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Responses to Questions Raised from the Peer Review of Canada's Eighth National Report for the Convention on Nuclear Safety

This document supplements the Canadian National Report for the Eighth Review Meeting of the Convention on Nuclear Safety. By offering additional and detailed information in response to 142 specific questions, or comments received from Contracting Parties, the document demonstrates how Canada has implemented its obligations under the Convention on Nuclear Safety. This document is produced by the Canadian Nuclear Safety Commission (CNSC) on behalf of Canada. Contributions to the document were made by CNSC staff and representatives from Ontario Power Generation, Bruce Power, NB Power, SNC-Lavalin Nuclear, Health Canada and Natural Resources Canada.

List of countries who provided questions and comments

United Kingdom

Posted By Article Reference Question Comment Answer
United Kingdom Article 8 8.1 (c) page 73 Section 8.1.(c) refers to a systematic approach for NPP-related knowledge and on the job training for NPP site inspectors in both technology and regulatory processes. It also refers to specific knowledge transfer training provided to senior inspectors who coach inspectors-in-training.

This should be a candidate as an area of Good Performance

Canada acknowledges the comment
United Kingdom Article 8 8.1 (c) page 73 Section 8.1.(c) refers to the anticipated attrition of CNSC staff and the integration of strategic workforce planning. Please provide further information on how such workforce planning is sufficiently agile to accommodate the range of demands from Canada’s nuclear industry, such as the ongoing CANDU refurbishment program; the potential proliferation of a range of SMR technologies across Canada and the ongoing regulation of existing liabilities. CNSC regularly conducts workforce planning allowing the organization to pivot and meet a range of changing demands in Canada’s nuclear industry. To ensure the CNSC is well-positioned to meet future industry needs, other internal business-planning processes are also integrated and considered as part of the workforce planning process, such as the annual strategic and operation planning cycles and quarterly program monitoring and the enterprise risk profile updates. All of these activities provide a complete and timely capture of industry pressures and drivers, which then feed into quarterly workforce planning discussions. Based on these discussions, the CNSC is able to adequately predict, plan and prepare for attrition and new staff in critical and emerging areas.
United Kingdom Article 8 8.2 (e) Page 75 Section 8.1.(e) refers to five recommendations from the Office of the Auditor General, following an audit conducted in 2016 into CNSC’s inspection program for nuclear power plants. The report states that CNSC completed corrective actions by March 2017. Please describe how the revised approach to inspection planning appropriately targets and prioritises NPP systems in a risk-informed manner. A risk-informed approach is used to develop the 5-year Baseline Compliance Plan. Major focus is given to areas of higher regulatory impact and risk is modelled to minimize bias and subjectivity. The risk-informed approach ensures that a comprehensive and balanced understanding of an NPP’s risk spectrum is used to allow for effective use of resources to address the more risk significant areas. 

A risk-informed approach is also used to develop the Reactive Compliance Plan. The Reactive Compliance Plan is determined based on areas that require, or appear to require, further licensee attention or are licensee-schedule-dependent. It consists of compliance verification activities to be conducted on an operating NPP over a 3 year period and may require modification at any time to react to emerging issues. Compliance verification activities in the Reactive Compliance Plan are prioritized based on the most risk-significant areas from a regulatory perspective.

The outcome of a risk-informed Baseline Compliance Plan and Reactive Compliance Plan is a set of compliance verification activities that focus regulatory attention on areas and issues commensurate with their importance to 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 uses a risk-informed approach, aligned with INSAG-25, to analyze, identify, prioritize and select compliance verification areas and the appropriate compliance verification activities, when developing both the 5-year Baseline Compliance Plan and 3-year Reactive Compliance Plan. CNSC governance identifies the following elements to be considered in compliance verification planning, which are identified in section 3 of INSAG-25:

  • CNSC Regulatory Requirements (e.g., Standards, codes and CNSC regulatory documents)
  • Licensee Performance History
  • Deterministic Considerations
  • Probabilistic Considerations
  • Organizational Considerations
  • Other Considerations (operating experience (OPEX), Periodic Safety Reviews, etc.)
United Kingdom Article 16 16.2(b) Page 158 Section 16.2.(b) refers to Canada’s working relationship with the United States in the exchange of emergency preparedness expertise.

Please provide further detail on the program of cross border emergency response drills associated with Canadian NPPs close to bordering US states including: periodicity; agencies involved and any areas of improvement successfully made in light of such cooperation.

Canada has participated in exercises organized by the United States, and vice versa, in order to promote cross-border cooperation.  Additionally, the provinces of Ontario and New Brunswick work with neighbouring States in the United States to test and exercise coordination. For example, in New Brunswick, the Maine Emergency Management Agency (MEMA) has participated in the planning and conduct of New Brunswick's full scale exercises (e.g. INTREPID 2018 and SYNERGY CHALLENGE 2018). As part of its involvement, MEMA participated in planning meetings/conferences leading up to the exercises, and tested notifications and activation of liaison staff to the New Brunswick Provincial Emergency Operations Centre (PEOC) during exercise play.

New Brunswick is also a member of the International Emergency Management Group (IEMG), which provides for the possibility of mutual assistance in managing an emergency or disaster among participating jurisdictions. Through a Memorandum of Understanding (The Compact), and in support of Resolution 23-5 of the Conference of New England Governors and Eastern Canadian Premiers, the IEMG also supports the process of planning, mutual cooperation, and emergency-related exercises, testing and other training activities.

In conjunction with its Provincial Nuclear Emergency Response Plan (PNERP), Ontario maintains the Implementing Plan For Fermi 2 Nuclear Generating Station, located in the neighbouring State of Michigan, USA.  The implementing plan describes the measures that shall be undertaken to deal with a nuclear emergency caused by an event at Fermi 2. The CNSC has an MOU with the USNRC to exchange information and expertise. Recently, the CNSC and USNRC facilitated an interchange of inspectors. This USNRC-CNSC interchange program brought valuable experience and knowledge related to inspector's response to an emergency, as well as other topics such as the regulatory approach to NPP oversight, inspection practices and reactor technology.

United Kingdom Article 14 14.(ii).(b) Page 130 Section 14.(ii).(b) refers to CNSC’s REGDOC 2.6.3 – Ageing Management, which provides comprehensive requirements for integrated ageing management programs, including organisational arrangements. Recognising the extended operation of Bruce and OPG units potentially up to 2064, there is no detail specified as to CNSC’s regulatory program for overseeing licensee ageing management programs.

Please provide further detail on CNSC’s regulatory program and basis for prioritising inspections of ageing structures, systems and components; both on the approach to end of operational life and to inform decisions on licence extension for long term operation.

CNSC's regulatory oversight program in this area is intended to confirm that the ageing management philosophy outlined in REGDOC 2.6.3 is effectively implemented. The systematic approach to ageing management involves performance of the activities in the PLAN-DO-CHECK-ACT cycle, outlined in Section 2.2 of REGDOC 2.6.3 (which is adapted from the IAEA Safety Guides/Reports on aging management and long term operation). To ensure regulatory oversight of the program, CNSC performs the following compliance verification activities, as a minimum (for additional description of lifecycle management plans and periodic inspection programs (PIPs), see response to Sequence 62):
  • Review of Licensee Ageing Management program documentation. This includes reviewing the plant-level Integrated Aging Management Program, condition assessments of structures, systems, and components (SSCs) and aging management plans, and "lifecycle management plans" for major pressure boundary components such as fuel channels, feeders and steam generators
  • Review of licensee periodic inspection program (PIP) plan documentation. PIP plans are produced to comply with the requirements of CSA Group Standards related to periodic inspection of NPP SSCs, and incorporated into the licence conditions handbooks (LCHs) for each NPP. An example is CSA N285.4 Periodic Inspection of CANDU Nuclear Power Plant Components.
  • Review of licensee inspection results, material surveillance and fitness for service assessments submitted in accordance with the PIP compliance verification criteria in the LCH
  • Conducting compliance inspections at licensee facilities to confirm that the requirements outlined in the documents described above are effectively implemented. This includes, for example, confirming that the implementation of aging management plans is effective, the personnel responsible for ageing management are equipped to perform their duties, and the regulatory requirements of the licensee's management system are met.
  • Review of event reports submitted by licensees in accordance with the requirements of REGDOC-3.1.1, Reporting Requirements for Nuclear Power Plants related to aging degradation of components.
To inform decisions related to long term operation, compliance verification activities are carried out to assess the adequacy of licensee's ageing management activities in accordance with the requirements of REGDOC-2.3.3, Periodic Safety Reviews, with particular attention paid to the following safety factors:      
  • Actual condition of SSC's important to safety
  • Equipment qualification 
  • Ageing

Compliance verification activities start with reviews of PSR submissions, including basis document and safety factor reports, and typically will include on-site inspections to verify elements of the reports, including SSC condition assessments, time-limited aging analyses and associated gap assessments and dispositions / recommendations, and confirm processes are in place to track required corrective actions to completion. Integrated Implementation Plan activities that are required to address any gaps related to ageing management are captured in LCHs and tracked to completion by CNSC staff.

United Kingdom Article 7 7.2.(i) (b) - Page 42 Section 7.2.(i) (b) and the associated Annex describe CNSC’s discussion paper on the regulatory strategy, approaches and challenges associated with small modular reactors. During this process, CNSC has undertaken a comprehensive review of the ability of its regulatory framework to accommodate innovative technology and has solicited early public feedback on potential licensing challenges. This approach has enabled early recognition, and therefore confidence to potential vendors and assurance to Natural Resources Canada, that the existing regulatory framework is broadly valid to accommodate SMRs. CNSC has recognised early that further refinement is needed to ensure a graded approach to future SMR siting, design, construction and operation.

This is a candidate for a good performance.

Canada acknowledges the comment
United Kingdom Article 8 8.2.(a) & 8.2.(b) Sections 8.1.(a) and 8.2.(a) describe the position of the CNSC in the government structure, and states that the commission reports to Parliament through the Minister of Natural Resources (NRCan), and through which it is required to submit its Departmental Plan. It is evident that individual regulatory decisions can only be reviewed by the Federal Court and that CNSC does not directly report to the Minister.

For activities not directly recoverable under the ‘Cost Recovery Fees Regulations’, it appears CNSC is required to secure the support of the Minister of Natural Resources to appropriate funding from Parliament. The minister also has responsibility for the promotion of nuclear energy. Thus, it is not clear how the balance of CNSC’s funding is sufficiently ring-fenced and assured in a manner that is sufficiently independent from any budgetary constraints and change in priorities that might arise from the promoting government ministry. Please clarify what safeguards exist to ensure that CNSC is able to confidently appropriate the balance of funding (i.e. outside of the cost recovery framework), in order to fully deliver its mission, independent of the sponsoring Government ministry.

As noted, the Minister of Natural Resources (NRCan) is responsible for the Canadian Nuclear Safety Commission.

In May 2000, the Nuclear Safety and Control Act came into effect and established the CNSC with responsibilities and authorities to regulate an industry that spans all segments of the nuclear fuel cycle and a wide range of industrial, medical and academic uses of nuclear substances.  In 2007/08, the CNSC received authority to re-spend the revenue it collects throughout the year on cost recoverable activities (see response to Sequence 88). With this authority, the CNSC has the 'statutory' capacity to re-spend funds collected.

Appropriation funding to support 'non-recoverable' activities requires 'voted' parliamentary approval on an annual basis. Over the years, through various governments of the day, the funding has remained stable, with increases granted due to inflationary pressures and wage adjustments.  The CNSC is an independent department that reports to Parliament through the Minister of NRCan.  The CNSC’s voted appropriation and statutory authority is independent of NRCan’s.

The appropriation is used to fund not-for-profit organizations, along with non-recoverable activities such as Canada's international commitments with respect to non-proliferation and nuclear safeguards.

If new appropriation funding is required, the CNSC has to demonstrate the need to the Government/Parliament and secure a source of funds via the Federal Budget and follow the normal Cabinet approval process.

United Kingdom Article 8 8.1.(C) - Page 73 Section 8.1.(c) refers to a graduate recruitment initiative that has brought 154 new graduates to the organisation since 2014. It is not clear how graduates acquire sufficient exposure and awareness to the industry that they will in future regulate (either as specialists or frontline inspectors), recognising they will have minimal operational experience.

Please provide further detail on how graduates acquire sufficiently diversified work experience that gives them the necessary exposure to engineering and technical issues associated with NPP operation prior to becoming fully competent in their chosen careers.

Many graduates are recruited through the CNSC Technical Co-Op Program. This 15-month co-op program is comprised of three rotations of five months in differing technical and regulatory areas, including but not limited to NPP sites.

New staff, including those hired through the graduate recruitment initiative, are required to undergo the inspector training and qualification program (ITQP) should they be in an inspector position in any service line. As mentioned in Section 8.1. (c) of the Canadian report, ITQP uses a systematic approach for NPP-related knowledge which includes CANDU courses and on-the-job training (OJT) and provides the necessary exposure to technical issues associated with NPP operations.

OJT teaches the skills, knowledge and competencies that are required to perform a specific job by relying on the experience of senior employees as mentors. Every regulatory program division has its own OJT manual and evaluation guide. The objective is to ensure that new inspectors experience the core inspection activities prior to receiving an inspector’s certificate. The OJT manuals identify the necessary knowledge and contain the learning tasks for doing inspections. The mentor and director are responsible for evaluating performance and training effectiveness by using the evaluation guide developed for their service line. OJT manuals provide a standardized approach for training new inspectors.

Because CNSC has site offices at the operating NPPs, OJT for NPP site inspectors is conducted in the context of full-time participation in regulatory and other activities at the licensed site, providing significant exposure to licensee operations.

Inspector cards are only issued once the regulatory program director and the issuing authority (Designated Officer) are confident that the inspector is deemed qualified.

United Kingdom Article 14 14.(ii).(b) - Page 130 Section 14.(ii).(b) refers to the Fuel Channel Lifecycle Management program, and the development of refined engineering methodologies and models for predicting material properties over the full operational life of fuel channel components. The report states that NPP licensees use models to support the underpinning safety case justifying their requests to extend limits on full power operation (EFPH limits)., Please provide further detail on how CNSC secures suitable independent validation / verification of licensees’ models on the behaviour of major life limiting components. The Fuel Channel Lifecycle Management program outlines the licensee's activities to confirm that the engineering methodologies and models are validated and verified. The licensees perform R&D through the CANDU Owner’s Group (COG). COG work packages are intended to obtain data to better understand and predict consequences of the complex degradation mechanisms, including pressure tube deuterium uptake, annulus spacer structural integrity and pressure tube fracture toughness. Researchers sponsored by COG, and staff from the licensees, notify CNSC staff how the results of this research are incorporated into the models. In many cases, CNSC staff hold meetings with industry to address CNSC concerns, to verify the licensees’ approaches are justified. If research results indicate that any aspects of the current models may be non-conservative, CNSC ensures that the new results are incorporated into revisions of the models so that they remain conservative. In many cases, formal review gates are established and enforced as licence conditions. These review gates require that licensees complete specific research and modelling activities, to better understand and model the degradation mechanisms, for the licensees to continue plant operation. In addition, COG provides bi-annual fuel channel seminars at which this research is presented. Through CNSC’s research support program, CNSC has requested the services of researchers at universities and the Chalk River Laboratories to perform third-party reviews or research projects to further validate the models and methodologies.

