealing with third-party carriers, the licensee should emphasize the need for
confidentiality and the trustworthiness of everyone involved in the transport of
nuclear material.
3.3.3 Transportation by road
If the Category III nuclear material is to be transported by road, arrange for a detailed
search of the vehicle before loading. The search should be performed by qualified
personnel to ensure that there has been no attempt to sabotage the vehicle. Following
the search, the vehicle should be kept in a secured area until it is loaded and ready to
move. Once loaded, the vehicle should be kept locked when not on the move. While
transporting Category III nuclear material, the vehicle should never be left
14
C-208(E) Transport Security for Category I, II and III Nuclear Material
unattended.
3.3.4 Transportation by rail
If all or part of the journey is by rail, the Category III nuclear material should be
transported by freight train or in a separate freight car attached to a passenger train.
The car should be locked and/or sealed.
3.3.5 Transportation by ship
If all or part of the journey is by ship, the Category III nuclear material should be
transported in a sealed and/or locked container.
3.3.6 Transportation by air
If all or part of the journey is by air, the Category III nuclear material should be
transported by designated cargo charter aircraft or scheduled cargo aircraft in a sealed
and/or locked container.
3.3.7 Communication
The shipper is responsible for notifying the receiver in advance of the planned
shipment, providing details of the nuclear material, the mode of transportation and the
estimated time of arrival. Immediately upon the arrival of the shipment, the receiver is
responsible for notifying the shipper of its arrival. It is important for the shipper and
the receiver to agree on a reasonable delay interval, after which the receiver will
notify the shipper in the event that the shipment does not arrive.
Prior to commencement of the transport of Category III nuclear material, the receiver
should confirm readiness to accept delivery at the expected time and place.
If for any reason the shipment is lost, stolen or diverted, the CNSC must be notified
immediately (General Nuclear Safety and Control Regulations, paragraph 29 (1)(a),
and NSC Act, paragraph 27(b)).
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Transport Security for Category I, II and III Nuclear Material C-208(E)
REFERENCES
Canadian Nuclear Safety Commission. General Nuclear Safety and Control Regulations. Ottawa:
CNSC, 2000.
Canadian Nuclear Safety Commission. Nuclear Safety and Control Act, Ottawa: CNSC, 2000.
Canadian Nuclear Safety Commission. Nuclear Security Regulations. Ottawa: CNSC, 2000.
Canadian Nuclear Safety Commission. Packaging and Transport of Nuclear Substances
Regulations. Ottawa: CNSC, 2000.
International Atomic Energy Agency. The Convention on the Physical Protection of Nuclear
Material, INFCIRC/274/Rev.1. Vienna, Austria: IAEA, May 1980.
International Atomic Energy Agency. Guidance and considerations for the implementation of
INFCIRC/225/Rev.4, The Physical Protection of Nuclear Material, IAEA-TECDOC-967. Vienna,
Austria: IAEA, September 1997.
International Atomic Energy Agency. The Physical Protection of Nuclear Material and Nuclear
Facilities, INFCIRC/225/Rev.4. Vienna, Austria: IAEA, 1997.
Transport Canada Transportation of Dangerous Goods Regulations. Ottawa: Transport Canada,
1992.
16
C-208(E) Transport Security for Category I, II and III Nuclear Material
GLOSSARY
handover point the time and location at which the responsibility for the shipment is
transferred from one licensee to another
IAEA International Atomic Energy Agency
inner area an area inside a protected area in which Category I nuclear material is used
or stored
protected area an area under constant surveillance by a guard or by electronic means,
surrounded by a physical barrier and having a limited number of controlled
access points
sabotage any deliberate act directed against a plant, facility, nuclear material
transport vehicle or nuclear material that could directly or indirectly
endanger the public health and safety by exposure to radiation
shipment any goods sent at one time from one person (the shipper) to another (the
receiver). For purposes of this guidance document, the term shipment is
used to indicate a cargo of Category I, II or III nuclear material being
transported from its point of origin through to its final destination
transit country any country whose borders are crossed by land, by air or through territorial
waters during the transport of nuclear material where the place of initial
loading and the final destination are outside that country
17
Transport Security for Category I, II and III Nuclear Material C-208(E)
APPENDIX A
Category I, II and III Nuclear Material SCHEDULE (Section 1)
Item Nuclear Form Quantity Quantity Quantity
Substance (Category I)1 (Category II)1 (Category III)1
1 Plutonium2 Unirradiated3 2 kg or more Less than 2 kg, but 500 g or less, but
more than 500 g more than 15 g
2 Uranium 235 Unirradiated3- 5 kg or more Less than 5 kg, but 1 kg or less, but
uranium enriched to more than 1 kg more than 15 g
20% 235U or more
3 Uranium 235 Unirradiated3- N/A 10 kg or more Less than 10 kg, but
uranium enriched to more than 1 kg
10% 235U or more, but
less than 20% 235U
4 Uranium 235 Unirradiated3- N/A N/A 10 kg or more
uranium enriched
above natural,but less
than 10% 235U
5 Uranium 233 Unirradiated3 2 kg or more Less than 2 kg, but 500 g or less, but
more than 500 g more than 15 g
6 Fuel consisting Irradiated N/A More than 500 g 500 g or less, but
of depleted or of plutonium more than 15 g of
natural uranium, plutonium
thorium or low-
enriched fuel
(less than 10%
fissile content)4
1. The quantities listed refer to the aggregate of each kind of nuclear substance located at a facility, excluding the
following (which are considered separate quantities):
(a) any quantity of the nuclear substance that is not within 1 000 m of another quantity of the nuclear
substance; and
(b) any quantity of the nuclear substance that is located in a locked building or a structure offering similar
resistance to unauthorized entry.
2. All plutonium except that with isotopic concentration exceeding 80% in plutonium 238.
3. Material not irradiated in a reactor or material irradiated in a reactor but with a radiation level equal to or less
than 1 Gy/h at 1 m unshielded.
4. Other fuel that by virtue of its original fissile content is classified as Category I or II before irradiation may be
reduced one category level while the radiation level from the fuel exceeds 1 Gy/h at 1 m unshielded.
18
DRAFT
REGULATORY
STANDARD
Maintenance Programs
for Nuclear Power Plants
C-210
Issued for public comments by the
Canadian Nuclear Safety Commission
August 2001
DRAFT REGULATORY STANDARD
Maintenance Programs for
Nuclear Power Plants
C- 210
Issued for public comments by the
Canadian Nuclear Safety Commission
August 2001
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in regulatory programs.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY STANDARD
Maintenance Programs for Nuclear Power Plants
C-210
August 2001
About this document
This draft Regulatory Standard describes the maintenance program, and the related provisions
concerning the training of maintenance workers and the keeping of maintenance records, that the
Canadian Nuclear Safety Commission (CNSC) may incorporate into operating licences for nuclear
power plants.
Comments
The CNSC invites interested persons to assist in the further development of this draft regulatory
document by commenting in writing on the document's content and potential usefulness. Please
respond by November 30, 2001. Direct your comments to the postal or e-mail address below,
referencing file 1-8-8-210.
The CNSC will take the comments received on this draft into account when developing it further.
These comments will be subject to the provisions of the federal Access to Information Act.
Document availability
This document can be viewed on the CNSC Internet site at www.nuclearsafety.gc.ca. To order a
printed copy of the document in English or French, please contact:
Operations Assistant
Regulatory Documents Group
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 996-9505
Facsimile: (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
i
Maintenance Programs for Nuclear Power Plants C-210
CONTENTS
About this document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1.1 Regulatory framework . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1.2 Licensing process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1.3 Relevant legislation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1.3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1.3.2 Maintenance requirements in regulations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2 The maintenance program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.1 Objective . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.2 Contents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.2.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.2.2 Policies, procedures, training, and work control processes . . . . . . . . . . . . . . . . . 6
2.2.3 Preventive maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
2.2.4 Monitoring and assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
2.2.5 In-service inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
2.3 Requirements to keep and retain maintenance-related records. . . . . . . . . . . . 10
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C-210 Maintenance Programs for Nuclear Power Plants
Purpose
The purposes of this regulatory standard of the Canadian Nuclear Safety Commission (CNSC)
are:
* To help applicants for operating licences for nuclear power plants to develop plant
maintenance programs and related training and record-keeping provisions that meet pre-
licensing and post-licensing requirements under the Nuclear Safety and Control Act
(NSC Act, Act) and regulations;
* To describe the maintenance program for the structures, systems and components of a
nuclear power plant that the licensee shall develop, submit or implement when required
to do so by a condition of the licence to operate the nuclear power plant; and
* To facilitate CNSC evaluations of the adequacy of proposed or actual maintenance
programs for nuclear power plants.
Scope
This document describes the elements of a maintenance program that the CNSC may require, on a
case-by-case basis, of licensees who operate nuclear power plants. It also describes
complementary requirements for the training of maintenance workers and the keeping of
maintenance records.
When appropriately incorporated into an operating licence for a nuclear power plant, this standard
and its contents are mandatory. In other situations, the standard and its contents constitute
guidance to interested persons, such as to licence applicants, licensees and CNSC staff, on what
maintenance programs for nuclear power plants should typically entail.
This maintenance standard pertains only to those structures, systems and components of nuclear
power plants that, if not maintained properly, could result in unreasonable risks to health, safety,
national security or the environment.
Definitions
In this Regulatory Standard:
* A nuclear power plant (NPP) is any fission-reactor installation that has been constructed
to generate electricity on a commercial scale. A nuclear power plant is a Class IA
nuclear facility, as defined in the Class I Nuclear Facilities Regulations.
1
Maintenance Programs for Nuclear Power Plants C-210
* A maintenance program for a NPP is a program that consists of measures, policies,
methods and procedures for maintaining those structures, systems and components
(hereinafter, "SSCs") of the plant that, if not maintained properly, could result in
unreasonable risk to the health and safety of persons, to national security or to the
environment.
For the purpose of this standard, the measures, policies, methods, and procedures that comprise a
typical maintenance program for a nuclear power plant pertain to both preventive and corrective
maintenance as well as to related monitoring, testing, assessment, inspection and verification
activities.
1 Background
1.1 Regulatory framework
The CNSC is the federal agency that regulates the uses of nuclear energy and materials to
protect health, safety, security and the environment, and to respect Canada's international
commitments on the peaceful use of nuclear energy.
The NSC Act requires persons or organizations to be licensed by the CNSC for carrying out
the activities referred to in Section 26 of the Act, unless otherwise exempted. The
associated regulations stipulate prerequisites for CNSC licensing, and the obligations of
licensees and workers.
1.2 Licensing process
The CNSC typically applies a phased process to its licensing of activities. For major
projects that are subject to the Canadian Environmental Assessment Act and regulations,
this process typically begins with an assessment of the environmental impacts, and then
proceeds progressively through site preparation, construction, operation, decommissioning
and abandonment phases.
The NSC Act and regulations require licence applicants to provide certain information at
each licensing stage. The type and level of detail of this information will vary to
accommodate the licensing stage and specific circumstances.
At all licensing stages, applications may incorporate (directly or by reference) new or
previously submitted information, in accordance with legislated requirements and the best
judgement of the applicant. An application that is submitted at one licensing stage can
become a building block for the next stage.
Upon receipt of an application that is complete, the CNSC reviews it to determine whether
the applicant is qualified to carry on the proposed activity, and has made adequate provision
for the protection of the environment, the health and safety of persons, and the maintenance
of national security and measures required to implement international obligations to which
Canada has agreed. If satisfied, the CNSC may issue, renew, amend or replace a licence that
contains relevant conditions. Typically, this licence will incorporate the applicant's
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C-210 Maintenance Programs for Nuclear Power Plants
undertakings and will contain other conditions that the CNSC considers necessary,
including those that reference or incorporate CNSC Regulatory Standards.
Upon receiving a CNSC licence, the recipient becomes the licensee, and is obliged to
comply with the conditions of the licence. Accordingly, if a condition concerning
maintenance is incorporated into a CNSC operating licence for a NPP as described above,
the licensee will be required to meet any resulting obligations.
1.3 Relevant legislation
1.3.1 Introduction
In the interests of health, safety, national security and protection of the environment, a
NPP must implement and maintain effective maintenance programs, tailored to case-
specific needs. By definition, such programs are likely to be necessary for those
structures, systems and components (SSCs) of a NPP that, if not maintained properly,
could result in unreasonable risk to the health and safety of persons, to national
security or to the environment.
The success of a NPP maintenance programs, and the demonstration of its success,
typically depend upon such factors as appropriate planning and organization,
adequate staffing and work procedures, quality assurance and control measures, on-
going monitoring, testing and assessment, and record-keeping.
1.3.2 Maintenance requirements in regulations
The inclusion of maintenance-related requirements in the CNSC regulations reflects
the importance of adequate maintenance to the safety and national security of nuclear
facilities. Provisions in the Class I Nuclear Facilities Regulations encompass essential
elements of effective maintenance programs, such as:
* the proposed measures, policies, methods and procedures for maintaining
the facilities [See paragraph 6(d)];
* the training and qualification of workers [Paragraph 6(m)]; and
* the keeping and retaining of records [Subsections 14(2), 14(3) and 14(4)].
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Maintenance Programs for Nuclear Power Plants C-210
In addition to specific references to "maintaining" and "maintenance", the regulations
under the NSC Act include other more general provisions that also concern
maintenance programs and workers at NPPs. For example, the General Nuclear
Safety and Control Regulations stipulate that every licensee shall:
* ensure the presence of a sufficient number of qualified workers to carry on
the licensed activity safely and in accordance with the Act, the regulations
made under the Act and the licence [Paragraph 12(1)(a)];
* train the workers to carry on the licensed activity in accordance with the
Act, the regulations made under the Act and the licence
[Paragraph 12(1)(b)]; and
* make available to all workers the health and safety information with respect
to their workplace that has been collected by the licensee in accordance with
the Act, the regulations made under the Act and the licence [Subsection
16(1)].
In addition, the operator of a NPP must meet the requirements of the Radiation
Protection Regulations during the conduct of licensed activities, including those
associated with plant maintenance.
2 The maintenance program
2.1 Objective
The objective of a maintenance program for a NPP is to assure that the SSCs of the plant
that are subject to the program function reliably and effectively over the operating life of the
plant, in accordance with regulations and licence requirements. This objective can be met
through effective design and implementation.
2.2 Contents
2.2.1 Overview
The maintenance program for a NPP shall clearly stipulate what is to be done as part
of the program, how it is to be done, who is to do it, and to what standards.
Accordingly, the maintenance program shall:
Applicability and supporting information
* identify and describe the SSCs to which the program applies and the
information (data, calculations, drawings, plans and reports ) that is relevant
to effective maintenance of the SSCs, including that relating to design,
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C-210 Maintenance Programs for Nuclear Power Plants
specifications, flow sheets, as-built states, pre-service inspections, in-service
inspections, tests, examinations, evaluations, performance, calibrations,
commissioning, acceptance and performance standards, stress analyses,
thermal analyses, and structural analyses.
Policies, methods, procedures and responsibilities
* set out policies, methods, procedures and organizational responsibilities for
planning and implementing the program, including the preventive
maintenance activities, the corrective maintenance activities, and the
relevant monitoring, testing, assessment, inspection and verification
activities. [See paragraph 3(1)(k) of the General Nuclear Safety and
Control Regulations, and paragraph 6(d) of the Class I Nuclear Facilities
Regulations.]
Configuration management
* set out the measures to comply with any requirements of the licensee's
program for managing the configurations of the nuclear power plant that are
relevant to the maintenance of the SSCs.
Quality assurance
* set out the quality assurance provisions that are to be applied to the
planning and implementation of maintenance-related activities.
Management of ageing
* set out the measures to be implemented to prevent, limit or mitigate the
physical degradation of SSCs due to such ageing mechanisms as corrosion,
erosion, vibration, mechanical forces, hydride blistering, heat, cold and
radiation.
Preventive and corrective maintenance
* set out the preventive and corrective maintenance activities to be
implemented to assure that the SSCs function reliably and effectively, in
accordance with regulatory requirements.
* specify performance targets for effectively managing and controlling the
accumulation of `backlogs' in preventive and corrective maintenance
activities.
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Maintenance Programs for Nuclear Power Plants C-210
Monitoring and assessment
* set out the measures to be implemented in order to detect deficiencies in the
maintenance program in an timely manner.
* provide for periodic and objective auditing by the licensee or other parties
of the program's effectiveness.
Training and qualification of workers
* set out a program for training maintenance staff and workers to assure that
they understand and adhere to work processes and procedures, and are
qualified to perform their duties safely.
Change control
* set out a structured process for correcting deficiencies in, or implementing
changes to, the maintenance program in an effective and timely manner.
2.2.2 Policies, procedures, training, and work control processes
The maintenance program for a NPP shall include relevant policies, procedures and
work control processes:
Planning, implementation and supervision
* for the planning, implementation and supervision by staff or contractors of
maintenance activities on SSCs, taking into account:
- operating experience,
- relevant codes, standards and regulatory requirements,
- the methods, resources, sequencing and approvals required,
- the impacts of the scheduling and conduct of the activities on health,
safety, national security and the environment.
- the need for verifications, environmental qualifications, seismic
qualifications, exclusion of foreign material, or post-maintenance testing,
- requirements for special tools and equipment, and
- the availability of spare parts for implementing the maintenance program.
Training and qualification of workers
* to ensure the presence of a sufficient number of qualified workers to
implement the maintenance program safely.
* to train the workers to implement the maintenance program safely.
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C-210 Maintenance Programs for Nuclear Power Plants
Monitoring and assessment
* that ensure independent reviews at specified frequencies of the following
aspects of the maintenance program:
- the adequacy of the schedule and its implementation,
- the adequacy of past responses to operational requirements for corrective
and preventive maintenance, monitoring, testing, assessment and
inspections,
- the adequacy of the control and management of preventive and corrective
maintenance backlogs,
- the control of radiation doses to maintenance workers,
- the availability of necessary resources and their effectiveness,
- the qualifications and competency of maintenance workers,
- compliance with quality assurance standards,
- the adequacy of work procedures,
- the effectiveness of monitoring and assessment provisions,
- the detected incidence of SSC failures and any resulting impacts on health,
safety, national security and the environment,
- the number and nature of omitted or deferred maintenance tasks,
- repetitions of similar corrective maintenance work on the same SSC or on
comparable SSCs, and
- the adequacy of maintenance tools, equipment or facilities.
* to ensure that SSCs are not taken out of service for the purpose of
conducting maintenance on them, except in accordance with the prior
approval of a designated individual, and within the prescribed limits and
conditions for safe operation of the NPP.
* to ensure, following maintenance, tests or in-service inspections of SSCs,
that the affected SSCs are not returned to service until their respective
configurations have been verified safe for a return to operation, in
accordance with documentation or the results of tests or assessments
performed to the satisfaction of a designated individual.
7
Maintenance Programs for Nuclear Power Plants C-210
* to ensure that when a procedure for carrying out a maintenance activity is
found to be inadequate, the activity is halted or safely managed until the
deficiency in the procedure is corrected.
* for identifying and responding to maintenance situations for which work
procedures must be at hand and followed in the interests of health, safety,
national security and protection of the environment.
* to manage and control any preventive maintenance or any corrective
maintenance backlog.
* to ensure that any necessary corrective maintenance on SSCs is performed
in a timely manner, taking into account health, safety, national security and
protection of the environment.
* for monitoring and assessing the results of maintenance activities, including
monitoring, testing, assessing, inspecting and verifying the state or
performance of SSCs.
* for determining and documenting the root causes of malfunctions or failures
of SSCs and for taking appropriate corrective actions to prevent
recurrences.
* for maintaining, managing, reviewing and updating procedures and records,
and making them available to workers in accordance with the Act, the
regulations and the licence.
2.2.3 Preventive maintenance
The maintenance program for a NPP shall include preventive measures that:
* derive from the application of a single conceptual approach to preventive
maintenance, or the application of any combination of conceptual
approaches to preventive maintenance, including those termed "reliability
centred maintenance", "periodic maintenance" or "predictive maintenance".
* are designed to limit and manage the physical degradation of SSCs due to
ageing mechanisms.
* take into account:
- the relative importance of the SSCs to health, safety, national security and
protection of the environment,
- the recommendations of the manufacturers and the vendors of SSCs,
- requirements for the reliability of SSCs, and the results of reliability
analyses on SSCs,
- operating experience,
8
C-210 Maintenance Programs for Nuclear Power Plants
- design and operating conditions, and
- the exposure of maintenance workers to radiation.
* specify the required frequency of preventive maintenance activities and
provide for the scheduling of those activities.
* specify and provide for the implementation of a policy to deal with the
deferred or omission of any necessary preventive maintenance activities.
2.2.4 Monitoring and assessment
The maintenance program for a NPP shall include provisions for on-going monitoring,
testing, assessment, inspection or verification that:
* are designed to determine whether the associated SSCs are functioning
competently and reliably, in accordance with design predictions and
performance requirements.
* are designed to detect, in a timely manner, conditions in the state or
operation of the associated SSCs that could result in unreasonable risk to
the health and safety of persons, to national security or to the environment.
* include appropriate allowances to compensate for or to remedy any adverse
effects on the associated SSCs, of the program's monitoring, testing,
assessment and inspection activities.
* provide for an appropriate frequency of monitoring, testing, assessment and
inspection of the associated SSCs, such as an increased frequency of these
activities after large-scale replacement, repair or modification of SSCs, or
until the SSCs are shown to meet performance or quality standards.
* provide for in-service inspection of the associated SSCs by means of non-
destructive testing.
2.2.5 In-service inspections
The maintenance program for a nuclear power plant shall include in-service inspection
activities for the subject SSCs. These in-service inspections shall be designed and
implemented so as to determine whether the SSCs are fit for continued safe operation,
or whether corrective maintenance measures or other remedial measures are required.
The inspection of the SSCs shall take into account their relative importance to the
safety objectives of the nuclear power plant. Accordingly, the maintenance program at
a nuclear power plant shall:
* include the in-service inspection activities that are required by the applicable
codes and standards, including in-service inspections of
9
Maintenance Programs for Nuclear Power Plants C-210
- pressure-retaining components in the reactor primary heat transport
system,
- components that are (a) part of, or connected to, a reactor primary heat-
transport system or a reactor moderator system, and (b) necessary, under
normal or abnormal operating conditions, including design-basis accident
conditions, in order to safely shutdown the reactor or to effectively cool
the reactor fuel,
- components that are (a) part of, or connected to, a steam or feed-water
line, and (b) necessary to safely shutdown a reactor unit or to effectively
cool the reactor fuel, and
- SSCs that are necessary, under normal, abnormal or postulated accident
conditions, to safely contain or to safely release radioactive materials or
hazardous materials.
* include a schedule of in-service inspections of the associated SSCs that,
while taking into account the rates of deterioration of individual SSCs, is
based on assumptions that are sufficiently conservative so as to assure that
the rate of any deterioration of a SSC is revealed before such deterioration
can lead to a failure or malfunction of the SSC.
2.3 Requirements to keep and retain maintenance-related records
As noted in section 4.3.2 above, the regulations under the NSC Act impose obligations upon
licensees with respect to records that are to be submitted, kept, retained and made available
to workers. Some of these obligations pertain to, or may pertain to, nuclear power plants
and the associated maintenance programs.
Examples include paragraph 14(2)(a), paragraph 14(2), paragraph 14(2)(e), subsection
14(4), and subsection 14(5) of the Class I Nuclear Facilities Regulations. To meet the
requirements associated with these examples, every licensee who operates a NPP must keep
a record of (i) the relevant maintenance procedures, (ii) the results of the inspection and
maintenance programs referred to in the licence, and (iii) the status of each worker's
qualifications, re-qualification and training, including the results of all tests and
examinations completed in accordance with the licence. Under these regulations the licensee
must also retain the plant maintenance procedures and the results of the inspection and
maintenance program referred to in the licence for 10 years after the expiry date of the
licence to abandon the NPP. Similarly, the regulations require that the licensee retain the
record of the status of each worker's qualifications, re-qualification and training, including
the results of all tests and examinations completed in accordance with the licence, for the
period that the worker is employed by the licensee and for 5 years after the worker ceases
to be so employed.
Additionally, section 27 of the General Nuclear Safety and Control Regulations requires
10
C-210 Maintenance Programs for Nuclear Power Plants
"every licensee", including a licensee that operates a NPP, to keep a record of all
information relating to the licence that is submitted by the licensee to the CNSC. Since
section 27 does not specify a period for retaining the record that is required by the section,
the information required by section 27 must be retained for the period ending one year after
the expiry of the licence (Section 28 of the General Nuclear Safety and Control
Regulations), or as otherwise specified in an applicable regulation under the NSC Act.
11
DRAFT
REGULATORY
GUIDE
C-218 (E)
PREPARING CODES OF PRACTICE
TO CONTROL RADIATION DOSES
AT URANIUM MINES AND MILLS
Issued for public comments by the
Atomic Energy Control Board
November 1999
Atomic Energy Commission de contrôle
Control Board de l'énergie atomique
DRAFT REGULATORY GUIDE
Preparing Codes of Practice to Control Radiation
Doses at Uranium Mines and Mills
C-218 (E)
Issued for public comments by the
Atomic Energy Control Board
November 1999
AECB Regulatory Documents
The Atomic Energy Control Board (AECB) operates within a legal framework that includes law and
supporting regulatory documents. Law includes such legally enforceable instruments as acts,
regulations, licences and directives. Regulatory documents such as policies, standards, guides,
notices, procedures and information documents support and provide further information on these
legally enforceable instruments. Together, law and regulatory documents form the framework for the
regulatory activities of the AECB.
The main classes of regulatory documents developed by the AECB are:
* Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the AECB in its regulatory program.
* Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the AECB accepts as meeting the regulatory
requirements.
* Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the AECB recommends for meeting regulatory requirements or improving
administrative effectiveness.
* Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be acted
upon in a timely manner.
* Regulatory Procedure: a document that describes work processes that the AECB follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not create
legally enforceable requirements. They support regulatory requirements found in regulations, licences
and other legally enforceable instruments. However, where appropriate, a regulatory document may
be made into a legally enforceable requirement by incorporation in an AECB regulation, a licence or
other legally enforceable instrument made pursuant to the Atomic Energy Control Act.
C-218(E) Preparing Codes of Practice to Control Radiation
Doses at Uranium Mines and Mills
DRAFT REGULATORY GUIDE
Preparing Codes of Practice to Control Radiation Doses at
Uranium Mines and Mills
C-218 (E)
November 1999
NOTICE
On March 20, 1997, Bill C-23, the Nuclear Safety and Control Act (NSC Act, the Act), received Royal
Assent. New regulations that are derived from this Act will become law and replace the existing regulations.
Draft Regulatory Guide C-218 references the NSC Act and new regulations, which will come into force in
2000 on a date to be fixed by order of the Governor in Council.
About this Document
Comments
In order for interested persons to determine this document's impact and value, public comments
are being solicited. At the end of the comment period, all comments will be studied to determine
how best to improve the document. Unless otherwise requested, a copy of all comments received
will be placed in the AECB Library, in Ottawa. Comments on this guide will be most helpful if
received in writing by December 31, 1999. Reference our file number 1-8-8-218, and direct
enquiries and/or comments to the address below.
Document availability
The document can be viewed on the AECB internet website at www.aecb-ccea.gc.ca. A copy of
C-218 may be ordered in English or French from:
Operations Assistant
Corporate Documents Section
Atomic Energy Control Board
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone (613) 996-9505
Facsimile (613) 995-5086
E-mail via Internet: reg@atomcon.gc.ca
i
Preparing Codes of Practice to Control Radiation C-218(E)
Doses at Uranium Mines and Mills
Contents
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
1. Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory framework and relevant legislation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Definition of a code of practice . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.3 The licensing process and codes of practice. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2. Developing and Using Codes of Practice . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2.1 Codes of practice and radiation protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2.2 Codes of practice and action levels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.3 Responding when action levels are reached . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.3.1 Legislation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
2.3.2 Responses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
2.4 Notifications and reporting procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
3. Substantiating a Code of Practice . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
ii
C-218(E) Preparing Codes of Practice to Control Radiation
Doses at Uranium Mines and Mills
Purpose
This draft regulatory guide helps applicants for uranium mine or mill licences develop codes of
practice in accordance with the Uranium Mines and Mills Regulations for the purpose of
controlling radiation doses to workers. In addition, this guide helps the Canadian Nuclear Safety
Commission (CNSC) determine whether proposed codes of practice are adequate.
Scope
This guide applies to all applicants for a CNSC licence to prepare a site and construct, operate or
decommission a uranium mine or mill. In addition to discussing how codes of practice may be
developed and used, this document describes the types of information that the CNSC typically
needs in order to evaluate the adequacy of the proposed codes of practice.
iii
C-218(E) Preparing Codes of Practice to Control Radiation
Doses at Uranium Mines and Mills
1. Background
1.1 Regulatory framework and relevant legislation
The Atomic Energy Control Board (AECB) is the federal agency that regulates
nuclear facilities and materials to prevent undue risk to health, safety, security and the
environment.
At present, the AECB operates under the authority of the Atomic Energy Control
(AEC) Act and regulations. However, these laws are soon to be replaced by new
legislation, the Nuclear Safety and Control (NSC) Act and associated regulations.
Under the NSC Act, the AECB will become the Canadian Nuclear Safety Commission
(CNSC), with continuing responsibilities for regulation of the nuclear industry.
The NSC Act and regulations will prohibit persons or organizations from possessing,
using, disposing or abandoning certain nuclear materials without a licence from the
CNSC, unless they are exempted from such requirements. The regulations will also
stipulate prerequisites for licensing, including the information that is to be, or may be,
included in applications for specific types of licences.
In particular, section 4 of the Radiation Protection Regulations will require all CNSC
licensees to implement a radiation protection program, and to measure or estimate
radiation doses to their workers. Additionally, the Uranium Mines and Mills
Regulations will stipulate that applications for a licence to construct, operate or
decommission a uranium mine or mill must contain a code of practice.
1.2 Definition of a code of practice
The legislation that pertains to uranium mines and mills will not contain an explicit
definition of the term "code of practice." Instead, subsection 4(2) of the Uranium
Mines and Mills Regulations will present the defining features of a code of practice in
terms of its required content:
"An application for a licence in respect of a uranium mine or mill, other than a
licence to abandon, shall contain a code of practice that includes
(a) any action level that the applicant considers appropriate for the
purpose of this subsection;
1
Preparing Codes of Practice to Control Radiation C-218(E)
Doses at Uranium Mines and Mills
(b) a description of any action that the applicant will take if an action level
is reached; and
(c) the reporting procedures that will be followed if an action level is
reached."
To understand subsection 4(2), a reader must also refer to subsection 4(1) for the
supporting definition of an `action level':
"In this section, `action level' means a specific dose of radiation or other
parameter that, if reached, may indicate a loss of control of part of a
licensee's radiation protection program or environmental protection
program, and triggers a requirement for specific action to be taken."
Accordingly, a proposed code of practice will consist of any indicators (that the
licence applicant considers appropriate) of a loss of control of any part of a radiation
protection program, descriptions of any actions the applicant will take, and the
reporting procedures to be followed if an indicator is reached.
1.3 The licensing process and codes of practice
The CNSC's licensing process for uranium mines and mills will follow the stages laid
out in law. This process will begin with an initial assessment of the environmental
impacts of the proposed project and will proceed progressively through site
preparation and construction, operating, decommissioning, and abandonment phases.
At each licensing stage, the CNSC, upon receipt of an application that is complete
and in the prescribed form, will review it to determine whether the applicant is
competent and has made adequate provision for the protection of the environment,
the health and safety of persons, and the maintenance of national security and
measures required to implement international obligations to which Canada has agreed.
If satisfied, the CNSC may issue a licence that contains relevant conditions.
The NSC Act and regulations will require licence applicants to provide certain
information at each licensing stage. Accordingly, the information required by the
CNSC at each licence application stage will depend upon the legislated requirements
and case-specific circumstances.
At all licensing stages, licence applications may incorporate (directly or by reference)
new or previously submitted information, in accordance with legislated requirements
and the best judgement of the applicant. An application that is submitted at one
licensing stage can thus become a building block for the next stage.
Under subsection 24(5) of the NSC Act, the CNSC will be empowered to include in a
2
C-218(E) Preparing Codes of Practice to Control Radiation
Doses at Uranium Mines and Mills
licence any condition that it considers necessary for the purpose of the Act.
Accordingly, a CNSC licence may incorporate an applicant's commitments, a
condition that incorporates an action level, or any other conditions that the CNSC
considers necessary in the interests of health, safety, security and protection of the
environment.
Upon receiving a CNSC licence, a licence applicant will become the licensee, and will
be obliged to comply with the conditions of the licence in performing the authorized
activities. If a code of practice is incorporated into a uranium mine or mill licence, the
licensee must meet any corresponding obligations regarding action levels, responses,
and reporting.
2. Developing and Using Codes of Practice
2.1 Codes of practice and radiation protection
When incorporated in a CNSC licence, a code of practice will become a regulatory
and administrative tool to control radiation doses at the associated uranium mine or
mill within the context of the licensee's radiation protection strategy. Accordingly, the
code of practice and the objectives and broader content of the associated radiation
protection program should be consistent.
The fundamental elements of an acceptable radiation protection program are qualified
and committed personnel who are authorized and equipped to ensure radiation safety,
appropriate procedures, and adequate resources. Appropriate combinations of these
core elements - people, plant and procedures - can help ensure the safe conduct of
CNSC-licensed activities.
At uranium mines and mills, the "plant" component of radiation protection programs
typically includes engineered controls; the "procedures" component encompasses
administrative controls and operating or work procedures. A key failure or deficiency
in either component could result in a loss of control of some part of the associated
radiation protection program.
To aid regulatory review, applications for a uranium mine or mill licence should
clearly describe how the licence applicants propose to integrate their respective codes
of practice and the associated radiation protection programs.
3
Preparing Codes of Practice to Control Radiation C-218(E)
Doses at Uranium Mines and Mills
2.2 Codes of practice and action levels
When developing action levels for proposed codes of practice, applicants for uranium
mine and mill licences should consult the AECB draft Regulatory Guide, C-228.
C-228 describes how action levels for nuclear facilities and activities may be
developed and used.
As noted in C-228, action levels for nuclear facilities should be expressed in terms of
parameters that are useful for controlling the associated radiation protection
programs.
Accordingly, action levels contained in codes of practice for uranium mines and mills
should be expressed in terms of relevant parameters, such as:
* total effective dose
* gamma radiation dose-rate
* exposure to radon progeny
* exposure to radon gas
* exposure to long-lived radioactive dust
* concentrations of uranium in urine
2.3 Responding when action levels are reached
2.3.1 Legislation
The Radiation Protection Regulations and the Uranium Mines and Mills
Regulations impose specific obligations on applicants for a uranium mine or
mill licence, and on uranium mine and mill licensees, when an action level is
reached. These obligations pertain to investigation, response and corrective
actions, and reporting and notification.
The Uranium Mines and Mills Regulations stipulate that the proposed code
of practice for a uranium mine or mill must describe:
* any action that the applicant will take if an action level is reached
* the reporting procedures that will be followed if an action level is
reached
4
C-218(E) Preparing Codes of Practice to Control Radiation
Doses at Uranium Mines and Mills
Additionally, the Radiation Protection Regulations require CNSC licensees
to:
* conduct an investigation when an approved action level is reached in
order to establish the cause for reaching the action level
* identify and take action to restore the effectiveness of the radiation
protection program
* notify the CNSC within the period specified in the licence
2.3.2 Responses
By definition, an action level, if reached, may indicate a loss of control of part
of a licensee's radiation protection program, and triggers a requirement for
specific action to be taken. Accordingly, a monitoring result that approaches,
equals or exceeds an action level, is often the first indication that certain
responses are or may be required.
When an action level in a uranium mine or mill code of practice is reached, the
licensee should first investigate to determine the cause. This investigation, and
the protective measures specified in the code of practice, should be
implemented as soon as the licensee becomes aware that an action level has
been reached.
If an investigation confirms that a loss of control of any part of the associated
radiation protection program has occurred, the licensee must identify and take
corrective actions to restore the effectiveness of the program. The substance,
severity and immediacy of this response will depend upon the specific
circumstances, such as the cause and nature of any loss of control, the actual
and potential consequences of any loss of control, and the licensee's
preferences. If the effectiveness of the radiation protection program cannot be
restored forthwith, the responsible licensee should propose appropriate interim
measures for CNSC consideration.
Typically, the greater the actual or potential radiation hazards when an action
level is reached, the more immediate, complex or rigorous the corresponding
response should be.
Accordingly, when an action level is reached, the appropriate response could
include:
* issuing direct reading dosimeters or other instruments
* posting radiation warning signs
5
Preparing Codes of Practice to Control Radiation C-218(E)
Doses at Uranium Mines and Mills
* requiring the use of protective equipment
* restricting access
* suspending all or some operations
* restricting work to essential activity
* implementing other measures
2.4 Notifications and reporting procedures
Paragraph 6(2)(c) of the proposed Radiation Protection Regulations requires that
licensees notify the CNSC within the periods specified in their licences if they become
aware that an action level has been reached.
Accordingly, when an applicant for a uranium mine or mill licence submits a proposed
code of practice that includes an action level, the applicant should also stipulate the
reporting procedures to be followed if the action level is reached.
To provide for the necessary investigation, fact-finding and follow-up required by
regulations when an action level in a code of practice is reached, the associated
reporting procedures should include appropriate protocols for notifying the
employees responsible for conducting investigations, implementing findings, and
notifying the CNSC and others. These protocols should specify who is to be notified
when an action level is reached, and how they are to be notified. The proposed
notification procedures should indicate whether notifications are to be oral or written,
and how they are to be delivered.
The urgency, frequency and level of reporting required in a proposed code of practice
should be commensurate with the anticipated consequences of reaching the associated
action level. For example, in many situations, the proposed notification of the CNSC
may be limited to a follow-up report (within a specified time frame) that clearly,
completely and accurately documents the event, the results of the investigation carried
out, and the corrective actions taken.
Situations involving greater exposures to radiation or radiation hazards may warrant
more immediate notifications of the CNSC. Where the licence applicant anticipates
circumstances that could warrant a more prompt notification of the CNSC, the
precipitating circumstances and the proposed follow-up should be clearly described in
the proposed reporting procedures.
3. Substantiating a Code of Practice
6
C-218(E) Preparing Codes of Practice to Control Radiation
Doses at Uranium Mines and Mills
Before issuing a licence under subsection 24(4) of the Nuclear Safety and Control Act, the
CNSC must be satisfied that the applicant is competent and has made adequate provision
for the protection of the environment, the health and safety of persons, and the maintenance
of national security and measures required to implement international obligations to which
Canada has agreed. The CNSC will arrive at these judgements by assessing the information
provided by the licence applicant.
When evaluating the adequacy of a proposed code of practice, the CNSC may take into
account other pertinent information provided in support of the associated licence
application. This information could include:
* a complete description of the facility, areas and activities to which the proposed code
of practice applies
* a description of the expected sources of radiation, and their anticipated
characteristics (type, magnitude, variability)
* an assessment of the total effective radiation dose that will be received by workers
during normal operations, and the related contributions from individual source terms,
including radon progeny, long-lived radioactive dust, and gamma radiation
* a description of the proposed engineering and administrative measures to monitor
and control radiation doses during the proposed operations
* a description of any areas of the proposed uranium mine or mill that are to be subject
to radiation control measures other than codes of practice, and a description of the
associated dose-control measures (such as work permits, bioassays and dosimetry)
* an evaluation of the radiological consequences, in terms of radiation levels, of any
failure of the proposed radiation control measures
* an explanation of how any proposed action level has been derived
* if no action level is proposed, an explanation as to why no action level is deemed
appropriate
* an assessment of the consequences of reaching or exceeding any proposed action
level
* an evaluation of the effectiveness of the measures to be taken if any proposed action
level is reached or exceeded
* a description of how the proposed code of practice integrates with the radiation
protection program.
7
DRAFT
REGULATORY
GUIDE
A Guide to Ventilation
Requirements for
Uranium Mines
and Mills
C-221
Issued for public comments by the
Canadian Nuclear Safety Commission
April 2001
DRAFT REGULATORY GUIDE
A Guide to Ventilation Requirements
for Uranium Mines and Mills
C-221
Issued for public comments by the
Canadian Nuclear Safety Commission
April 2001
Regulatory Documents
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in regulatory programs.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
A Guide to Ventilation Requirements
for Uranium Mines and Mills
C-221
April 2001
About this Document
The purpose of C-221 is to help persons address the requirements for the submission of
ventilation-related information when applying for a licence to site and construct, operate or
decommission a uranium mine or mill. This guide is also intended to help applicants for a uranium
mine or mill licence understand their operational and maintenance obligations with respect to
ventilation systems, and to help CNSC staff evaluate the adequacy of applications for uranium
mine and mill licences.
Comments
The CNSC invites interested persons to assist in the further development of this draft regulatory
document by commenting in writing on the document's content and potential usefulness. Please
respond by June 29, 2001. Direct your comments to the postal or e-mail address below,
referencing file 1-8-8-221.
The CNSC will take the comments received on this draft into account when developing it further.
These comments will be subject to the provisions of the federal Access to Information Act.
Document availability
This document can be viewed on the CNSC website www.nuclearsafety.gc.ca. To order a printed
copy of the document in English or French, please contact:
Operations Assistant
Corporate Documents Section
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 996-9505
Facsimile: (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
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C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
Contents
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory framework . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Regulatory and licensing process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2 Overview of Ventilation Requirements in Legislation . . . . . . . . . . . . . . . . . . . . . . . . 2
3 Information Requirements at Prescribed Licensing Stages . . . . . . . . . . . . . . . . . . . . 3
3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.2 All classes of licences except a licence to abandon . . . . . . . . . . . . . . . . . . . . . . . . 4
3.2.1 Summary of information requirements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.2.2 Discussion of information requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.3 Licence to prepare a site for and construct . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.3.1 Summary of information requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.3.2 Discussion of information requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
3.4 Licence to operate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
3.4.1 Summary of information requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
3.4.2 Discussion of information requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
3.5 Licence to decommission . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
3.5.1 Summary of information requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
3.5.2 Discussion of information requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
4 Operational and Maintenance Ventilation-related Requirements. . . . . . . . . . . . . . 10
4.1 Requirement to post codes of practice . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
4.2 Requirement for written procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
4.3 Requirement for workers to be trained . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
4.4 Requirement to audit compliance with operating procedures . . . . . . . . . . . . . 10
4.5 Design, designation and prevention measures. . . . . . . . . . . . . . . . . . . . . . . . . . 10
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4.6 Requirement for operating and maintenance contingencies . . . . . . . . . . . . . . 11
4.7 Requirement to keep records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
4.8 Requirement to make records available . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
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C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
Purpose
The purposes of C-221 is to help persons address the requirements for the submission of
ventilation-related information when applying for a Canadian Nuclear Safety Commission (CNSC)
licence to site and construct, operate or decommission a uranium mine or mill.
This guide is also intended to help applicants for a uranium mine or mill licence understand their
operational and maintenance obligations with respect to ventilation systems, and to help CNSC
staff evaluate the adequacy of applications for uranium mine and mill licences.
Scope
This guide is relevant to any application for a CNSC licence to prepare a site for and construct,
operate or decommission a uranium mine or mill. In addition to summarizing the ventilation-
related obligations of uranium mine and mill licensees, the guide describes and discusses the
ventilation-related information that licence applicants should typically submit to meet regulatory
requirements.
The guide pertains to any ventilation of uranium mines and mills for the purpose of assuring the
radiation safety of workers and on-site personnel. This ventilation may be associated with any
underground or surface area or premise that is licensable by the CNSC as part of a uranium mine
or mill. These areas and premises typically include mine workings, mill buildings, and other areas
or premises involving or potentially affected by radiation or radioactive materials. Some examples
of the latter include offices, effluent treatment plants, cafeterias, lunch rooms and personnel
change-rooms.
1 Background
1.1 Regulatory framework
The Canadian Nuclear Safety Commission (CNSC) is the federal agency that regulates
nuclear energy and materials to protect health, safety, security and the environment and to
respect Canada's international commitments on the peaceful use of nuclear energy.
The Nuclear Safety and Control Act ("the Act") requires persons or organizations to be
licensed by the CNSC for carrying out the activities referred to in Section 26 of the Act,
unless otherwise exempted. The associated regulations stipulate prerequisites for CNSC
licensing, and the obligations of licensees and workers.
1.2 Regulatory and licensing process
The Act obliges the CNSC to determine, before granting or refusing to grant a licence,
whether the applicant for the licence is qualified and has made adequate provision for the
health and safety of persons, national security and protection of the environment. To make
these determinations, the CNSC needs credible and relevant information from applicants.
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Upon receipt of a licence application, or formal notice of intent to file an application
containing an adequate description of the project, the CNSC determines whether the
application involves a project that requires an environmental assessment pursuant to the
Canadian Environmental Assessment Act (CEA Act) and its regulations. If an
environmental assessment is required under the CEA Act, the CNSC may not exercise any
authority that would permit the project to be carried out in whole or part until the
environmental assessment process is complete. When CEA Act legislation does not apply to
the project, the CNSC may proceed with routine processing of the associated licence
application.
The CNSC's licensing process for uranium mines and mills follows the stages laid out in the
Uranium Mines and Mills Regulations, proceeding progressively through site preparation
and construction, operating, decommissioning and abandonment phases. At each licensing
stage, the CNSC determines whether the licence applicant is qualified and has made
adequate provision for the protection of the environment, the health and safety of persons,
and the maintenance of national security and measures required to implement international
obligations to which Canada has agreed. If satisfied, the CNSC may issue a licence that
contains appropriate conditions.
Typically, a CNSC licence incorporates the applicant's commitments and any other
conditions that the CNSC considers necessary in the interests of health and safety of
persons, national security and protection of the environment.
The information required by the CNSC at each licence application stage is influenced by
case-specific circumstances. Typically, the information supplied at one stage serves as a
building block for the next. An application for a CNSC licence may include new
information, or in accordance with section 7 of the General Nuclear Safety and Control
Regulations, it may incorporate by reference any information that is contained in another
licence issued by the CNSC.
2 Overview of Ventilation Requirements in Legislation
The Uranium Mines and Mills Regulations contain both direct and indirect references to the
ventilation of uranium mines or mills. For example, section 3 of the Regulations is a
comprehensive summary of the general types of information to be included in an application for a
uranium mine or mill licence, except a licence to abandon. The section includes requirements that
pertain exclusively to the ventilation of uranium mines or mills, as well as others that encompass,
but are not limited to, ventilation-related matters.
Sections 4, 5, 6 and 7 of the Regulations also address a range of topics that encompass or
concern, but are not limited to, ventilation systems and related matters. These subjects include:
* codes of practice;
* the design of the mine or mill;
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C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
* the results of commissioning work;
* the design of equipment, systems and components;
* quality assurance;
* commissioning plans; and
* operating, maintenance and decommissioning policies, methods and procedures.
Section 10 of the Regulations pertains to both ventilation-related and non-ventilation-related
matters. It requires uranium mine and mill licensees:
* to establish, implement and maintain written operating procedures for their licensed
activities;
* to train their workers to perform work in accordance with operating procedures; and
* to audit their workers for the purpose of verifying compliance with operating
procedures.
Sections 11 and 12 specify the actions to be taken by CNSC licensees with respect to the
operation or malfunction of ventilation systems. Section 16 of the Uranium Mines and Mills
Regulations requires licensees to keep certain records, including those for ventilation systems and
activities, and to make these records available to workers and workers' representatives.
The documents listed in the References section of this guide provide information on developing
and using action levels and codes of practice to control radiation doses to uranium mine and mill
workers. This guidance may be relevant to the operation of ventilation systems, in order to keep
radiation doses to workers and the public as low as reasonably achievable (ALARA), social and
economic factors being taken into account.
3 Information Requirements at Prescribed Licensing Stages
3.1 Introduction
The ventilation-related requirements discussed in this guide appear in the Uranium Mines
and Mills Regulations because radiation safety in uranium mines and mills depends in part
on the provision of adequate ventilation in the workplace. Historically, uranium mines and
mills have used active or passive ventilation measures to limit concentrations of airborne
radioactivity in workplaces. When properly designed, constructed, monitored and
maintained, such systems have proven to be both practical and effective in reducing
radiation hazards.
Under the Uranium Mines and Mills Regulations, applicants for all classes of uranium mine
and mill licences, except a licence to abandon, will be responsible for submitting prescribed
information pertaining to any proposed ventilation systems or activities. These information
requirements will vary with the licensing stage. For example, although an application for a
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A Guide to Ventilation for Uranium Mines and Mills C-221
uranium mine or mill operating licence should typically include a description of the finalized
proposal (i.e., the "policies, methods and procedures") for operating and maintaining any
proposed ventilation system and "as-built" design details, more preliminary information on
operating and maintenance policies, methods and procedures might suffice at the siting and
construction stage. Conversely, those proponents who plan to use unproven or non-
conventional technologies or methods might need to provide more rigorous substantiations
at an earlier licensing stage than might be required of the proponents of more conventional
(i.e., proven) approaches.
At all stages of the uranium mines and mills licensing process, the ventilation-related
measures and activities proposed by individual mines or mills under the NSC Act and
regulations will depend in part upon unique combinations of legislated requirements and
case-specific factors. These factors will reflect the options open to the applicants or
proponents and their respective preferences, and include site, environmental and
technological constraints, such as ore-body characteristics, mining and processing
technologies, facility designs and operating methods.
In some situations, proponents of uranium mine or mill projects may need to submit detailed
information on their proposed undertakings for reasons that are not directly linked to CNSC
licensing stages or requirements. For example, to obtain the necessary approvals for a
proposed project, the proponent may need to address recommendations resulting from
environmental assessments or hearings, or conditions set by government or other
authorities.
At all licensing stages, the CNSC will review any proposed ventilation systems or activities
against regulatory requirements, and will take into account relevant information that
pertains both directly or indirectly to the systems or activities.
The Act and regulations do not prescribe the form of applications for uranium mine or mill
licences, only the type of information that is to be included. However, to aid regulatory
review the information contained in licence applications should be organized and presented
clearly and logically.
The following sections summarize and discuss information requirements for uranium mine
and mill ventilation systems.
3.2 All classes of licences except a licence to abandon
3.2.1 Summary of information requirements
Subparagraphs 3(d)(vii) and 3(d)(viii) of the Uranium Mines and Mills Regulations
require an application for any CNSC licence in respect of a uranium mine or mill,
other than a licence to abandon, to contain descriptions of:
* the proposed ventilation and dust control methods;
* the proposed equipment for controlling air quality; and
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C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
* the proposed level of effectiveness of and inspection schedule for the
ventilation and dust control systems.
3.2.2 Discussion of information requirements
The description of the proposed level of effectiveness of the ventilation and dust
control systems should explain how the system will be or has been optimized in
accordance with the ALARA (as low as reasonably achievable) principle of dose
limitation.
If, in an applicant's opinion, ventilation systems, either active or passive, are not
necessary to meet regulatory requirements, this conclusion should be stated and
substantiated in the licence application.
If alternative radiation protection measures are to be substituted for engineered
ventilation measures, these measures should be described and justified within the
context of the applicant's radiation protection program.
The licence application should include all documentation needed to objectively verify
the appropriateness of proposed designs and the validity of performance predictions.
The necessary documentation could include descriptions of supporting assumptions,
criteria, calculations, research , modelling results, drawings, plans or diagrams.
3.3 Licence to prepare a site for and construct
3.3.1 Summary of information requirements
In addition to the information required by subparagraphs 3(d)(vii) and 3(d)(viii) of the
Uranium Mines and Mills Regulations, an application for a licence to prepare a site
for and construct a uranium mine or mill shall contain the following information, as it
pertains to the ventilation of the mine or mill:
* a proposed code of practice that includes:
(a) any action level that the applicant considers necessary for purposes of
subsection 4(2) of the Uranium Mines and Mills Regulations,
(b) a description of any action that the applicant will take if an action level
is reached, and
(c) the reporting procedures that will be followed if an action level is
reached (subsection 4(2));
* a description of the proposed design of the mine or mill (paragraphs 5(1)(a),
5(2)(a));
* a description of the components, systems and equipment proposed to be
installed at the mine or mill, including their design operating conditions
(paragraphs 5(1)(c), 5(2)(c));
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* the proposed quality assurance program for the mine or mill (paragraphs
5(1)(d), 5(2)(d));
* the results of a process-hazard analysis and a description of how those
results have been taken into account (paragraphs 5(1)(e), 5(2)(e));
* the proposed commissioning plan for the ventilation components, systems,
and equipment to be installed at the mine or mill (paragraphs 5(1)(i),
5(2)(i)).
3.4.2 Discussion of information requirements
The information that is submitted to meet the information requirements listed in
section 3.3.1 above for a licence to prepare a site for and construct a uranium mine or
mill should typically include such supporting details as:
* a description of any alarm system or component, including a main fan
warning device, that will be installed to ensure that the ventilation system
operates safely (subsection 11(a)).
* the dimensions, location and layout of ventilation ducts;
* the location, type and use of all ventilation system controls and regulators;
* the design, location and operation of any equipment or devices to measure
air quality or air quantity;
* the location of system air intakes and exhausts;
* a description of any design provisions to ensure effective separation of
primary air intakes and exhausts;
* a description of any proposed auxiliary ventilation systems;
* the preliminary operating and maintenance procedures for the ventilation
system (paragraphs 10(a), 16(a));
* the preliminary programs for monitoring air quality and quantity;
* a description of the quantity and quality of air that is to be supplied to each
workplace area;
* a description of the expected rate of air exchange in the workplace after
installation of any proposed ventilation systems;
* a description of the expected air quality in the workplace after installation of
any proposed ventilation systems;
* a description of any administrative provisions to ensure effective operation
of the ventilation system;
* proposed operating parameters for winter and summer; and
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C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
* any measures to control the movement of radiation from unventilated to
ventilated areas of underground mines.