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Indonesia General ii It is mentioned in Executive Summary, that Canada issues one license for multi-unit NPPs. If one of them is in trouble and legal action needs to be taken by CNSC, how does CNSC impose legal action for that trouble unit only, which is under one license? In Canada, to date, there are no multi-unit NPPs which have different licensees (operators) for units under the same licence. The fact there there is more than one unit does not in any way change the authority of the CNSC to impose legal action against the operator and the action could be clearly directed to a single, specific unit with no impact on the operation of other units. The extent of the action taken would be commensurate and aligned with the nature of the infraction. The CNSC could impose limits and even shut down a single, multiple or all units at one NPP if needed. n286
Indonesia General ii It is mentioned in Executive Summary, that Canada issues one license for multi-unit NPPs. If one of them is in trouble and legal action needs to be taken by CNSC, how does CNSC impose legal action for that trouble unit only, which is under one license? Duplicate question.  See response to Sequence 10.
Indonesia General ii It is mentioned in Executive Summary, that Canada issues one license for multi-unit NPPs. If one of them is in trouble and legal action needs to be taken by CNSC, how does CNSC impose legal action for that trouble unit only, which is under one license? In Canada, to date there are no multi-unit NPPs that have different licensees for units under the same licence. The fact there there is more than one unit does not in any way change the authority of the CNSC to impose legal action against the operator and the action could be clearly directed to a single, specific unit with no impact on the operation of other units. The extent of the action taken would be commensurate and align with the nature of the infraction. The CNSC could impose limits and even shut down a single, multiple or all units at one NPP, if needed. In fact, the legal action would be directed to a specific non-compliance or infraction and would not necessarily apply to all units.
Indonesia Article 7 page 59 Canada conducts  surveillance and monitoring activities to collect real-time information about licensee performance and possible emerging issues. Does CNSN  acquire  real-time data from NPP  e.g power reactor, in and out temperature cooling system,  flow rate of primary system etc?

Duplicate question - see Response to Sequence 15.

Indonesia Article 7 page 59 Canada conducts  surveillance and monitoring activities to collect real-time information about licensee performance and possible emerging issues. Does CNSN  acquire  real-time data from NPP  e.g power reactor, in and out temperature cooling system,  flow rate of primary system etc?

Duplicate question - see Response to Sequence 15.

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Slovenia Article 13 13(a), p.114 Could you inform when the CSA Group standard, N286-12, Management system requirements for nuclear facilities will have to be respected by all nuclear facilities. (What is the period for the alignment with the provisions of N286-12)?

How often the management system is inspected by CNSC?

Does the RB conduct specific inspections on Leadership and Safety Culture?

At present, the licensees of all Class IA (includes NPPs) and IB facilities and uranium mines and mills have aligned their management systems with CSA N286-12 and have standard compliance verification criteria in their licence conditions handbooks.

Management system inspections are incorporated in the CNSC baseline inspection plans. There are an average of 2/year for NPP's that are specific to some management system topics. All NPP inspections include routine observations on some management system elements. For other Class I and UMM licensees, these are done according to the inspection plan, which is based on the risk profile of the licensed activities. The CNSC conducts reactive inspections on management systems as required and involve management system specialists in other more technical inspections.

Inspections on management review are conducted periodically as part of the 5-year baseline compliance plan. This is where the leaders of the organization review data to assess the overall effectiveness of the management system and all programs/activities associated with it.

CNSC staff review and assess the process and results of each safety culture assessment to verify the licensee’s compliance with the regulatory requirements (CNSC REGDOC-2.1.2, Safety Culture requires the licensees to conduct safety culture self-assessments at least once every five years).


Article 13

13(a), p.115

Are the requirements for self-assessment and independent assessment related to management system as well as to safety culture and leadership for safety?

The CNSC expects licensees to assess their performance routinely via self and independent (internal and external) assessments in all activities and to all license requirements, since this is the scope of the management system. The CSA standard N286-12, Management System Requirements for Nuclear Facilities,  is explicit with respect to these activities for safety culture as well as human performance. It also requires the leadership of the organization to continuously assess the effectiveness of the management system in meeting all of its objectives. The CNSC has also published REGDOC-2.1.2, "Safety Culture," which details expectations and guidance regarding the means of assessing safety culture.

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Pakistan Article 14 14(i)(d) Reference section 14(i) (d), Canada may like to share its good practices of Pickering facility with regards to obsolescence management identified by OSART in 2016.

Obsolescence management takes into consideration long-term aging management assessments and the requirements associated with transitioning to decommissioning. The plant’s obsolescence management program supports long-term operation plans and the transition to decommissioning. The obsolescence team consists of individuals dedicated to identification and resolution of obsolescence issues. These individuals implement obsolescence solutions by providing both procurement and design engineering support. This arrangement allows the plant to have a more focused and consistent approach and reduce the number of hand offs between various engineering work groups. This arrangement allows the team to look at wider application strategically to select the most cost effective solution. The co-operation between these groups with different mandates enables the preparation of Condition Assessments which consider full component life span needs. This approach leads to improved safety when approaching the transition to final decommissioning.

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Ukraine Article 8.1 Page 81 According to 8.1 (i), the strategy for ensuring regulatory readiness for SMRs is built on a robust but flexible regulatory framework that provides a sound legal basis. Usually, "robust" and "flexible" are conflicting terms. But is the regulatory framework for SMRs more "robust" or more "flexible" in comparison with the regulatory framework for existing nuclear facilities? The CNSC’s existing regulatory framework is applicable to a wide range of activities with respect to the operation and decommissioning of existing reactor facilities, while providing clear regulatory expectations to applicants applying for site preparation, construction and operation of new reactors such as SMRs.

The framework’s robustness comes from the use of safety principles and objectives drawn from decades of operational experience around the world. These principles and objectives remain a fundamental part of a safety case regardless of technology type or facility size. However, the regulatory framework recognizes that there are many acceptable ways to demonstrate that these principles and objectives have been met. The CNSC has articulated requirements and guidance to provide for the flexibility to demonstrate regulatory requirements are met, and appropriate and defensible safety margins and defence-in-depth are achieved. Regulatory requirements and guidance put the onus on the applicant to demonstrate that proven practices are being used or established and that uncertainties associated with their proposal are being addressed commensurate with a graded approach.

Ukraine Article 7.2 Page 213 One of the topics covered by DIS-16-04 is the licensing approach to a new demonstration reactor. Would you please clarify the term "demonstration reactor" and features of its licensing approach? The term 'demonstration reactor' has no legal meaning in the Canadian regulatory framework. A demonstration reactor typically refers to a full-scale a first-of-a-kind (FOAK) facility at a higher maturity level than a prototype/test reactor but designed to support certain specific demonstration activities established by a licensee.  These may include gathering operating experience and related data for new commercial applications, performing long term tests/data gathering/R&D to support refinements to the generic design in view of an nth of a kind /fleet deployment.  CNSC uses the same licensing process for Class I Nuclear Facilities whether it is a FOAK or an nth-of-a-kind (NOAK). CNSC acknowledges that the information available to support a FOAK vs. a NOAK may vary and require appropriate safety and control measures commensurate with the risk arising from any gaps in operating experience at the time of licensing and throughout the facility lifecyle.
Ukraine General Annex 18, page 278 "The CNSC process for pre-project vendor design review is divided into three distinct phases. …Phase 2: The CNSC confirms that submissions for the specific design demonstrate that the proposed design complies with REGDOC-2.5.2 and related documents. The scope of the review is fixed by the CNSC and usually involves assessment in 19 focus areas: ... fuel design and qualification". Could you please clarify how does CNSC license nuclear fuel (is it deep comprehensive safety analysis of the fuel or there are criteria that must not be exceeded by the fuel and CNSC checks if the nuclear fuel complies with these criteria)? CNSC specialists assess fuel design during Phase 1 and Phase 2 pre-licensing reviews according to the high-level requirements specified in CNSC REGDOC-2.5.2, "Design of Reactor Facilities: Nuclear Power Plants", Section 8.1.1 "Fuel elements, assemblies and design". The vendor of an advanced reactor design is expected to describe the fuel design assembly, such as the fuel matrix, cladding, spacers, support plates, moveable rods inside the assembly, etc. The vendor is also expected to identify all interfacing systems in the reactor core, as well as auxilliary systems for the handling of fresh and spent reactor fuel, and systems for the detection of failed fuel. Per the requirements, the fuel design shall account for all known failure mechanisms, including failure criteria used to support comprehensive deterministic safety analysis.

Beyond the practices for vendor design review, the CNSC requires the fuel design to be qualified for operation, prior to loading it in a reactor. The vendor must demonstrate a verified and auditable knowledge base to confirm that fuel assembly design requirements are met, including the consideration of fuel failure mechanisms. This can be achieved either through documented operational experience with the same type of fuel in other reactors or through a program of experimental testing and analysis. The CNSC authorizes the use of a new fuel design in a reactor based on a thorough review of the safety case presented by the applicant, which addresses all the reactor’s operational states (normal operation and anticipated operational occurrences (AOOs)) and accident conditions. Per REGDOC-2.5.2, the safety case must:

  • Identify, for operational states (normal operations and AOOs) and accident conditions, all the degradation mechanisms that could threaten the capacity of the fuel to perform its design function (note: for operational states, degradation mechanisms are referred to as damage mechanisms; for accident conditions, they are referred to as failure mechanisms)
  • Formulate, for each damage mechanism occurring under operational states, fuel design criteria that, if met, ensure that the mechanism will not prevent the fuel from remaining fit for service;
  • Formulate, for each fuel failure mechanism occurring under accident conditions, fuel design criteria that, if met, ensure that the mechanism is prevented from occurring.
  • Demonstrate, using fuel design criteria for operational states, that 1) the fuel is not damaged and 2) the damage that the fuel may cause to the fuel channel components (such as in CANDUs) is acceptable in the sense that these components remain fit for service.
  • Include a deterministic safety analysis that demonstrates, for the new fuel design, the CNSC dose limits for each of the plant’s design basis accidents, and design extension conditions without severe fuel degradation, are met with sufficient margin. Fuel design criteria associated to fuel failure mechanisms that might occur during those accident scenarios are key inputs to that demonstration. These fuel design criteria must have their foundation in experiments, operational evidence and analyses that identify the limitations of the material properties of the fuel when subjected to operational states and accident conditions. Requirements and guidance regarding these criteria are provided in the CNSC regulatory documents REGDOC 2.4.1, “Deterministic Safety Analysis” and REGDOC 2.5.2.

Depending on the degree of novelty of a new fuel design, CNSC’s regulatory review may include topics such as:

  • reactor physics;
  • fuel thermalhydraulics;
  • thermal, mechanical and material design of the new fuel;
  • Fuel-fuel channel interactions, fuel-heat transport system interactions and seismic qualification;
  • Deterministic safety analysis and probabilistic safety analysis
  • Quality assurance, human factors and training
  • Nuclear criticality safety
Ukraine Article 7 Section 7.2 (ii), page 43 "The CNSC’s regulatory regime defines NPPs as Class IA nuclear facilities and the regulatory requirements for these facilities are found in the Class I Nuclear Facilities Regulations. These regulations require licenses for each of the five types of activities in the lifecycle of a Class IA nuclear facility:
  • license to prepare a site
  • license to construct
  • license to operate
  • license to decommission
  • license to abandon"

Could you please clarify how does CNSC conduct regulatory review on issued license to abandon? What kinds of criteria are used by CNSC for review?

(We would appreciate if you give us a link to the regulation associated with the issue we asked).

Release from CNSC regulatory control occurs when the licensee has successfully decommissioned the facility and restored the site to a state in which it can be released for future use (e.g., green field or brown field [industrial]). The CNSC requires a submission for a licence to abandon or exemption from licensing. This submission must be supported by reports on the results of the decommissioning and site restoration activities, as well as the results of the radiological and environmental monitoring, to demonstrate that the site no longer needs to be licensed under the NSCA.

If unrestricted release is yet to be achieved, perpetual licensing from the CNSC may be required, or oversight by another regulatory or governmental body allows the Commission to exempt the site indefinitely from CNSC licensing (determined on a case-by-case basis) if the risks are determined to be acceptable by the Commission. In order to be granted an exemption, the licensee must present a safety case that demonstrates long-term safety.

Section 4 of the General Nuclear Safety and Control Regulations (GNSCR) outlines the information that must be submitted to obtain a licence to abandon a nuclear substance, nuclear facility, prescribed equipment or prescribed information. Additional information for an application for a licence to abandon is provided under section 8 of the Class I Nuclear Facilities Regulations (CINFR) and under section 8 of the Uranium Mines and Mills Regulations (UMMR).  CNSC regulations can be found at

Section 5 of the Class II Nuclear Facilities and Prescribed Equipment Regulations (CIINFPER) outlines the information that is required in an application for a licence to decommission a Class II nuclear facility. All licensees must submit policies and procedures regarding decommissioning as part of their application.

In reviewing a submission for a licence to abandon, the CNSC must be satisfied that the abandonment of the site, nuclear substance, prescribed equipment or information does not pose an unreasonable risk to the environment, the health and safety of persons, or national security. The abandonment must also not result in a failure to comply with Canada’s international obligations.

Ukraine General Summary, Chapter II, page 19 "During the reporting period, the CNSC continued its progress in enhancing the regulatory framework – which included various regulatory documents relevant to existing NPPs and new-build projects, including emerging small modular reactor (SMR) technologies – and aligning the regulatory framework with international standards (as a minimum). These changes have been introduced into the regulatory framework in a risk-informed way by focusing on technology neutral fundamental safety objectives. Renewals of operating licenses for NPPs (which occur approximately every ten years) were used to introduce new standards and requirements, with provisions for implementation of the new requirements over predefined time periods". Could you please clarify the meaning of word “neutral” when you saying “neutral fundamental safety objective”? The lack of correct punctuation in this sentence likely caused confusion with regards to what the word 'neutral' is in reference to.  It should read "technology-neutral, fundamental safety objectives."  This implies that the fundamental safety objectives would apply regardless of the type of reactor built (technology-H22neutral).
Ukraine Article 15 Page 135; page 255 (annex 15 (a)) According to the information on page 135, for Canadian NPPs, the effective personnel radiation dose limit is set at 50 mSv per year with a 100 mSv limit for a five-year period. From the table "Occupational dose summary for Canadian NPPs, 2016–2018" on page 255, it follows that the maximum individual effective dose of personnel exposure for 2016-2018 is 23.05 mSv and 22.19 mSv (both values ​​refer to the Bruce NPP).

Question: What has been the maximum annual individual effective dose of personnel exposure recorded at Canadian NPPs over the last 10 years (2009-2018)?

The maximum individual annual effective dose of a worker at a Canadian NPP over the last ten years was 29.0 mSv (received in 2012).
Ukraine Article 15 Page 256, Annex 15 (a), second table Canadian radiation protection rules require the licensees to provide an effective radiation dose for NPP personnel for a five-year period beginning on 1 January 2001. The table shows the maximum values ​​for an individual radiation dose over three years (2016-2018) for the five-year period 2016-2020.


  1. Have there been any cases when the individual effective dose of 100 mSv was exceeded over five-year periods since 1 January 2001?
  2. What has been the maximum individual effective dose for the five-yearperiod from 1 January 2001 to 31 December 2018?
  3. Is the effective dose of 100 mSv exceeded for any consecutive five years? Does this take into account five calendar years or any consecutive 60 months?
  1. There have not been any cases in NPPs in Canada where an individual worker has exceeded the five-year effective dose limit since January 2001.
  2. The maximum individual effective dose for the five-year dosimetry period since January 1, 2001 was 74.44 mSv, received in the 2011-2015 dosimetry period.
  3. The Radiation Protection Regulations define the five year dosimetry period and the dose limit is based on the fixed period. The effective dose for NPP workers has been maintained below the 100 mSv limit for the fixed five-year dosimetry period since January 2001. The effective dose has also been maintained below 100 mSv for five-year periods, taking into account consecutive years.
Ukraine Article 7 Section 7.2 (ii), page 43 If the necessary applications are filed with the required information, the Commission may, at its discretion, issue a license that includes multiple classes of licenses (e.g. a license to prepare a site and construct, or a license to construct and operate). A single license may also be issued for multiple facilities, each at a different stage in their lifecycle.

Q.4: The Report states the following:  “Commission may issue a license that includes multiple classes of licenses…”  and in the document “Licensing Process for New Nuclear Power Plants in Canada, May 2008”[hereinafter - Document] (p.4 Chapter 4. Licensing Process for New Nuclear Power Plants): “Separate licenses are granted for each phase in the lifecycle of the nuclear power plant, and would be issued in sequence. However, the applications to prepare a site, to construct and to operate a new nuclear power plant may be assessed in parallel.”

The statement in the Report contradicts the statement in the Document.

Could you please clarify the difference between the information on multiply classes of licenses given in the Report and the information on separate licenses given in the Document?