3.5 Licence to operate
3.5.1 Summary of information requirements
In addition to the information required by subparagraphs 3(d)(vii) and 3(d)(viii) of the
Uranium Mines and Mills Regulations, an application for a licence to operate a
uranium mine or mill shall contain the following information, as it pertains to the
ventilation of the mine or mill:
* a proposed code of practice that includes:
(a) any action level that the applicant considers necessary for purposes of
subsection 4(2) of the Uranium Mines and Mills Regulations,
(b) a description of any action that the applicant will take if an action level
is reached, and
(c) the reporting procedures that will be followed if an action level is
reached (subsection 4(2));
* the proposed policies, methods and procedures for operating and
maintaining the ventilation systems (paragraphs 6(1)(c), 6(2)(c));
* a description of the structures, components, systems and equipment that
have been constructed or installed at the mine or mill, and their design
operating conditions as a result of commissioning (paragraphs 6(1)(b),
6(2)(b));
* the results of any commissioning work (paragraph 6(1)(a));
3.5.2 Discussion of information requirements
To meet the above requirements as they pertain to ventilation systems or related
measures, an application for a licence to operate a uranium mine or mill should
include or incorporate the relevant information, whether new or previously submitted.
Accordingly, an application to the CNSC for a uranium mine or mill operating licence
should demonstrate that any engineered ventilation system will be operated,
monitored and maintained in accordance with regulatory requirements. Typically, the
application should:
* describe the measures to ensure that a person is designated to receive and
respond to a warning signal provided by a main-fan warning device
(subsection 11(b));
* describe the measures that are to be implemented to prevent any person or
activity from interfering with the proper operation of the ventilation system
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(subsection 11(c);
* describe the measures that the applicant proposes to implement to protect
the health and safety of workers if the ventilation system in the licensed
workplace fails to function in accordance with the licence (paragraph
12(1)(a));
* describe the measures that the licence applicant proposes to implement to
ensure, in the event of failure of a ventilation system, that only the work that
is immediately necessary to restore the system is performed in the affected
workplace (paragraph 12(1)(b));
* describe how the licensee will inform a worker of the protective measures
that have been taken and are to be taken in connection with any work
necessary to restore a ventilation system (subsection 12(2));
* describe how the ventilation system has been constructed to meet any
relevant conditions of the associated CNSC licence to prepare a site for and
construct the facility;
* describe the results from the monitoring of the performance of the
ventilation system during commissioning;
* describe any planned changes with respect to the design, operation,
monitoring, maintenance or performance of the engineered ventilation
system;
* describe any proposed code of practice with respect to the ventilation
system; and
* describe the finalized policies, methods and procedures for operating,
maintaining and controlling the ventilation system.
3.7 Licence to decommission
3.7.1 Summary of information requirements
In addition to the information required by subparagraphs 3(d)(vii) and 3(d)(viii) of the
Uranium Mines and Mills Regulations, the regulations stipulate that an application
for a licence to decommission a uranium mine or mill shall contain the following
information as it pertains to the ventilation of the mine or mill.
* a proposed code of practice that includes:
(a) any action level that the applicant considers necessary for purposes of
subsection 4(2) of the Uranium Mines and Mills Regulations,
(b) a description of any action that the applicant will take if an action level
is reached, and
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C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
(c) the reporting procedures that will be followed if an action level is
reached (subsection 4(2));
* the proposed schedule for the decommissioning work, including the
proposed starting date and the expected completion date of the
decommissioning work and the rationale for the schedule (subsection 7(a));
* a description of the land, buildings, structures, components, systems,
equipment, nuclear substances and hazardous substances that will be
affected by the decommissioning (subsection 7(b));
* the proposed measures, methods and procedures for carrying on the
decommissioning (subsection 7(c)); and
* a description of the planned state of the site upon completion of the
decommissioning work (subsection 7(d)).
3.8.2 Discussion of information requirements
The information submitted in support of an application for a licence to decommission
should address the requirements listed in 3.5.1 above to a level of detail and accuracy
that demonstrates that the applicant is qualified and has made adequate allowance for
the health and safety of persons, national security and protection of the environment
during decommissioning.
One example of making adequate allowance might be to provide enhanced ventilation
during some or all decommissioning activities. This could involve the continued use,
with or without modifications, of a ventilation system that was used during the
operating phase, or the use of new measures. Accordingly, a licence applicant's
preferred measures will typically be influenced by case-specific circumstances, such as
whether previously installed ventilation systems have become radioactively
contaminated during use, or whether the proposed ventilation systems are likely to
become similarly contaminated upon use. To expedite regulatory review and
licensing, the application for a decommissioning licence should address any such
possibilities.
The CNSC may need to know the details of an applicant's proposed decommissioning
plans in order to evaluate the adequacy of any proposed use or decommissioning of
ventilation systems and equipment.
To help control radiation doses to workers and the public during decommissioning
activities, action levels that involve ventilation activities or results can be incorporated
into codes of practice at uranium mines or mill.
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4 Operational and Maintenance Ventilation-related
Requirements
4.1 Requirement to post codes of practice
If a code of practice is referred to in a uranium mine or mill licence, the licensee is required
by section 9 of the Uranium Mines and Mills Regulations to post a copy of the code of
practice at a location within the uranium mine or mill that is accessible to all workers and
where it is most likely to come to their attention.
4.2 Requirement for written procedures
Paragraph 10(a) of the Uranium Mines and Mills Regulations obliges every CNSC licensee
to establish, implement and maintain written procedures for the conduct of licensed
activities. Accordingly, uranium mine and mill licensees that use ventilation systems to help
protect their workers and the public must establish, implement and maintain written
procedures to ensure that these systems operate effectively. These operating procedures
should include provisions, such as inspection, surveillance or sampling programs, for
purposes of evaluating, controlling and demonstrating the effectiveness of the associated
systems.
4.3 Requirement for workers to be trained
Paragraph 10(b) of the Uranium Mines and Mills Regulations further obliges licensees to
train their workers to perform work in accordance with operating procedures. To meet this
obligation as it relates to a ventilation system at a uranium mine or mill, the licensee must
ensure that the workers who are responsible for following the ventilation-related operating
procedures receive training to perform their work.
4.4 Requirement to audit compliance with operating procedures
Paragraph 10(c) of the Uranium Mines and Mills Regulations also obliges licensees to
audit their workers for the purpose of verifying compliance with the relevant operating
procedures for the conduct of licensed activities, including those for ventilation systems at
uranium mines and mills.
4.5 Design, designation and prevention measures
Section 11 of the Uranium Mines and Mills Regulations requires every licensee to:
* ensure that each main fan of the ventilation systems established in accordance with
the licence is equipped with a device that provides a warning signal when the main
fan is not functioning properly (paragraph 11(a));
* ensure that a person is designated to receive and respond to any warning signal
provided by the main fan warning device (paragraph 11(b)); and
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C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
* implement measures to prevent any person or activity from interfering with the
proper operation of the ventilation systems (paragraph 11(c)).
4.6 Requirement for operating and maintenance contingencies
Where a ventilation system in a uranium mine or mill workplace is not functioning in
accordance with the licence, section 12 of the Uranium Mines and Mills Regulations
requires the licensee to:
* implement alternative measure to protect the health and safety of the workers
(paragraph 12(1)(a)); and
* ensure that only the work necessary to restore the ventilation system is performed
in the workplace (paragraph 12(1)(b)).
Subsection 12(2) of the Uranium Mines and Mills Regulations stipulates that before a
worker performs any work that is necessary to restore a ventilation system at a uranium
mine or mill, the uranium mine or mill licensee shall inform the worker of the protective
measures that have been taken, and are to be taken in connection with the work.
4.7 Requirement to keep records
Section 16 of the Uranium Mines and Mills Regulations requires every uranium mine or
mill licensee to keep the following records that pertain to, or could pertain to, mine or mill
ventilation systems:
* operating and maintenance procedures (paragraph 16(1)(a));
* the design of the components and systems installed at the mine or mill
(paragraph 16(1)(e));
* the method and relevant data used to ascertain the doses of radiation received by
workers at the uranium mine or mill and the intake of radioactive nuclear
substances by those workers (paragraph 16(1)(f));
* any measurement made in accordance with the licence or the regulations made
under the Act (paragraph 16(1)(g));
* the inspections and maintenance carried out in accordance with the licence or the
regulations made under the NSC Act (paragraph 16(1)(h));
* the quantity of air delivered by each main fan identified in the licence
(paragraph 16(1)(i));
* the performance of each dust control system (paragraph 16(1)(j)); and
* the training received by each worker (paragraph 16(1)(k)).
4.8 Requirement to make records available
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A Guide to Ventilation for Uranium Mines and Mills C-221
Section 16 of the Uranium Mines and Mills Regulations also requires every uranium mine
or mill licensee to:
* make the prescribed records available at the uranium mine or mill to the workers
or workers' representative (subsection 16(2));
* retain a record of the training received by workers employed at the uranium mine
and mill (subsection 16(3)); and
* post at a location within the uranium mine or mill that is accessible to all workers,
and where it is most likely to come to their attention, a record of the
measurements made in respect of every workplace in accordance with the licence
and the Uranium Mines and Mills Regulations (subsection 16(4)).
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C-221 A Guide to Ventilation Requirements for Uranium Mines and Mills
References
1. "Developing and Using Action Levels," Regulatory Guide G-228, Atomic Energy Control
Board (March 2001).
2. "Preparing Codes of Practice to Control Radiation Doses at Uranium Mines and Mills," Draft
Regulatory Guide C-218, Atomic Energy Control Board (November 1999).
13
DRAFT
REGULATORY
GUIDE
C-273 (E)
MAKING, REVIEWING AND
RECEIVING ORDERS UNDER
THE NUCLEAR SAFETY
AND CONTROL ACT
Issued for public consultation and trial use by the
Atomic Energy Control Board
May 2000
Atomic Energy Commission de contrôle
Control Board de l'énergie atomique
DRAFT REGULATORY GUIDE
Making, Reviewing and Receiving Orders
Under the Nuclear Safety and Control Act
C-273 (E)
Issued for public consultation and trial use by the
Atomic Energy Control Board
May 2000
Regulatory Documents
The Atomic Energy Control Board (AECB) operates within a legal framework that includes law
and supporting regulatory documents. Law includes such legally enforceable instruments as acts,
regulations, licences and directives. Regulatory documents such as policies, standards, guides,
notices, procedures and information documents support and provide further information on these
legally enforceable instruments. Together, law and regulatory documents form the framework for
the regulatory activities of the AECB.
The main classes of regulatory documents developed by the AECB are:
* Regulatory Policy: a document that describes the philosophy, principles and
fundamental factors used by the AECB in its regulatory program.
* Regulatory Standard: a document that is suitable for use in compliance assessment
and describes rules, characteristics or practices which the AECB accepts as meeting
the regulatory requirements.
* Regulatory Guide: a document that provides guidance or describes characteristics
or practices that the AECB recommends for meeting regulatory requirements or
improving administrative effectiveness.
* Regulatory Notice: a document that provides case-specific guidance or information
to alert licensees and others about significant health, safety or compliance issues that
should be acted upon in a timely manner.
* Regulatory Procedure: a document that describes work processes that the AECB
follows to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in an
AECB regulation, a licence or other legally enforceable instrument made pursuant to the Atomic
Energy Control Act.
DRAFT REGULATORY GUIDE
Making, Reviewing and Receiving Orders
Under the Nuclear Safety and Control Act
C-273 (E)
May 2000
NOTICE
On March 20, 1999, Bill C-23, the Nuclear Safety and Control Act (NSC Act, the Act), received Royal
Assent. New regulations that are derived from this Act will become law and replace the existing regulations.
Regulatory Guide C-273 references the NSC Act and new regulations, which will come into force in 2000
on a date to be fixed by order of the Governor in Council.
About this Document
Comments
In order for interested persons to determine this document's impact and value through working
experience, public comments are being solicited over a trial-use period. At the end of this
12-month trial period, comments will be studied to determine how best to improve the document.
Unless otherwise requested, a copy of all comments received will be placed in the AECB Library,
in Ottawa.
Comments on this guide will be most helpful if received in writing during the trial period, with
final comments accepted by May 3, 2001. Reference our file number 1-8-8-273, and direct
enquiries and/or comments to the address below.
Document availability
The document can be viewed on the AECB internet website at www.aecb-ccea.gc.ca. Copies of
C-273 may be ordered in English or French using the contact information below.
Operations Assistant
Corporate Documents Section
Atomic Energy Control Board
P.O. Box 1046, Station B, 280 Slater Street
Ottawa, Ontario K1P 5S9 CANADA
Telephone (613) 996-9505 Facsimile (613) 996-5086
E-mail via Internet: reg@atomcon.gc.ca
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C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
Contents
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
Purpose and Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Pertinent sections of the NSC Act . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2 Making Orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.1 Roles and responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.1.1 Inspectors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.1.2 Designated officers. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.2 Requirements for making orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2.2.1 Order made by an inspector . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
2.2.2 Order made by a DO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3 Reviewing Orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.1 Roles and responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.1.1 Commission: ways of reviewing an order . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.1.2 Commission responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.1.3 Designated officers. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.2 Procedural requirements for reviewing orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.2.1 Review by the Commission . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
3.2.2 Commission responsibilities: appeal or redetermination . . . . . . . . . . . . . . . . . . . 6
3.2.3 Review by a designated officer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
4 Receiving Orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
4.1 Rights and responsibilities. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
4.1.1 Licensees or other persons to whom an order is directed have the right to: . . . . . 8
4.1.2 Persons named in, or subject to, or those directly affected by an order have the
right to: . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
4.1.3 Persons named in an order are responsible for: . . . . . . . . . . . . . . . . . . . . . . . . . . 8
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Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
4.2 Procedural requirements for receiving orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
4.2.1 Upon receipt of an order . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
4.2.2 Upon an appeal or redetermination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
Appendix A: Excerpts from the CNSC Rules of Procedure . . . . . . . . . . . . . . . 12
Making of Orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
Appeal or Redetermination of an Order . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
Decision . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
Appendix B: Form for Making Orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
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C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
Abbreviations and Acronyms
AECB Atomic Energy Control Board
CNSC Canadian Nuclear Safety Commission, the Commission
DO Designated Officer
NSC Act Nuclear Safety and Control Act, the Act
Rules CNSC Rules of Procedure
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Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
Purpose and Scope
The purpose of this document is to describe the roles and responsibilities of inspectors and
designated officers (DOs) charged with making and reviewing orders under the NSC Act, and the
steps required to carry out these activities. It also advises persons subject to orders about their
rights and responsibilities, and the steps required to respond to an order.
This document does not repeat all relevant subsections and paragraphs of the NSC Act and the
CNSC Rules of Procedure (the Rules); rather, it summarizes and interprets them.
1. Introduction
1.1 Background
An order is one of the tools used by the Canadian Nuclear Safety Commission (CNSC, the
Commission) in carrying out its responsibilities under the Nuclear Safety and Control Act
(NSC Act). It is a powerful legal instrument used to compel someone to do something in the
interests of health, safety, the environment, national security or compliance with Canada's
international obligations. An order must be obeyed by the recipient; failure to comply can
lead to further regulatory measures, including prosecution or licensing actions. The NSC Act
describes the circumstances under which an order can be given and sets out a procedure for
the review and appeal of an order.
The proposed CNSC Compliance Policy includes making orders as a CNSC enforcement
action aimed at securing compliance with CNSC requirements by licensees and regulated
persons. The policy and its requirements will apply to officers and employees of the CNSC
who are involved in developing and carrying out compliance activities.
1.2 Pertinent sections of the NSC Act
The following sections, subsections and paragraphs of the NSC Act pertain to making,
reviewing, and receiving orders:
* Power to Make Orders: subsections 35(1) and (2); paragraph 37(2)(f)
* Duty to Assist and Comply: sections 36, 41, 42
* Review of Orders: subsections 35(3) and 37(6)
* Relevant Procedures: sections 38-40
* Appeals and Redeterminations: section 43
* Offences: subsections 48(d), (e) and (f); and (5l(1) and (3)
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C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
2. Making Orders
This section describes the roles and responsibilities of inspectors and DOs with respect to making
orders, and the steps required to carry out this activity.
2.1 Roles and responsibilities
2.1.1 Inspectors
As part of their role, inspectors are authorized under the NSC Act to order licensees
to take measures necessary to protect the environment or health and safety of
persons, or to maintain national security and comply with international obligations to
which Canada has agreed (NSC Act subsection 35(1)). Inspectors may also make
orders to any person during the course of specific inspections (NSC Act subsection
35(2)). With respect to making an order, inspectors are responsible for:
* carrying their certificate at all times (NSC Act subsection 29(2));
* ensuring that the subject of the order is within the inspector's area of authority
as defined in the certificate (NSC Act subsections 29(1) and (2));
* using all reasonable means to consult with qualified colleagues about a
situation before making an order;
* ensuring that the criteria set out in NSC Act S35 for making the order have
been met;
* discussing the order with the persons required to comply with the order to
ensure that they understand the nature of it (Rule 33) and the timeframe for
compliance, see appendix B, section 7;
* referring any order to the Commission for review (NSC Act subsection 35(3)).
(The review function may be delegated to a DO under NSC Act 37(2)(g).)
2.1.2 Designated officers
As part of their role, DOs may be authorized under the NSC Act to make any order
that an inspector may make under subsections 35(1) or (2) of the Act.
With respect to making an order, a DO is responsible for:
* referring orders made by the DO to the Commission for review (NSC Act
subsection 37(6)), and
* all the responsibilities listed for inspectors in subsection 2.1.1 of this document.
2.2 Requirements for making orders
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Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
Orders given by inspectors and DOs are given in accordance with section 35 of the NSC Act
and the CNSC Rules of Procedure. Sections of these Rules of Procedure that pertain to
making and reviewing orders can be found in Appendix A.
Note : The extent of any consultation with persons to whom an order will be directed prior
to making the order is a matter for the discretion and judgement of the inspector or DO,
depending upon the urgency of the situation and the surrounding circumstances. In all cases,
the inspector or DO will comply with the requirements of CNSC Rules of Procedure Rule
33.
2.2.1 Order made by an inspector
Note : See Appendix C for an Inspector's Checklist for Making Orders. For
inspectors, the procedure for making an order is:
(a) Enter and inspect the vehicle or place in question (NSC Act subsections 30(1), (2)
and (3); and sections 31 and 32).
Note : Under subsection 35(2) the power of the inspector is restricted in cases given
in subsection 30(2) and (3)
(b) If a deficiency is identified that may require action, draw it to the attention of the
person responsible for it and explain the concerns. If an order is appropriate,
consult with qualified colleagues and inform any person who will be required to
comply with the order of the nature and basis of the order, and the proposed time
limit for compliance (subsection 33(1) of the Rules).
(c) Confirm the order, in writing, with any person who will be required to comply
with the order (subsection 33(2) of the Rules); keep a record of the order and
your inspection notes.
Note : See Appendix B for an example of the form used to give orders. Although no
legal requirement exists to use this form, its use is strongly recommended. Orders that
do not make use of this form must be clearly identified as an order and must include
all relevant information as set out on the form.
(d) Refer the order to the Commission for review no more than 10 days after an order
has been given (subsection 34(1) of the Rules); keep a copy of the order.
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C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
2.2.2 Order made by a DO
For DOs, the procedure for making an order is:
(a) After appropriate consultation with qualified colleagues, inform any person who
will be required to comply with the order of the nature and basis of the order, and
the proposed time limit for compliance (subsection 33(1) of the Rules).
(b) Confirm the order, in writing, with any person who will be required to comply
with the order (subsection 33(2) of the Rules); and keep a record of the order.
Note : Refer to Appendix B for an example of the form used to give orders. Although
no legal requirement exists to use this form, its use is strongly recommended. Orders
that do not make use of this form must be clearly identified as an order and must
include all relevant information.
(c) Refer the order to the Commission for review no more than 10 days after an order
has been given (subsection 34 (1) of the Rules).
3. Reviewing Orders
This section describes the roles of the Commission or a DO in reviewing orders made by
inspectors or DOs, and the necessary steps required to carry out this activity.
3.1 Roles and responsibilities
3.1.1 Commission: ways of reviewing an order
As part of its role, the Commission may review an order in one of several ways:
* Review an order made by an inspector (NSC Act subsection 35(3)).
Note : The Commission may assign a DO to review the order (NSC Act paragraph
37(2)(g)).
* Review an order made by a DO (NSC Act subsection 37(6)).
* Review, on its own initiative or as requested by staff, an order reported to the
Commission by a DO (NSC Act subsection 37(5)); or an order made by the
Commission, an inspector or a DO (NSC Act subsection 43(3)).
* Review, on appeal by any person who is directly affected by the decision, a
DO's confirmation, amendment or revocation, or replacement of an order
made by an inspector (NSC Act paragraph 43(1)(d)).
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Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
3.1.2 Commission responsibilities
With respect to reviewing an order, the Commission is responsible for:
* providing the persons required to comply, or persons named in an order, with
the opportunity to be heard in accordance with the prescribed Rules of
Procedure before confirming, amending, revoking, or replacing an order made
by an inspector or a DO (NSC Act paragraphs 40(1)(c) and (d));
* hearing new evidence or rehearing evidence as it considers necessary for the
confirmation, amendment, revocation or replacement of an order (NSC Act
paragraphs 43(2)(e) and (f) and paragraphs 43(4)(g), (h), (i) and (j));
* providing the persons required to comply, or persons named in an order, with
the opportunity to be heard in accordance with the prescribed Rules of
Procedure before amending, revoking, or replacing an order; and before
reconfirming, canceling, amending, revoking or replacing a confirmation,
amendment or revocation of an order as the result of an appeal or
redetermination (NSC Act paragraph 40(1)(h)).
3.1.3 Designated officers
As part of their role, DOs may be authorized by the Commission to review orders
given by an inspector (NSC Act subsection 37(2)).
If authorized by the Commission to review an order given by an inspector, a DO is
responsible for:
* confirming, amending, revoking, or replacing any order made by an inspector
(NSC Act paragraphs 37(2)(g));
* providing the persons required to comply with, or persons named in, an order
with a reasonable opportunity to be heard before confirming, amending,
revoking, or replacing an order (NSC Act paragraph 39(1)(c)).
3.2 Procedural requirements for reviewing orders
Orders must be reviewed by the Commission, or if authorized by the Commission, by a DO,
in accordance with the CNSC Rules of Procedure. Sections of the Rules that are relevant to
reviewing an order can be found in Appendix A.
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C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
3.2.1 Review by the Commission
For the Commission, the procedure for reviewing an order is:
(a) Receive the order from an inspector or DO (as a result of NSC Act subsections
35(3) and 37(6) respectively, and subsection 34(1) of the Rules).
(b) Notify the persons named in, or subject to, the order of their right to be heard,
and the requirement to inform the Commission, within 10 days of receipt of this
notice, of whether they have any information to present with respect to the order
(paragraphs 34(2)(a) and (b) of the Rules).
(c) Notify the persons named in, or subject to, the order of the time and manner in
which the person may be heard and whether information and written submissions
are required to be filed with the Commission and sent to other parties, and if so,
the time limits for filing and sending the names and addresses of other parties
(paragraphs 34(3)(a) and (b) of the Rules).
(d) Review the information provided.
(e) Confirm, amend, revoke or replace the order (NSC Act subsections 35(3) and
37(6)).
(f) Provide, in writing, the decision to confirm, amend, revoke, or replace an order to
the persons required to comply with, or persons named in, the order, and to
persons who intervened in the proceeding, within 10 days of the decision
(subsection 34(4) of the Rules).
3.2.2 Commission responsibilities: appeal or redetermination
In the case of an appeal or redetermination of an order, the Commission will:
(a) Receive the appeal or application for redetermination (as a result of NSC Act
paragraphs 43(1)(d) and (2)(e) and (f); and subsection 35(2) of the Rules).
(b) Determine how the appeal or redetermination will be considered ,for example by
oral or written submission in a public or private manner (subsection 35(3) of the
Rules).
(c) After determining the manner in which the appeal or redetermination will be
heard, notify in writing, the participants in the proceeding related to the order of:
* the manner in which the appeal or redetermination will be considered and
the way in which the appellant may present information, evidence and
submissions (paragraph 35(4)(a) of the Rules);
* the date, place and time limits for giving or presenting information and
submissions (paragraph 35(4)(b) of the Rules);
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Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
* the names and addresses of the persons to whom copies of the
information and submissions must be sent (paragraph 35(4)(c) of the
Rules).
(d) Decide on the matter of the appeal or redetermination (NSC Act paragraphs
43(2)(e) and (f) and subsection (3)). For an oral hearing, hold the hearing and
determine the matters in question. If a written submission is received, review the
submission and determine the matters in question.
Note : In the case of a redetermination, a decision about whether to proceed with a
redetermination must be made.
(e) Notify, in writing, the parties and any other persons who intervened in the
proceeding of the decision (Rule 36) by sending a copy of the decision to the
parties and to any other persons who intervened in the proceeding.
3.2.3 Review by a designated officer
For a DO, the procedure for reviewing an order is:
(a) Be authorized, by the Commission, to review an order made by an inspector (NSC
Act subsection 37(1) and paragraph (2)(g)).
(b) As soon as practicable after receiving a notice from a person referred to in
section 2.1.1 of this document, notify the persons named in, or subject to, the
order of their right to be heard and the requirement to notify the DO, within 10
days of receipt of this notice, of any information and submissions they have to
present with respect to the order (NSC Act paragraph 39(1)(c) and paragraphs
34(2)(a) and (b) of the Rules).
(c) Notify the persons named in, or subject to, the order of the time and manner in
which the person may be heard and whether the information and written
submissions are required to be filed with the DO and sent to other parties, and if
so, the time limits for filing and sending the names and addresses of other parties
(paragraphs 34(3)(a) and (b) of the Rules).
(d) Review the information provided.
(e) Confirm, amend, revoke or replace the order (NSC Act paragraph 37(2)(g)).
(f) Provide, in writing, the decision to confirm, amend, revoke, or replace an order to
the persons required to comply with, or persons named in, the order, and to
persons who intervened in the proceeding, within 10 days of the decision
(subsection 34(4) of the Rules).
(g) Report to the Commission on a confirmation amendment, revocation or
replacement of the order (NSC Act paragraph 37(5)(d)).
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C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
4. Receiving Orders
This section describes the rights and responsibilities of the persons required to comply with
orders, and the actions required to comply with such orders.
4.1 Rights and responsibilities
4.1.1 Licensees or other persons to whom an order is directed have the right
to:
* be informed of the nature of the order, the basis for it and any proposed
time limit for compliance, before the order is made (subsection 33(1) of
the Rules).
* receive the order in writing (subsection 33(2) of the Rules)
4.1.2 Persons named in, or subject to, or those directly affected by an order
have the right to:
* be heard on the matter before the order is confirmed, amended, revoked
or replaced (NSC Act paragraphs 39(1)(c) and 40(1)(c) and (d));
* appeal decisions made by a DO, confirmation, amendment, revocation or
replacement of an order given by an inspector (NSC Act paragraph
43(1)(d));
* apply for a redetermination of an order by the Commission (NSC Act
paragraph 43(2)(e));
* apply for a redetermination of a confirmation, amendment, revocation or
replacement of an order by the Commission (NSC Act paragraph
43(2)(f)); and
* be heard by the Commission on an appeal or redetermination (NSC Act
paragraph 40(1)(h)).
Note : All appeals are considered provided that an application for an appeal has been
submitted and the person making the appeal is named in, or subject to, the order (NSC
Act paragraphs 43(2)(e) and (f)).
4.1.3 Persons named in an order are responsible for:
* giving an inspector all reasonable assistance to enable the inspector to
carry out his or her duties (NSC Act section 36); and
* complying with the order within the time limit specified in the order, or
immediately if no time limit is specified, even if the person has not had
the opportunity to make representations with respect to the order (NSC
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Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
Act section 41);
Note : Information on liability for costs is provided in section 42 of the NSC Act.
4.2 Procedural requirements for receiving orders
4.2.1 Upon receipt of an order
Upon receipt of an order, a person named in, or subject to, an order must:
(a) Comply with the order within the time limit specified in the order, or if no time
limit is specified, immediately, whether or not the person has had the opportunity
to make representations with respect to the order (NSC Act section 41).
(b) Notify the Commission or DO of their intent, if any, to present information and
submissions with respect to the order within 10 days of being notified of this
requirement (paragraph 34(2)(b) of the Rules).
(c) If notified by the Commission or a DO, file information and submissions with the
Commission or DO and send them to other parties within the time limits specified
by the Commission or DO (paragraph 34(3)(b) of the Rules).
(d) Participate in the review in the manner determined pursuant to paragraph 34(3)(a)
of the Rules.
4.2.2 Upon an appeal or redetermination
In the event that a person named in, or subject to, the order wishes to appeal or apply
for redetermination of a decision made by the Commission or a DO, the appellant
shall:
(a) File, within 10 days of receiving the review decision, an appeal or application for
redetermination with the Commission (NSC Act paragraph 43(1)(d)). The appeal
must include the following information:
* a reference to the paragraph of subsections 43(1) or (2) of the Act under
which the appeal or application is being made (paragraph 35(2)(a) of the
Rules);
* a reference to the order that is the subject of the appeal (paragraph
35(2)(b) of the Rules);
* the grounds for appeal including a statement showing how the appellant
is directly affected by the order (paragraph 35(2)(c) of the Rules);
* a statement of the action that the appellant advises the Commission to
take under subsection 43(4) of the Act (paragraph 35(2)(d) of the Rules);
* a statement indicating whether or not the appellant wishes to present
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C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
new evidence (paragraph 35(2)(e) of the Rules);
* a description of the manner in which the appellant wishes to participate
in the proceeding (paragraph 35(2)(f) of the Rules);
* the name, address, and telephone and fax numbers of the appellant
(paragraph 35(2)(g) of the Rules); and
* a statement indicating whether the appellant intends to be represented by
counsel or an agent, and if so, the name, address, and telephone and fax
numbers of this person (paragraph 35(2)(h) of the Rules).
(b) Send copies of the information and written submissions to the people indicated by
the Commission in their notice (paragraph 35(4)(c) of the Rules).
(c) Present information and submissions on the day and time specified by the
Commission (paragraph 35(4)(b) of the Rules).
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Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
Glossary
Canadian Nuclear Safety Commission
also called "the CNSC" or "the Commission," established by section 8 of the Nuclear
Safety and Control Act
designated officer (DO)
any person whom the Commission considers qualified and who possesses a certificate
setting out the activities that a designated officer is authorized to carry out (NSC Act
section 37).
Note : A DO may be a CNSC employee or a non-CNSC employee under an arrangement
with the CNSC.
inspector
any person whom the Commission considers qualified and designates as an inspector (NSC
Act section 29)
Note : An inspector may be a CNSC employee or a non-CNSC employee under an
arrangement with the CNSC.
order
an order of an inspector under subsection 35(1) or (2) of the NSC Act, or of a DO under
paragraph 37(2)(f), or of the CNSC under sections 43, 16 or 47
11
C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
Appendix A: Excerpts from the CNSC Rules of Procedure
PART 6
ORDERS OF INSPECTORS AND DESIGNATED OFFICERS
Making of Orders
33. (1) Before making an order under section 35 or paragraph 37(2)(f) of the Act, an
inspector or a designated officer, as the case may be, shall inform, either orally or in writing,
the person to whom the order is given of its nature and the basis for it, as well as any
proposed time limit for compliance.
(2) The inspector or designated officer shall give the order referred to in subrule (1) in
writing.
Proceedings to Confirm, Amend, Revoke or Replace an Order
34. (1) As soon as practicable and, in any event, not later than 10 days after an inspector
or a designated officer, as the case may be, gives an order under subrule 33(2), the
inspector or designated officer shall refer it to the Commission for confirmation,
amendment, revocation or replacement in accordance with subsection 35(3) or 37(6) of the
Act, as the case may be.
(2) Where the Commission acts under subsection 35(3) of the Act or a designated officer
is authorized by the Commission to act under paragraph 37(2)(g) of the Act, the
Commission or designated officer, as the case may be, shall notify the persons named in or
subject to the order of
(a) their opportunity to be heard under paragraph 39(1)(c) or 40(1)(c) or (d) of the
Act, as the case may be; and
(b) the requirement that those persons who intend to present information and
submissions in respect of the order so notify the Commission or designated officer
within 10 days after receipt of the notice.
(3) As soon as practicable after receiving a notice referred to in paragraph (2)(b), the
Commission or designated officer shall notify the person giving the notice
(a) of the manner in which and time when the person may be heard; and
(b) whether information and written submissions are required to be filed with the
Commission or designated officer and sent to other parties and, if so, the time limits
12
Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
for filing and sending and the names and addresses of the other parties.
(4) A decision by the Commission or designated officer, as the case may be, to confirm,
amend, revoke or replace an order shall be in writing and must be sent, within 10 days after
being made, to the persons who are named in or subject to the decision, and to any other
persons who intervened in the proceeding.
Appeal or Redetermination of an Order
35. (1) This rule and rule 36 apply in respect of
(a) an appeal, under subsection 43(1) of the Act, of a confirmation, amendment,
revocation or replacement, by a designated officer, of an order of an inspector; and
(b) a rehearing and redetermination, by the Commission,
(i) under paragraph 43(2)(e) of the Act, of an order of the Commission, and
(ii) under paragraph 43(2)(f) of the Act, of a confirmation, amendment,
revocation or replacement by the Commission, of an order of an inspector or a
designated officer.
(2) An appeal or an application for a redetermination referred to in subrule (1) may be
made by sending to the Commission a notice that includes the following information:
(a) a reference to the paragraph of subsection 43(1) or (2) of the Act under which the
appeal or application is being made;
(b) a reference to the order that is the subject of the appeal or application;
(c) the grounds for the appeal or application, including, in the case of an appeal, a
statement showing how the appellant is directly affected by the order being appealed;
(d) a statement of the action that the appellant or applicant submits that the
Commission should take under subsection 43(4) of the Act;
(e) a statement indicating whether or not the appellant or applicant wishes to present
new evidence;
(f) a description of the manner in which the appellant or applicant proposes to
participate in the proceeding;
(g) the name, address and telephone and facsimile numbers of the appellant or
applicant; and
13
C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
14
Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
(h) a statement indicating whether the appellant or applicant intends to be represented
by counsel or an agent and, if so, the name, address and telephone and facsimile
numbers of the counsel or agent.
(3) After receiving the information referred to in subrule (2), the Commission shall decide
whether the appeal or redetermination will be by way of public hearing under paragraph
40(5)(b) of the Act or written submissions or by another manner that will enable the
Commission to determine the matter before it in a fair, informal and expeditious manner.
(4) After determining the manner in which the appeal or redetermination will be
considered, the Commission shall send to the appellant or applicant and to those persons
who were participants in the proceeding related to the order being appealed or
redetermined, a notice that includes
(a) a description of the manner in which the redetermination will be considered and a
description of the manner in which the appellant or applicant and any other participant
may present information, evidence and submissions to the Commission;
(b) the date, place and time limits for giving or presenting information and
submissions; and
(c) the names and addresses of the persons to whom copies of the information and
written submissions must be sent.
Decision
36. The Commission shall give notice of its decision in respect of the matter that has been
the subject of appeal or rehearing and redetermination dy sending a copy of the decision to
the parties and to any other persons who intervened in the proceeding.
15
C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
Appendix B: Form for Making Orders
Canadian Nuclear Safety Commission INDEX: (Pre-printed number)
Commission canadienne de sûreté nucléaire
ORDER UNDER SECTION 35 OR PARAGRAPH 37(2)(f) OF THE
NUCLEAR SAFETY AND CONTROL ACT
ORDRE EN VERTU DE L'ARTICLE 35 OU DE L'ALINÉA 37(2)f) DE LA
LOI SUR LA SÛRETÉ ET LA RÉGLEMENTATION NUCLÉAIRES
1 2
CNSC Licence No.: (if applicable) Date of Order: (y/m/d)
No de permis de la CCSN : (s'il y a lieu) Date de l'ordre : (a/m/j)
3 Company/Licensee (if applicable) and address: 4 Name (and title or position) of person receiving the Order:
Nom de l'entreprise/du titulaire de permis (s'il y a lieu) et adresse : Nom (et titre ou poste) de la personne destinataire de l'ordre :
(signature and title - signature et titre)
(print/en majuscules)
5 Actions or measures required to be taken by licensee and/or other person(s) (specify) in respect of any facility, place, substance, vehicle,
equipment or information (specify): / Mesures correctives à prendre par le titulaire de permis ou autre(s) personne(s) (préciser) en rapport
avec une installation, un endroit, une substance, un véhicule, un équipement ou des renseignements (préciser) :
6 Information on which Order is based, including any inspections, reports, evaluation, contraventions, with relevant dates: / Renseignements
servant à établir l'ordre, y compris tout rapport et toute inspection, évaluation et contravention, avec dates pertinentes :
CERTIFICATE NUMBER__________________
Page 1 of/de
16
Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
7 Time limit for complying: Immediately: 9 Time specified: 9 _________________(y/m/d)
Date limite pour se conformer : Immédiatement : 9 Date précise : 9 _________________(a/m/j)
8 CNSC Inspector or Designated Officer making the Order/Inspecteur ou fonctionnaire désigné de la CCSN qui donne l'ordre :
Name/nom : _____________________________________________ Tel./téléphone : _____________________
Address/adresse : _____________________________________________ Fax/télécopieur : _____________________
_____________________________________________
_____________________________________________ Signature:___________________________
9 Method of transmitting the Order: Personal delivery 9 Mail 9 Fax 9 Other (specify)
9
Mode de transmission de l'ordre : Livrée en personne 9 Poste 9 Télécopieur 9 Autre (préciser) 9
SEE REVERSE SIDE FOR ADDITIONAL INFORMATION/VOIR VERSO POUR RENSEIGNEMENTS SUPPLÉMENTAIRES
CNSC Form SEC01(0899)
A SUMMARY OF SOME RELEVANT SECTIONS OF THE ACT
ORDERS OF AN INSPECTOR
35. (1) An inspector may order that a licensee take any measure considered necessary to protect the environment, health or safety of persons or
to maintain national security or compliance with international obligations to which Canada has agreed.
35. (2) Refer to this section for Orders issued to persons who are not licensees.
DESIGNATED OFFICERS
37. (2) The Commission may authorize a designated officer to
(f) make any order that an inspector may make under subsections 35(1) or (2).
PROCEDURES
38. Every order of an inspector and every order of a designated officer under paragraph 37(2)(f) shall be made, and every measure under
paragraph 37(2)(c), (d) or (g) shall be taken, in accordance with prescribed CNSC Rules of Procedure.
COMPLIANCE WITH ORDER
41. Every person named in, or subject to, an order of the Commission, an inspector or a designated officer shall, whether or not the person has
had an opportunity to make representations with respect to the order, comply with the order within the time specified in it or, if no time is
specified, immediately.
OPPORTUNITY TO BE HEARD: Refer to sections 39 and 40 of the Act.
LIABILITY FOR COSTS: Refer to section 42 of the Act.
OFFENSES AND PUNISHMENT: Refer to sections 48 to 65 inclusive of the Act.
17
C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
SOMMAIRE DE QUELQUES ARTICLES PERTINENTS DE LA LOI
ORDRES DE L'INSPECTEUR
35. (1) L'inspecteur peut ordonner à un titulaire de licence ou de permis de prendre les mesures qu'il estime nécessaires à la préservation de la
santé ou de la sécurité des personnes, à la protection de l'environnement, au maintien de la sécurité nationale ou au respect par le Canada de ses
obligations internationales.
35. (2) Veuillez vous référer à ce paragraphe pour tout Ordre donné aux personnes qui ne sont pas titulaires de permis.
FONCTIONNAIRES DÉSIGNÉS
37. (2) La Commission peut autoriser le fonctionnaire désigné à :
f) donner les ordres qu'un inspecteur peut donner en vertu des paragraphes 35(1) ou (2).
RÈGLES DE PROCÉDURE APPLICABLES
38. Les ordres de l'inspecteur, les décisions du fonctionnaire désigné visées aux alinéas 37(2)c), d) ou g) et les ordres du fonctionnaire désigné
visés à l'alinéa 37(2)f) sont donnés ou pris en conformité avec les règles de procédure réglementaires.
CARACTÈRE OBLIGATOIRE DES ORDRES ET DES ORDONNANCES
41. Les destinataires des ordres des inspecteurs et des fonctionnaires désignés et des ordonnances de la Commission ainsi que toutes les autres
personnes qui y sont visées sont tenus de s'y conformer avant l'expiration du délai qui y est fixé ou, à défaut, sans délai, même s'ils n'ont pas eu
la possibilité de présenter leurs observations au préalable.
POSSIBILITÉ D'ÊTRE ENTENDU : veuillez consulter les articles 30 et 40 de la Loi.
RESPONSABILITÉS DES FRAIS : veuillez consulter l'article 42 de la Loi.
INFRACTIONS ET PEINES : veuillez consulter, inclusivement, les articles 48 à 65 de la Loi.
CNSC Form SEC01(0899)
18
Making, Reviewing and Receiving Orders under the NSC Act C-273 (E)
Appendix C: Inspector's Checklist for Making Orders
Inspector:
Inspection Date:
Location:
Licensee:
Place a checkmark in the Action column for activities that are successfully completed.
Activity Action
1. Inspector's Certificate is carried/shown.
2. Check scope of certificate
3. Inspected the place/vehicle in question.
4. Identified the deficiency.
5. Explained concerns to the person(s) responsible.
6. Heard person's explanation for deficiency.
7. Consulted with qualified colleagues (if applicable).
8. Informed the person that an order is being issued and that the person
is required to comply with it in a specified time frame.
Date licensee to comply by: / /
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C-273 (E) Making, Reviewing and Recieving Orders under the NSC Act
9. Confirmed order (in writing) with the person required to comply
with it.
Order was made by:
Order was received by: Name_______ Position_______ Date____
Form for Making Orders _______
Other _______
10. Referred order to the Commission for review.
Date: / /
This date is within 10 days of making the Order? Yes No
11. Retained a copy of the Order.
12. Retained copy of background notes/inspection report.
Notes:
20
DRAFT
REGULATORY
GUIDE
Preparing a Security
Report for Licence
Applications
C-274
Issued for public consultation by the
Canadian Nuclear Safety Commission
February 2001
DRAFT REGULATORY GUIDE
Preparing a Security Report for Licence Applications
C-274 (E)
Issued for public consultation by the
Canadian Nuclear Safety Commission
February 2001
Regulatory Documents
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY GUIDE
Preparing a Security Report for Licence Applications
C-274 (E)
February 2001
About this Document
This regulatory guide, C-274 Preparing a Security Report for Licence Applications, provides
guidance for licensees when preparing and submitting documentation to meet the security
requirements pursuant to the Nuclear Safety and Control Act (NSC Act, the Act) and its
regulations.
Comments
The CNSC invites interested persons to assist in the further development of this draft regulatory
document by commenting in writing on the document's content and potential usefulness. Please
respond by May 15, 2001. Direct your comments to the postal or e-mail address below,
referencing file 1-8-8-274.
The CNSC will take the comments received on this draft into account when developing it further.
These comments will be subject to the provisions of the federal Access to Information Act.
Document availability
This document can be viewed on the CNSC Internet site at www.nuclearsafety.gc.ca. To order a
printed copy of the document in English or French, please contact:
Operations Assistant
Corporate Documents Section
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 996-9505
Facsimile: (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
i
Preparing a Security Report for Licence Application C-274 (E)
Contents
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory framework . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Security report content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2 General Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2.1 Licensee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2.2 Geographic location . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.3 Corporate security policy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2.4 Licence conditions and the Nuclear Security Regulations . . . . . . . . . . . . . . . . . . 2
2.5 Site plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
3 Security Organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.1 Role of security in the facility organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.2 Structure of the security organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.3 Security guard selection criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.4 Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.5 Drills . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.6 Security personnel equipment and vehicles. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.7 Records and reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.8 Prescribed information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
4 Requirements Concerning Protected Areas, Inner Areas and
Security Monitoring Rooms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
4.1 Protected areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
4.1.1 Physical barriers and intrusion detection devices for protected areas . . . . 4
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C-274 (E) Preparing a Security Report for Licence Application
4.2 Inner areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.2.1 Physical barriers and intrusion detection devices for inner areas. . . . . . . . 5
4.3 Other areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.4 Security monitoring room . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.4.1 Internal communications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.4.2 External communications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
4.5 Emergency exits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
5 Entry and Exit Controls for Protected and Inner Areas . . . . . . . . . . 6
5.1 Entry and exit controls for protected areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
5.1.1 Unescorted entry and exit for protected areas . . . . . . . . . . . . . . . . . . . . . 6
5.1.2 Escorted entry and exit for protected areas . . . . . . . . . . . . . . . . . . . . . . . 6
5.2 Entry and exit controls for inner areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
5.2.1 Unescorted entry and exit for inner areas . . . . . . . . . . . . . . . . . . . . . . . . . 7
5.2.2 Escorted entry and exit for inner areas . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
5.3 Access system . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.3.1 Identification badge/access card type. . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.3.2 Identification badge/access card control . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.3.3 Identification badge/access card usage. . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.4 Vehicle entry and exit control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.5 Package and equipment access control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
5.6 Keys to access protected and inner areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
6 Security Systems and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
6.1 Design and performance characteristics of security systems and technical
devices . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
6.2 Maintenance programs for new and existing security systems, technical devices
and communications equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
6.2.1 Equipment operation tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
6.2.2 Preventative maintenance programs. . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
6.2.3 Repairs and maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
7 Contingency Planning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
7.1 Contingency plans . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
7.2 Security personnel availability and responsibility. . . . . . . . . . . . . . . . . . . . . . . 11
iii
Preparing a Security Report for Licence Application C-274 (E)
8 Sabotage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
9 Arrangements with Off-site Response Force . . . . . . . . . . . . . . . . . . 11
10 Employee Awareness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
Appendix A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
iv
C-274 (E) Preparing a Security Report for Licence Application
Purpose
This document provides guidance for licensees when preparing and submitting documentation to
meet the security requirements pursuant to the Nuclear Safety and Control Act (NSC Act, the
Act) and its regulations.
Scope
The guide gives the headings and explanatory text to lead the user through the steps needed to
produce documentation that meets the security requirements of the Nuclear Security Regulations,
the General Nuclear Safety and Control Regulations, and the Class 1 Nuclear Facilities
Regulations. The appendix provides the format for standardized documentation for licence
applications. The resulting documentation is referred to as the "Security Report".
The guide applies to licensees and licence applicants for Category I and II nuclear materials, and
facilities with a nuclear reactor that may exceed 10 megawatts thermal power during operation.
1 Introduction
1.1 Regulatory framework
The Canadian Nuclear Safety Commission (CNSC, the Commission) is the federal agency
that regulates the use of nuclear energy and materials to protect health, safety, security and
the environment, and to respect Canada's international commitments on the peaceful use of
nuclear energy.
The NSC Act requires persons or organizations to be licensed by the CNSC for carrying out
the activities referred to in section 26 of the Act, unless otherwise exempted. The associated
regulations stipulate prerequisites for CNSC licensing, and the obligations of licensees and
workers.
1.2 Security report content
The document headings 2 through 10 and descriptive text throughout the body of this
document were prepared by Commission staff to assist the licensee and licence applicants
when preparing a Security Report for the CNSC licence application and renewal process.
2 General Information
2.1 Licensee
This section of the report should include, as indicated in the licence application, the name of
the licensee, address, postal code, telephone and fax numbers, and the legal ownership. The
licensee should include three emergency management response positions with names and
duty contact numbers.
1
Preparing a Security Report for Licence Application C-274 (E)
2.2 Geographic location
This section of the report should provide a scaled drawing showing the geographic location
(longitude/latitude) and topography surrounding the site. The topographical illustrations
should include main roads, all secondary access roads, access by rail/water/air and the
locations of the nearest towns and/or cities, as well as the natural features of the area.
2.3 Corporate security policy
This section of the report should provide or summarize the licensee's corporate security
policy.
2.4 Licence conditions and the Nuclear Security Regulations
This section of the report should describe the licensee's security obligations under
* the Nuclear Security Regulations, the General Nuclear Safety and Control Regulations
and the Class 1 Nuclear Facilities Regulations; and
* its licence conditions.
2.5 Site plan
The report must include a site plan, pursuant to section 16 of the Nuclear Security
Regulations. The site plan should include
* the perimeter of the nuclear facility referred to in paragraph 2(b) of the Nuclear Security
Regulations
* the barrier enclosing every protected area
* the protected areas
* the unobstructed areas
* the structure or barrier enclosing every inner area
* the inner areas
* fixed security posts
* the location of the Security Monitoring Room(SMR) and, if applicable, of the secondary
SMR
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C-274 (E) Preparing a Security Report for Licence Application
3 Security Organization
3.1 Role of security in the facility organization
This section of the report should give a description of the duties and responsibilities of
managers who have direct responsibility for the security program and operational decisions.
It should include a facility organization chart showing the facility management to security
interface.
3.2 Structure of the security organization
This section of the report should describe the structure of the security organization, levels
of authority and accountability for the security organization pursuant to paragraph 3(e) of
the Nuclear Security Regulations. It should include a security organization chart.
The duties and responsibilities of the various levels of security guards should be described.
Also, the size of the security force should be given, including shift crews (day and off-
hours), minimum complement for each shift, number and location of guard posts.
3.3 Security guard selection criteria
This section of the report should describe the criteria and procedures for recruiting,
screening, and selecting new security personnel.
3.4 Training
This section of the report should describe, pursuant to section 34 of the Nuclear Security
Regulations, the orientation and the training program for new security personnel. It should
describe the refresher training program for existing security personnel.
A training plan which should be provided, should include the course content, hours of
training per subject and the testing methodology.
3.5 Drills
Section 36 of the Nuclear Security Regulations requires licensees to conduct security drills
that are required to achieve and maintain competence in security duties at least once every 6
months. In order that the Commission may fully appreciate the scope and effectiveness of
the drills, the report should describe how drill scenarios are developed, document their
frequency, and the audit process used to determine the effectiveness of the drills and the
lessons learned.
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Preparing a Security Report for Licence Application C-274 (E)
3.6 Security personnel equipment and vehicles
This section of the report should describe the number and specifications for all equipment,
including vehicles, provided to security personnel. This information should include vehicle
related equipment; portable communications; night-vision aids; and physical and radiation
protective equipment, search devices, and conditions for use.
3.7 Records and reports
This section of the report should describe, pursuant to section 37 of the Nuclear Security
Regulations, the system used to record the name of every person authorized to enter a
protected or inner area, the duties and responsibilities of its nuclear security guards, and the
training given to each nuclear security guard. The licensee should also maintain a record of
current security procedures, event reporting and security guard performance monitoring.
3.8 Prescribed information
This section of the report should describe the methods by which the licensee will protect
prescribed information, in any form, pursuant to section 21 of the General Nuclear Safety
and Control Regulations.
4 Requirements Concerning Protected Areas, Inner Areas and
Security Monitoring Rooms
4.1 Protected areas
This section of the report should describe the location and function of the facility's
protected areas.
4.1.1 Physical barriers and intrusion detection devices for protected areas
Pursuant to sections 9, 10 and 11 of the Nuclear Security Regulations, this section of
the report should describe all physical barriers, including entry/exit portals and /or
guard posts, and intrusion detection devices located at the boundaries to and in the
protected areas. It should describe the devices, equipment, illumination and
assessment capability to detect and assess the cause of an annunciating alarm in any
protected area. A description of the unobstructed area surrounding the protected
areas should be included.
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4.2 Inner areas
The inner area should be located within a protected area as prescribed in section 12 of the
Nuclear Security Regulations. This section of the report should describe the location and
function of the facility's inner areas.
4.2.1 Physical barriers and intrusion detection devices for inner areas
Pursuant to sections 13 and 14 of the Nuclear Security Regulations, this section of
the report should describe the physical barriers and intrusion detection devices located
at the boundaries to and in the inner areas. It should describe also the devices,
equipment, illumination, and assessment capability to detect and assess the cause of
an annunciating alarm in any inner area.
4.3 Other areas
This section of the report should be used to describe additional locations where security
measures are in place to enhance the overall physical protection program required by the
Nuclear Security Regulations, e.g., controlling access to the site, pursuant to paragraph
3(1)(g) of the General Nuclear Safety and Control Regulations.
4.4 Security monitoring room
This section of the report should describe how the SMR meets the requirements given in
section 15 of the Nuclear Security Regulations and, where applicable, how the secondary
SMR meets the requirements given in subparagraph 14(a)(iii) of the Nuclear Security
Regulations. The description should include the location and function of the SMR, its
design and construction, access control, security equipment and staffing. It should describe
also the internal and external communication systems and type of communication equipment
provided for the security personnel.
4.4.1 Internal communications
Pursuant to paragraph 3(d) and subparagraphs 15(c)( iii) and (iv) of the Nuclear
Security Regulations, this section of the report should describe how security
personnel communicate with one another and with the SMR It should include the type
and specification of the communication equipment. For non-portable communication
equipment, the licensee should describe procedures for maintaining the operation of
such equipment in the event of loss of normal power.
4.4.2 External communications
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Preparing a Security Report for Licence Application C-274 (E)
Pursuant to paragraph 3(d) and subparagraphs 15(c)(i) and (ii) of the Nuclear
Security Regulations, this section of the report should describe how the SMR and
security personnel, as applicable, communicate with off-site agencies, particularly
emergency services such as off-site armed response forces. For non-portable
communication equipment, the licensee should describe procedures for maintaining
the operation of such equipment in the event of loss of normal power.
4.5 Emergency exits
This section of the report should describe the location of all emergency exits from
designated protected and inner areas, and the security provided in these locations both in
emergency and non-emergency conditions.
5 Entry and Exit Controls for Protected and Inner Areas
5.1 Entry and exit controls for protected areas
This section of the report should describe the access control points, the type of records and
the procedures for entry to and exit from the protected areas for employees, contractors,
visitors and others, e.g. students and attached staff.
5.1.1 Unescorted entry and exit for protected areas
This section of the report should describe entry and exit control procedures applicable
to employees or other persons authorized for unescorted access to a protected area
pursuant to subsection 17(2) of the Nuclear Security Regulations. It should include
details on identification requirements, badging and records.
The licensee should describe how it will meet the requirements of sections 25, 26 and
27 of the Nuclear Security Regulations, including but not limited to procedures for
searching persons for explosives and any weapon such as a concealed firearm. The
search procedure should specify when a search may be conducted e.g., reasonable
suspicion.
5.1.2 Escorted entry and exit for protected areas
This section of the report should describe entry and exit control procedures applicable
to persons authorized for escorted access to a protected area pursuant to subsection
17(3) of the Nuclear Security Regulations. It should include information on
identification, badging and records. There should also be a description of the
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C-274 (E) Preparing a Security Report for Licence Application
procedures used for escorting and any escort training programme. Visitor-escort
ratios should be specified.