CNSC replaced the 2008 CNSC publication "Licensing Process for New Nuclear Power Plants" (the Document) with REGDOC-3.5.1, "Information Dissemination: Licensing Process for Class I Nuclear Facilities and Uranium Mines and Mills." The CNSC published Version 2 of REGDOC-3.5.1 in 2017. The Canadian CNS report cites information from the new REGDOC on the approach to licensing, which has evolved since 2008. However, note that the wording in REGDOC-3.5.1 and the CNS report still lacks precision - this will be corrected in the next revision of REGDOC-3.5.1 so that it is consistent with the Nuclear Safety and Control Act and CNSC regulations.

To clarify, the Nuclear Safety and Control Act and CNSC regulations use the term “class” to define types of activities and facilities that are governed by CNSC licences. However, they do not prohibit the issuance of a single licence to cover more than one life-cycle phase of a given type of licenced facility. For example, the Commission could issue a single licence that covers both site preparation and construction of an NPP (which is a Class IA facility).

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Posted By Article Reference Question Comment Answer
Finland General Section II VDNS, in the list of planned activities to improve safety, installation of a passive containment filtered venting system at Bruce B is mentioned. What was the reasoning behind this action and will it be implemented in other NPPs as well? The Bruce units already have a Emergency Filtered Air Discharge System; however, this system was designed for a single unit design basis accident. Analysis has shown that the system has capacity to handle severe accidents in two units. Given that the four units in the Bruce stations have a common containment, a decision was made to enhance the capability by installing a passive containment filtered venting system to protect the containment in a four-unit event. Similar systems will be installed at both the Bruce A and Bruce B stations. The Point Lepreau station has had a passive containment filtered venting system installed since 2012 and OPG is installing a similar system in the Darlington station. There are no plans to install a passive containment filtered venting system at the Pickering station due to the planned shutdown in 2024. Alternative strategies are in place to protect the containment at Pickering.
Finland Article 14 Summary Concerning the challenge 3 from the 7th review meeting, its said in the summary that Canada's current fleet of multi-unit NPPs are currently undergoing refurbishment or life-extension activities, which has delayed the necessity of this initiative." What about the plans concerning the Pickering NPP where Units 2 and 3 will be maintained in safe storage states (there is currently no decision for life-extension acitivities)? We agree that this challenge remains open. The safe storage of Pickering Units 2 and 3 is regulated under the same CNSC 'Licence to Operate' as the six operating units at Pickering.  The robust requirements currently in place ensure a safe and orderly shutdown of additional units.  For example, the licence require OPG to implement a strategy for the end of commercial operations and transition to safe storage. The licence is valid until 2028 and so will be in place during the period when all the currently operating units at Pickering are planned to be shut down permanently.
Finland Article 7 Introduction Sections  D and E, VDNS, Articles 7 and 18 Articles 7 and 18, Introduction Sections D and E, VDNS; How do you define regulatory safety goals and criteria for SMRs? In which conditions it would be OK to relax the requirements for severe accident management, i.e., DiD level 4, solely based on PSA numbers or are there also some deterministic requirements? In which conditions it would be OK to decrease the planning and protection zones around a SMR, solely based on PSA numbers or are there some deterministic requirements?

The term 'SMR' has no legal meaning in the Canadian regulatory framework.  All reactor facilities are regulated as Class 1A Nuclear Facilities with regulatory requirements to be applied commensurate with a graded approach taking into account novelties, complexity of activities and potential for harm to workers and the public.  Further information about the graded approach is articulated in two regulatory documents:  REGDOC-3.5.3, Regulatory Fundamentals and REGDOC-1.1.5, Supplemental Information for Small Modular Reactor Proponents. The use of a graded approach is not a relaxation of requirements, but rather the application of requirements in a manner commensurate with the risks and characteristics of a facility or activity (taking uncertainties into account).

  1. Safety goals for all reactor facilities are found in REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants and RD-367, Design of Smal Reactor Facilities. 
  2. CNSC design requirements articulate that a safety analysis of the plant design is expected to include hazard analysis, deterministic safety analysis, and probabilistic safety assessment (PSA) techniques. The safety analysis is expected to demonstrate achievement of all levels of defence in depth, and confirm that the design is capable of meeting the applicable expectations, dose acceptance criteria and safety goals.   
  3. As stated in REGDOC-1.1.3: Licence Application Guide: Licence to Operate a Nuclear Power Plant, an application for a Licence to Operate is expected to describe an emergency preparedness program that meets the requirements of REGDOC-2.10.1, Nuclear Emergency Preparedness and Response. Section 1.3 of REGDOC 2.10.1 further describes accident management and its links with emergency preparedness and the principle of defence in depth for power reactors.
  4. In Canada, planning and protection zones are not prescribed.  The appropriateness of proposed exclusion zone sizes are based on dose acceptance criteria complemented by consideration of factors including but not limited to: security requirements, environmental factors, evacuation needs and land use. The dose acceptance criteria are used to determine the minimum acceptable size of an exclusion zone.  Additional factors can subsequently be used to determine if the zone size needs to be increased. Ultimately, evaluation of the acceptability  of the zone size is an integrated consideration of the factors mentioned above.  More information about the process to establish project-specific zones can be found in CNSC Discussion Paper DIS 16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges available at:
Finland General Chapter II - summary Canada proposes that the suggestion from the 7th Review meeting would be closed, although work will continue in the next reporting period to address the remaining three Category 3 CSIs. We are not sure that the suggestion can be closed and this needs to be discussed during the 8th Review Meeting. Canada acknowledges the comment
Finland General Chapter II It is mentioned that Canada developed a policy "roadmap" for SMRs and Canada continued to refine its regulatory approach to ensure suitability for SMRs, including the establishment of additional guidance information, review of regulatory process and adjustment of regulatory capacity." This could be considered as as good performance for adjusting regulatory framework for enabling new technologies. Canada acknowledges the comment
Finland Article 7 7.2 (i) (b), page 39

The publication of discussion papers which are used to solicit early public feedback on regulatory initiatives could be considered as a good performance. Some interesting topics published during this reporting period include for example SMRs and human performance.

Canada acknowledges the comment

Finland Article 7 Chapter II, VDNS It is said that The Canadian regulatory framework has been aligned with the IAEA safety standards, which themselves have been demonstrated to fulfill the principles of the VDNS." IAEA safety standards are consensus documents which leave some expected safety levels open including for example safety criteria. For certain topics there is a need to make a decision on national implementation of IAEA Safety Standards. Have you made any further interpretations in the national nuclear safety regulations what the VDNS principle 1 means in practice? As noted in the Canadian report, during the reporting period, Canada published amendments to the Class I Nuclear Facilities Regulations and the Radiation Protection Regulations that addressed lessons learned from Fukushima, including those linked to measures that would reduce the likelihood and/or consequences of severe accidents.

The CNSC can typically publish/revise its REGDOCs faster than it can amend the regulations.  Consequently, many REGDOCs with content related to defence-in-depth were published or revised in the 2014 and 2015 timeframe and the content took into account relevant lessons learned from Fukushima:

  • REGDOC-2.3.2, Accident Management (version 2)
  • REGDOC-2.3.3, Periodic Safety Reviews
  • REGDOC-2.4.1, Deterministic Safety Analysis
  • REGDOC-2.4.2, Probabilistic Safety Assessment (PSA) for Nuclear Power Plants
  • REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants
  • REGDOC-2.6.3, Aging Management

More recently, in the reporting period for the 8th review meeting, CNSC also published the following REGDOCs containing significant information related to defence-in-depth:

  • REGDOC-1.1.2, Licence Application Guide: Licence to Construct a Nuclear Power Plant
  • REGDOC-1.1.3: Licence Application Guide: Licence to Operate a Nuclear Power Plant
  • REGDOC-1.1.5, Supplemental Information for Small Modular Reactor Proponents
  • REGDOC-2.6.1, Reliability Programs for Nuclear Power Plants

In recent years, the CSA Group has also published revisions to standards related to defence-in-depth:

  • CSA N290.0, General Requirements for Safety Systems of Nuclear Power Plants
  • CSA N290.19, Risk-informed decision making for nuclear power plants
  • CSA N290.17, Probabilistic Safety Assessment for Nuclear Power Plants
  • CSA N290.18, Periodic safety review for nuclear power plants
Finland Article 14 Article 14, VDNS It is said that the operating NPPs in Canada are in line with the Principle 2 of VDNS. What is the severe accident management strategy of operating CANDU reactors for cooling the molten core? The CANDU design, being a pressure tube reactor, has the benefit of having a separate low pressure moderator that provides a large source of water. In the event of a severe accident, the strategy is to maintain the molten core within the calandria vessel (moderator tank). This vessel contains a large amount of water that will be allowed to steam off into the containment structure. All CANDU reactors in Canada have moderator makeup capabilities to provide ongoing cooling capability. In the event of a failure of the calandria vessel, the molten core would move into the shield tank, which is a large biological shielding tank full of light water that will provide cooling to the molten core. Again, this tank has relief capacity to the containment structure and water make-up capability to allow for long-term cooling.
Finland General Whole report Canadian national report doesn't list any challenges. What do you think are currently the biggest challenges in Canada related to nuclear safety? CNSC:
  1. Consolidate all elements of CNSC safety policy into a single document
  2. Establish dose constraints for all Class I facilities and standardize regulatory practice for derived release limits
  3. Formalize the practice of CNSC inspector exchanges between licensee locations
  4. Document existing practices used to ensure a comprehensive, regular review of the objectivity and independence of CNSC inspectors
  5. Ensure the radiation protection requirements are consistent with IAEA General Safety Requirements (GSR) Part 3

Industry: Establishing a consistent approach to evaluating alternative approaches that may be adopted in SMR designs (particularly advanced reactor SMRs) against existing regulatory expectations that were developed for more traditional large reactor designs.

Finland Article 10 Article 10 It is mentioned that in April 2018, the CNSC published REGDOC-2.1.2, Safety Culture. What requirements for safety culture existed before that and how the publication of the new REGDOC has influenced the safety culture programs of the licensees? Since the REGDOC is relatively new, are there any CNSC review/inspection findings yet concerning licensees' safety culture? If there are, could you give some examples?

Prior to the CNSC's publication of REGDOC 2.1.2, Safety Culture, there were no requirements to monitor or assess safety culture.  However, licensees of NPPs and other major  facilties in Canada have been conducting safety culture assessments for a number of years.  The priority for, and monitoring of, safety culture at NPPs was well-established as a result of lessons learned from industry events at places such as Chernobyl and Davis Besse and Fukushima.  Examples of guidance included the following.

WANO Guideline on How to Evaluate the SOER 2003-2 Davis Besse states “The self-assessment should be conducted periodically (such as every cycle, biennially, etc.) as determined by the station.”

INPO Guideline on SOER 2002-4 Davis Besse states – “While a frequency for conducting the self-assessment is not specified, a nominal frequency of 2 years has become the norm. If the decision is made to conduct the self-assessment less frequently than biennially, determine what means are used by senior managers and corporate staffs to measure and monitor the health of the safety culture at the station between self-assessments.”

An example of licensee provisions that existed before the publication of REGDOC-2.1.2 can be found in OPG's Nuclear Safety Oversight Standard, which specifies that each site has to perform a nuclear safety culture assessment every 3 years. This frequency was based on the CNSC and CANDU Owners Group recommendations at the time of 5 years and 3 - 5 years, respectively. 

To date, the main influence of the REGDOC has been for licensees to integrate security culture in their assessments in addition to safety culture. For clarity, REGDOC 2.1.2 concerns how licensees monitor and assess their own safety (and security) culture, thus the primary CNSC review/inspection findings would be of this process (not of the state of licensee's own safety culture).

Finland Article 10 Article 10 It is said that during the reporting period, the CNSC completed a regulatory safety culture self-assessment. what were the most significant findings and actions decided? The Regulatory Safety Culture Self-Assessment resulted in 5 recommendations. Each of the 5 recommendations were reviewed and accepted by CNSC management and assigned a response lead with supporting actions.  The 5 reccomendations were:
  1. Providing ongoing coaching and mentoring to supervisors, managers, and executives on the leadership characteristics necessary for maintaining a health regulatory safety culture.
  2. Developing a problem identification, resolution, reporting, and communication tool.
  3. Develop an overarching regulatory safety culture policy.
  4. Develop strategies to sdure critical technical and regulatory knowledge is actively managed and readily available.
  5. Conduct a follow-up regulatory safety culture assessment in 3-5 years.

They are described in more detail in the response to Sequence 120.

Finland Article 10 Page 91. It is said: "The foundation of safety culture is further established by constantly examining nuclear safety, cultivating a “what if?” approach to safety planning and preparation....” Can you give an example on of how “cultivating a What if? approach” is done in practice? The Traits of a Healthy Nuclear Safety Culture include a questioning attitude. Individuals avoid complacency and continuously challenge existing conditions and activities in order to identify discrepancies that might result in error or inappropriate action. All employees are watchful for assumptions, anomalies, values, conditions, or activities that can have an undesirable effect on plant safety. This trait is applicable to staff in all areas including safety planning and preparation whether it's in a meeting, during field work, knowledge worker tasks, etc.

As an example, per OPG governance, all major elements of the nuclear emergency plan are tested/drilled every five years, with a full-scale integrated exercise once every three years, per site.  The OPG drill and exercise program improves the ability to prepare for and respond to potential future nuclear emergencies that are identified usinga "what if" approach. Through simulated drills/exercises, OPG staff and external partners gain an increased understanding of critical emergency procedures, and learn how to conduct themselves more appropriately during various nuclear emergency incidents.

Finland Article 10 Page 91. It is said that “Senior management is …track the effectiveness of safety programs…” (p.91) Can you give a practical example on a  safety program developing safety culture and how the senior management in practice track the effectiveness of this safety culture development? At OPG, senior management has developed processes to encourage and track the effectiveness of safety programs and demonstrate through action that safety is of overriding priority. Some of these programs include regular World Association of Nuclear Operator (WANO) reviews, Nuclear Safety Review Board (NSRB) reviews, Nuclear Safety Culture Assessments, and Independant Assessments. The results of these are rolled up to the Nuclear Executive Commitee, Nuclear President and Chief Nuclear Officer.
Finland Article 10 Page 92. A safety culture monitoring panel (NSCMP) process is mentioned. Can you explain how this NSCMP panel is formed and how does it work? The Nuclear Safety Culture Monitoring Panel (NSCMP) is a team of experienced individuals with diverse backgrounds who meet periodically to review inputs most indicative of the health of the nuclear safety culture. For example, OPG's NSCMP monitors the process inputs that are indicative of the health of the organization’s nuclear safety culture to identify strengths and potential concerns that merit additional attention by the organization. The panel has two major functions:
  1. Review emergent issues that could impact nuclear safety culture health to ensure the issues are appropriately addressed.
  2. Periodically assess nuclear safety culture trends or potential issues and provide a report to the Site Leadership Team.

NEI 09-07 provides a framework to monitor nuclear safety culture. It includes and augments the current periodic nuclear safety culture assessments. NEI 09-07 “describes the industry approach to assessing and addressing nuclear safety culture issues. The goal is to provide an ongoing holistic, objective, transparent and safety-focused process, which uses all of the information available to provide an early indication of potential problems, develop effective corrective actions and monitor effectiveness of the actions.”

The NSCMP meets at least quarterly to assess the sites nuclear safety culture.  The panel uses, at a minimum, the following process inputs for its assessment:

  • Station Condition Records and Corrective Action Plans
  • Performance Trends
  • CNSC inspection results
  • Nuclear Safety Culture Self Assessments
  • Industry Evaluations
  • Operating Experience
  • Self Assessments / Benchmarking
  • Behavioural Observations
  • Employee Concerns
  • Workforce issues
  • Nuclear Oversight Audits

Site-specific data across the organization is collected, compiled and reviewed by the panel. Specific recommendations and trends are presented to the Site Leadership Team.