The licensee should describe how it will meet the requirements of sections 25, 26 and
27 of the Nuclear Security Regulations, including but not limited to procedures for
searching persons for concealed firearms, explosives and any other weapon that could
be used to commit a crime.
If applicable, the licensee should include a description of conditions when special
escorting privileges have been granted, for example, escorts provided for contractors
or construction crews.
5.2 Entry and exit controls for inner areas
This section of the report should identify the access control points, records and procedures
for inner areas.
5.2.1 Unescorted entry and exit for inner areas
This section of the report should describe entry and exit control procedures applicable
to employees or other persons who have been authorized to enter inner areas
pursuant to section 18 of Nuclear Security Regulations. It should include information
on identification, badging and records.
The licensee should describe how it will meet the requirements of sections 25, 26 and
27 of the Nuclear Security Regulations, including but not limited to procedures for
searching persons for concealed firearms, explosives and any other weapon that could
be used to commit a crime.
5.2.2 Escorted entry and exit for inner areas
This section of the report should describe entry and exit control procedures as they
apply to persons authorized for escorted access to an inner area pursuant to section
20 of Nuclear Security Regulations. This should include information on identification,
badging and records. Escort arrangements, including identification of individuals who
may be authorized to escort and a description of the procedures used for escorting,
and escort training should be provided. Visitor-escort ratios should be specified.
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Preparing a Security Report for Licence Application C-274 (E)
The licensee should describe how it will meet the requirements of sections 25, 26 and
27 of the Nuclear Security Regulations, including but not limited to procedures for
searching persons for concealed firearms, explosives and any other weapon that could
be used to commit a crime.
5.3 Access system
5.3.1 Identification badge/access card type
This section of the report should describe the identification badging and/or access
card system for employees, contractors and visitors entering the protected or inner
areas. This should include a description of the information on each type of
identification badge/access card, such as colour code, photo, security clearance,
name, description, expiration date, restrictions, and other applicable information.
5.3.2 Identification badge/access card control
This section of the report should describe the system for issuing, accountability and
storage of identification badges/access cards and related records.
5.3.3 Identification badge/access card usage
This section of the report should describe the requirements for wearing and displaying
identification badges/access cards while on the site. It should describe the procedure
for surrendering the identification badges/access cards on termination of employment
or when leaving the site.
5.4 Vehicle entry and exit control
This section of the report should describe methods used to control all points of vehicle
movement to and from protected and inner areas under both normal operating and
emergency conditions. It should include methods for establishing and maintaining written
procedures that permit security personnel to identify vehicles that are authorized to enter
protected and inner areas. Procedures for vehicle badging and vehicle entry/exit searches
should also be described.
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C-274 (E) Preparing a Security Report for Licence Application
5.5 Package and equipment access control
This section of the report should describe methods used to control all points of access for
packages and equipment to protected and inner areas under both normal operating and
emergency conditions. It should include methods for establishing and maintaining written
procedures that permit security personnel to identify packages and equipment that are
authorized to enter protected and inner areas. If applicable, equipment and package badging
or authorization as well as entry/exit searches should also be described.
5.6 Keys to access protected and inner areas
This section of the report should describe licensee procedures for controlling all keys,
including any devices such as, locks, combinations, card keys, passwords, biometric
identification systems and other related equipment used for access to or egress from
protected and inner areas. It should briefly describe the make, type, design, manipulation
and pick-resistant features of each type.
The section should also describe how the keys are issued and controlled for each area.
Procedures for recording loss or theft of keys should be explained. This section should also
describe circumstances under which lost or stolen keys are changed and procedures
followed when an employee with access to keys terminates employment.
6 Security Systems and Equipment
6.1 Design and performance characteristics of security systems and
technical devices
This section of the report should explain the purpose, function, design, performance and
specifications of all technical devices that contribute to facility security. It should include a
detailed description of equipment and systems with technical specifications of the various
devices and block diagrams of system integration, as prescribed by paragraph 12(1)(d) of
the General Nuclear Safety and Control Regulations and paragraph 6(b) of the Class 1
Nuclear Facilities Regulations. This information may be submitted by supplying
manufacturers' data on the various devices. Equipment operation procedures and security
operating procedures should be submitted.
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Preparing a Security Report for Licence Application C-274 (E)
6.2 Maintenance programs for new and existing security systems,
technical devices and communications equipment
This section of the report should describe the maintenance program for security
systems. Information on scheduled maintenance procedures should be limited to a
brief description of the work being performed, a listing of the technical service
references used and the service schedule. Equipment service manuals should not be
provided. The maintenance information should include, but not be limited to intrusion
alarms, detection devices, emergency exit alarms, technical devices, lighting devices,
communication equipment, and other physical protection-related devices and
equipment.
6.2.1 Equipment operation tests
This section of the report should describe the testing and inspection program for
security-related equipment during routine operation. The description should include
the purpose, frequency and extent of testing and inspecting.
6.2.2 Preventative maintenance programs
This section of the report should describe the preventative maintenance program
established to ensure that all security-related subsystems and components are
maintained in operating condition. This section should also describe corrective actions
or compensatory measures used in the event of component failure within the security
equipment system.
6.2.3 Repairs and maintenance
This section of the report should describe the procedures, including schedules, for
performing repairs and maintenance of security-related equipment and systems
pursuant to paragraph 12(d) of the General Nuclear Safety and Control Regulations.
7 Contingency Planning
7.1 Contingency plans
This section of the report should describe the licensee's planned response to security
incidents, such as intrusions, threats, theft of nuclear material, sabotage or civil
disturbance. This section should describe the method for assessing a security breach,
procedures for response including the command structure of the security force during
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C-274 (E) Preparing a Security Report for Licence Application
all stages of the incident, and the procedures in place to transfer responsibility/
command to a responding off-site response force.
A copy of the security contingency plan and the procedures used to assess and
respond to specific security incidents should be provided. Reference should also be
made to the licensee's other contingency plans requiring support by the security force.
7.2 Security personnel availability and responsibility
This section of the report should describe the availability of security personnel, the
method for calling in additional personnel and the duties they will be assigned during
site emergencies.
8 Sabotage
This section of the report should describe how the licensee meets the requirements of paragraph
12(1)(h) General Nuclear Safety and Control Regulations and subsection 6(l) of the Class 1
Nuclear Facilities Regulations.
9 Arrangements with Off-site Response Force
This section of the report should describe the documented response arrangements, e.g.,
memorandum of understanding (MOU) or similar documentation of commitment, which the
licensee has made with a response force pursuant to section 35 of the Nuclear Security
Regulations. A copy of the MOU or document of commitment should be attached to the Security
Report.
In addition to providing detailed information regarding section 35 requirements, the report should
include an estimate of the response time and the strength of the response force ranging from the
initial response to the build-up of full response resources. The licensee, in consultation with the
response force, should determine the types of threats to the nuclear facility that the response force
is capable of handling, and what mechanisms are in place for the response force to request and
receive support from other police agencies.
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Preparing a Security Report for Licence Application C-274 (E)
10 Employee Awareness
This section of the report should describe the employee awareness program that is in place to
meet the requirements of subsection 24 (2) of the Nuclear Security Regulations and paragraphs
12(1)(j) and 17(b), (c)(ii), (iv) and (e) of the General Nuclear Safety and Control Regulations.
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C-274 (E) Preparing a Security Report for Licence Application
APPENDIX A
Recommended Format for the Report
The following guidelines are provided to assist licensees in preparing and submitting a
standardized Security Report for licence applications.
Terms used in the Report should conform with the definitions/terminology contained in section 1
of the Nuclear Security Regulations, e.g., protected areas, inner areas, security monitoring room
(SMR), etc.
1. Style and Composition
The report should contain a table of contents, subject to the revision procedures described in
section 4 below.
Information contained in the report should be clear and concise.
To the extent possible, duplication of information should be avoided. Information included in
other sections of the Report may be covered by specific reference to those sections.
2. Physical Specifications of Submissions
Reports should be submitted on standard 8 ½ x 11" paper. Pages should be punched for standard
loose-leaf 3-ring binders. Text should be single-spaced and pages numbered sequentially
throughout the main part of the document. Page numbers in the appendices may be numbered
separately. Each page of the report should contain the security classification level, page number, a
revision number if applicable, and a date.
3. Procedures for Updating and/or Revising
The report includes information to be submitted with a licence application for Category I and II
nuclear materials, and facilities with a nuclear reactor that may exceed 10 megawatts thermal
power during operation. Therefore, any changes to the Security Report during the licensing
period should follow a change control protocol. Changes submitted to the Commission should be
on a replacement-page basis and clearly indicate that the changes have been authorized within the
facility or organization. The changes or revised portions should be clearly identified by underlining
the revised/changed portions. All pages submitted to update or revise the report should show the
date of the change. A letter that includes an index page, listing the pages to be inserted and the
pages to be removed, should accompany the revisions (e.g. Document History and Revision
Sheet). When major changes or additions are made a revised table of contents should be provided.
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Preparing a Security Report for Licence Application C-274 (E)
4. Number of Copies
The Commission requires a single copy of the Security Report. Additional copies of the Report
may be requested by Commission staff on a case by case basis.
5. Classification and Public Disclosure
The report falls under the category of Prescribed Information, pursuant to section 21 of the
General Nuclear Safety and Control Regulations. Prescribed Information shall be protected to
prevent any transfer or disclosure that is not authorized by the Act and the regulations made
under the Act pursuant to subsection 23(2) General Nuclear Safety and Control Regulations.
This document is not for release to the general public, and its distribution is strictly on a need-to-
know basis. The report will be protected from public disclosure subject to the exemption
provisions of the Access to Information Act.
All submissions/correspondence of a prescribed nature between the Commission and the licensees
should be clearly labelled as such on the top right-hand side on each page and double enveloped
when mailed to authorized recipients at the Commission. The inner envelope should carry an
appropriate classification designation, e.g., PROTECTED SECURITY, CONFIDENTIAL, and
the name of the Commission Officer concerned, whereas the outer envelope should have the usual
Commission mailing address only.
6. Definitions and Abbreviations
Definitions and abbreviations should be consistent throughout the report and, if not defined by the
Nuclear Security Regulations, should be consistent with generally accepted usage to ensure
common understanding. The definitions and abbreviations should be given as an appendix.
14
DRAFT
REGULATORY
GUIDE
Human Factors
Engineering
Program Plans
C-276
Issued for public consultation by the
Canadian Nuclear Safety Commission
March 2001
DRAFT REGULATORY GUIDE
Human Factors Engineering Program Plans
C-276
Issued for public consultation by the
Canadian Nuclear Safety Commission
March 2001
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY GUIDE
Human Factors Engineering Program Plans
C-276
March 2001
About this Document
The draft regulatory guide Human Factors Engineering Program Plans guides licensees and
licence applicants in their preparation of an effective Human Factors Engineering Program Plan
that adequately incorporates human factors elements into licensable activities. In addition, this
document supports P-119, Policy on Human Factors.
Comments
The CNSC invites interested persons to assist in the further development of this draft regulatory
document by commenting in writing on the document's content and potential usefulness. Please
respond by June 29, 2001 . Direct your comments to the postal or e-mail address below,
referencing file 1-8-8-276.
The CNSC will take the comments received on this draft into account when developing it further.
These comments will be subject to the provisions of the federal Access to Information Act.
Document availability
This document can be viewed on the CNSC Internet site at (www.nuclearsafety.gc.ca). To order
a printed copy of the document in English or French, please contact:
Operations Assistant
Corporate Documents Section
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 996-9505
Facsimile: (613) 995-5086
E-mail:reg@cnsc-ccsn.gc.ca
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Human Factors Engineering Program Plans C-276
CONTENTS
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory Framework . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Licensing Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
3 Basic Elements of the Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.1 Scope of the Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.2 Background of the Activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3 Human Factors Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3.1 Roles and Responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3.2 Training Needs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3.3 Related Groups . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.4 Goals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.5 Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.5.1 Rationale . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4 Technical Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.1 Technical Basis of the Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.2 Technical Elements for Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.2.1 Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4.2.2 Verification and Validation Plan. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
4.3 Methods for Addressing the Technical Elements . . . . . . . . . . . . . . . . . . . . . . . . 7
4.4 Intended Tools . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
4.5 Technical Guides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
5 Processes and Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
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C-276 Human Factors Engineering Program Plans
5.1 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.2 Timelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.3 Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.4 Disposition of Human Factors Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.5 CNSC Contact . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
Bibliography . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
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Human Factors Engineering Program Plans C-276
iv
C-276 Human Factors Engineering Program Plans
Purpose
This document guides licensees and licence applicants in their preparation of an effective Human
Factors Engineering Program Plan that adequately incorporates human factors elements into
licensable activities. In addition, this document supports P-119, Policy on Human Factors.
Scope
This guide is directed towards the activities of Class I and Class II nuclear facilities, and uranium
mines and mills. However, every applicant and licensee should consider preparing a Human
Factors Engineering Program Plan that takes into account the risk, complexity, and the potential
impact on the health and safety of persons and the environment of the licensed activity.
This guide presents those technical elements that might be covered in such a Plan. Documentation
of these elements will typically depend upon the type of licensed or licensable activity, and
consequently will address factors such as the type of risks to workers, public safety and
environmental impact. The CNSC will also consider the special circumstances of small business
when assessing the human factors approach taken by applicants and licensees.
1 Background
1.1 Regulatory Framework
The CNSC is the federal agency that regulates the use of nuclear energy and materials to
protect health, safety, security, and the environment, and to respect Canada's international
commitments on the peaceful use of nuclear energy.
The NSC Act requires persons or organizations to be licensed by the CNSC in order to
carry out the activities referred to in Section 26 of the Act, unless otherwise exempted. The
associated regulations stipulate prerequisites for CNSC licensing, and the obligations of
licensees and workers.
The NSC Act and its regulations contain several provisions that are intended to ensure that
interfaces between humans and items involving nuclear substances, prescribed equipment,
or nuclear facilities will occur without unacceptable impacts on persons, the environment,
or national security.
References can be found in the following paragraphs of the General Nuclear Safety and
Control Regulations:
* An application for a licence shall contain (a description of) "the applicant's
organizational management structure insofar as it may bear on the applicant's
compliance with the Act and the regulations made under the Act, including the
internal allocation of functions, responsibilities and authority" (paragraph 3(l)(k)).
1
Human Factors Engineering Program Plans C-276
* An application for a licence shall contain, "at the request of the Commission, any
other information that is necessary to enable the Commission to determine whether
the applicant (i) is qualified to carry on the activity to be licensed or (ii) will, in
carrying our that activity, make adequate provision for the protection of the
environment, the health and safety of persons and the maintenance of national
security and measures required to implement international obligations to which
Canada has agreed" (paragraph 3(n)).
* "Every licensee shall ensure the presence of a sufficient number of qualified workers
to carry on the licensed activity safely and in accordance with the Act, the regulations
and the licence" (paragraph 12(l)(a)).
* Every licensee shall take all reasonable precautions to protect the environment and
the health and safety of persons and to maintain security" (paragraph 12(l)(c))."
1.2 Licensing Process
The CNSC applies a phased process to its licensing of nuclear facilities and activities. For
major facilities, this process begins with an assessment of the environmental impacts of the
proposed project, and proceeds progressively through site preparation, construction,
operation, decommissioning and abandonment phases.
The NSC Act and its regulations require licence applicants to provide certain information at
each licensing stage. The type and level of detail of this information will vary to
accommodate the licensing stage and specific circumstances.
Upon receipt of a complete application, the CNSC reviews the application to determine
whether the applicant is qualified to carry on the proposed activity, and has made adequate
provision for the protection of the environment, the health and safety of persons, and the
maintenance of national security and the measures required to implement international
obligations to which Canada has agreed. Safe and reliable human performance plays a major
role in overall system safety. As part of licence application reviews, CNSC staff assess
whether the applicant has made adequate provision for human capabilities and limitations
(human factors) as they relate to the safe conduct of the proposed activity.
2 Introduction
As described in P-119, Policy on Human Factors, it is the policy of the CNSC to consider human
factors issues in the nuclear facilities and activities licensed by the Commission. Human factors are
defined within the CNSC policy as
...factors that influence human performance as it relates to the safety of a nuclear facility
or activity over all phases, including design, construction, commissioning, operation,
maintenance, and decommissioning.
A Human Factors Engineering Program Plan documents the means by which human factors
2
C-276 Human Factors Engineering Program Plans
considerations are integrated into activities licensed by the CNSC. While this Plan is necessary to
ensure the proper development, execution, management, and documentation of the human factors
aspect of any licensable activity, it is not the intention of this guide to create unique human factors
work methods or processes. Such processes should have already been integrated into the normal
design process wherever possible.
The Human Factors Engineering Program Plan should describe the human factors technical
elements to ensure that the system or licensable activity is designed and evaluated according to
established human factors principles and practices. The technical program elements given in the
Plan should be supported by a comprehensive plan for verification and validation of the resulting
design or activity (C278, Guide to Human Factors Verification and Validation Plans) .
For a given project, the applicant should demonstrate that each of the human factors technical
elements (described in subsection 4.2, "Technical Elements for Review") has been addressed and
either built into the Plan or deemed not applicable. A preliminary outline of a proposed Human
Factors Engineering Program Plan may be submitted to the CNSC to initiate discussion. The need
for, and the technical elements to be included in, a Human Factors Engineering Program Plan,
would be established though discussion between the licensee and CNSC staff.
3 Basic Elements of the Plan
An effective Human Factors Engineering Program Plan includes information about
* the scope of the Plan,
* the background of the activity,
* human factors input,
* goals of the plan, and
* criteria for determining areas of consideration.
The format presented in this guide is only a suggestion. It is recognized that the format of Human
Factors Engineering Program Plans may vary considerably due to different focuses and levels of
activity.
3.1 Scope of the Plan
The scope of the Human Factors Engineering Program Plan should specify areas, systems
and components involved, and the phases in which human factors engineering will be
incorporated. Adequate justification for any exclusions should also be provided in this
section and discussed in the "Criteria" section, as described in subsection 3.5.
The Plan should include documentation on any constraints, limitations, and assumptions that
apply to the human factors program of work. These may relate to level of technology,
3
Human Factors Engineering Program Plans C-276
resource limitations, time constraints, consistency and compatibility with existing design or
operational features, or any other restrictions or requirements imposed on the Human
Factors team or the plan.
3.2 Background of the Activity
Provide a brief description of the overall project or activity being designed, including
purpose, scope, and time frames.
3.3 Human Factors Input
3.3.1 Roles and Responsibilities
Clearly define the role of any human factors representative associated with the
licensable activity for which the plan is being prepared. Expand on that role definition
with a statement about any part of the project which will require human factors
involvement and input.
3.3.2 Training Needs
Familiarity of the project or activity team with established human factors principles,
benefits, techniques and guidelines is important to successful implementation of the
Human Factors Engineering Program Plan. If training in these matters is required by
the project team members, indicate those training needs and the plans for addressing
them.
3.3.3 Related Groups
To varying degrees, the human factors technical elements addressed in the plan will
overlap and interface with other functions and disciplines within the project. Identify,
at a high level, all groups that may be impacted by the plan, and indicate how their
input will be considered or incorporated.
3.4 Goals
Provide concise statements about the objectives of the plan. Goal definition early in
development is vital to the plan's effectiveness and validity.
4
C-276 Human Factors Engineering Program Plans
3.5 Criteria
Provide a description of the type of criteria that will be used to determine which aspects of
the activity or project warrant human factors consideration. It is recommended that criteria
be based on function, task importance, or risk, and that criteria statements be clear, concise,
and objective.
3.5.1 Rationale
Indicate the rationale behind different levels of human factors effort, with an
explanation of how such levels of effort reflect established criteria. Some examples of
the types of decisions for which rationales would be helpful include:
* The human factors program being limited to certain areas in a facility,
* Task analysis being restricted to selected tasks,
* The human factors program being limited to certain project phases.
4 Technical Considerations
Consideration should be given to the following technical aspects of the Human Factors
Engineering Program Plan:
* Technical basis
* Technical elements for review
* Methods for addressing the technical elements
* Intended tools
* Technical guides
4.1 Technical Basis of the Plan
Clearly state the technical basis for the plan, including specific licence applicant's policies
and procedures for human factors, regulatory documents, and industry documents.
4.2 Technical Elements for Review
The following technical elements should be included in the plan.
4.2.1 Elements
human-machine interface system: any region or point at which a person interacts
with a machine
human-machine allocation of function: assigning system functions to human and
machine agents
5
Human Factors Engineering Program Plans C-276
human reliability: addressing issues pertaining to the probability that an individual or
group will adequately perform a given task at the appropriate time
job design: determining how tasks will be grouped together and how work will be
coordinated
operating experience review: the review and use of knowledge gained from nuclear
industry operating experience to improve future performance
physical working environment: the total physical environment within which a
worker performs his or her tasks
procedures development: the systematic process for the development of work
instructions or instruction sets used to accomplish a given task
shift-work systems: all of the schedules implemented in a given workplace to meet
the requirements of a given plant or process
staffing: for the purpose of the Human Factors Engineering Program Plan, the
process for determining numbers and placement of appropriate personnel for a given
job
validation: the process of determining the degree to which the human-machine
system design and supporting mechanisms facilitate the achievement of overall safety
and operational goals (see subsection 4.2.2)
verification: the process of demonstrating that equipment and systems have been
designed as specified and that adherence to human factors guidelines has been
maintained (see subsection 4.2.2)
4.2.2 Verification and Validation Plan
It is understood that a Verification and Validation Plan cannot always be submitted
concurrently with the Human Factors Engineering Program Plan. However, it is
expected that, as dictated by the type of project or activity, a commitment will be
made in the Human Factors Engineering Program Plan to submit a Verification and
Validation Plan at a later date. The Verification and Validation Plan would be
expected to follow C278, Guide to Human Factors Verification and Validation
Plans.
Provide justification for any omissions of the technical elements listed in subsection
4.2.1. It is expected that additional human factors issues may be identified and may
warrant assessment on a case-by-case basis.
4.3 Methods for Addressing the Technical Elements
6
C-276 Human Factors Engineering Program Plans
Describe the methods and techniques that will be used to address each of the technical
elements for review. Examples of methods and techniques might include:
* functional analysis
* task analysis
* human error analysis
* timeline analysis
* operations analysis
* physical demands analysis
* verification and validation activities
Provide a statement for each method, indicating how the output from each analysis and
activity will be used. For example: "Task analysis data is used as input to the specification
of human-machine interface features."
4.4 Intended Tools
Indicate the human factors facilities, equipment and tools that will be used to support the
design. These may include such items as
* simulators,
* laboratories,
* software packages.
4.5 Technical Guides
During development of the detailed design phase of a project, it is expected that various
human factors guides will be used to address such topics as
* alarm annunciation, abbreviations and acronyms;
* panel device selection and layout;
* colour usage; and
* procedure writing.
Whether guidelines are developed specifically for the activity to standardize operational
practices and conventions, or selected from applicable published material, they should be
relevant to the current facility and activity, level of technology, and user population. In
addition, all guides should be comprehensive and up to date.
7
Human Factors Engineering Program Plans C-276
5 Processes and Procedures
5.1 General
To ensure consistency across the various work elements of the Human Factors Engineering
Program Plan, identify the steps required for its implementation.
5.2 Timelines
On a timeline, plot the activities related to human factors to show their place within the
project development cycle. Reference to the master project schedule may be appropriate if it
incorporates information relevant to the purposes of the timeline.
5.3 Documentation
Specify how human factors data will be incorporated into the existing design documentation
structure for the project. For large projects, a document hierarchy diagram should be
included to illustrate this incorporation.
5.4 Disposition of Human Factors Issues
Determine a reasonable method for recording, categorizing, tracking, and responding to the
issues and recommendations that arise during implementation of the plan. Development of
the processes and procedures for this aspect of the plan should take into account the
ultimate goals of the human factors program, as well as any anticipated limitations to those
goals.
Provide a description of how tracking of unanticipated human factors issues will be
conducted to ensure consideration in development of future Human Factors Engineering
Program Plans. It is anticipated that project groups affected by the recommendations of the
human factors team may, at times, disagree with those recommendations. The process for
resolving differences of opinion that might be generated by human factors issues should
include an explanation of the authority structure to clarify how and by whom final decisions
are to be made.
5.5 CNSC Contact
Include a proposal for maintaining contact with CNSC staff during plan implementation,
listing proposed submissions, meetings, and communications processes.
8
C-276 Human Factors Engineering Program Plans
BIBLIOGRAPHY
1. Nuclear Safety and Control Act and its regulations
2. P119, Policy on Human Factors
3. IEEE 1023-1988, Institute for Electrical and Electronics Engineering Guide for the
Application of Human Factors Engineering to Systems, Equipment and Facilities of
Nuclear Power Generating Stations
4. IEC 964, Design for Control Rooms of Nuclear Power Plants, 1989
5. CNSC Compliance Program 12280
6. Stramler, J. H. Jr. Dictionary for Human Factors/Ergonomics, Boca Raton: CRC Press,
1992
7. Wierenga, D. et al. Procedure Writing: Principles and Practices, Columbus: Battelle
Press, 1993
8. Goldstein, I. L. Training in Organizations: Needs Assessment, Development, and
Evaluation (2nd ed.) Pacific Grove: Brooks/Cole Pub. Co., 1986
9. AECB INFO0605, Human Factors Guides
9
DRAFT
REGULATORY
GUIDE
Guide to Human
Factors Verification
and Validation Plans
C-278
Issued for public consultation by the
Canadian Nuclear Safety Commission
March 2001
DRAFT REGULATORY GUIDE
Guide to Human Factors Verification
and Validation Plans
C-278
Issued for public consultation by the
Canadian Nuclear Safety Commission
March 2001
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory Policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory Standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the
regulatory requirements.
Regulatory Guide: a document that provides guidance or describes characteristics or
practices that the CNSC recommends for meeting regulatory requirements or improving
administrative effectiveness.
Regulatory Notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be
acted upon in a timely manner.
Regulatory Procedure: a document that describes work processes that the CNSC follows
to administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
DRAFT REGULATORY GUIDE
Human Factors Verification and Validation Plans
C-278
March 2001
About this Document
The draft regulatory guide Human Factors Verification and Validation Plans assists licensees and
licence applicants in preparing an effective Human Factors Verification and Validation Plan. The
goal of this Plan is to establish that the human factors elements of a project or activity, that is
licensed or licensable by the Canadian Nuclear Safety Commission, have been adequately
addressed pursuant to document P119 , Policy on Human Factors.
Comments
The CNSC invites interested persons to assist in the further development of this draft regulatory
document by commenting in writing on the document's content and potential usefulness. Please
respond by June 29, 2001. Direct your comments to the postal or e-mail address below,
referencing file 1-8-8-278.
The CNSC will take the comments received on this draft into account when developing it further.
These comments will be subject to the provisions of the federal Access to Information Act.