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Posted By Article Reference Question Comment Answer
Norway Article 6 6(b) It is stated that ““Canada has a mature nuclear industry with an excellent safety record spanning several decades.” Could you provide some data that support this statement? The maturity of the industry members can be summarized by considering the tenure of the NPP licensees at their respective sites.  OPG has operated Darlington and Pickering NPPs since its inception in 1999.  Prior to that, effectively the same staff operated those NPPs (as well as Bruce A and B) under the organization Ontario Hydro, which was the operational predecessor of OPG.  Bruce Power was formed in 2001, drawing on OPG staff who were already working at Bruce A and B.  Bruce Power has operated Buce A and B since that time.  NB Power has effectively been the operator of Point Lepreau since it came online in 1983.

The nuclear safety record can be substantiated using data, including those in recent CNS reports.  The following statements are true for NPPs in Canada durng the reporting periods for the 6th, 7th, and 8th review meetings (similar statements could be made regarding earlier periods).

  • Maximum doses to workers were well below annual regulatory limits (very few worker doses were above licensee action levels).
  • Radiological releases were very low (generally less than 1% of the derived release limtis, with very few above licensee action levels).
  • The estimated maximum dose to the public from any of the NPPs was very low (highest of all annual values was 0.0090 mSv).
  • There were no accidents or serious process failures.
  • The operational events that occurred were resolved through expected elements of defence-in-depth, such as design elements and proceduralized operator actions.
  • The record of conventional worker safety was very good (typically much better than that of comparable industries).
Norway Article 7 7.2(iii) (b) Inspections How many inspections are performed annually at the NPPs? This is dependent on the number of units at the NPP as well as the status of major projects.  Point Lepreau (a single unit station) was scheduled for 84 inspections in 2019/20 (a combination of Type II and field inspections), while Darlington (a four-unit station undergoing refurbishment) was scheduled for 125 inspections in 2019/20.
Norway Article 7 7.2(iii) (b) Inspections What are the criteria when developing the inspection plan? The CNSC's risk-informed approach, and associated criteria, for planning inspections are described in the answer to Sequence 3.

The annual compliance plan is based on the 5-year compliance baseline plan, activities that were not completed during the previous year and any additional compliance activities that may be required (e.g., compliance activities to address recent licensee performance history).

The planning for stations with units that are undergoing refurbishment or shutdown must also consider additional compliance activities unique to these changes in operating states.

Norway Article 7 7.2(iii) (b) Inspections Is the inspection plan shared with the licensees? The annual plan is sent to the licensee towards the beginning of the fiscal year, however it does not contain proposed dates, merely a list of all inspections planned to be carried out in that fiscal year.
Norway Article 7 7.2(iii) (b) Inspections How long time in advance is the licensee notified about an inspection? Are there also unannounced inspection? The notice period depends on the type of inspection. For a Type II inspection that requires a high level of licensee interaction/preparation, the licensee is notified 60 days prior to the start of the inspection. 

The length of notification for a field inspection can vary substantially based on the amount of licensee involvement necessary. Should the field inspection need to observe specific work in the field, the licensee will receive reasonable notice in order to coordinate the observation with the inspectors. Some field inspections, however, would fall under the category of unannounced inspection should they not need licensee involvement. This built-in flexibility allows inspectors to react to developing situations at the NPP.

Norway General Annex 11.2(a) Which is Unit 0? In a multi-unit station, the main control room will have a station for Unit 0, which monitors safety and process systems that are common to all of the reactor units.

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Republic of Korea

Posted By Article Reference Question Comment Answer
Korea, Republic of General 14 With reference to page 14 of the Canada national report, the decommissioning of Gentilly-2 is scheduled between 2059 and 2064. When will the decommissioning process start and what is the expected duration of the actual decommissioning activities? The preparation and final decommissioning are planned for 2059-2064. On December 28, 2012 Gentilly-2 was shutdown and stabilization activities began in order to prepare the plant for decommissioning. In December of 2014, sate storage state (wet) was reached, with all spent fuel removed from the reactor and stored in the irradiated fuel bay.

From the summer of 2015 to 2021, all fuel is being moved to dry strage in CANSTOR units where it is passively cooled by natural air convection.

All of the used fuel will remain in the CANSTOR until 2050 when the fuel transfer to a deep geologic repository will begin.

Korea, Republic of General 14 Until when will the spent fuel dry storage facility be operated at the Gentilly-2 site? What will be the legal position of the dry storage facility if the spent fuel is not delivered to the disposal facility even after the decommissioning of the nuclear power plant is complete? The spent fuel dry storage facility will be operated until 2050 when the fuel will be transferred to a deep geological repository (DGR). All spent fuel will be transferred to the DGR before decommissioning begins in 2059. The facility will continue to be regulated by a licence to decommission, while the DGR (or other facility if, for some reason, the spent fuel is not delvered to the DGR) will be regulated under its own licence.
Korea, Republic of Article 7 46 With reference to article 7.2(ii)(a), the table of page 46 of the Canada national report states that REGDOC-1.1.5 was drafted for licensing of small modular reactors. As Korea is also expecting the safety review for an SMR (SMART), could you provide the draft version of REGDOC-1.1.5? REGDOC-1.1.5 was published in August 2019 and is now available on the CNSC external website for viewing. The address is:
Korea, Republic of Article 7 57, 80 The CNSC uses various means for compliance promotion and has high-level communication channels between the NPP licensees and the CNSC (e.g. CNSC executive forum).
  • Is there any high-level meeting between the CNSC and licensees to discuss nuclear safety culture and its enhancement?
  • How does the CNSC promote licensees' compliance of regulatory requirement on safety culture (e.g. REGDOC-2.1.2, Safety Culture)?
  • A working level meeting was held between members of the CANDU Owners' Group (COG) and members from the Human and Organizational Performance Division (HOPD) of the CNSC in late 2019 to begin discussions regarding implementation of CNSC REGDOC 2.1.2, Safety Culture.  In addition, CNSC staff periodically observe safety culture monitoring panel meetings at the NPPs, where safety culture and its enhancement are discussed.
  • For clarity, REGDOC 2.1.2 concerns how licensees monitor and assess their own safety culture; thus the primary CNSC review/inspection findings would be of this process (not of the state of the licensee's safety culture).  Working level discussions between the CNSC and COG are helpful in promoting understanding of the requirements in REGDOC 2.1.2, as are CNSC staff's observations of licensee's safety culture monitoring panel meetings.
Korea, Republic of Article 7 59, 70 The CNSC site inspector generally leads the inspection team.
  • How many CNSC site inspectors are working at each site? Is there any principles regarding the assignment (designation and rotation) of site inspectors?
  • Is there an internal procedure or guidance which describes site inspector's role, inspection procedure and protocols with licensees?
  • What is the proportion of site inspectors in the annual man-year for NPP inspections?
  • The number of site inspectors depends on the number of units at the site and any major activities taking place, such as refurbishment. A single unit station, such as Point Lepreau, has three to four inspectors, in addition to a site supervisor who is also a designated inspector.  A site such as Bruce, which has eight units, has eight inspectors plus a site supervisor.  The CNSC site office at Bruce is currently set to add two additional inspectors during upcoming refurbishment activities. When site inspectors are hired, they are assigned to a specific location. The CNSC encourages inspectors to participate on inspections at other licensee locations, and twice a year all of the inspectors gather to share OPEX and lessons learned. It was noted by the IRRS mission, that these practices are not currently formalized in the management ystem. The CNSC has accepted the suggestion to formalize these protocols.
  • The CNSC management system has procedures that are followed by inspectors when it comes to conduct of an inspection as well as the protocol for communication with the licensee. "Conducting an Inspection" is the corporate procedure followed by all inspectors, regardless of service line, however the Directorate of Power Reactors also has procedures that dictate how each type of inspection is to be performed. A site inspector's role is detailed in their work description, including a summary of their duties and responsibilities, as well as all of the skills and knowledge requirements they are expected to have for each level of inspector.
  • There are 35 CNSC NPP site inspectors that spend approximately 35% of their time conducting inspections (Type I, Type II, and field inspections).  Their other duties include surveillance and monitoring, event review, and compliance assessments.  That effort of site inspectors represents 45% of the total CNSC effort on NPP inspections.
Korea, Republic of Article 8 77 The CNSC has recently developed the Mobile Inspection Kit to automate inspection and reporting. Please provide more details on the overview and implementation of the Mobile Inspection Kit. The Mobile Inspection Kit project was launched to support inspectors from divisions whose inspectors are not stationed at the licensee's site.  The project provided them with tablets that enabled them to conduct their inspections efficiently and issue a preliminary summary of findings to the licensee prior to leaving the site.
Korea, Republic of Article 10 91 It is described in the report that Canadian NPPs are promoting "just culture." Is there specific policy, procedures, processes, or educational programs supporting the promotion of just culture? The Traits of a Healthy Nuclear Safety Culture include a safety-conscious work environment where personnel feel free to raise safety concerns without fear of retaliation, intimidation, harassment, or discrimination. The station creates, maintains, and evaluates policies and processes that allow personnel to raise concerns freely.

Organizations also have problem identification and resolution programs that identify, analyze, document, and respond to problems with a level of effort and timeliness commensurate with their significance.

Regarding educational programs/promotion, there are multiple areas where "just culture" is promoted, including internal communications, training, station handbooks and associated roll outs, all-hand sessions/town-hall discussions, various meetings, etc.

Korea, Republic of Article 10 93 Bruce power included security culture in its 2016 self-assessment on safety culture.
  • What are the pros and cons of integrated assessment on security culture and safety culture based on experience?
  • Where there any contradictions found between safety and security culture?
  • Does Bruce Power plan to have an integrated assessment on safety and security culture in 2020?
  • Did Bruce Power identify synergy effects, limitations and recommended approaches through integrated questionnaire and assessment from the safety and security interface perspective?
  • The pros of conducting the safety and security culture assessments together are: 1) good synergy between the topics 2) better participation than running the 2 separate due to "assessment fatigue" 3) it allows for a joint action plan to address any common issues in safety and security culture, while the cons include 1) it results in a much larger survey which takes additional time to analyze 2) it requires people that understand both security and afety culture for the assessment team, people knowledgable in both areas are few in numbers
  • There were no contradictions found between safety and security culture, however; it was noted that the staff that work in the security field responded much different to the security culture related questions than the general staf population
  • Yes, Bruce Power intends to continue with integrated Safety and Security Culture Assessments
  • One of the greatest benefits was that the integrated assessment resulted in most staff realizing that there is a lot of synergy between safety and security culture. We assessed 62 characteristics of a health safety culture and 20 security culter characteristics, in doing so we discovered that that there was a gap between how the general population and the security staff responded to the security culture characteristics, while the safety culture questions resulted in general alignment between work groups. We also found that using our safety culture assessment methodology worked quite well for teh integrated assessment with security and intend to follow a similar methodology for our 2020 assessment. We would recommend others use a similar process as staff are generally familiar with it and it is proven to give the desired results.
Korea, Republic of Article 10 95 In April 2018, the CNSC published REGDOC-2.1.2 and formalize requirements and expectations for licensee safety culture. REGDOC-2.1.2 describes that when safety culture in the NPP is assessed, safety leadership process is checked.
  • What is the safety leadership process?
  • What area does the CNSC focus in the assessment?
  • The characteristics in Appendix A (the normative framework) of REGDOC 2.1.2  are based on IAEA GS-G-3.5, IAEA NSS-7, both of which recognize the importance of leadership in safety / security culture.  The indicators listed with the characteristic give more detail on the characteristic.  In this case, the safety leadership process characteristic has indicators addressing, for example, communication, teamwork, management practices (e.g. visibility in the field, not abusing authority) and corporate practices  (e.g. change management, decision making).  The ensemble of indicators can be used to provide insight on the characteristic, and the ensemble of characteristics can give some insight into an organization's safety culture.
  • The CNSC focuses on the monitoring and assessment processes in REGDOC 2.1.2 Safety Culture.  That is, once a safety policy is written, how is that safety policy lived?  For Nuclear Power Plants, more fulsome assessments following the guidance in the REGDOC- the process of a safety culture assessment is the focus.
Korea, Republic of Article 10 96 The CNSC conducted a self-assessment on safety culture in 2018.
  • Please describe in detail the assessment methods and process?
  • How does the CNSC implement and manage the follow-up actions?
  • The assessment methodology was developed to align with the International Atomic Energy Agency’s (IAEA) Safety Reports Series No. 83: Performing Safety Culture Self-assessments, to the extent practicable. The assessment used a multi-method approach to gather and collect data from document reviews, focus group discussions and the administration of a questionnaire on safety oversight culture perceptions. The CNSC also utilized the Nuclear Energy Agency's (NEA’s) The Safety Culture of an Effective Nuclear Regulatory Body.
  • The implementation and management of follow-up actions was developed and approved by CNSC management and tracked to completion using a management action plan (MAP). The MAP identified individuals accountable for delivery of actions addressing each of the recommendations from the self-assessment.
Korea, Republic of Article 11 100 The importance of managing “Human and Organizational Factors (HOF)” is increasing to ensure safe operation of nuclear power plant. Please provide detailed explanation regarding operating plants:
  • Roles of the regulator in ensuring safety management of HOF;
  • HOF under the scope of regulation (human performance program, training, fatigue management, organizational change management, etc.);
  • Departments (branches, teams, groups, etc.) of the regulatory organization responsible for HOF verification (inspection and assessment);
  • Number and qualifications of those regulators working on HOF verification/inspection/assessment;
  • How to distinguish regulatory oversight areas between HOF and safety culture; and
  • How to reflect the close relationship of HOF and safety culture in safety issues into an analysis on human and organizational credibility.
  • During licensing and compliance activities, CNSC staff review and inspect licensee provisions in the context of the regulatory framework for HOF, which is described in the answer to question ii). CNSC staff also extensively review and inspect the licensees’ management systems, which include the governance, processes and procedures that are used to manage HOF.
  • The CNSC’s safety and control area (SCA) framework covers activities that enable effective Human Performance Management through the development and implementation of processes that ensure a sufficient number of licensee personnel are in all relevant job areas and have necessary knowledge, skills, procedures and tools in place to carry out their duties safely. The key domains in the SCA framework related to human and organizational factors (HOF) are listed below. HOF considerations may also apply to other aspects of the SCA framework, such as outage management performance, severe accident management and recovery, maintenance, waste management practices, security facilities and equipment:

Management System

  • organization
  • performance assessment, improvement and management review
  • operating experience
  • change management
  • safety culture
  • management of contractors

Human Performance Management

  • human performance program
  • personnel training
  • personnel certification
  • initial certification examinations and requalification tests
  • work organization and job design
  • fitness for duty

Operating Performance

  • procedures
  • reporting and trending

Safety Analysis

  • human actions in deterministic and probabilistic safety analysis

Physical Design

  • design governance (human factors in design)
  1. and iv) There are four divisions, within the Directorate of Safety Management, that conduct regulatory activities related to HOF at the CNSC (in parentheses are the numbers of staff conducting regulatory activities relating to HOF in each division):
    1. Human and Organizational Performance Division (12)
    2. Training Program Evaluation Division (7)
    3. Personnel Certification Division (9)
    4. Management Systems Division (12)

The CNSC staff with responsibilities relating to HOF considerations include graduates with qualifications and experience relevant to the specific aspects of HOF that they review and assess. A number of them have higher degrees in human factors/ergonomics and psychology and professional certifications related to HOF. In addition, resident site inspectors at the NPPs conduct regulatory oversight, which includes the work arrangements and activities carried out by licensee's staff and contractors.

  1. Human performance relates to work activities that a person or team carries out and the results of those work activities. HOF include the factors that influence human performance, i.e., that help people to perform their tasks successfully, and the influences that can detract from the desired performance, including that related to safety. One aspect of HOF is safety culture, which CNSC defines as “the characteristics of a work environment, such as values, rules and common understandings that influence worker perceptions and attitudes about the importance that the organization places on safety”. NPP licensees are required to document their commitment to fostering safety culture and to carry out comprehensive, systematic and rigorous safety culture assessments at least every five years. Consideration of safety culture by an organization promotes introspection, reflection and understanding of the underlying values that are driving the organization, to recognize the current status and to identify where the organization wants to be. Besides safety culture, there are other HOF that influence human performance, such as the human performance program, personnel training and fitness for duty. These relate more directly and tangibly to people’s work, through considering the relationships between humans, the technology and the organization that provide a context to the specific work activities.
  2. vi) Clause 3d.1 of the Class I Nuclear Facilities Regulations requires an application for a licence, other than a licence to abandon, to contain “the proposed human performance program for the activity to be licensed, including measures to ensure workers’ fitness for duty”. CNSC staff are currently developing a regulatory document on human performance, which will provide additional requirements and guidance that will help distinguish the various elements that are considered in assessments of HOF. CNSC staff considers that safety culture is one aspect of HOF, as described above. In the CNSC safety and control area (SCA) framework, there are 14 SCAs, and each of them is broken down into specific areas. Under SCA 1, Management System, one of the specific areas is Safety Culture. Under SCA 2, Human Performance Management, there are six specific areas: Human Performance Program, Personnel Training, Personnel Certification, Initial Certification Examinations and Requalification Tests, Work Organization and Job design, and Fitness for Duty. While some of these specific areas clearly correspond to HOF, the intent is for the specific area Human Performance Program to consider an integrated overview of how HOF are understood and managed by the licensee. The approach being considered for regulatory oversight of HOF draws on IAEA-TECDOC-1846 Regulatory Oversight of Human and Organizational Factors for Safety of Nuclear Installations (2018) and Daniellou, F., Simard, M. and Boissières, I. (2011). Human and organizational factors of safety: a state of the art. Number 2011-01 of the Cahiers de la Sécurité Industrielle, Foundation for an Industrial Safety Culture, Toulouse, France (ISSN 2100-3874) (Available at
Korea, Republic of Article 12 110 With reference to Article 12, page 110 of the Canada national report, it is stated that all Canadian NPPs have implemented human performance programs that emphasize detection and correction of human errors with a focus on monitoring individual behaviours. Also, Canadian NPP licensees have developed a coding scheme to effectively identify and track the causes of human error related events. With respect to the provided information in the article, Korea would like to inquire the following questions:
  1. Are there any other additional methods utilized to detect human errors besides monitoring individual's behaviour?
  2. With regard to the coding system, what type of codes are defined in the system and what is the criteria for determining and classifying human-error codes?
  3. How is the coding scheme applied to investigate and analyze human-error events?
  1. The individual can self-monitor through a questioning attitude to detect off-normal situations. Three specific tools to detect an error prior to an event are independent, concurrent and peer checks. The part of a pre-job brief that forsees credible as well as worst-case consequences on the plant, on personnel and on the environment if error goes undetected also sensitizes workers to potential situations which may help detect them before they become problems. Trending of smaller issues with human performance tools is also a way to help detect broader issues. 
  2. Human performance station governance describes the criteria for determining and classifying human-error codes. Coding includes error precursors (distractions, high-workload, repetitive actions, etc.), human performance tools and performance mode (skill, rule or knowledge-based).
  3. The coding helps to determine what broke down in order to determine what actions are required to address the event.
Korea, Republic of Article 12 111 With reference to Article 12, page 111 of the Canada national report, it is stated that CNSC regulatory document RD-204 requires licensees to have a documented fitness-for-duty program for certified workers. With respect to the provided information in the article in question, Korea would like to inquire the following questions:
  1. Would it be possible to provide the requirements described and regulated in the RD-204 document?
  2. To comply with the requirement related to fitness for duty, which activities are enforced in Canadian NPPs? And how does the CNSC inspect the fitness for duty for personnel of NPPs in Canada?
  1. RD-204 contains a high-level requirements for NPPs to have a documented fitness-for-duty program for certified workers ( More detailed requirements for fitness for duty of NPP workers are found in:
    1. REGDOC 2.2.4, Fitness for Duty: Managing Worker Fatigue
    2. REGDOC 2.2.4, Fitness for Duty, Volume II: Managing Alcohol and Drug Use
    3. REGDOC 2.2.4, Fitness for Duty, Volume III: Nuclear Security Officer Medical, Physical and Psychological Fitness
  2. Fitness for duty requirements apply to all high-security sites in Canada, which includes NPPs. When assessing compliance of licensee programs and their implementation against fitness for duty requirements, the CNSC uses standard compliance assessment techniques and enforcement strategies (e.g. on-site inspections and technical assessments of licensee submissions). Typically when a new  regulatory document is published, CNSC conducts a type I inspection on site to ensure that the licensees’ programs meet the requirements. Type II inspections and small field inspections are then conducted to assess the output of the licensees' programs against the requirements. As REGDOC-2.2.4, Fitness for Duty: Managing Worker Fatigue was recently published, Type I inspections are occurring. During these inspections, CNSC staff verify many aspects, such as the licensees' measures to manage fatigue, including measures to control the number of hours of worked, the number of consecutive shifts, and rest periods between shifts. With respect to REGDOC-2.2.4, Fitness for Duty, Volume II: Managing Alcohol and Drug Use, future Type I inspections will verify that licensees have clear policy statements related to alcohol and drug use and impairment at work and support programs for workers to seek help; these inspections will also verify that alcohol and drug testing is being conducted for safety-sensitive and safety-critical positions according to the requirements in the regulatory document. With respect to REGDOC-2.2.4, Fitness for Duty , Volume III: Nuclear Security Officer Medical, Physical and Psychological Fitness, CNSC staff will verify that all Nuclear Security Officers have passed medical, psychological, and physical fitness testing by verifying their certificates; CNSC staff also conduct inspections and oversight to ensure, for example, that the physical fitness testing (layout, equipment used etc.) meets the requirements in the regulatory document.
Korea, Republic of Article 12 112 With reference to Article 12, page 111 of the Canada national report, it is stated that the CNSC continues to work with licensees conducting PSR through review against modern standards, and modifications in response to the Fukushima accident have included human factors in design considerations. With respect to the provided information in the article in question, Korea would like to inquire the following questions:
  1. How does the CNSC evaluate, from the perspective of human factors, the modifications of design in response to the Fukushima accident in the results of PSR?
  2. Does CNSC also evaluate the design modification of the remote shutdown room and emergency response room in the results of PSR?
  1. In 2015-16, CNSC staff reviewed samples of activities to verify that NPP licensees were properly taking into account human factors in design for specified processes and artefacts that resulted from Fukushima action items. CNSC staff also verified that the licensees were adequately developing and modifying their procedures that resulted from these action items. Following these inspections, CNSC staff observed the execution of operator field actions necessary to supply water from the pumper (emergency mitigating equipment) to the boilers and observed SAMG validation exercises in the simulator. These activities were effectively independent of PSR.
  2. CNSC staff evaluate the emergency support facilities in PSR safety factor 1 against requirements in CNSC REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants (Section 8.10.2 for secondary control room and Section 8.10.3 for emergency support facilities). The facilities include a technical support centre and on-site emergency response facilities. CNSC staff focus on the installation of the safety parameter display system). CANDU designs have a secondary control area outside of the main control room that can shut down the reactor.
Korea, Republic of Article 13 114 13 (a) General management system requirements

The Class I Nuclear Facilities Regulations include decommissioning as one of the activities that license applicants should establish and implement in their management system

In this regard, Korea would like to inquire following questions

  1. How is Canada's regulatory framework around decommissioning organized?
  2. Does N286-12, N286.7-16 or REGDOC-2.1.1 described in Canada's QA-related regulatory documents contain any QA-related information that may be applied during the decommissioning phase? If so, how is it described?
  1. The CNSC requires that planning for decommissioning take place throughout a licensed activity’s life-cycle, and that both a preliminary decommissioning plan (PDP) and a detailed decommissioning plan (DDP) be prepared for approval by the CNSC. Specific requirements for a licence to decommission a Class I or Class II nuclear facility or uranium mine and mills facility are listed in section 7 of the Class I Nuclear Facilities Regulations (CINFR), section 5 of the Class II Nuclear Facilities and Prescribed Equipment Regulations (CIINFPER), and section 7 of the Uranium Mines and Mills Regulations (UMMR). Information listed in section 3 of the General Nuclear Safety and Control Regulations (GNSCR) and the general requirements section of the CINFR and UMMR are also required. Examples of the requirements for an application for a licence to decommission a Class I nuclear facility include:
    1. effects on the environment and the health and safety of persons that may result from the decommissioning, and the measures that will be taken to prevent or mitigate those effects
    2. proposed measures to control releases of nuclear substances and hazardous substances into the environment
    3. 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 national security, including an emergency response plan

In addition to the NSCA and the regulations made under it, the CNSC has developed regulatory documents, which are a key part of its regulatory framework for nuclear activities in Canada. They provide additional clarity to licensees and applicants by explaining how to meet the requirements set out in the NSCA and the regulations made under it.

Regulatory guide G-219, Decommissioning Planning for Licensed Activities (currently under revision as REGDOC 2.11.2), provides regulatory expectations to licensees regarding the preparation and content of preliminary and detailed decommissioning plans for activities licensed by the CNSC.

Additional requirements and guidance on decommissioning of nuclear facilities and other locations where nuclear substances are managed, possessed, or stored are set out in CSA Group standards N294.0, Decommissioning of facilities containing nuclear substances. For example, this document stipulates that safety assessment of the decommissioning shall be performed to identify hazards to workers and the public from both routine decommissioning activities and credible accidents during decommissioning. In such a case where the end-state for decommissioning results in a waste disposal site, a licensee must satisfy all regulatory requirements for a radioactive waste disposal facility and demonstrate safety via a safety case and post-closure safety assessment of a waste disposal facility. Further requirements on the safety case and safety assessment can be found in CNSC REGDOC-2.11.1, Waste Management Volume III, Assessing the Long Term Safety of Radioactive Waste Management.

Regulatory documents and industry standards may be referenced in the information supplied by an applicant in support of its licence application, and are used by CNSC staff to evaluate the application. These regulatory documents and standards become part of the licensing basis for the facility once the licence is issued.

All of these regulatory requirements are aimed at ensuring licence applicants demonstrate that they are qualified to carry out the activity and that they will make adequate provisions for health, safety, environmental protection, and security. CNSC staff rigorously review all submissions to determine if the proposed decommissioning safety and control measures in a licence application, and the documents that support the application, are adequate and meet the applicable requirements.

  1. Yes, the requirements for quality assurance would be established when the Commission issues a licence to decommission. That licence would cite CSA Group standard N286-12 directly (or, it would be included in the licensing basis for the facility by reference).
    1. N286-12 states “This Standard applies to the life cycle of the nuclear facility from the initial conception through completion of decommissioning”.  So, its quality assurance requirements extend to decommissioning.
    2. CSA Group standard N294-19, Decommissioning of Facilities Containing Nuclear Substances requires a quality assurance program to be implemented and states “The quality assurance program should be commensurate to CSA N286 or other recognized and accepted quality assurance program”.
    3. REGDOC-2.1.1 also refers to CSA N294.
Korea, Republic of Article 13 115 According to Canada's regulatory framework, licensees perform a self-assessment of their management system, and then report the result to the CNSC, which the CNSC periodically reviews.

In this regard, Korea would like to inquire following questions

  1. Does the regulatory body conduct direct inspection or audits on major licensees? If not, how reliable are their self-assessments?
  2. How often does the CNSC review and does CNSC review all licensees or some?
  1. The CNSC conducts routine inspections of the licensees' management systems, including those at NPPs. Top licensee management conducts broad self-assessments on management system effectiveness. The CNSC inspects this process as well as all other types of assessments. Licensee self-assessments are reliable based on historic CNSC observations and they are conducted in adherence to the licensees' schedules.
  2. CNSC staff inspect and review the assessment processes and results, on average, twice every five years for NPPs.
Korea, Republic of Article 14 121,  123~124, Annex 14 (i) (b) With reference to article 14 (i) (c), pages 123-124 of the Canadian national report, it is described that the Pickering pilot whole-site PSA results met the CNSC’s expectations. Whereas, it seems that the deterministic analyses for Level 4 defence-in-depth (severe accidents), required by REGDOC-2.4.1, are being done case-by-case for the existing reactors, to the extent of updating certain analyses (article 14 (i) (b), page 121). Furthermore, this document requires best-estimate analysis for those accidents, without considering uncertainties to the extent for the case of Level 3 defence-in-depth. With respect to this, what is the regulatory position on handling of the uncertainties involved in the deterministic and probabilistic safety assessments for severe accidents, including relevant decision making? CNSC REGDOC-2.4.2, specifically requires the conduct of sensitivity, importance and uncertainty analyses. In Canadian PSA practice for the severe accident analysis conducted as part of the Level 2 PSA, major assumptions and key parameters are assessed for importance through sensitivity analysis. The most important parameters are selected to be included in uncertainty analysis for severe accidents, which generates results to support systematic uncertainty and sensitivity analysis of the Level 2 PSA.

For deterministic safety analyes, severe accident analysis is best estimate, and no estimation of uncertainties are requested. The degree of uncertainty in this analysis is high when compared with the analysis performed for design basis events.

Korea, Republic of Article 14 131~132
  1. According to page 131, some part of developed assessment methodologies for operation beyond 210,000 EFPH have already been incorporated into CSA standard N285.8. Is there any plan to incorporate the whole methodologies into CSA standard? If yes, when do you expect it to be?
  2. Does EFPH limit in the table on page 132 mean that utilities Bruce and Pickering Units 5 to 8 were approved by CNSC to operate to the EFPH limit?
  1. The CSA Group standard N285.8, Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors, specifies mandatory technical requirements and acceptance criteria for fitness-for-service assessments to support the safe operation of pressure tubes from a structural perspective. it also includes non-mandatory (optional) engineering procedures and material properties that may be used in evaluations to demonstrate compliance with the technical requirements and acceptance criteria.
    CSA N285.8 is maintained and updated by the responsible committee of technical experts with the objective of keeping it up-to-date and technically valid. As relevant industry R&D activities progress, other engineering procedures and material properties may be included in future updates of the standard. However, some engineering procedures and material properties might not be incorporated into the CSA Group standard, for example due to the complexity and evolving nature of the subject area, and would instead be documented in industry R&D reports.
  2. Yes, the limits shown in the table on page 132 are the effective full power hour limits for pressure tube operation approved for those NPPs by the CNSC. 
Korea, Republic of Article 14 251 Regarding the implementation of Article 14 of the Nuclear Safety Convention, Annex 14(b) of the CANADA's national report explained that along with the aging management programs required by REGDOC-2.6.3, Canadian licensees have developed a series of periodic inspection programs and plans that expand the minimum inspection and testing program requirements to address operational and safety issues."  In addition, the most significant of these programs and plans are described.

With respect to this, please explain the details of the regulatory body's review and inspection activities on the licensee's periodic inspection program and plans (the aging management programs required by REGDOC-2.6.3) provided in the Annex 14(b).

To clarify, compliance verification criteria for periodic inspections are established in CSA Group Standards that are referenced in the licence conditions handbook for each licensee. In addition, licensees are required to establish aging management programs in accordance with REGDOC-2.6.3, Aging Management. The aging management programs include life cycle management plans (LCMPs) for major components including fuel channels, steam generators, feeders and reactor internals, which may contain in-service inspection plans that are greater in scope than the periodic inspection programs. The compliance verification criteria for a licensee’s periodic inspection programs reference the standards CSA N285.4, Periodic Inspection of CANDU Nuclear Power Plant Components, CSA N285.5, Periodic Inspection of CANDU Nuclear Power Plant Containment Components and CSA N287.7, In-Service Examination and Testing Requirements for Concrete Containment Structures. CSA N285.4 includes supplemental Clauses 12, 13 and 14 for fuel channels, feeders, and steam generator tubing, respectively. CNSC oversight activities include regulatory reviews of:
  • PIP plan revisions to verify compliance against the requirements of the CSA Group standards, including updated editions of the standards, as applicable;
  • reports prepared by licensees following each outage to verify compliance with PIP plans; and
  • dispositions of inspection findings to verify compliance with established fitness-for-service guidelines.