Document availability
This document can be viewed on the CNSC Internet site at (www.nuclearsafety.gc.ca). To order
a printed copy of the document in English or French, please contact:
Operations Assistant
Corporate Documents Section
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 996-9505
Facsimile: (613) 995-5086
E-mail: reg@cnsc-ccsn.gc.ca
i
Guide to Human Factors Verification and Validation Plans C-278
CONTENTS
About this Document . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Document availability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1 Regulatory Framework . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2 Licensing Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
2 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
3 Basic Elements of the Verification and Validation Plan . . . . . . . . . . . . . . . . . . . . . . 3
3.1 Basis and Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.1.1 Scope and objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
3.1.2 Background Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.2 Verification of Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3 Validation of Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3.1 Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3.2 Location . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3.3 Techniques and Tools . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
3.3.4 Participants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.3.5 Participant Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.3.6 Performance Measurement in Validation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
3.3.7 Data Collection and Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
Bibliography . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
ii
C-278 Guide to Human Factors Verification and Validation Plans
Purpose
This document assists licensees and licence applicants in preparing an effective Human Factors
Verification and Validation Plan. The goal of this Plan is to establish that the human factors
elements of a project or activity, that is licensed or licensable by the Canadian Nuclear Safety
Commission (CNSC), have been adequately addressed.
Scope
This guide provides the elements of an effective plan for the verification and validation of the
human factors elements of activities licensed or licensable by the CNSC. It is primarily directed
towards the activities of Class I and Class II nuclear facilities, and uranium mines and mills.
This guide does not directly prescribe the process for developing a Human Factors Verification
and Validation Plan. The information provided here, used in conjunction with the C276, Guide
to Human Factors Engineering Program Plans, will assist in the creation of effective Verification
and Validation Plans.
1 Background
1.1 Regulatory Framework
The CNSC is the federal agency that regulates the use of nuclear energy and materials to
protect health, safety, security, and the environment, and to respect Canada's international
commitments on the peaceful use of nuclear energy.
The NSC Act requires persons or organizations to be licensed by the CNSC in order to
carry out the activities referred to in Section 26 of the Act, unless otherwise exempted. The
associated regulations stipulate prerequisites for CNSC licensing, and the obligations of
licensees and workers.
The Nuclear Safety and Control (NSC) Act and its regulations contain several provisions
that are intended to ensure that interfaces between humans and items involving nuclear
substances, prescribed equipment, or nuclear facilities will occur without unacceptable
impacts on persons, the environment, or national security.
References can be found in the following paragraphs of the General Nuclear Safety and
Control Regulations:
* An application for a licence shall contain (a description of) "the applicant's
organizational management structure insofar as it may bear on the applicant's
compliance with the Act and the regulations made under the Act, including the
internal allocation of functions, responsibilities and authority" (paragraph 3(l)(k)).
* An application for a licence shall contain, "at the request of the Commission, any
other information that is necessary to enable the Commission to determine whether
the applicant (i) is qualified to carry on the activity to be licensed or (ii) will, in
carrying our that activity, make adequate provision for the protection of the
1
Guide to Human Factors Verification and Validation Plans C-278
environment, the health and safety of persons and the maintenance of national
security and measures required to implement international obligations to which
Canada has agreed" (paragraph 3(n)).
* "Every licensee shall ensure the presence of a sufficient number of qualified workers
to carry on the licensed activity safely and in accordance with the Act, the regulations
and the licence" (paragraph 12(l)(a)).
* "Every licensee shall take all reasonable precautions to protect the environment and
the health and safety of persons and to maintain security" (paragraph 12(l)(c)).
1.2 Licensing Process
The Canadian Nuclear Safety Commission typically applies a phased process to its licensing
of nuclear facilities and activities. For major facilities, this process begins with an
assessment of the environmental impacts of the proposed project, and proceeds
progressively through site preparation, construction, operation, decommissioning, and
abandonment phases.
The NSC Act and Regulations require licence applicants to provide certain information at
each licensing stage. The type and level of detail of this information will vary to
accommodate the licensing stage and specific circumstances.
Upon receipt of a complete application, CNSC staff review the application to determine
whether the applicant is qualified to carry on the proposed activity, and has made adequate
provision for the protection of the environment, the health and safety of persons, and the
maintenance of national security and the measures required to implement international
obligations to which Canada has agreed. Safe and reliable human performance plays a major
role in overall system safety. As part of licence application reviews, CNSC staff assess
whether the applicant has made adequate provision for human capabilities and limitations
(human factors) as they relate to the safe conduct of the proposed activity.
2 Introduction
The extent to which human factors aspects have been considered, and the effectiveness of that
consideration, is best demonstrated through the implementation of a Verification and Validation
Plan. For the purposes of this guide, the terms verification and validation are defined as follows:
Verification
The process of demonstrating that equipment and systems have been designed as specified,
and that adherence to human factors guidelines has been maintained.
2
C-278 Guide to Human Factors Verification and Validation Plans
Validation
The process of determining the degree to which the human-machine system design and
supporting mechanisms facilitate the achievement of overall safety and operational goals.
A Verification and Validation Plan documents the set of activities within a specific project that
will be carried out to demonstrate that the human factors elements of activity and project designs
pertaining to nuclear facilities conform to human factors design principles. This type of plan is
expected in order to support a Human Factors Engineering Program Plan as described in C-276.
This will ensure that the design enables personnel to perform their tasks safely and to meet
operational goals. A well designed and implemented Verification and Validation Plan is the most
effective indicator of overall performance short of an undesirable "lessons learned" outcome.
3 Basic Elements of the Verification and Validation Plan
An effective Verification and Validation Plan includes comprehensive information about:
* the basis and objectives of the Plan,
* verification of design, and
* validation of design.
3.1 Basis and Objectives
The basis for the Verification and Validation Plan depends on its objectives. In order to
facilitate review by CNSC staff, provide clear definitions of the basis and objectives in the
Plan, including scope and background information.
3.1.1 Scope and objectives
This section of the Verification and Validation Plan will reflect the overall scope and
objectives of the project, but general considerations for this section should include:
* impacted facility areas (e.g., main control rooms, instrument rooms, secondary
control rooms)
* human-machine interface systems and components involved
* allocation of function (i.e., to humans, to automated systems, and between
team members)
* the design phase at which the Verification and Validation Plan will be
implemented
3
Guide to Human Factors Verification and Validation Plans C-278
3.1.2 Background Information
Describe relevant background information about the design, such as any previous
Verification and Validation Plans that have been completed or review activities that
have already taken place.
3.2 Verification of Design
Provide an outline of the approach that will be used to conduct human factors design
verification. Typically, this activity will involve a comparison of each human-machine
system component against appropriate human factors principles, guidelines, and standards.
3.3 Validation of Design
Provide information about the following elements of the validation process:
* Approach
* Location
* Techniques and Tools
* Participants
* Participant Training
* Performance Measurement in Validation
* Data Collection and Analysis
3.3.1 Approach
Provide an outline of the approach that will be used to conduct validation of the
integrated systems associated with the design.
3.3.2 Location
Identify the location of the validation trials.
3.3.3 Techniques and Tools
Validation of integrated systems is accomplished by evaluating task accomplishment
using appropriate validation tools.
Tabletop analysis, walk-throughs using comprehensive drawings, photographs,
prototypes, mock-ups, full-scale simulators, or other techniques appropriate to the
nature of the project may be used.
4
C-278 Guide to Human Factors Verification and Validation Plans
3.3.4 Participants
Identify the participants by job type (i.e., operator, engineer, shift supervisor) who
will be involved in the validation exercises. The number of participants should be
indicated.
3.3.5 Participant Training
It is expected that some training of participants will be necessary. Provide information
about the level and nature of training that will be provided.
3.3.6 Performance Measurement in Validation
Clearly state the technical basis for performance measures and acceptance criteria.
Performance Measures
Present a general discussion of the categories of performance measures that will be
used for validation activities (e.g., time, accuracy, frequency, amount achieved or
accomplished, etc.).
Acceptance Criteria
For both objective and subjective performance measures, include clear statements of
how acceptance criteria relevant to those measures will be derived.
3.3.7 Data Collection and Analysis
Effective validation requires appropriate collection and analysis of data. Describe the
data collection methods that will be used and how the results will be analyzed.
5
Guide to Human Factors Verification and Validation Plans C-278
BIBLIOGRAPHY
1. Nuclear Safety and Control Act and its Regulations
2. P119, Policy on Human Factors
3. IEEE 1023-1988, Institute for Electrical and Electronics Engineering Guide for the
Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear
Power Generating Stations
4. IEC 964, Design for Control Rooms of Nuclear Power Plants, 1989
5. CNSC Compliance Program 12280
6. Stramler, J. H. Jr. Dictionary for Human Factor/Ergonomics, Boca Raton: CRC Press, 1992
7. Wierenga, D. et al. Procedure Writing: Principles and Practices, Columbus: Battelle Press,
1993
8. Goldstein, I. L. Training in Organizations: Needs Assessment, Development, and Evaluation
(2nd ed.) Pacific Grove: Brooks/Cole Pub. Co., 1986
9. AECB INFO0605, Human Factors Guides
6
Canadian Nuclear Commission canadienne
Safety Commission de sûreté nucléaire
REGULATORY
GUIDE
Computer Programs
Used in Design and
Safety Analyses of
Nuclear Power Plants
and Research Reactors
G-149
October 2000
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the regulatory
requirements.
Regulatory guide: a document that provides guidance or describes characteristics or practices
that the CNSC recommends for meeting regulatory requirements or improving administrative
effectiveness.
Regulatory notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be acted
upon in a timely manner.
Regulatory procedure: a document that describes work processes that the CNSC follows to
administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
REGULATORY GUIDE
Computer Programs Used in
Design and Safety Analyses of
Nuclear Power Plants and Research Reactors
G-149
Published by the
Canadian Nuclear Safety Commission
October 2000
Computer Programs Used in Design and Safety Analyses of
Nuclear Power Plants and Research Reactors
Regulatory Guide G-149
Published by the Canadian Nuclear Safety Commission
© Minister of Public Works and Government Services Canada 2000
Extracts from this document may be reproduced for individual use without permission provided
the source is fully acknowledged. However, reproduction in whole or in part for purposes of
resale or redistribution requires prior written permission from the Canadian Nuclear Safety
Commission .
Catalogue number CC173-3/2-149E
ISBN 0-662-29526-9
Ce document est également disponible en français.
Document availability
The document can be viewed on the CNSC website. Copies in English or French may be ordered
using the contact information below:
Communications Division
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 995-5894 or 1-800-668-5284 (Canada only)
Facsimile: (613) 992-2915
E-mail: info@cnsc-ccsn.gc.ca
Website: www.nuclearsafety.gc.ca
October 2000 G-149
TABLE OF CONTENTS
1.0 PURPOSE........................................................................................................................... 1
2.0 SCOPE ................................................................................................................................ 1
3.0 BACKGROUND ................................................................................................................ 1
3.1 Regulatory framework ........................................................................................................ 1
3.2 Licensing process ............................................................................................................... 1
3.3 Relevant legislation ............................................................................................................ 2
3.4 Commitment to quality ....................................................................................................... 3
4.0 DEVELOPING A COMPUTER PROGRAM .................................................................... 3
4.1 Developing a computer program in phases ........................................................................ 3
4.2 Development phases ........................................................................................................... 3
4.3 Guidelines for verification and design review of development phases .............................. 3
4.4 Guidelines for validation .................................................................................................... 4
4.5 Overview of verification and design review....................................................................... 4
4.6 Verification ......................................................................................................................... 4
4.7 Verifying the requirements specifications .......................................................................... 4
4.8 Verifying the program design ............................................................................................. 5
4.9 Verifying the coding ........................................................................................................... 5
4.10 Verifying the program integration ...................................................................................... 5
4.11 Performing design review ................................................................................................... 6
4.12 Performing validation ......................................................................................................... 7
5.0 MAINTENANCE ............................................................................................................... 9
5.1 Configuration management ................................................................................................ 9
5.2 Change control.................................................................................................................. 10
6.0 USING COMPUTER PROGRAMS ................................................................................ 10
6.1 Overview........................................................................................................................... 10
6.2 Uncertainty analysis ......................................................................................................... 11
6.3 User prerequisites ............................................................................................................. 11
6.4 User support...................................................................................................................... 12
6.5 User options ...................................................................................................................... 12
6.6 Verifying the application process ..................................................................................... 12
7.0 DOCUMENTATION ........................................................................................................ 13
7.1 Overview........................................................................................................................... 13
7.2 Creating development documentation .............................................................................. 13
7.3 Application documentation .............................................................................................. 14
III
G-149 October 2000
8.0 EXISTING COMPUTER PROGRAMS .......................................................................... 15
9.0 PROCUREMENT OF COMPUTER PROGRAMS ........................................................ 15
GLOSSARY.................................................................................................................................. 16
IV
October 2000 G-149
COMPUTER PROGRAMS USED IN
DESIGN AND SAFETY ANALYSES OF
NUCLEAR POWER PLANTS AND RESEARCH REACTORS
1.0 PURPOSE
This regulatory guide is intended to provide guidance to licensees involved in the
development, maintenance and use of computer programs used in the design and safety
analysis of nuclear power plants and research reactors so that a high degree of confidence
may be placed in both the programs and the results of their application.
2.0 SCOPE
This guide applies to licensees whose computer programs are used in:
designing or supporting the design of a nuclear power plant or research reactor
analyzing operational transients, incidents or accidents.
This guide does not apply to operational control systems software.
For computer programs developed before the effective date of this regulatory guide, the
degree of applicability is specified in section 8 of this guide.
3.0 BACKGROUND
3.1 Regulatory framework
The Canadian Nuclear Safety Commission (CNSC) is the federal agency that regulates
the use of nuclear energy and materials to protect health, safety, security and the
environment, and to respect Canada s international commitments on the peaceful use of
nuclear energy.
The Nuclear Safety and Control Act ( the Act ) requires persons or organizations to be
licensed by the CNSC for carrying out the activities referred to in Section 26 of the Act,
unless otherwise exempted. The associated regulations stipulate prerequisites for CNSC
licensing, and the obligations of licensees and workers.
3.2 Licensing process
The CNSC typically applies a phased process to its licensing of nuclear facilities and
activities. For major facilities, this process begins with a consideration of the
environmental impacts of the proposed project, and proceeds progressively through site
preparation, construction, operation, decommissioning and abandonment phases.
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The Nuclear Safety and Control Act and regulations require licence applicants to provide
certain information at each licensing stage. The type and level of detail of this
information will vary to accommodate the licensing stage and specific circumstances.
At all licensing stages, applications may incorporate (directly or by reference) new or
previously submitted information, in accordance with legislated requirements and the
best judgement of the applicant. An application that is submitted at one licensing stage
can become a building block for the next stage.
Upon receipt of an application that is complete, the CNSC reviews it to determine
whether the applicant is qualified to carry on the proposed activity, and has made
adequate provision for the protection of the environment, the health and safety of
persons, and the maintenance of national security and measures required to implement
international obligations to which Canada has agreed. If satisfied, the CNSC may issue,
renew, amend or replace a licence that contains relevant conditions. Typically, this
licence will incorporate the applicant s undertakings, and will contain other conditions
that the CNSC considers necessary.
3.3 Relevant legislation
The CNSC uses this guide to assess information submitted as part of a licence
application. Specifically, the General Nuclear Safety and Control Regulations, under
paragraph 3(l)(i), requires that applicants provide a description and the results of any
test, analysis or calculation performed to substantiate the information included in the
application. In addition, section 6 of the Class I Nuclear Facilities Regulations requires
that the application for a licence to operate a Class 1 nuclear facility contains (b) a
description of the systems and equipment at the nuclear facility, including their design
and their design operating conditions; and (c) a final safety analysis report
demonstrating the adequacy of the design of the nuclear facility.
It is the licensee s responsibility to ensure that the computer programs used in design and
safety analyses of nuclear power plants and research reactors and the output of these
programs are reliable and adequate for their intended applications. These objectives can
be attained either by:
(a) using the methods of development, maintenance and use of computer programs
suggested in this guide, or
(b) applying other methods that are equally, or more, effective.
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3.4 Commitment to quality
An organization s commitment to software quality is important, and it is the
responsibility of licensees to develop a quality-assurance program for the development,
maintenance and use of computer programs.
Adherence to relevant standards referred to in this guide, where appropriate, is one of the
important elements in producing and maintaining high-quality computer programs.
4.0 DEVELOPING A COMPUTER PROGRAM
4.1 Developing a computer program in phases
Develop computer programs in phases, using the output of one phase to help set the
requirements for one or more of the phases that follow. Typical development phases (see
Glossary for definitions) include the following:
problem definition
requirements specification
program design
coding
program integration
validation
4.2 Development phases
Each development phase should consist of the following elements:
define the input and requirement specifications the functions to be performed
and the expected outputs based on the preceding phase;
design, develop or test, as stated in the specifications;
perform verification and design review, where appropriate, to ensure that the
product matches the requirement specifications defined for the phase. If a phase
cannot be verified until after more than one phase is complete, each phase should be
verified iteratively.
4.3 Guidelines for verification and design review of development phases
With the exception of the validation phase, guidelines for the verification and design
review of development phases are provided in sections 4.5 through 4.11 of this guide.
Although there are no explicit guidelines for other development activities for these
phases, the important elements to be considered in each development phase may be
obtained from these guidelines.
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4.4 Guidelines for validation
Guidelines for validation are provided in section 4.12 of this guide. Although validation
has been described as a development phase, the activities of this phase may require
extension beyond development phases to ensure that the computer program is validated
for specific applications.
4.5 Overview of verification and design review
Perform verification and design review to show whether or not a program meets its
requirements specifications. Specifically, these activities show the following:
The mathematical equations adequately reflect the phenomena and processes of the
physical system.
The program design correctly reflects the mathematical equations.
The program accurately reflects the design.
The program s results satisfy its requirements specifications such that the results
mirror the behaviour of the physical system.
Prepare a plan of the activities for verification and design review early during program
development, for example, in the requirements specification phase. The plan will
typically include the objectives, verification and design review approach, the schedule,
and the project organization and management.
Review the verification and design review plan on a set schedule, updating the plan as
necessary.
4.6 Verification
Assign verification tasks to the developer of a computer program and independent,
qualified personnel.
Place the outputs to be verified under configuration control (refer to section 5 of this
guide) before starting verification.
4.7 Verifying the requirements specifications
Demonstrate that:
the specifications conform to the standards adopted for the program development so
that, for example, all required sections are present and each section contains all
required information;
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the specifications include the following detail:
- all required functions;
- all program input and output requirements, described in enough detail for a
program to be designed;
- identification of development standards;
- acceptance criteria;
the specifications are clear and unambiguous.
4.8 Verifying the program design
Demonstrate that:
the design conforms to the standards adopted for the development of the program so
that, for example, all required sections are present and each section contains all the
required information;
the design can be traced to the requirements specifications, so that all requirements
are reflected in the design and all design features stem from these requirements;
the design is complete, showing the required functions and specifying the program
inputs and outputs, the operational environment and the processing steps;
the design is internally consistent;
the design is clear and unambiguous.
4.9 Verifying the coding
Demonstrate that the source program:
conforms to programming and language standards adopted for the development of
the computer program;
does not include unnecessarily complex coding;
does include error-checking capabilities;
does include logic that is consistent with the design specifications.
4.10 Verifying the program integration
Demonstrate that:
the integrated program conforms with the requirements of the operating system;
the integrated program interfaces properly with external files;
the program links correctly for example, all modules and subroutines are properly
linked;
the control language is correct;
the processing and transmission of data between modules are correct.
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4.11 Performing design review
Perform a design review in at least the following phases:
requirements specifications
program design
Include representatives of the following groups in the design review team:
developers
programmers
end-users
a quality-assurance group
The design review team should include independent members with well-recognized
expertise in review areas, for example, experts in the physical sciences, mathematics,
computer science and software programming.
In the design review of the requirement specifications, demonstrate that:
the technical description of the problem to be solved is correct, complete and
consistent with the statement of the problem;
the solution methods are appropriate, based on state-of-the-art knowledge.
During the design review of the requirement specifications, review the following:
the scope of the problem, the specifications of the physical system and the
identification of the initial and boundary conditions relevant to the phenomena to be
modelled;
knowledge of the physical system;
identification and understanding of the phenomena;
the modelling concept and the models of the physical system and phenomena,
including assumptions, methods of approximation and simplification;
the limitations of the models, and what these limitations imply;
the analytical and numerical solution methods, including the numerical techniques
and algorithms;
the requirement specifications and instructions for coding the computer program.
The review of the computer program design aims to demonstrate that all program
requirements are implemented correctly in the design, that all aspects of the design can
be traced to the requirements and that the design is feasible, consistent and based on
state-of-the-art knowledge.
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During the design review of the computer program, review the following:
program conceptual design
basic logic
system flow diagrams
numerical techniques and algorithms
operational environment and program interface
In the design review of the computer program, demonstrate that:
the numerical techniques are appropriate for the types of problems to be solved, as
well as for the treatment of boundary or inter-phase conditions and special
situations, such as singularities;
the mathematical equations have been transformed correctly into a numeric scheme
of solutions, and the steps in the solution design are in the correct sequence;
the user options and their restrictions are clearly described.
4.12 Performing validation
Perform validation to determine the accuracy of the computer program s predictions, and
to help determine the uncertainties discussed in section 6.2 of this guide.
Develop a validation plan early in the development of a computer program. The
validation plan should include:
the objectives and scope of the validation;
the validation approach, including the method of validation, requirements to be
validated, acceptance criteria for each requirement, and method for evaluation of
validation results;
the basis for the selection of validation cases, and the specifications for each case;
the validation database;
the validation procedure, including the validation result reporting procedure.
Review the validation plan on a set schedule, updating the plan as necessary.
Assign program validation tasks to qualified persons, specifically developers and those
who are familiar with the program, the sources of the data being used to validate the
program and the program s intended use.
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Validation tests are meant to determine:
a program s accuracy;
how well a program and its conceptual models reflect physical processes or systems;
a program s capabilities and limits, including the range of parameters to which the
program may properly be applied.
During validation, systematically test the performance of conceptual models, empirical
correlations and the integrated program. Relevant are tests against analytical solutions,
operational data, and data from separate effects and integral effect tests. Comparisons
against validated programs may be included.
To perform systematic testing:
develop a test plan;
identify and rank important phenomena and parameters;
identify the models, empirical correlations or components of the program that are to
be tested;
identify suitable existing experimental data, as well as data to be obtained from new
experiments;
assess the program using separate effect and integral effect tests, operational
measurements, analytical solutions or the results of one or more validated programs;
assess the sensitivity to the input options;
evaluate validation results.
During validation, examine the following key areas when applicable:
the assumptions made to simplify the physical system;
the theoretical and experimental bases for the models and empirical correlations, as
well as their applicability range;
the compatibility of the ranges of the parameters, as well as the geometrical and
phenomenological similarities between the system being simulated and the
experiments;
the ability of the code to predict behaviour during integrated tests, to show that (1)
no unwarranted interaction occurs between various models, and (2) the code
accurately simulates how the system s parts interact;
the sensitivity of individual models and empirical correlations, as well as the
integrated code, to variations in key parameters or factors, particularly near the end
of their allowable ranges and near-critical values such as singularities or
discontinuity;
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October 2000 G-149
potential errors arising from the use of the models and empirical correlations outside
their intended ranges;
the variation in code predictions when using various options of the alternative
models and empirical correlations as recommended in the code documentation (refer
to section 7 of this guide).
Note all validation results, clearly identifying the program s abilities, limits and ranges of
applicability.
During validation, identify improvements that should be made to the program. Identify,
as well, any need to enhance the validation data, including experimental data.
Write a report, based on validation results, that shows the correct and appropriate use of
the program.
Verify that the validation process has been conducted in accordance with the validation
plan, and that the validation results are accurately reported.
5.0 MAINTENANCE
5.1 Configuration management
Establish procedures for computer program maintenance and change control.
Adopt practices that protect program integrity. Clearly define the roles and
responsibilities of those who are responsible for a program s integrity, as well as those
who maintain and use it.
Maintain a configuration management system to help ensure program integrity and to
keep track of a program s versions and components.
Typical components of a program to be configured include:
source program
object program
executable program
processors and operating systems
files that control the generating of executable code from source code
input data files
documentation
tools that support program development or maintenance
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Configuration management involves:
using a naming convention to uniquely identify the functional and physical
characteristics of each configuration component
controlling changes to the configuration components
recording and reporting change processing and implementation status
verifying compliance with configuration requirements
5.2 Change control
Set up a change control system to control changes to each configuration item. A change
control system should include:
a designated person responsible for all changes to a computer program;
a change approval group whose members represent developers, users and those who
performed the program s verification and validation;
procedures to:
- review proposed changes to a configuration item;
- approve changes;
- implement changes;
- document changes and their rationales;
- verify that a change was implemented correctly;
- assess the impact of each change on the use of the program and on the quality
of its predictions;
- make any necessary revisions to existing documentation.
6.0 USING COMPUTER PROGRAMS
6.1 Overview
Establish guidelines and procedures for the use of computer programs. These should
include guidelines for user prerequisites and the designation of persons responsible for
ensuring the correct and appropriate use of computer programs.
Apply a program only to a problem or set of problems for which the program has been
validated and designed to solve (only within the range of applicability identified in the
verification and validation reports). If the program is used outside its range of validation,
the validity of the extrapolation should be justified.
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October 2000 G-149
Minimize as far as possible the program s user effects. An illustration of user effects is: if
two experienced users, with the same version of a computer program, perform a test or
analyze a problem and obtain significantly different results, then the program may be
user-dependent. The differences can be attributed to the use of input options, such as
those for time step control, nodalization schemes, correlations, coefficients, or
parameters.
Report errors and deficiencies in a program or its application to the person who is
responsible for the maintenance of the program. Correct the errors and deficiencies, using
the change control procedures listed in section 5.2 of this guide.
6.2 Uncertainty analysis
Evaluate a computer program s results to determine the uncertainties in each type of
application of the program.
Account for potential uncertainties identified in the development and use of the program,
including those due to:
simplifications and assumptions made to compensate for lack of knowledge or to
render a problem solvable, including assumptions that simplify equations, empirical
correlations, and physical and system models;
inadequate knowledge of the problem that the computer program will be used to
analyze, such as complex phenomena, scaling effects, and initial and boundary
conditions;
limits to the ability to represent a physical system, such as those presented by
nodalization techniques;
inadequacy of the data used to validate the program, for example, instrumentation
errors and test repeatability;
compensating errors or uncertainties;
propagation of uncertainties, i.e. the uncertainty of the results increases as the
transient progresses.
6.3 User prerequisites
Take steps to ensure that each user of a computer program possesses experience that
matches or exceeds the safety profile of the program s application.
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G-149 October 2000
Important factors in the knowledge and experience needed by the program s user are:
good knowledge and understanding of the information in all application documents
and the verification and design review reports, as well as familiarity with other
information in the design documents;
sufficient experience in using the program for the intended application, and similar
ones, with a good knowledge of program responses to (1) various system and
phenomenon models and to (2) assumptions and changes in important parameters;
extensive knowledge of the problem to be analyzed, for example, knowledge of the
physical system and the phenomena being modelled.
6.4 User support
User support should be available to advise the user on the correct use of a computer
program and proper modellings or simulations for important or large and complex
programs, such as thermal hydraulics system programs or multi-field coupling programs.
Set up a user support team if warranted, and include the following members:
the developer or those who know the program well;
those who performed, or who know a great deal about, the program s verification
and validation;
those who have a thorough knowledge of the system to be modelled and the problem
to be solved.