Regulatory reviews of the LCMPs focus on compliance with the nine attributes of an effective aging management program (listed in Appendix A of REGDOC 2.6.3). As the LCMPs are periodically updated, subsequent CNSC reviews focus on the significant changes to the inspection scope, frequency, and timing. Any changes must be shown to be technically justifiable, based on the results of previous inspections and updated understanding of degradation mechanisms and OPEX. The feedback of OPEX, and other information, into the inspection scope, is a part of the PLAN-DO-CHECK-ACT approach, from REGDOC 2.6.3. These routine regulatory reviews may require meetings and/or CNSC requests for information from the licensee. CNSC also has inspection activities for both PIPs and LCMPs. For example, CNSC staff verify that the LCMPs and PIPs are adequately implemented, the personnel conducting the inspections are qualified to the applicable CSA requirements, the inspection techniques are able to reliably obtain the required data, and the related documentation is up-to-date and meets the requirements in the plans.

Korea, Republic of Article 15 134 With reference to Article 15 in page 134, it is stated that Paragraph 3(1)(f) of the General Nuclear Safety and Control Regulations requires that an application for a license contain proposed action levels.
  1. Please provide examples of the action levels that have been contained in the license application documents.
  2. Please provide reporting levels in terms of radiation dose, radioactivity or radioactivity concentration in the REGDOC-3.1.1 Reporting Requirements for Nuclear Power Plants.
  1. Here are some examples of OPG's action levels (Pickering NPP).
    1. 2 mSv for external whole body dose above the planned dose (dosimeter set point) for a job
    2. 2 mSv for internal tritium-oxide exposure above the planned exposure level
    3. 20 mSv for whole body (effective) dose for a calendar year
    4. 37 and 3.7 kBq/sqm for alpha and beta surface contamination, respectively, in Zone 1
  2. CNSC REGDOC-3.1.1, Reporting Requirements for Nuclear Power Plants, requires NPP licensees to report events where the limits in the Radiation Protection Regulations are exceeeded (see response to Sequence 96 for a list of those limts).
Korea, Republic of Article 19 184 With reference to Article 19.6, Korea would like to inquire the following questions: Is there a process to permit or approve the restart of the plant when a reportable event with reactor trip occurs? If so, please explain the following items.
  1. overall process from event reporting to approval of plant restart
  2. legislation or procedures which are the basis of the process
  3. screening criteria (e.g. significance of events) for the process
  4. decision makers (who decide to apply the process to an event and who approve the restart) and participants to the decision making
When an event is found to be a serious process failure or where the determination as to the cause and/or extent of the condition has proved inconclusive (i.e., a serious process failure cannot be ruled out), a request for restart of the reactor shall be submitted in writing and approval to restart the reactor must be obtained from the CNSC in accordance with REGDOC-3.1.1, "Reporting Requirements for Nuclear Power Plants". If there is sufficient assurance that the cause of the serious process failure has been resolved and it is now safe to return the facility to service, a CNSC authorized person has the authority to give the consent to the licensee to proceed with the restart of the reactor.
  • The written request for restart of the reactor shall include the following information:
  • description of the event
  • causes of the event
  • consequences and safety significance of the event
  • recovery plan including corrective actions and fitness for service assessment on the systems/components impacted from the failure, if applicable (this shall be completed prior to restart)
  • a statement regarding plant readiness to resume safe operation (this shall include any conditions that the licensee proposes to impose upon reactor restart and/or subsequent reactor operation to ensure safe operation of the nuclear facility)
  • extent of completion of the conditions mentioned in the statement regarding plant readiness to resume safe operation
Korea, Republic of Article 18 180(169) REGDOC 2.5.2 (Design of Reactor Facilities: Nuclear Power Plants) only considers a single failure when designing a plant. How does the CNSC regulate or manage the multiple failures? For accident analysis, events are classified into abnormal operational occurrencess, design basis accidents, and beyond-design basis accidents (which include design extension conditions and severe accidents), depending mostly on their frequencies. It is mostly design basis accidents that are used for design, and this is not limited to single failure events; it may include multiple failures. In addition, the analysis of dual failures under design extension conditions is also required by CNSC REGDOCs 2.4.1, Determinisitci Safety Analysis and 2.4.2, Probabilistic Safety Assessment (PSA) for Nuclear Ppower Plants. These dual failure analyses must meet acceptance criteria (mostly related to dose limits) and demonstrate that the event does not progress to a severe accident.

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China General Appendix C/P194 August 2, 2017 and March 4, 2018, Failure of primary heat transport (PHT) pump seals at Bruce A  Unit 3 and 4, the report just has the incident description for Bruce A  Unit 3.

Question: Are the causes and consequences of these two incidents the same ?After the incident of Unit 3, had the actions been taken in-time by Bruce A to prevent the incident happened at Unit 4?

The causes and consequences where essentially the same for both events, althought the duty crew response was quicker for the Unit 4 event due to additional specific training as a result of the Unit 3 event. As a result, the impact of the Unit 4 event was slightly less than of the Unit 3 event. The root cause was determinned to be rubbing of the tertiary seal on the pump shaft due to insufficent clearances. A design change to the same design as the Bruce B units has been implemented in all Bruce A units. This was not implemented in the time between the Unit 3 and Unit 4 events.
China General Summary/P24 It is mentioned that " Installation of auxiliary shutdown cooling system at Darlington during the next reporting period".

Question: Is it a new design modification? If yes, which auxiliary functions of the shutdown cooling system are added or supplemented by the system? Do other CANDU units need to add this system?

The auxiliary shut down cooling  system is a new design modification for Darlington NGS only.  The system adds two new circulating pumps and associated piping to increase the station’s flexibility in managing heat sink configurations during unit outage conditions.  Other CANDU units already have base designs that provide greater heat sink flexibility than did Darlington’s base design, so similar installations are not planned at other facilities.

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Belgium Article 14 14 (ii), p. 130 CNSC staff members review and approve certain operational changes or other changes to items in the licensing basis. Which operational changes or other changes to items in the licensing basis are subject to review and approval? Some licensee changes are simply reviewed by CNSC staff.  For example, the licence conditions handbook (LCH) identifies a certain set of licensee documents (roughly 200 documents for an NPP licensee) for which the licensee must notify the CNSC in writing when it makes changes.  This "written notification" mechanism is mentioned at the end of Section 7.2 (ii) (a).  For some documents, the licensee must notify the CNSC before making the change; for others, the licensee can notify the CNSC at the time of making the change.  In either case, CNSC staff reviews the change to confirm that the licensee will remain within the licensing basis.  Examples of licensee documents that require written notification of change to the CNSC include management system manuals, ALARA program documents, periodic inspection plans for critical components, and maintenance program documents. As noted in the report, certain other licensee changes are subject to approval.  For example, CNSC REGDOC-2.4.2, Probabilisitc Safety Assessment (PSA) for Nuclear Power Plants requires NPP licensees to obtain the acceptance of CNSC staff for the PSA methodologies and computer codes it plans to use.  As another example, CSA standard N285.4-14, Periodic Inspection of CANDU Nuclear Power Plants requires NPP licensees to obtain the acceptance of CNSC staff for changes to periodic inspection program documents.
Belgium Article 19 19 (ii), p. 176 Operating limits for Canadian NPPs are identified in the safe operating envelope (SOE) documentation. Changes to these limits that may negatively affect safety require appropriate justification by operations support staff and are reviewed by the CNSC. Do these changes to the limits require approval by the CNSC as well? See response to sequence 70. The Licence Conditions Handbook (LCH) elaborates the licence conditions in details and requires that the licensee notifies the CNSC of any changes to safe operating limits at least 30 days before their implementation. Namely, the licensee submits the modified SOE document to CNSC at least 30 days before the revised SOE document is implemented. If the change to safe operating limits has a negative impact on safety, CNSC staff informs the licensee and CNSC (Commission) approval is required.
Belgium Article 8 8.1 p. 70 The Directorate of Nuclear Substance Regulation is responsible for some licences related to NPPs that fall outside the scope of the operating licence (e.g., licences for nuclear substances and radiation devices, transport). Does this mean that an NPP has a separate licence for nuclear substances and radiation devices? Yes, in some cases, NPP operators hold multiple licences if the licence to operate does not cover all activities that need to be licensed by the CNSC.
Belgium General D.2 page 21 Several plans are drafted for the Long Term Operation of NPPs, such as the Integrated Implementation Plan (IIP), the Continued Operations Plan (COP), the Sustainable Operations Plan (SOP). How do these plans relate to one another? Is there an integrated approach? OPG has a strategic plan that describes the purpose of these various documents (SOP, stabilization actvities plan (SAP) and IIP) and their relative timelines. However, there is currently no published regulatory document that describes the regulatory framework.  OPG has requested and received regulatory expectations from the CNSC that are specific to the end of commercial operations.
Belgium Article 18 Article 18 The CNSC was engaged in many pre-licensing vendor design reviews for SMRs. These reviews may require specialized expertise that is not always the same expertise as is needed for operating reactors. How does CNSC ensure that it has the necessary expertise for these reviews? CNSC's internal technical support rrganization, the Technical Support Branch (TSB), has a broad skillset of generalist and specialist capabilities to support regulatory activities across all classes of activities regulated under the NSCA.  Fundamental to their skills is not only technical knowledge but also an understanding of  the underpinning objectives of the requirements and guidance  and how they can be applied for different technological approaches.  For example, it is important to understand the fundamental objectives of fuel qualification in order to critically assess a new fuel being proposed by a vendor.  The fuel may be different, but fundamental safety onbectives still need to be met to a high degree of certainty.  The vendor design review (VDR) process, which has been used for more than 10 years, enables CNSC staff to maintain these critical thinking skills and apply them to novel designs. Where specific technical proficiencies are needed, CNSC has processes to obtain assistance and or training through international and domestic  arrangements with, for example research facilities, regulatory cooperation agreements etc.  The VDR process provides an early opportunity to assess where additonal skills training or support may be needed before the level of review depth requires those skills.

The CNSC has described its Strategy for Readiness to Regulate Advanced Reactor Technologies (/eng/reactors/research-reactors/other-reactor-facilities/readiness-regulate-advanced-reactor-technologies).  It includes informing technical resources of the upcoming challenges, identifying specific training requirements and developing and delivering training to appropriate staff and management in a timely manner.

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Lithuania General Page 4 Please provide practical examples of safety measures that licensees implemented to introduce new standards and requirements during renewals of operating licences for NPPs. During the last licence renewal cycle, periodic safety reviews (PSR) were conducted at the Bruce, Darlington and Pickering NPPs. This was the first time that PSR has been tied to the renewals of the operating licences. Point Lepreau is currently conducting a PSR in preparation for it's next licence renewal. The intgrated implementation plans (IIPs) from the PSRs became part of the licensing basis upon the renewal of the operating licences. As part of the IIPs, a number of improvements, both physical and in governance, were undertaken to improve safety. For example, at Bruce Power, updates to the final safety report were undertaken to meet the requirements of REGDOC-2.4.1, Deterministic Safety Analysis. This resulted in a new section ot the report for common mode events. Examples of physical improvements, include upgrades to fire protection systems to meet requirements in CSA Group standard N293, Fire Protection for CANDU Nuclear Power Plants and seismic-related improvements to meet requirements in the CSA N289 series.
Lithuania General Page 4 Please clarify what kind of safety measures were implemented by licensees to strength defence in depth and enhance emergency response. Licensees installed additional means to provide water to the steam generators, primary heat transport system, moderator and shield tank (suppplied by portable pumps through hoses into these systems using quick connect fittings). Darlington and Point Lepreau have also installed passive containment filtered venting systems. These additional systems provide additional defence in depth beyond the normal, back up and emergency systems already installed in the NPPs.  Se the response to Sequence 75 for additional information.
Lithuania Article 7 Page 20 The CNSC supplements each licence with a more detailed licence conditions handbook (LCH), which describes the requirements and expectations for the NPP and provides a comprehensive list of the codes, standards, and regulatory documents that form part of its licensing basis. Please clarify what is the practical benefit of licence conditions handbook? The introduction of the LCH has allowed the Commission to standardize (to a great extent) the licences it issues for NPPs (and other facilities), while allowing CNSC staff flexibility to document more detailed compliance verification criteria that can be unique to each licensee.  CNSC staff can select the criteria and organize the information in the LCH in a manner that suits the licensee, and is relatively free to change the LCH following its change control process.  Major changes, including amendments to the licence itself, are still executed with the involvement of the Commission, but these are rare.  CNSC include licensee-specific details in the LCH, such as details of when the licensee is implementing new requirements (e.g., a newly-published CNSC REGDOC) or lists of licensee documents that require notification to CNSC staff if the licensee changes them.  The LCH also provides CNSC staff's non-mandatory guidance that is organized according to the licence conditions - such guidance would be inappropriate in the licence itself.   
Lithuania Article 7 Page 42 The CNSC have the authority to include in licences any term or condition that it deems necessary for the purposes of the NSCA, including the requirement for a financial guarantee. Please clarify the term „financial guarantee“? A financial guarantee is a condition in the licence that ensures that a licensee makes adequate provision for the safe decommissioning of existing or proposed nuclear facilities (the provision is typically in the form of guaranteed funds). Operationally, the Commission may also require financial resources be available for termination of licensed activities other than for decommissioning of nuclear facilities.
Lithuania Article 15 Page 88 Ins certain countries the annual equivalent eye lense dose was reduced to 20 mSv  (50 mSv in a single year or 100 mSv in five years period).  How is the equivalent eye lense dose monitored in your country? Do you use special Hp3 dosimeters (or Hp(0,07) or Hp(10)) for monitoring of the eye lense dose or do you use dose assessment according to the workplace monitoring results? If the eye lense dose is measured with dosimeters, what principles are used to select workers for that (e.g., individual annual dose should exceed 0.3 of the annual dose limit of 20 mSv, or special workplace conditions)? Which period of monitoring is selected? Where do you store the results of monitoring of workers eye lens dose (National dose registry or...)?
  1. The current dose limit for the lens of the eye as specified in the Radiation Protection Regulations for Nuclear Energy Workers is 150 mSv per one year dosimetry period. NPP licensees are currently using an estimation method to determine lens of eye doses.
  2. Currently, NPP licensees are not using Hp3 dosimeters for direct monitoring of the lens. Licensees are investigating the use of Hp(0.07) for estimating dose to the lens. Workplace monitoring has been undertaken to assess hazard profiles and to validate modeling.
  3. Not applicable. NPP licensees are not currently routinely using dosimeters to directly monitor for doses to lens of the eye.
  4. Not applicable, as per the response to Sequence 2.
  5. In Canada, the National Dose Registry is the repository for dose records. Licensees are also required to maintain records of worker exposures.

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Japan Article 8.1 p 69, 8.1(b)

Organization of CNSC staff

2nd  paragraph

CNSC has technical support branch internally. How are the technical support work such as safety research, analytical evaluation, etc. shared with organizations outside CNSC? Upon completion of a research project, and following the final report, an abstract is created and posted on CNSC's website. The abstract provides the public (and any interested organizations outside the CNSC) with an idea of what research was conducted and what conclusions were reached. The abstract then has a link to the final report. The following is the link on the CNSC's website:
Japan Article 10 p 96,10(d)

last paragraph

What kind of education and training is CNSC conducting for fostering and developing the safety culture of regulatory body?

If possible, please provide the summary presented at the public Commission meeting in October of 2018?

Training and education in support of safety culture cuts across a multitude of practices and courses offered to CNSC staff including:
  • CNSC speaker series for all CNSC staff on the leadership characteristics necessary for a healthy regulatory safety culture.
  • New Director’s program and new Directors’ Community of Practice (CoP)
  • Internal and external learning and development activities that help create a collaborative work environment

A presentation was made to the Commission members on the self-assessment of regulatory safety culture at the CNSC.  This included background information, a summary of methods, findings including areas of strength and improvement, recommendations, and follow-up actions. the fndings included 6 strengths, which noted that CNSC staff:

  • perceive safety as the overarching priority in decision making
  • generally feel comfortable using existing mechanisms to raise issues and concerns
  • display a willingness to collaborate and share expertise across the organization
  • feel safe reporting mistakes and are able to express their professional opinions at work
  • believe they receive the required training, and possess the appropriate competencies to fulfill their duties
  • believe that leaders and managers take action based on the results of regular self-assessments and audits

The summary also noted, in 7 areas for improvement, that CNSC staff:

  • feel there is a need for improvement in communicating the rationale behind decisions. This does not refer to Commission decisions, which are well documented in the CNSC`s record of proceedings, but rather line management decisions
  • perceive that at times managers do not solicit challenges to their own assumptions or seek varied perspectives
  • feel that the timeliness of management decisions could be improved and that the decision making process (other than that of the Commission`s) should be more open, particularly to challenge varied perspectives
  • perceive the ability of leaders and managers to resolve differences of opinion requires increased attention
  • feel that the flow of information could be improved between the hierarchical levels within the organization
  • believe there is a need to have a common understanding of safety oversight culture amongst CNSC staff
  • believe there is a need to manage and transfer the technical and regulatory knowledge of long-standing employees

The assessment identified 5 recommendations to address the areas for improvement; CNSC management developed an action to respond to each of them  (see response to Sequence 120 for details).  