6.5 User options
User input options should be avoided as far as practicable.
Write guidelines for each user option, and make them available to users.
6.6 Verifying the application process
Verify that the application process is correct by demonstrating that it satisfies the
guidelines for the use of computer programs provided in this section.
If an application process cannot be satisfactorily verified, identify exceptions, and
provide a justification for each.
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7.0 DOCUMENTATION
7.1 Overview
The documentation of a computer program and its associated components should
conform with documentation standards adopted in the development of the program.
Verify the program documents to demonstrate that they are complete, consistent, clear
and unambiguous.
7.2 Creating development documentation
In the development documentation include:
problem definition
program development plan
requirements specifications
design specification
programmer s manual
verification and design review report
In the problem definition, document in detail the statement of the problem and the
rationale and objectives of the program.
In the program development plan, describe the organization, schedule and activities
related to the development, design review, verification and validation of the program.
Also document procedures for updating the program development plan.
In the requirements specifications, clearly identify all program requirements. In addition,
provide the basis for verifying the program design and evaluating the performance of the
program through validation.
In the design specification, specify the elements needed for program coding, including
the logical structure, information flow, models, numerical solution techniques,
discretization method, data structure, supporting software and hardware, and the
operating environment.
In the programmer s manual, describe the program flow and structure, the method of
translating theory into coding, instructions for maintaining and modifying the computer
program and conventions on programming practices, such as variable naming and
program commenting.
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In the verification and design review report, document verification activities, including
the description of the method and results of each verification or design review, the range
of applicability and recommendations to users regarding the capability and limits of the
computer program. Also include the input data used in each verification.
7.3 Application documentation
Application documentation should include:
program abstract
theory manual
user s manual, including specific user guidelines
one or more validation reports
source code
sample input and output
In the program abstract, provide the following information about the program:
the name and version of the program and applicable configuration items;
the program s purpose, capabilities, limitations and operating environment;
a summary of the problem(s) the program is designed to solve;
the names of the organization and key individuals responsible for code
development, support and maintenance of the program.
In the theory manual, describe the basis of the computer program including:
the physical systems to be modelled and their models;
the phenomena to be modelled and their models or empirical correlations;
mathematical formulations of the problem and solution techniques;
assumptions and constraints, and what they imply about the limits of the program s
capabilities and the range of its applicability;
references.
In the user s manual include enough information to run the computer program. Provide
details on the required input data, techniques for restarting the code, diagnostic
information and options available to the user, including:
basic information on compiling, linking and executing the code;
detailed description of all input parameters indicating type and format;
a discussion of pre- and post-processing of the code, including code restarts;
sample input and output files that show representative problems;
a description of possible execution error messages and termination messages.
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October 2000 G-149
In addition, to minimize the risk of inappropriate use of a program, provide user
guidelines on the following topics:
the correct and proper use of the program for each of its applications;
the range of applicability;
the program s capabilities and limitations;
the options available to users, such as model options or choice of nodalization
schemes.
In the validation report, include the following:
validation plan;
validation results;
evaluation of the validation results, including a determination of the computer
program s accuracy and range of uncertainty for each test;
recommendations for improvement to the program;
specification of further experiments;
recommendations for the correct and appropriate use of the program for each
application, range of applicability, and program capability and limitations.
8.0 EXISTING COMPUTER PROGRAMS
Existing computer programs are those used by the CNSC licensee before the effective
date of this document.
Assess and document the extent to which existing programs conform with this guide.
Improve, to the extent practicable, the areas that do not conform with relevant sections of
this guide so that there is an increase in the level of confidence in a program consistent
with the program s impact on safety.
Provide justification for any non-conformance with the guidelines in this guide.
9.0 PROCUREMENT OF COMPUTER PROGRAMS
Establish procedures for the procurement and maintenance of computer programs.
Ensure that all computer programs destined for use by a CNSC licensee, but developed
outside a CNSC licensee s organization, meet or exceed the criteria of this guide.
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GLOSSARY
coding
The step in which a computer program design is transformed into a sequence of
machine-readable instructions suitable for processing by a computer.
computer program
A sequence of machine-readable instructions suitable for processing by a computer; also referred
to in this document as a program.
developer
A person who writes the code for a computer program, or a portion of one, and who works on its
associated documents.
discretization
A method of approximation of the true mathematical function to be integrated.
problem definition
A detailed description of the problem and the rationale for the decision to develop a computer
program.
program design
The elements required for program coding, including the logical structure, information flow,
models, numerical solution techniques, discretization method, data structure, supporting
software and hardware, operating environment, and other features required to satisfy the
requirements specifications.
program integration
The step in which a source program is integrated with other components, such as library
routines, system linking specifications, operating system control language, external data libraries
and hardware environment to form an integrated, executable program.
requirements specifications
The requirements the program must satisfy and the basis for verifying the program design and
then evaluating the program through validation.
source program
A program that must be compiled, assembled or interpreted before it can be executed.
user
A person who uses a computer program.
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validation
A process by which the results of a computer program are compared with operational
measurements, experimental data, or analytical solutions, to determine the program s accuracy
and uncertainty.
verification
The step in which it is determined whether the products of a given phase of the program
development cycle fulfil the requirements set during the previous phase.
17
Canadian Nuclear Commission canadienne
Safety Commission de sûreté nucléaire
REGULATORY
GUIDE
Emergency Planning at
Class I Nuclear Facilities
and Uranium Mines
and Mills
G-225
August 2001
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the regulatory
requirements.
Regulatory guide: a document that provides guidance or describes characteristics or practices
that the CNSC recommends for meeting regulatory requirements or improving administrative
effectiveness.
Regulatory notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be acted
upon in a timely manner.
Regulatory procedure: a document that describes work processes that the CNSC follows to
administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
REGULATORY GUIDE
Emergency Planning at
Class I Nuclear Facilities and
Uranium Mines and Mills
G-225
Published by the
Canadian Nuclear Safety Commission
August 2001
Emergency Planning at Class I Nuclear Facilities and Uranium Mines and Mills
Regulatory Guide G-225
Published by the Canadian Nuclear Safety Commission
© Minister of Public Works and Government Services Canada 2001
Extracts from this document may be reproduced for individual use without permission provided
the source is fully acknowledged. However, reproduction in whole or in part for purposes of
resale or redistribution requires prior written permission from the Canadian Nuclear Safety
Commission.
Catalogue number CC173-3/2-225E
ISBN 0-662-30688-0
Ce document est également disponible en français.
Document availability
This document can be viewed on the CNSC website (www.nuclearsafety.gc.ca). To order a
printed copy of the document in English or French, please contact:
Communications Division
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 995-5894 or 1-800-668-5284 (Canada only)
Facsimile: (613) 992-2915
E-mail: publications@cnsc-ccsn.gc.ca
August 2001 G-225
TABLE OF CONTENTS
1.0 PURPOSE........................................................................................................................... 1
2.0 SCOPE ................................................................................................................................ 1
3.0 DEFINITION OF EMERGENCY PLAN ....................................................................... 1
4.0 BACKGROUND ................................................................................................................ 2
4.1 Overview of Canada s emergency planning framework .................................... 2
4.2 Regulatory framework ........................................................................................ 2
4.3 Relevant legislation for this document ............................................................... 3
4.4 The CNSC licensing process and emergency planning...................................... 3
5.0 CONTENTS OF EMERGENCY PLANS .......................................................................... 5
5.1 Introduction ........................................................................................................ 5
5.2 Documentation of the emergency plan ............................................................... 5
5.3 Basis for emergency planning ............................................................................ 6
5.4 Personnel selection and qualification ................................................................. 6
5.5 Emergency preparedness and response organizations ........................................ 6
5.6 Staffing levels ..................................................................................................... 6
5.7 Emergency training, drills and exercises ............................................................ 7
5.8 Emergency facilities and equipment .................................................................. 9
5.9 Emergency procedures........................................................................................ 9
5.10 Assessment of emergency response capability ................................................. 11
5.11 Assessment of emergencies .............................................................................. 11
5.12 Activation and termination of emergency responses........................................ 11
5.13 Protection of facility personnel and equipment................................................ 12
5.14 Interface with off-site organizations ................................................................. 12
5.15 Recovery program ............................................................................................ 13
5.16 Public information program ............................................................................. 14
5.17 Public education program................................................................................. 14
III
August 2001 G-225
EMERGENCY PLANNING AT
CLASS I NUCLEAR FACILITIES AND
URANIUM MINES AND MILLS
1.0 PURPOSE
This regulatory guide is intended to help (1) applicants for operating licences for Class I
nuclear facilities and (2) applicants for licences in respect of uranium mines and mills to
develop emergency measures that satisfy the following:
paragraph 6(k) of the Class I Nuclear Facilities Regulations and subparagraph
3(c)(x) of the Uranium Mines and Mills Regulations; and,
the requirements of subsection 24(4) of the Nuclear Safety and Control Act, by
demonstrating that the applicant will, in carrying on the proposed activity, make
adequate provision for the protection of the environment, the health and safety of
persons, and the maintenance of national security and measures required to
implement international obligations to which Canada has agreed.
The guide is also intended to aid the Canadian Nuclear Safety Commission (CNSC) in its
evaluations of the adequacy of the emergency measures proposed by the above-
mentioned applicants.
2.0 SCOPE
This guide applies to applicants for a CNSC licence to operate a Class I nuclear facility,
and to applicants for uranium mine and mill licences.
The guide describes and discusses the elements of emergency preparedness and response
that licence applicants should typically consider when they are developing plans to
prevent or mitigate the effects of accidental releases from a Class I nuclear facility or a
uranium mine or mill.
3.0 DEFINITION OF EMERGENCY PLAN
In this guide, the term emergency plan refers to the documented measures required of
applicants and licensees under paragraph 6(k) of the Class I Nuclear Facilities
Regulations and subparagraph 3(c)(x) of the Uranium Mines and Mills Regulations.
Accordingly, an emergency plan for a Class I nuclear facility or a uranium mine or mill
consists of a description of a proposed or actual program to cope with accidental releases.
This program encompasses both emergency preparedness and emergency response
measures. It ensures that appropriate emergency response capabilities are developed and
maintained available for use.
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G-225 August 2001
Emergency plans for Class I nuclear facilities or uranium mines and mills should be
commensurate with the complexity of the associated undertakings, and the probability
and potential severity of the emergency scenarios associated with operation of these
facilities.
An emergency plan may consist of one or several documents. It may incorporate
pertinent information directly or by reference. Emergency plans may cover a broad range
of issues. Specifics may vary to accommodate facility-specific needs and circumstances,
regulatory requirements, and the individual preferences of licence applicants.
4.0 BACKGROUND
4.1 Overview of Canada s emergency planning framework
In Canada, the respective roles of the various levels of government in nuclear
emergency preparedness and response derive from legislated responsibilities.
Provincial and territorial governments have the primary responsibility for
protecting public health and safety, property and the environment within their
borders. The federal government regulates the peaceful uses of nuclear energy
in Canada, manages nuclear liability, and supports the responses of provinces
to nuclear emergencies within their boundaries.
The federal government is also responsible for liaisons with the international
community and their diplomatic missions in Canada, for assisting Canadians
abroad, and for coordinating the national response to nuclear emergencies that
occur in foreign countries, but impact on Canada.
Under the administrative framework of the Federal Nuclear Emergency Plan,
all levels of government and various agencies and organizations have
responsibilities for developing and implementing emergency plans to deal with
nuclear emergencies that have impacts outside the bounds of the nuclear
facility licensed by the CNSC.
4.2 Regulatory framework
The CNSC is the federal agency that regulates the use of nuclear energy and
materials to protect health, safety, security and the environment, and to respect
Canada s international commitments on the peaceful use of nuclear energy.
The Nuclear Safety and Control Act ( the Act ) requires persons or
organizations to be licensed by the CNSC for carrying out the activities
referred to in section 26 of the Act, unless otherwise exempted. The associated
regulations stipulate prerequisites for CNSC licensing, and the obligations of
licensees and workers.
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August 2001 G-225
4.3 Relevant legislation for this document
As noted above, the Class I Nuclear Facilities Regulations and the Uranium
Mines and Mills Regulations require that applications for operating and other
licences include certain information related to emergency planning.
Collectively, they stipulate that an application for a licence to operate a Class I
nuclear facility or a uranium mine or mill shall contain a description of the
proposed measures to prevent or mitigate the effects of accidental releases of
nuclear substances and hazardous substances on the environment, the health
and safety of persons, and the maintenance of security, including measures to:
assist off-site authorities in planning and preparing to limit the effects of
an accidental release;
notify off-site authorities of an accidental release or the imminence of an
accidental release;
report information to off-site authorities during and after an accidental
release;
assist off-site authorities in dealing with the effects of an accidental
release; and
test the implementation of the measures to prevent or mitigate the effects
of an accidental release.
4.4 The CNSC licensing process and emergency planning
The CNSC typically applies a phased process to its licensing of nuclear
facilities and activities. For major facilities, this process begins with a
consideration of the environmental impacts of the proposed project, and
proceeds progressively through site preparation, construction, operation,
decommissioning and abandonment phases.
The Act and its regulations require licence applicants to provide certain
information at each licensing stage. The type and level of detail of this
information will vary to accommodate the licensing stage and specific
circumstances.
At all licensing stages, applications may incorporate (directly or by reference)
new or previously submitted information, in accordance with legislated
requirements and the best judgement of the applicant. An application that is
submitted at one licensing stage can become a building block for the next
stage.
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G-225 August 2001
Upon receipt of an application that is complete, the CNSC reviews it to
determine whether the applicant is qualified to carry on the proposed activity,
and has made adequate provision for the protection of the environment, the
health and safety of persons, and the maintenance of national security and
measures required to implement international obligations to which Canada has
agreed. If satisfied, the CNSC may issue, renew, amend or replace a licence
that contains relevant conditions. Typically, this licence will incorporate the
applicant s undertakings, and will contain other conditions that the CNSC
considers necessary, including a condition that incorporates or relates to
emergency planning.
The regulatory reviews by the CNSC of the adequacy of emergency plans will
typically cover the following:
documentation of the emergency plan,
basis for emergency planning,
personnel selection and qualification,
emergency preparedness and response organizations,
staffing levels,
emergency training, drills and exercises,
emergency facilities and equipment,
emergency procedures,
assessment of emergency response capability,
assessment of accidents,
activation and termination of emergency responses,
protection of facility personnel and equipment,
interface with off-site organizations,
recovery program,
public information program, and
public education program.
In accordance with the regulations, licensees may revise their emergency plans
to take into account relevant factors, such as operating experiences or changed
needs or circumstances.
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August 2001 G-225
5.0 CONTENTS OF EMERGENCY PLANS
5.1 Introduction
Although the Class I Nuclear Facilities Regulations and the Uranium Mines
and Mills Regulations require that applications include, for specified purposes,
certain information on emergency planning and response, the pertinent
legislation does not define the required information in detail, nor does it
prescribe a form in which it is to be organized and submitted. The applicant is
responsible for submitting information that is relevant to its needs and
circumstances, that meets the intent of the legislation and that enables
regulatory review.
The following sections illustrate, under key subject headings, the factors that
applicants should take into account when developing emergency plans to
address their specific circumstances.
5.2 Documentation of the emergency plan
The regulations stipulate that information pertaining to the proposed
emergency measures is to be included in the respective licence applications,
and that these applications are also to describe the proposed facility, activities,
substances and circumstances to which the emergency plan and requested
licence are to apply.
Each emergency plan should indicate how its ongoing maintenance and
revision will be controlled. These document control procedures should be
suitably rigorous so as to ensure that the quality and relevance of the plan is
maintained. These procedures should be consistent with control procedures for
other licensing documents.
Any relevant agreements with other agencies or parties regarding emergency
preparedness and response should also be referenced in or annexed to the
emergency plan.
To provide assurance that a proposed emergency plan is supported by the
licence applicant s senior management, the plan should be formally and
explicitly approved by management prior to submission for regulatory review.
The submitted plan should indicate that this approval has been received.
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G-225 August 2001
5.3 Basis for emergency planning
Emergency plans for Class I nuclear facilities and uranium mines and mills
should be based on accidental release scenarios that have, or could have,
adverse impact on the environment and the health and safety of on-site staff or
the public. The plans should also be based on those scenarios assumed in the
safety analysis submitted in support of the licensing of the respective facility.
5.4 Personnel selection and qualification
The success of emergency preparedness and response initiatives depends in
part on the competence and actions of the persons involved. To be effective,
these persons must be adequately qualified through training or experience,
must be empowered with the necessary authority, and must be equipped with
adequate resources. Accordingly, emergency plans should describe how the
competence and effectiveness of those persons who are to be involved in
emergency preparedness and response will be assured, such as by application
of selection criteria or qualification measures.
5.5 Emergency preparedness and response organizations
In this document, emergency planning encompasses both emergency
preparedness and emergency response activities. Accordingly, the emergency
plans for Class I nuclear facilities and uranium mines and mills should assign
and define formal responsibilities for developing, maintaining, and
implementing emergency preparedness and emergency response activities. For
both the emergency preparedness and emergency response organizations, the
plan should clearly describe the qualifications, duties, authorities, and
accountabilities of the persons involved, and their respective organizational
and reporting relationships. These descriptions should include all persons with
a significant role, including the emergency response teams involved in first aid,
fire fighting and radiation surveys.
5.6 Staffing levels
To ensure and maintain a credible state of emergency preparedness, nuclear
facilities should be adequately staffed at all times. Emergency plans for Class I
nuclear facilities, or uranium mines and mills, should ensure that sufficient
numbers of qualified personnel are available at all times to maintain the
facilities in a safe condition and to respond effectively to emergencies.
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August 2001 G-225
To provide this assurance, emergency plans should clearly define the levels and
nature of staffing to be maintained, and how this staffing will be assured. In
particular, emergency plans should specify the minimum and optimum levels of
staffing to be maintained for key positions or functions related to emergency
preparedness and response. They should include appropriate arrangements to
compensate for scheduled and unscheduled absences of emergency response
personnel. Typically contingency arrangements consist of designating
responsible alternates, or retaining backup personnel or services from
contractors or other external agencies.
5.7 Emergency training, drills and exercises
Adequate training and testing can help ensure that individuals and
organizations are prepared for emergencies. Accordingly, emergency plans for
Class I nuclear facilities and uranium mines and mills should provide for any
training and testing of individuals or organizational units necessary to assure
and demonstrate that they are qualified and able to completely fulfill their
assigned emergency preparedness and response roles.
An emergency training program for a nuclear facility delivers relevant training
to participants on the substance of the emergency plan, individual and
organizational responsibilities, the use of emergency equipment and facilities,
radiation and personnel protection, emergency communications and exchange
of information, emergency procedures, and cooperation and interaction with
other on-site and off-site authorities.
Emergency training may consist of both formal and informal instruction,
including workplace and classroom instruction, and emergency drills and
exercises.
Drills and exercises are used to train participants as well as test and measure
the effectiveness of emergency training, the quality of emergency response
programs, and the capabilities and performances of people, facilities and
equipment. Emergency drills and emergency exercises usually differ in
complexity and purpose.
7
G-225 August 2001
Emergency exercises simulate emergency events over a minimum of several
hours in order to test the integrated performance of the emergency response
program. When correctly designed and conducted, these exercises
simultaneously measure and demonstrate the preparedness and competence of
participants, the quality of the associated procedures, and the effectiveness of
the administrative framework. Deficiencies that are identified during exercises
can be rectified in a timely manner to provide greater assurance that the
emergency plan can and will be implemented successfully in the event of a real
emergency.
Exercises normally involve a large number of on-site and off-site stakeholders,
including regional, provincial, federal and, where appropriate, international
authorities.
Emergency drills are more limited in scope and purpose, typically involving
testing a procedural or physical component of the emergency response
program. Drills may be conducted as an initial or periodic test, as a supervised
training session, or as an evaluation of a remedial event. For example, after
steps are taken to correct a weakness identified by an emergency exercise, the
effectiveness of the remedial measures may be further evaluated by a drill.
Accordingly, an emergency plan for a Class I nuclear facility or a uranium mine
or mill should clearly describe any proposed training and relevant supporting
information, including:
the objectives and content of the planned training;
how the training is to be delivered to meet objectives;
the qualifications required of training instructors;
the staff positions for which incumbents will be required to undertake
periodic or on-going training;
the training requirements for contractors and off-site organizations (e.g.
firefighters, police personnel, ambulance drivers, hospital staff) that
support or participate in on-site activities insofar as they relate to the
training that is outside their normal professional duties but required for
their role and interface in an on-site emergency (e.g. training on access
requirements or radiation protection);
the objectives, plans, schedules, procedures and assessment criteria for the
conduct of emergency drills and exercises;
the positions that are responsible for managing, planning, and evaluating
drills and exercises;
8
August 2001 G-225
procedures and criteria for evaluating the results of emergency drills and
exercises and for taking follow up actions; and
how the results of training (courses, drills and exercises) will be recorded,
and the records maintained.
5.8 Emergency facilities and equipment
Emergency plans should describe the services, equipment, supplies and
facilities that are to be available to cope with emergencies. These needs will be
determined by facility-specific circumstances.
The facilities that could be needed in an emergency include:
administration facilities,
technical support and control centres,
personnel/public assembly areas,
an emergency operations coordination centre,
a centre to integrate on-site activities with off-site programs,
first aid or medical facilities, and
laboratory facilities.
The following equipment and materials might also be needed:
an emergency source of electrical power;
reference materials, such as accurate versions of charts, maps, plans,
drawings, diagrams, specifications and procedures;
safety and personnel protection equipment and supplies (e.g., fire-fighting,
physical and respiratory protection);
administrative aids, such as status boards and reference materials; and
fixed or portable instruments or equipment, as required, to warn, detect,
measure, monitor, survey, analyze, record, assess, process, treat, transport,
announce, communicate or compute.
If emergency facilities, equipment and materials are to be useful when needed,
they must be in suitable condition. Accordingly, emergency plans should
include provisions to assure that the emergency equipment, facilities or
materials remain in acceptable condition at all times. These provisions could
include inspection, testing, maintenance or replacement, within formal systems
of quality control and inventory control and accounting.
5.9 Emergency procedures
Procedures are important elements of emergency plans, and are fundamental to
the success of emergency preparedness and emergency response programs.
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G-225 August 2001
Emergency preparedness programs typically include procedures for:
conducting emergency exercises;
testing, maintaining and assuring the availability of emergency facilities
and equipment (e.g., sirens, telecommunications equipment, monitoring
equipment);
tracking developments and actions;
educating the public; and
updating the emergency plan and procedures.
Emergency response programs typically include procedures for:
assigning responsibilities and accountabilities;
assessing and classifying emergencies;
assessing source terms and consequences;
activating emergency responses;
implementing emergency responses;
notifying and alerting site personnel and other stakeholders (on-site and
off-site communications);
protecting on-site and off-site emergency response personnel;
assembling, protecting and evacuating personnel;
controlling exposures to radiation, and radioactive and hazardous
substances;
limiting the occurrence and spread of radioactive contamination;
responding to over-exposures, contamination incidents, injuries or
fatalities;
post-accident monitoring and assessments of systems, effluents and
conditions (e.g., observations, tests, measurements, collection of samples,
sample preparation and analysis, reporting of sampling, measurement and
test results);
documenting and controlling the exchange of information;
effecting scheduled shift changes and workplace turnovers;
controlling vehicular and human traffic;
directing, controlling and supporting emergency responses;
implementing corrective actions or remedial measures; and
maintaining the security of nuclear materials.
The emergency plan may incorporate emergency preparedness and response
procedures directly, or it may reference pertinent documents, such as the
facility procedures manual.
10
August 2001 G-225
5.10 Assessment of emergency response capability
CNSC licensees should review their emergency response programs at regular
intervals to ensure that these programs remain updated. Accordingly,
emergency response programs should include document control procedures that
specify who (position or unit) is to review and update the program on a
ongoing basis, and how this is to be done.
The emergency plan should identify the responsible persons and empower them
with the sufficient autonomy and authority. It should specify the frequency of
emergency preparedness audits, how deficiencies are detected and reported,
and how corrective actions are tracked and implemented.
5.11 Assessment of emergencies
When an emergency is detected or suspected, facility personnel must determine
its implications or consequences. Accordingly, emergency plans for Class I
nuclear facilities and uranium mines and mills should describe how the relative
and absolute severity of emergencies are to be determined and categorized. The
plans should describe the methods and procedures to be followed when
assessing pertinent conditions, needs and parameters, such as:
the status, integrity and stability of the associated facilities and their
components;
quantities, concentrations, or release-rates of radiation, contaminants, or
hazardous substances;
on-site and off-site impacts on or threats to health, safety, national security
and the environment;
requirements for routine and contingency supplies, equipment or other
services or resources; and
requirements for compilation and maintenance of records.
5.12 Activation and termination of emergency responses
The emergency plan should describe the procedures for initiating and
terminating responses to both on-site and off-site emergencies associated with
facility operations.The plan should identify the organizations, positions or
individuals who are to be notified in the event of a suspected or actual
emergency, as well as those responsible for activating emergency responses;
for completing the notifications required as part of the emergency response
phase; and for terminating the emergency response phase.
The process for notifying off-site stakeholders of an on-site emergency at a
Class I nuclear facility or a uranium mine and mill should be compatible with
those of any complementary provincial and federal off-site emergency plans.
11
G-225 August 2001
5.13 Protection of facility personnel and equipment
During an emergency, facility personnel and essential equipment must be
adequately protected.
To protect on-site and off-site personnel during a nuclear emergency, a
combination of normal and abnormal measures may be necessary.
Accordingly, emergency plans for Class I nuclear facilities and uranium mines
and mills should provide for both routine administrative controls and any
special personnel-protection measures that could be necessary in the event of
emergencies. These emergency provisions could include:
establishing or designating areas for the emergency assembly of site
personnel;
implementing special administrative measures, such as action levels, to
control radiation doses;
accounting for site personnel;
conducting routine or special radiation surveys;
providing routine or special dosimetry services;
providing search and rescue, decontamination, and first aid services; and
providing any other emergency equipment, instruments, materials,
facilities and services that are necessary to assure that on-site and off-site
personnel are adequately protected.
In addition to providing for the protection of people during emergencies,
emergency plans should identify essential equipment, and describe how its
operation and effectiveness during emergencies are assured. Essential
equipment consists of the equipment required to detect, assess, or cope with
potential emergencies. The emergency plan should describe the proposed
procedures or systems for protecting essential equipment.
5.14 Interface with off-site organizations
To assure that responses to emergencies will be consistent, efficient and
effective, it is desirable that all jurisdictions, organizations and persons
involved in the administration and delivery of emergency preparedness and
response programs cooperate and coordinate with each other. Accordingly, the
operators of Class I nuclear facilities and the operators of uranium mines and
mills may need to coordinate and cooperate with other jurisdictions and
organizations in the event of a facility emergency with off-site safety
implications. To assure effective interfaces between facility personnel and
external stakeholders in such emergencies, the facility emergency plan and any
regional, provincial and national emergency preparedness and response plans
12
August 2001 G-225
and programs dealing with any off-site implications of facility emergencies,
must be suitably compatible. Thus, where necessary, facility emergency plans
for Class I nuclear facilities and uranium mines and mills should be compatible
with the off-site emergency preparedness and response programs of regional,
provincial and federal jurisdictions and organizations.