Japan Article 15 p135, 6th paragraph Please explain in more detail about "Planning for unusual situation". In terms of planning for unusual situations during normal routine work, radiation exposure permits are used to control worker dose. These permits contain back out criteria where workers would leave the area should dose conditions change, which would be an unusual situation. There is also the area of infrequently performed tasks where dose needs to be strictly managed. An example of this would be the case of a fuelling machine stuck on a fuel channel. Maintenance personnel would need to enter containment to perform maintanence to allow the fueling machine to be safely removed from the fuel channel. In some cases, there will be irradiated fuel in the fuelling machine. Given this unsual situation, detailed radiation protection plans would be prepared to ensure the workers exposures are ALARA.
Japan Article 15 p140, 6th paragraph According to that paragraph, environmental monitoring is duty for licensees. However, with reference to page 142, it is stated that the monitoring is also implemented by CRMN. How do they avoid the duplication in their monitoring programs? Health Canada's Canadian Radiological Monitoring Network (CRMN) is a national network that routinely collects air particulate, precipitation, external gamma dose, drinking water, atmospheric water vapour, and milk samples for radioactivity analysis to serve as the basis for health risk assessment and event identification. As part of this, the cross-Canada Fixed Point Surveillance (FPS) network monitors radiation dose to the public in real-time due to radioactive materials in the terrestrial environment, whether they are airborne or on the ground. It includes contributions from both natural and man-made sources.  The CRMN and FPS have a much larger scope than the environmental monitoring performed by licensees, and as an independent data set, the data are publicly available on the Government of Canada website.

Industry and Health Canada have memoranda of understanding to allow sharing of data collected in the monitoring plans. While there may be some duplication of parameters monitored, there is some benefit in that there are independent data sets that can be used to ensure accurate data is being collected. The CNSC also conducts independent environmental monitoring to confirm the licensees programs are adequately monitoring environmental impacts.

Japan Article 16 p151, 3rd paragraph It is mentioned that provinces ensure arrangements for evacuation centres in their own municipal. How is the support plan for people who evacuated to other provinces formulated? In Canada, the Province of Ontario and the Province of New Brunswick host NPPs. Neither Ontario nor New Brunswick have emergency planning zones that border or overlap neighboring Provinces or Territories or US States and therefore it is unnecessary to plan to evacuate members of the public to evacuation centres outside of the affected Province.
Japan Article 16 p151, 3rd paragraph There appears to be no mention of complex disaster (such as radiation accident caused by earthquake). How do they prepare for it? Nuclear emergency management in Canada is effectively integrated at the national level into an all-hazards framework, as described in the Emergency Management Act. This Act and its implementing policies and procedures focus on mitigation/prevention, preparedness, response and recovery to all types of emergencies including both natural and human-induced hazards and disasters.  The National Emergency Response System (NERS) provides for the harmonization of all-hazards response. The all-hazards Federal Emergency Response Plan (FERP) is intended to coordinate federal response and planning, and to support the provinces and territories.

The linkage between nuclear emergency preparedness and response and the all-hazards NERS is described in the Federal Nuclear Emergency Plan (FNEP).  The FNEP works within the NERS and augments the FERP for a nuclear or radiological emergency. Health Canada, administers the FNEP, an annex to the FERP, to coordinate federal technical response and support the provinces and local authorities in managing radiological consequences of an emergency. Partner federal agencies maintain their own all-hazards and radiological plans as applicable, integrated with the FERP and FNEP. The FNEP includes annexes for each relevant province linking the FNEP concepts to the provincial arrangements.

Ontario and New Brunswick further coordinate their respective nuclear emergency plans with the NERS all-hazards concepts through Ontario’s Provincial Nuclear Emergency Response Plan (PNERP) and New Brunswick’s (NB ) Point Lepreau Nuclear Nuclear Generation Station (PLNGS) Off-Site Emergency Plan. FNEP annexes provide the link between the FNEP and the provincial plans. Both the PNERP and the Point Lepreau Nuclear Off-Site Emergency Plan refer to the roles and responsibilities of federal organizations. Considered together, the FNEP and the provincial plans ensure that the emergency management system is generally based on the results of the hazard assessment, and that complex disasters can be managed in a consistent manner.

The Ontario PNERP and the NB PLNGS Off-site plan planning basis is informed by the safety assessments conducted by the operators of each of their nuclear facilities. The safety assessment considers a range of low probability events and usually categorizes them as follows:

  • internal events: events caused by random component failures and human error originating in the plant;
  • internal fires: fires originating in the plant;
  • internal floods: floods originating in the plant;
  • external natural hazards: events like earthquakes, high winds, floods, freezing rain, meteorites, geomagnetic storms and solar figures; and
  • external human-induced hazards: events such as airplane crashes and accidents at nearby industrial facilities

The planning basis considers the impacts of a wide range of accident progressions and the plans provides detailed planning and preparedness to mitigate the off-site effects of Design Basis Accidents as well as provides additional tools and mechanisms to mitigate the effects of Beyond Design Basis Accidents (BDBAs), including severe accidents, which are considered even more unlikely than Design Basis Accidents.

Japan Article 18.3 p 174, 18(iii)

last paragraph

The National Report says that the safety parameter display system contributes to enhance safety by improving human factors and human-machine interface.

If any, please provide example of improvement of human factors and human-machine interface after installation of the safety parameter display system.

The application of human factors engineering (HFE) processes provides a robust human-system interface that supports efficient and safe operations, guards against human error and allows for error detection and recovery.  For the safety parameter displays, HFE’s goals are:
  • to ensure that the display presents all of the required parameters in the most appropriate format and organization to support correct diagnosis of accidents,
  • to provide alerts to the operators when actions are required, and
  • to ensure that the safety parameter display system (SPDS) is an integrated aspect of the control room operator interface and supports overall operating goals and strategies.

The success of the SPDS, including the human factors considerations in its design, would be evident in its effectiveness in diagnosis tasks.  The application of human factors in the SPDS design would provide assurance that operator requirements, capabilities and limitations were accounted for and that the design is sufficiently tolerant and robust against human error.

Japan Article 15 p139, 2nd paragraph How is the risk of the radioactive waste (both high level and low level)  evaluated in the environmental risk assessment? Further to the information provided in response to Sequence 141, the previous question, major facilities including NPPs and waste management facilities are required to conduct an environmental risk assessment (ERA) in accordance with CSA Group standard N288.6, Environmental risk assessment at Class I nuclear facilities and uranium mines and mills. The ERA characterizes the risk posed by releases of nuclear and hazardous substance and physical disturbances (stressors) on representative human and non-human biota.

ERAs for waste management facilities consider the risk from managing and storing radioactive waste at the facility. The type of waste that may be produced or stored onsite includes used fuel, radioactive solid waste, radioactive liquid waste, radioactive gaseous waste as well as non-radioactive solid, liquid and gaseous waste.

The emissions from the storage of the various types of waste at waste management facilities are considered in the ERA which includes the ecological risk assessment for the environment and the human health risk assessment for members of the public.

Japan Article 15 p139, 1st paragraph In the line 15th in that paragraph, it is mentioned that "ERA is updated with the data collected with effluent and environmental monitoring programs, special studies and new science". Please provide additional information about "special studies and new science". The ERA is a systematic process that identifies, quantifies and characterizes the risk posed by contaminants and physical stressors in the environment. It provides science-based information to support decision-making. The ERA is reviewed and revised periodically and is informed from operational experience, monitoring and special investigations, the incorporation of advances in scientific knowledge and, if available, Indigenous traditional knowledge. As a result, the initial ERA evolves through the lifecycle of the facility and activity and becomes more site specific.

Special studies (or special investigations) are conducted to:

  • provide data required to reduce uncertainty and confounding factors in the ERA
  • increase knowledge of the behavior of contaminants and physical stressors in the environment (i.e., refining environmental parameters)
  • investigate monitoring results that indicate potential deviation from the transport or effects predictions in the ERA or the licensing basis

Examples of the types of special studies that may be conducted to serve as inputs to ERAs include:

  • avian or ungulate studies (e.g., telemetry) to document local habitat utilization patterns previously assessed at the regional level to refine exposure and risk estimates
  • studies to develop site-specific transfer factors or BAFs to reduce uncertainties associated with generic values (example, BAFs for multiple aquatic trophic levels)
  • population genetic studies of Great Lakes fishes to determine whether fish losses via impingement and/or entrainment are associated with a local (with respect to licensed facility) fish populations or lake-wide populations (i.e., local populations would be at greater risk)
  • atmospheric and/or thermal plume monitoring to confirm original predictions with respect to plume behavior around nuclear facilities

The results of these special investigations serve as inputs along with the other environmental monitoring data for the next revision of an ERA.

New Science

Another input in revising an ERA includes the incorporation of recent developments in scientific knowledge (or “new science”). New science may include revisions to toxicological reference values as well as bioaccumulation factors (BAF). Example:

  • incorporation of recent scientific development in the role of OBT in tritium environmental transport and exposure

The evolution of the ERA through accumulated monitoring data and new scientific knowledge decreases uncertainty, thereby enhancing the licensee’s ability to assess the significance of deviations from predictions of environmental performance.

Japan Article 8.1 p 79, 8.1(f) 1st paragraph The National Report says that Members of public can participate in the Commission hearings and meetings, and those can be viewed online on the CNSC website and transcript of public hearings and meetings are made available.

To what extent the contents of hearings and meetings between CNSC staff and licensee(s) where Commission Members do not participate are made publicly available?

The CNSC makes all formal correspondence related to licensing publicly available. A list of correspondence is presented on the website with contact information to request copies of the documents.

In addition, all documentation associated with Commission proceedings are made available on the website.  The webcast is made available for 90 days following the Commission proceeding and full transcripts of the proceedings are archived on the website.

In addition, federal legislation  (Access to Information Act and Privacy Act) allows for interested parties to obtain of interest to them (e.g., related to meetings, other regulatory dealings), subject to restrictions such as those related to privacy, security and confidentiality.

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France Article 8 § 8.1 p.65 Could Canada precise how the designation process for the President and Chief Executive Officer of the Canadian Nuclear Safety Commission CNSC is organized?

Details are provided in Section 10 of the NSCA:

Members Constitution
10 (1) The Commission consists of not more than seven permanent members to be appointed by the Governor in Council.

(3) The Governor in Council shall designate one of the permanent members to hold office as President.

Full- or part-time members
members may be appointed as full-time or part-time members.

Tenure and term of appointment
(5) Each permanent member holds office during good behaviour for a term not exceeding five years and may be removed at any time by the Governor in Council for cause.

(7) A member is eligible to be re-appointed to the Commission in the same or another capacity.

Section 12 deals with the duties of the President in her capacity as both President and Chief Executive Officer:

Duties of President
12 (1) The President is the chief executive officer of the Commission and has supervision over and direction of the work of the members and officers and employees of the Commission, including the apportionment of work among the members and, where the Commission sits in a panel, the assignment of a member or members to the panel and of a member to preside over the panel.

Absence, etc. of President
(2) If the President is absent or incapacitated or if the office of President is vacant, such other member as may be designated by the Commission has all the powers and functions of the President during the absence, incapacity or vacancy, but no person may so act for a period exceeding ninety days without the approval of the Governor in Council.

All Orders in Council are made on the recommendation of the responsible Minister of the Crown and take legal effect only when signed by the Governor General. The “Governor in Council” mentioned in section 10 of the Act is the Governor General of Canada, acting with the advice and consent of, the Queen’s Privy Council for Canada (i.e., Cabinet). Officials from the sponsoring Minister’s Office must consult with the Director of Appointments in the Prime Minister’s Office prior to transmitting a recommendation for appointment to the Governor in Council. The actual Order in Council (OIC) normally specifies the tenure of the appointment and the term of office. Holding office “during good behaviour”, mentioned in subsection 10(5) of the NSCA, means that the President (as well as other Commission members) may only be removed for cause. This is quite different from appointments “during pleasure”. This second option that may be a condition of appointment means that removal is at the discretion of the Governor in Council. For transparency, the Orders in Council are publicly available on the third working day following approval by the Governor General, but in exceptional circumstances, they may be posted on the web prior to the third working day if requested by the responsible Minister.

The following link is to a page on the Privy Council Office website that deals specifically with Governor in Council appointments.

At the top of this page you will see reference to the Government of Canada’s commitment to an open and transparent process for selecting Governor in Council appointees. It also notes  “These open, transparent, and merit-based selection processes reflect the fundamental role that Governor in Council appointees play in our democracy as they serve on commissions, boards, Crown corporations, agencies, and tribunals across the country.”

France Article 8 § 8.1(a) p.67 Could Canada precise how annual appropriations from Parliament accounting for 30 percent of the Canadian Nuclear Safety Commission CNSC’s funding are decided? The CNSC’s funding and cost recovery activities are guided by the Cost Recovery Fees Regulations (Canada Gazette Part II, Vol. 137, No. 13 / Regulatory Impact Analysis Statement / pages 1689-1709). They provide the legislative terms in which the CNSC recovers costs from licensees, as well as those activities that receive funding from the Government of Canada through annual funding appropriations.  Regarding activities that are funded through cost recovery, for-profit organizations in Canada pay a fee in support of their licence to recover costs that ‘provide identifiable recipients with direct benefits beyond those received by the general public’. On an annual basis, the CNSC recovers approximately 70% of its expenditures. 

On an annual basis, the CNSC receives approximately 30% of its funding through government appropriations. The appropriation covers activities for non-for-profit organizations (e.g. hospitals, universities), which do not pay a fee in support of their licences. In addition, annual appropriations fund non-recoverable activities that do not “provide identifiable recipients with direct benefits beyond those received by the general public”. They include:

  • i) International Obligation and Cooperation - Fulfillment of the federal government’s international nuclear policies and obligations (e.g. non-proliferation arrangements, international safeguards)
  • ii) Development and Maintenance of Legislation and Regulations - Development and amendment work to the NSC Act and its regulations to protect health, safety, security and the environment and respect Canada’s international commitments on the peaceful uses of nuclear energy
  • iii) Government Cooperation - Sharing information with other federal and provincial bodies including clarification of roles to avoid or reduce regulatory overlap
  • iv) Information Services - Provision of information to the public
France Article 8 § 8.1(f) p.79 The Funding Program established by Canada in 2017 is a very innovative measure to involve the Stakeholders in the decision-making. Could Canada evaluate the efficiency of it in terms of project acceptance and in terms of increasing trust in the Authorities? The CNSC’s participant funding program (PFP) was established in 2011 and as part of its terms and conditions it measures the effectiveness of the program by providing surveys to all funding recipients to determine if the funding effectively supported them in enhancing their participation in CNSC regulatory processes in order to bring unique, value-added information to the Commission. 90% of funding respondents to the surveys indicated that without the funding they would not have been able to participate in the CNSC’s regulatory processes as fully, or at all. 

In addition, a majority of the recipients indicated that they felt that the CNSC’s hearing and regulatory processes were meaningful and that they felt that they were heard by the Commission. Finally, the PFP has been an invaluable tool for the CNSC in building relationships and trust with Indigenous groups with an interest in CNSC regulated facilities and activities.