In particular, facility emergency plans should:
identify the jurisdictions, organizations, organizational units or persons
that could be formally involved in emergency preparedness and response
activities pertaining to facility emergencies with off-site impacts;
describe the administrative procedures, processes and resources whereby
facility personnel and personnel of external organizations, units and
jurisdictions will cooperate and interact in emergency preparedness and
response activities pertaining to off-site impacts from facility
emergencies;
reference or incorporate any agreements between the facility operator and
other jurisdictions, organizations or personnel, regarding cooperation and
interaction in emergency preparedness and response activities to address
off-site impacts from facility emergencies; and
describe how the facility operator will ensure that the resources required to
cooperate in off-site emergency preparedness and response matters will be
available and provided when needed.
5.15 Recovery program
An emergency plan for a Class I nuclear facility or a uranium mine or mill
should include provisions to restore the facility to normal operations after the
termination of an emergency, including, as one of its provisions, the
requirement to establish a recovery organization and to develop a recovery
plan. The recovery plan should:
identify the organizational unit individual or group that is
responsible for directing and assuring effective recovery;
define the responsibilities of the organization unit that is responsible for
assuring recovery;
identify and describes the resources (personnel, facilities, equipment) that
are to be available for recovery purposes;
describe how personnel will be protected when implementing or assessing
the recovery program (e.g., personnel protection measures for entry into
hazardous areas); and
provide for post-accident assessments of the causes, details, impacts or
consequences of accidents.
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G-225 August 2001
5.16 Public information program
The emergency plan should include, directly or by reference, appropriate
provisions to communicate pertinent information to the public during an
emergency. For complex facilities such as nuclear power plants, these
provisions should consist of communications policies and strategies that
formally define the roles, responsibilities and essential qualifications of
communications personnel. These arrangements should require those
responsible for communicating key information to the public to possess
appropriate training, experience, and skills in public relations. During an
emergency at a nuclear facility, the duties of its communications personnel can
range from preparing or editing communications materials to coordinating and
controlling the release of pertinent information to external stakeholders, such as
the public, the media and other emergency organizations.
5.17 Public education program
The emergency plan for a Class I nuclear facility or a uranium mine or mill
should take into account any need for education of the public with respect to
emergencies at the facility, and their implications. Where a public education
program appears warranted to assure public understanding of how to
participate and cooperate effectively in the event of an accident at the facility,
the facility emergency plan should describe such a program. This description
should stipulate who will be responsible for delivery of this program, what their
responsibilities are, how the program will be delivered, and what the program
will entail.
14
Canadian Nuclear Commission canadienne
Safety Commission de sûreté nucléaire
REGULATORY
GUIDE
Developing and
Using Action Levels
G-228
March 2001
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the regulatory
requirements.
Regulatory guide: a document that provides guidance or describes characteristics or practices
that the CNSC recommends for meeting regulatory requirements or improving administrative
effectiveness.
Regulatory notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be acted
upon in a timely manner.
Regulatory procedure: a document that describes work processes that the CNSC follows to
administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in
regulations, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
REGULATORY GUIDE
Developing and Using
Action Levels
G-228
Published by the
Canadian Nuclear Safety Commission
March 2001
Developing and Using Action Levels
Regulatory Guide G-228
Published by the Canadian Nuclear Safety Commission
© Minister of Public Works and Government Services Canada 2001
Extracts from this document may be reproduced for individual use without permission provided
the source is fully acknowledged. However, reproduction in whole or in part for purposes of
resale or redistribution requires prior written permission from the Canadian Nuclear Safety
Commission.
Catalogue number CC173-3/2-228E
ISBN 0-662-29638-9
Ce document est également disponible en français.
Document availability
The document can be viewed on the CNSC website. Copies in English or French may be ordered
using the contact information below:
Communications Division
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 995-5894 or 1-800-668-5284 (Canada only)
Facsimile: (613) 992-2915
E-mail: publications@cnsc-ccsn.gc.ca
Website: www.nuclearsafety.gc.ca
March 2001 G-228
TABLE OF CONTENTS
1.0 PURPOSE........................................................................................................................... 1
2.0 SCOPE ................................................................................................................................ 1
3.0 BACKGROUND ................................................................................................................ 1
3.1 Regulatory framework ........................................................................................................ 1
3.2 Relevant legislation for this document ............................................................................... 1
3.3 The CNSC licensing process and action levels .................................................................. 3
4.0 ACTION LEVELS FOR RADIATION PROTECTION ..................................................... 3
5.0 UNDERSTANDING ACTION LEVELS ........................................................................... 4
6.0 DEVELOPING, USING AND REVISING ACTION LEVELS ......................................... 5
7.0 MONITORING ................................................................................................................... 7
8.0 RESPONDING WHEN AN ACTION LEVEL IS REACHED .......................................... 8
9.0 EXAMPLES OF THE USE OF ACTION LEVELS .......................................................... 9
III
March 2001 G-228
DEVELOPING AND USING
ACTION LEVELS
1.0 PURPOSE
This regulatory guide is intended to help applicants for Canadian Nuclear Safety
Commission (CNSC) licences develop action levels in accordance with paragraph 3(1)(f)
of the General Nuclear Safety and Control Regulations and section 6 of the Radiation
Protection Regulations.
2.0 SCOPE
This guide applies to all applications for a CNSC licence, other than an application for a
licence to abandon. It describes how the licence applicant can develop action levels that
provide for the radiation protection of workers and the public during the conduct of
activities licensed by the CNSC. This document does not deal with action levels for the
purpose of environmental protection at a uranium mine or mill.
3.0 BACKGROUND
3.1 Regulatory framework
The CNSC is the federal agency that regulates the use of nuclear energy and materials to
protect health, safety, security and the environment, and to respect Canada s international
commitments on the peaceful use of nuclear energy.
The Nuclear Safety and Control Act ( the Act ) requires persons or organizations to be
licensed by the CNSC for carrying out the activities referred to in section 26 of the Act,
unless otherwise exempted. The associated regulations stipulate prerequisites for CNSC
licensing, and the obligations of licensees and workers.
3.2 Relevant legislation for this document
The following CNSC legislation is pertinent to an understanding of this regulatory guide:
Paragraph 3(1)(f) of the General Nuclear Safety and Control Regulations which
requires that an application for a CNSC licence contain any proposed action level
for the purpose of section 6 of the Radiation Protection Regulations.
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Subsection 4(a) of the Radiation Protection Regulations which states that:
Every licensee shall implement a radiation protection program and shall, as part of
that program,
(a) keep the amount of exposure to radon progeny and the effective dose and
equivalent dose received by and committed to persons as low as reasonably
achievable, social and economic factors being taken into account, through the
implementation of
(i) management control over work practices,
(ii) personnel qualification and training,
(iii) control of occupational and public exposure to radiation, and
(iv) planning for unusual situations.
Subsection 6(1) of the Radiation Protection Regulations which defines an action
level to be a specific dose of radiation or other parameter that, if reached, may
indicate a loss of control of part of a licensee s radiation protection program, and
triggers a requirement for specific action to be taken.
Subsection 6(2) of the Radiation Protection Regulations which stipulates that:
When a licensee becomes aware that an action level referred to in the licence for
the purpose of this subsection has been reached, the licensee shall:
(a) conduct an investigation to establish the cause for reaching the action level;
(b) identify and take action to restore the effectiveness of the radiation protection
program implemented in accordance with section 4; and
(c) notify the Commission within the period specified in the licence.
Subsection 4(1) of the Uranium Mines and Mills Regulations which defines an
action level to be:
A specific dose of radiation or other parameter that, if reached, may indicate a loss
of control of part of a licensee s radiation protection program or environmental
protection program, and triggers a requirement for specific action to be taken.
Subsection 4(2) of the Uranium Mines and Mills Regulations which stipulates that:
An application for a licence in respect of a uranium mine or mill, other than a
licence to abandon shall contain a code of practice that includes
(a) any action level that the applicant considers appropriate for the purpose of this
subsection;
(b) a description of any action that the applicant will take if an action level is
reached;
(c) the reporting procedures that will be followed if an action level is reached.
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3.3 The CNSC licensing process and action levels
The CNSC typically applies a phased process to its licensing of nuclear facilities and
activities. For major facilities, this process begins with a consideration of the
environmental impacts of the proposed project, and proceeds progressively through site
preparation, construction, operation, decommissioning and abandonment phases.
The Nuclear Safety and Control Act and regulations require licence applicants to provide
certain information at each licensing stage. The type and level of detail of this
information will vary to accommodate the licensing stage and specific circumstances.
At all licensing stages, applications may incorporate (directly or by reference) new or
previously submitted information, in accordance with legislated requirements and the
best judgement of the applicant. An application that is submitted at one licensing stage
can become a building block for the next stage.
Upon receipt of an application that is complete, the CNSC reviews it to determine
whether the applicant is qualified to carry on the proposed activity, and has made
adequate provision for the protection of the environment, the health and safety of
persons, and the maintenance of national security and measures required to implement
international obligations to which Canada has agreed. If satisfied, the CNSC may issue,
renew, amend or replace a licence that contains relevant conditions. Typically, this
licence will incorporate the applicant s undertakings, and will contain other conditions
that the CNSC considers necessary, including a condition that incorporates or relates to
an action level.
4.0 ACTION LEVELS FOR RADIATION PROTECTION
The Radiation Protection Regulations and the Uranium Mines and Mills Regulations
contain different definitions of an action level .
Under the Radiation Protection Regulations, an action level is defined to be a specific
dose of radiation or other parameter that, if reached, may indicate a loss of control of part
of a licensee s radiation protection program, and triggers a requirement for specific
action to be taken.
However, under the Uranium Mines and Mills Regulations, an action level for a uranium
mine or mill is a specific dose of radiation or other parameter that, if reached, may
indicate a loss of control of part of a licensee s radiation protection program or
environmental protection program, and triggers a requirement for specific action to be
taken.
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Thus, the definition of an action level in the Uranium Mines and Mills Regulations
encompasses both radiation protection and environmental protection. However, for the
purposes of section 6 of the Radiation Protection Regulations and this guide, the two
definitions are consistent.
5.0 UNDERSTANDING ACTION LEVELS
Action levels are designed to alert licensees before regulatory dose limits are reached. By
definition, if an action level in a licence is reached, a loss of control of some part of the
associated radiation protection program may have occurred, and specific action is
required. The specified action under the Radiation Protection Regulations consists of
establishing the cause for reaching the action level, restoring the effectiveness of the
radiation protection program, and notifying the CNSC within the period specified in the
licence.
Accordingly, CNSC licensees may use action levels to help them monitor and maintain
the effectiveness of the radiation protection programs that they must implement under
subsection 4(a) of the Radiation Protection Regulations. In particular, licensees may set
action levels, and monitor related parameters, so as to provide for early warnings of any
actual or potential losses of control of the parts of the radiation protection program to
which the action levels apply; thereby maximizing their opportunities for follow-up
investigations and any interventions that may be necessary in order to restore control.
Where a radiation protection program consists of several parts, action levels may be
appropriate for each or any part of the program.
Action levels may be expressed in terms of any parameters that, if reached, may indicate
a loss of control of an associated part of the licensee s radiation protection program.
Some examples of such parameters are:
the quantity of radiation exposure or dose that an individual receives ( individual
dose ),
a radiation level within a work area ( ambient dose rate ),
radioactivity per unit surface area ( surface contamination level ),
an air-exchange rate in a work place ( ventilation rate ),
a rate at which nuclear substances are released to the environment ( emission rate,
discharge rate ), and
a concentration or a quantity of a nuclear substance in a workplace or in an effluent
( concentration , loading ).
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Action levels are typically site and facility specific. An action level at one facility could
lie within the normal operating range of another facility. Over the lifetime of a facility or
activity, an action level may be dynamic or static. That is, it may be revised upwards or
downwards to accommodate the prevailing circumstances. For example, an action level
for a new facility or activity may warrant refinement once significant operating
experience is gained. Similarly, if conditions at a facility change (e.g., changes in
geological conditions at a uranium mine), a related action level may also need to be
changed.
All parts of a licensee s radiation protection program are considered to be under control
when the radiation doses to persons as a consequence of the licensed activities are kept
as low as reasonably achievable, social and economic factors being taken into account
(ALARA), through the implementation of the measures specified in subsection 4(a) of
the Radiation Protection Regulations.
Figure 1 on the next page illustrates, graphically, the typical relationship among a
regulatory limit, the relevant action levels, and the maintenance of the amount of
exposure to radon progeny and the effective dose and equivalent dose received by and
committed to persons as low as reasonably achievable, social and economic factors being
taken into account.
6.0 DEVELOPING, USING AND REVISING ACTION LEVELS
Typically, an action level for a nuclear facility or activity will be developed as part of the
CNSC licensing process, in accordance with paragraph 3(1)(f) of the General Nuclear
Safety and Control Regulations. However, some CNSC licensed activities may not
warrant the use of action levels. Examples include some uses of fixed gauges, static
eliminators or gas chromatography.
If it is to be useful and credible, an action level must be a meaningful indicator over a
defined time period of the state of a radiation protection program. Accordingly, the action
level must be measurable to accepted standards of accuracy.
Where possible, an action level for a nuclear facility should take into account the facility
design and relevant operating experience. A licence applicant who lacks such experience,
as in the case of new activities or operations, may be able to draw upon the experience of
comparable designs and operations. To facilitate regulatory review of any proposed
action level, the licence applicant should thoroughly and clearly explain the rationale for
the level and its planned use.
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G-228 March 2001
Figure 1: An example of the typical relationship among regulatory limits, action levels
and ALARA1 doses
Dose Regulatory limit = A dose limit, as per section 13, 14 or 15 of the Radiation Protection
Regulations.
Dose Action level `n' = The dose to persons when a designated action level2, `n', is reached. The
occurrence may indicate a loss of control (of a part of the radiation
protection program) that typically has more serious consequences than
when a lower action level is reached.
Dose Action level 2 = The dose to persons when a designated action level, `Action level 2', is
reached. The occurrence may indicate a loss of control (of a part of the
radiation protection program) that typically has more serious consequences
than when a lower action level is reached.
Dose Action level 1 = The dose to persons when a designated action level, `Action level 1', is
reached. When this action level is set near a dose that is ALARA, the
occurrence typically has relatively minor consequences.
Dose ALARA = A dose that is ALARA, as a result of (i) management control over work
practices3, (ii) personnel qualification and training, (iii) control of
occupational and public exposure to radiation, and (iii) planning for
unusual situations.
Notes
1. A radiation dose is ALARA if it is as low as reasonably achievable, social and economic factors being taken into account.
2. An action level may be expressed in units of radiation dose, or in any terms of any other parameter that could be indicative
of a loss of control of a part of the associated radiation protection program.
3. Management control over work practices may, where appropriate, include levels for the routine release of liquid or gaseous
effluents. Typically, such levels are set sufficiently low that control of the parts of the radiation protection program to keep
doses ALARA state will not be jeopardized if these control levels are reached during normal operations.
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March 2001 G-228
Accordingly, the following steps for developing and using action levels may be helpful to
licence applicants and licensees:
From the design, identify those processes and activities that could result in doses to
workers or the public.
For activities and processes that could result in doses to workers or the public,
identify the measurable parameters that will indicate, directly or indirectly, whether
the radiation protection program is adequately controlled.
On the basis of realistic assumptions, select appropriate action levels, expressed in
terms of the appropriate parameters, for all key processes and activities.
Incorporate use of the selected action levels into the proposed radiation protection
program.
Implement the radiation protection program and the associated action levels in
accordance with the CNSC licence.
As operating experience accumulates, revise action levels accordingly.
To revise an action level that is referred to in a licence, the licensee must obtain an
appropriate licensing action from the CNSC. When applying for this action, the applicant
should demonstrate that the proposed revision is appropriate for the purposes of section 6
of the Radiation Protection Regulations and any relevant requirements of the licence.
7.0 MONITORING
To serve as an effective indicator of a possible loss of control of a part of a radiation
protection program, an action level must be supported by a monitoring program that can
accurately detect when the action level is reached. Accordingly, licence applications that
include any proposed action level should also describe the monitoring program that the
applicant plans to implement in order to detect when the action level is reached.
Since the purpose of monitoring action levels is to provide timely warning of any
potential or actual loss of control of part of the radiation protection program, a
corresponding monitoring proposal should consist of an appropriate methodology and
frequency of sampling or measurement. This selection of methodology and frequency
will be influenced by case-specific factors, and should be commensurate with the
probability and consequences of a loss of control of a part of the radiation protection
program. For example, as the probability that regulatory dose limits could be approached
or exceeded as a consequence of a loss of control of part of a radiation protection
program increases, more rigorous action level monitoring programs may be appropriate.
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G-228 March 2001
When a proposal for monitoring an action level is accepted and incorporated into a
CNSC licence, the licensee must ensure that the program is implemented and maintained
in accordance with the licence.
8.0 RESPONDING WHEN AN ACTION LEVEL IS REACHED
When an action level referred to in a licence is reached, specific responses are required
pursuant to subsection 6(2) of the Radiation Protection Regulations. The licensee must
conduct an investigation to determine the cause, identify and take action to restore the
effectiveness of the radiation protection program, and notify the CNSC within the time
period specified in the licence.
Although an action level is not an enforceable dose limit, a failure to meet the above
obligations would contravene the Radiation Protection Regulations, and would
constitute an offence under the Nuclear Safety and Control Act.
The reaching of an action level could be due to any number of causes. An action level
could be reached repeatedly as a consequence of chronic deficiencies in the associated
radiation protection program. Ongoing occurrences could be triggered by a shift in
normal operating conditions. Occasional or more frequent occurrences could be triggered
by transient conditions that might not relate to a significant loss of control of the
radiation protection program, or to a significant change in the radiation doses associated
with normal operating conditions.
Accordingly, case-by-case assessments and commensurate remedies may be required in
response to each situation where an action level is reached. The appropriate response will
depend in part upon the results of the assessment, as well as any other relevant factor
such as the hazards associated with the action level of concern.
The investigation that a licensee undertakes to determine why an action level referred to
in a licence has been reached may need to first confirm whether the evidence (e.g.,
measurements, observations or calculations) that indicates that the action level has been
reached is valid i.e., whether the action level has indeed been reached.
Further to determining the cause for reaching an action level, the licensee must identify
and take actions to restore the effectiveness of the radiation protection program. These
actions should be appropriate to the circumstances and commensurate with the level of
risk associated with the reaching of the action level. If the licensee cannot restore the
effectiveness forthwith, the licensee should propose interim measures for CNSC
consideration. The measures to restore the effectiveness of the radiation protection
program, whether interim or final, should be based on credible experience, data or
analyses, and should take into account the consequences of the loss of control.
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March 2001 G-228
Typically, the greater the radiation hazards that result when an action level is reached, the
more immediate, complex or rigorous will be the measures to restore the effectiveness of
the radiation protection program.
In addition to the above responses when an action level is reached, paragraph 6(2)(c) of
the Radiation Protection Regulations requires the licensee to notify the CNSC within the
period specified in the licence. This period will take into account the consequences of
reaching the action level. The greater the radiation hazards when an action level is
reached, the shorter the specified notification period is likely to be. Accordingly, the
specified period for notifying the CNSC when an action level referred to in a licence is
reached will typically be longer (i.e days, weeks or months) than the immediate
notification required pursuant to paragraph 16(a) of the Radiation Protection
Regulations when the licensee becomes aware that a dose limit may have been exceeded.
9.0 EXAMPLES OF THE USE OF ACTION LEVELS
In the past, some operators of nuclear facilities, and some persons engaged in nuclear
activities, used various indicators of radiation dose to monitor, control and assure that
their radiation protection programs were effective. Traditionally, these dose indicators
were arrived at on the basis of facility-specific or industry experience. In some cases,
these indicators were used in the same way that action levels are to be used under the
current regulations. In other situations, the uses of such indicators were not expressly
linked to any loss of control of a radiation protection program, and consequently were a
form of the management control over work practices that is referred to in paragraph
4(a)(i) of the Radiation Protection Regulations. In still other situations, licensees used a
combination of the two approaches.
Some examples of the historical and current application of dose indicators follow:
Uranium mines and mills
Under the former Uranium and Thorium Mining Regulations, the operators of
uranium mines and mills used dose indicators that were also termed action levels
to help assure the radiation safety of workers, on-site personnel and the public. The
use of these action levels differed from that envisaged under section 6 of the
Radiation Protection Regulations, in that their application was not limited to
situations involving a possible loss of control of a radiation protection program.
Instead, they were used as administrative mechanisms to track and control radiation
doses at much lower levels.
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Under the current Uranium Mines and Mills Regulations, applications for licences
shall include a code of practice that contains any action level that the applicant
considers appropriate, as well as the proposed responses and reporting procedures if
a proposed action level is reached.
At a given mine or mill, different action levels may be needed for different
operations. Accordingly, any action level and related responses contained in a
proposed code of practice should be tailored to the specific situation and its needs.
Medical and research institutions
Medical and research institutions commonly use open sources of radiation. To
address the possibility of significant intakes of nuclear substances during such uses,
the associated radiation protection programs may require that certain actions be
taken if the results of precautionary monitoring to screen for significant intakes of
radionuclides reach specific levels.
For example, some medical and research institutions that use radioiodine in their
operations have in the past adopted criteria for taking remedial actions in response
to the results of thyroid monitoring programs. When these results reached or
exceeded defined criteria, specified responses were implemented. Typically, these
responses consisted of such actions as repeating the thyroid counting procedure to
verify the initial result, performing supplementary bioassays on co-workers, or
implementing interim measures to prevent further exposure of the workers until the
cause was identified and remedied.
Radioisotope use
In licences that authorized radioisotope use, the former Atomic Energy Control
Board (now the Canadian Nuclear Safety Commission) routinely included
conditions that obliged the licensee to carry out certain actions if specified surface
contamination criteria were reached or exceeded. Further, if preventative and
confirmatory monitoring indicated that a surface was radioactively contaminated in
excess of a predetermined criterion, the licensee was typically required to
decontaminate the affected work surfaces to an acceptable level.
The above use of contamination criteria was an integral part of the licensee s
radiation protection program, and the associated contamination criteria served a
purpose similar to that of action levels under the existing regulations. Accordingly,
some applicants for licences under the Nuclear Safety and Control Act and
regulations may choose to propose the use of action levels that are expressed in
terms of surface contamination levels.
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Thus, under the Nuclear Safety and Control Act and regulations, licensees may continue
to use a combination of action levels and management controls to help keep radiation
doses below regulatory limits, and as low as reasonably achievable, social and economic
factors being taken into account.
11
Canadian Nuclear Commission canadienne
Safety Commission de sûreté nucléaire
REGULATORY
POLICY
Policy on
Human Factors
P-119
October 2000
REGULATORY DOCUMENTS
The Canadian Nuclear Safety Commission (CNSC) operates within a legal framework that
includes law and supporting regulatory documents. Law includes such legally enforceable
instruments as acts, regulations, licences and orders. Regulatory documents such as policies,
standards, guides, notices, procedures and information documents support and provide further
information on these legally enforceable instruments. Together, law and regulatory documents
form the framework for the regulatory activities of the CNSC.
The main classes of regulatory documents developed by the CNSC are:
Regulatory policy: a document that describes the philosophy, principles and fundamental
factors used by the CNSC in its regulatory program.
Regulatory standard: a document that is suitable for use in compliance assessment and
describes rules, characteristics or practices which the CNSC accepts as meeting the regulatory
requirements.
Regulatory guide: a document that provides guidance or describes characteristics or practices
that the CNSC recommends for meeting regulatory requirements or improving administrative
effectiveness.
Regulatory notice: a document that provides case-specific guidance or information to alert
licensees and others about significant health, safety or compliance issues that should be acted
upon in a timely manner.
Regulatory procedure: a document that describes work processes that the CNSC follows to
administer the regulatory requirements for which it is responsible.
Document types such as regulatory policies, standards, guides, notices and procedures do not
create legally enforceable requirements. They support regulatory requirements found in regula-
tions, licences and other legally enforceable instruments. However, where appropriate, a
regulatory document may be made into a legally enforceable requirement by incorporation in a
CNSC regulation, a licence or other legally enforceable instrument made pursuant to the Nuclear
Safety and Control Act.
REGULATORY POLICY
Policy on
Human Factors
P-119
Published by the
Canadian Nuclear Safety Commission
October 2000
Policy on Human Factors
Regulatory Policy P-119
Published by the Canadian Nuclear Safety Commission
© Minister of Public Works and Government Services Canada 2000
Extracts from this document may be reproduced for individual use without permission provided
the source is fully acknowledged. However, reproduction in whole or in part for purposes of
resale or redistribution requires prior written permission from the Canadian Nuclear Safety
Commission .
Catalogue number CC173-3/1-119E
ISBN 0-662-29521-8
Ce document est également disponible en français.
Document availability
The document can be viewed on the CNSC website. Copies in English or French may be ordered
using the contact information below:
Communications Division
Canadian Nuclear Safety Commission
P.O. Box 1046, Station B
280 Slater Street
Ottawa, Ontario K1P 5S9
CANADA
Telephone: (613) 995-5894 or 1-800-668-5284 (Canada only)
Facsimile: (613) 992-2915
E-mail: info@cnsc-ccsn.gc.ca
Website: www.nuclearsafety.gc.ca
October 2000 P-119
POLICY ON HUMAN FACTORS
1.0 PURPOSE
The purpose of this regulatory policy is to assure that the Canadian Nuclear Safety
Commission (CNSC) takes human factors issues into account in its regulatory activities.
2.0 SCOPE
This policy describes how the CNSC will take human factors into account during its
licensing, compliance and standards-development activities.
3.0 DEFINITION AND EXAMPLES OF HUMAN FACTORS
For purposes of this policy, the term human factors means factors that influence human
performance as it relates to the safety of a nuclear facility or activity over all phases,
including design, construction, commissioning, operation, maintenance and
decommissioning.
Some examples of human factors are: organizational and management structures, policies
and programs; the allocation of functions to humans and machines; the design of user
interfaces; staffing provisions; job-design features; work schedules; the design of written
procedures; training, and the physical work environment.
4.0 POLICY STATEMENT
The Canadian Nuclear Safety Commission recognizes that human factors can affect the
performance of the facilities and activities that it regulates. Accordingly, it is the policy
of the Commission that:
When reviewing applications for CNSC licences in accordance with any applicable
laws, procedures and guidelines, the Commission will take into account human
factors that could impact upon the Commission s mandate for protection of the
environment, the health and safety of persons, the maintenance of national security
and the implementation of international obligations to which Canada has agreed.
The Commission will evaluate the measures proposed by licence applicants, and the
measures implemented by licensees to address human factors, to determine whether
the measures provide for protection of the environment, the health and safety of
persons, the maintenance of national security and the implementation of
international obligations to which Canada has agreed.
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P-119 October 2000
The CNSC will, where needed, provide licence applicants and licensees with written
guidance on how to address human factors that could affect the safety of CNSC-
regulated facilities and activities.
The CNSC will cooperate with other organizations and jurisdictions to foster
consistent national and international standards with respect to human factors.
5.0 EVALUATION
The CNSC internal audit group will evaluate the CNSC s adherence to this policy, and
the policy s effectiveness, during periodic program reviews in accordance with
management priorities.
6.0 POLICY AUTHORITY
This regulatory policy is issued under the authority of the Nuclear Safety and Control
Act.
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