France Article 10 § 10 (b) to (d) p.91 to 96 and § 12 p.108 to 113 Could Canada precise which kind of enforcement action could be applied to a licensee in case the Regulatory Body observes weaknesses and shortcomings for example during inspections on human factors areas such as safety culture, implementation of human performance programs, organizational performance, etc... and please provide examples of the application of such enforcements?  Weaknesses and shortcomings of safety culture and human/organizational performance are typically addressed through discussion and/or  correspondance. (For clarity, REGDOC 2.1.2, Safety Culture concerns how licensees monitor and assess their own safety culture, thus the primary CNSC review/inspection findings would be of this process, not of the state of licensee's safety culture). If the issue amounted to a non-compliance with a requirement in the licensing basis, one of the stronger regulatory enforcement tools (as described in Section 7.2(ii) is used to bring the licensee into compliance (as would be the case for any non-compliance). 
France Article 11 § 11.2(b) p.106 Various knowledge management and risk mitigation plans exist in the plants for critical and "at risk" positions due to the departure of a significant number of workers with knowledge and experience. Could Canada indicate if the same situation is true for the regulatory body? As CNSC faces anticipated attrition of senior and technical staff, workforce planning integrates succession planning for leadership and critical roles. As part of these regular discussions, critical and at-risk positions are identified, along with interim mitigation strategies. To ensure knowledge and experiences are captured and shared, the CNSC has developed an Effective Knowledge Transfer course and actively works with managers, experts and successors to develop knowledge transfer plans prior to retirements/departures. Additionally, each CNSC staff member has an individual learning plan to identify development needs, which can also capture current or anticipated knowledge gaps that have surfaced through workforce planning and succession discussions.
France Article 13 § 13 p.114 to 116 Could Canada precise procedures to manage the diminished number of suppliers holding nuclear grade certification? Could Canada precise the procedures to manage the availability of components to replace those which are ageing, due to the non-availability of identical replacement parts from original manufacturers, obsolescence or developments in technology? Industry has implemented a few strategies for dealing with the diminished number of suppliers holding nuclear grade certification. These include reaching out to certified suppliers in other regions that had not been used in the supply chain in the past. Licensees are also using a commercial grade dedication process to qualify parts. Lastly, the licensees are working together through the CANDU Owners Group to source common parts that are no longer availble from the original manufacturers.
France Article 14 Summary § D.3 p.14 Could Canada precise if they have taken into account their experience feedback of the decommissioning of Gentilly 2 Nuclear Power Plant for defining even better the ageing management programs of the Nuclear Power Plants still in operation? Yes, CNSC has considered experience and feedback of the decommissioning of Gentilly-2 NPP in the evolution of requirements and expectations for the aging management programs of the NPPs still in operation. For instance, CNSC sponsored a research project, Analysis of Degradation Mechanisms of Cable Insulation due to Aging in a Decommissioned Nuclear Power Plant. Its objective was to assess the degradation of sample cable insulation removed from Gentilly-2. The project also included a comparison of the test data received from the real aged cables versus samples aged in a laboratory, following practices specified in modern codes and standards.

The report was forwarded to NPP licensees so that they could implement the recommendations and conclusions at their NPPs, if applicable.

Furthermore, CNSC has additional cable samples (removed from Gentilly-2) for potential use in a future cable research project (e.g., in CNL/AECL's research project, Aging of electrical cables - environmental stressors and condition assessment studies, to be completed in March 2022.

France Article 16 Summary § II p,20 How does the update of the Canadian Guidelines for Protective Actions During a Nuclear Emergency address the harmonization of preparedness and especially with  neighbouring countries? The updated guidelines, Generic Criteria and Operational Intervention Levels for Nuclear Emergency Planning and Response, provide guidance on criteria for protective actions, and are largely based on recommendations from international organizations, such as the International Commission on Radiological Protection (ICRP) and IAEA. The guidance was developed at the federal level in consultation with the provinces to ensure their acceptability. The guidance values were adopted by both Ontario and New Brunswick (the provinces with operating NPPs) in their provincial emergency arrangements. 

Canada maintains bilateral agreements involving the exchange of emergency preparedness expertise and information with the United States of America, the only nation with whom Canada shares a land border. Both Canada and the United States operate NPPs having emergency planning zones that cross the international border between the two nations. Preparedness and response arrangements between the federal and provincial levels are further coordinated through inter-jurisdictional all-hazards and nuclear emergency preparedness committees as part of Canada's national coordinating mechanism.  Bilateral agreements have been established with the United States for nuclear emergencies having transboundary consequences. Provincial and federal agencies are required to inform the Government of Canada of consultations with their United States counterparts during a nuclear emergency. Canada has participated in exercises organized by the United States, and vice versa, in order to promote cross-border cooperation.

Canada has also participated in international exercises organized by the IAEA and OECD/NEA. Canada participates in international emergency preparedness activities under the Convention on Nuclear Safety, the Convention on Early Notification of a Nuclear Emergency and the Convention on Assistance in Case of a Nuclear Accident or Radiological Emergency (Assistance Convention).

France General summary In his report, the President of the 7th review meeting had recommended that Contracting Parties consider the implementation of the good practices that where identified during the meeting. Could your country provide information on the actions carried out with regards to the implementation of those good practices in your country ? One of the four Good Practices was actually awarded to Canada for the CNSC's participant funding program. Two of the Good Practices were awarded to Euratom for items that are specifically related to the European Union (investment in the EU for all steps of nuclear cycle and safety objectives in the amended Nuclear Safety Directive) and are therefore not applicable to Canada.

The fourth Good Practice was awarded to Saudi Arabia for extensive outreach to members of the public and to neighbouring and other countries.  Canada conducts extensive outreach with members of the public as evidenced by the Good Practice awarded to Canada for participant funding to enable the public to participate in Commission hearings, as well as the use of discussion papers to seek public input prior to issuance of documents. The CNSC and licensees also organize public information sessions on nuclear issues and engage heavily in indigenous outreach and consultation. In terms of outreach to neighbouring countries, Canada maintains close ties with the United States of America, the only nation with whom Canada shares a land border. Canada works closely with the United States (e.g., through recent inspector exchanges), as well as having cross-border emergency plans in place. Canada also engages closely with the international community, contributing to peer review missions, assisting developing countries with their nuclear programs (including Saudi Arabia) and participating in many IAEA working groups.

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Article 15

Are exposed workers in your country categorized based on expected effective/ equivalent doses? If yes, what are the effective and equivalent dose limits of exposed workers of „lower” categories?

As required by the Radiation Protection Regulations, workers are either categorized as Nuclear Energy Workers (NEW) or non-NEW workers (categorized as persons who are not NEWs).  There are dose limits prescribed in the Regulations for NEWs and for persons who are not NEWs as follows:

Equivalent Dose:

  1. Lens of an eye - Nuclear energy worker - One year dosimetry period - 150 mSv, Any other person - One calendar year - 15 mSv
  2. Skin - Nuclear energy worker - One year dosimetry period - 500 mSv, Any other person - One calendar year - 50 mSv
  3. Hands and feet - Nuclear energy worker - One year dosimetry period - 500 mSv,Any other person - One calendar year - 50 mSv

Effective Dose:

  1. Nuclear energy worker (including a pregnant nuclear energy worker) - a) One year dosimetry period - 50 mSv b) Five year dosimetry period - 100 mSv
  2. Pregnant nuclear energy worker - balance of pregnancy - 4 mSv
  3. A person who is not a nuclear energy worker - One calendar year - 1 mSv


Article 15

ICRP No. 103 recommends the limit on the equivalent dose for the lens of the eye 20 mSv in a single year or 100 mSv in any five consecutive years subject to a maximum dose of 50 mSv in a single year. Does Canada intends to decrese your current limit for eye lens (150 mSv in a single year)?

Canada is currently in the process of amending the Radiation Protection Regulations. The future revision of the Regulations will include revised dose limits for the lens of the eye. The amended Regulations are anticipated to be published in 2020.


Article 15

How often do exposed workers have to undergo preventive medical examinations?

There is currently no requirement for preventative medical examinations in Canada.


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Poland General Summary - page 19 How the REGDOC-2.3.3 has been implemented to licensing basis and what is the legal status of other regulatory documents (REGDOCS). Does licensee or applicant must comply with REGDOCS before they are implemented to licensing basis or it is possible to propose different means to fulfill requirements from act and regulations? Please briefly describe process by which CNSC decide to incorporate RECDOSC into particular licensees’ licensing basis. Periodic safety review (PSR) is being introduced gradually for NPPs.  Upon renewal of each operating licence, the Commission has introduced a licence condition to conduct a PSR in preparation for the following licence period (typically ten years in duration).  If a PSR has already been coducted, the licence has also been modified to include a condition requiring the licensee to complete the Integrated Implementation Plan that was developed as part of the PSR.  The detailed requirements for PSR are catpured in the licence conditions handbook (LCH) for the licensee, where REGDOC-2.3.3 and other compliance verification criteria and guidance are cited, as needed.

In general, when new regulatory documents (or CSA Group standards) are published, the CNSC determines if they should become part of the operating basis for new and/or existing NPPs.  If that is the case, CNSC staff determine appropriate timeframes for each affected licensee to implement the new requirements, consulting the licensees as needed.  The timeframe for each NPP may be unique, depending on details of the facility and operation, other ongoing developments impacting that licensee, and the timing of the next licence renewal and/or PSR.   Once determined, CNSC staff summarizes the implementation plan for the new regulatory document (or CSA Group standard) in the LCH.

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Germany Article 6 p. 26 Could Canada please elaborate the public participation process in the development of the regulatory framework? Consultation with the public, licensees and other stakeholders is an integral component of developing the CNSC's regulatory framework. The CNSC uses a variety of means to seek input from licensees, the public, indigenous peoples, non-governmental organizations, all levels of government and international stakeholders. Meetings and workshops may be ogranized to engage stakeholders and solicit early feedback on the development of regulatory policies, requirements and guidance, and on what regulatory instruments are appropriate. Regulations, regulatory documents (REGDOCs) and discussion papers published by the CNSC are generally subject to a formal public consultation process.

For regulations, CNSC's regulatory plan describes anticipated regulatory changes or actions that CNSC intends to bring forward or undertake over the next two years. It identifies public consultation opportunities and a departmental contact point for each regulatory initiative. Draft regulations are pre-published in the Canada Gazette, Part I for comments for a period of 30 to 75 days.

For REGDOCs and discussion papers, CNSC seeks feedback and public comments through the CNSC's website and the Government of Canada's Consulting with Canadians website. The consultation period for REGDOCs and discussion papers ranges from 60 to 120 days. All comments received during public consultation are posted on the CNSC's website and further feedback is requested for another 15 days. Comments received during public consultation are reviewed and addressed by staff. If comments received during public consultation are substantial and warrant significant changes, revised REGDOCs may be provided to interested stakeholders for additional input. This second round of public consultation would generally be 60 days.

All final documents are released through the CNSC's website, for use by stakeholders and licensees.

Germany Article 8 p. 73 According to the National Report the CNSC supports the building of leadership capabilities at all levels by offering such learning activities as emotional intelligence, influence and persuasion (without authority), building high performance teams, building resilience and critical thinking among others. Could Canada please share the experiences gained from this program? CNSC recognizes that leadership capability at all levels contributes to regulatory excellence and the well-being of its workforce. CNSC has described and communicated the key leadership competencies and key behavioural competencies that are expected of staff. CNSC has incorporated the behavioural competencies into its human resource management policies, practices and processes such as executive selection, assessment, development and performance, as well as employee assessment, selection and development and recognition.  Communication in the organization continues to emphasize that technical knowledge and abilities are critical, but not sufficient to the performance of a world-leading regulator. 

CNSC is only beginning to see the impact of shifting expectations for leadership in the organization.  Recognizing informal leadership and being able to communicate specifically what behaviours management or peers are acknowledging is important.   Demonstrating that appointees to executive positions are demonstrating their ability to perform and develop their leadership capabilities is also important.  Providing employees and managers with feedback that is evidence-based and connects to a development plan is proving beneficial.

Germany Article 9 p. 85 The National Report states that “The licensees’ processes also require independent assessments to confirm the effectiveness of the management systems in achieving the expected results”. Could Canada please clarify who exactly carries out these independent assessments? Independent assessments of the effectiveness of the licensee management systems are carried out by the licensee's independent audit group. These staff are independent of the operations reporting line and report directly to the Chief Nuclear Officer or Chief Executive Officer.

Independent assessments also includes all of those done by external groups including INPO, WANO, ISO, IAEA, CNSC. The CNSC ensures through its compliance activities that the licensees take actions to improve based on all independent and self-assessments.

Germany Article 10 p. 91 According to the National Report, REGDOC-2.1.2 “Safety Culture”, published in April 2018, refers to a “healthy safety culture”. Could Canada please elaborate the differences between “safety culture” and “healthy safety culture”? Safety Culture has traditionally been described with binary terms such as strong/weak, good/bad, positive/negative, etc.  Unlike a binary pairing, the CNSC views safety culture more like health - for a living being, health might not mean everything is excellent, but is overall able to live, perhaps even thrive.  Furthermore, health is understood as a complex continuous concept, with many external and internal influencing factors.  Describing safety culture in terms of health attepts to capture these nuances that a binary classification misses.
Germany Article 12 p. 110 The National Report states that some licence holders have broadened the focus of their human performance programs in order to consider managing defences against human error and that the CNSC recognizes the benefit of this approach. Could Canada please explain in more detail how such defence methods are being implemented? In the past decade, the NPP licensee’s approach to human performance was dominated by the INPO approach which is described in U.S. Department of Energy’s, DOE-HDBK-1028-2009, Human Performance Improvement Handbook: Volume 1: Concepts and Principles and Volume 2: Human Performance Tools for Individuals, Work Teams and Management, Washington: U.S. Department of Energy, (2009).  The emphasis had been on implementing human performance tools / event-free tools and observation and coaching in the field to reinforce the use of the tools.  Defences serve to prevent undesirable and unsafe work activities and results, through measures that guide, co-ordinate or control the work and its context in order to control or reduce the harm from a nonconformance or error. Defence methods are identified through risk assessment, where risk is characterized as: Probability of the Event X Consequences of the Event.

A risk assessment is influenced by individual and collective knowledge, experience, perception and facts. For jobs that have elevated risk to either personnel or the plant, licensees implement additional controls and defences. For jobs that are lower risk to personnel and the plant, licensees emphasize the importance of following processes and procedures.

Managing defences aims to reduce the likelihood that people make errors through considering the characteristics and needs of the humans, design of the technology to fit the system to the humans, and developing the organization to provide the leadership and the needed resources and provisions for safe and effective human performance. Managing defences also includes identifying and implementing measures to reduce the negative impacts of errors or other system failures to minimize harm to people and to the plant.

Inspections and discussions with licensee staff about human performance identified that NPP licensees have been updating their approaches to human performance, and a broader, more integrated understanding is evident, through recognizing that the majority of human error can be directly attributed to process, programmatic or organizational issues. There is increased emphasis on organizational learning from a range of sources to understand the context of the work. This involves improving the work system through understanding the perspectives of the people who conduct the work, and identifying the systemic drivers of performance.  Learning teams /focus groups have been used in response to certain events and in some cases to proactively understand the issues associated with work activities. Observation and coaching in the field is generally using an “ask, not tell” approach, with the aim of learning about how the work is done. Based on the principles that “people come to work wanting to do a good job” and “the people who do the work are the ones who have the answers” one licensee requested and implemented improvement ideas in a successful campaign, which engaged and empowered the workforce to identify opportunities and drive the changes. NPP management increasingly understands and recognizes the importance of human performance to ensure successful and safe operation, and the importance of leadership and organizational factors to shape and support human performance. CNSC staff has observed that management practices are starting to reflect this understanding.

Germany Article 13 p. 116 According to the National Report, Licence holders are required to maintain effective supply chain management and procurement quality assurance programs that discover and mitigate the intrusion of counterfeit, fraudulent and suspect items (CFSIs) into their operations. To further improve the effectiveness of their programs, the licence holders implemented a variety of enhancements to increase surveillance of suppliers' and sub-suppliers' quality programs and to enhance awareness and training of supply chain staff with respect to CFSI issues. Could Canada please explain in more detail what enhancements are being applied here? The Canadian industry has implemented a number of measures to enhance awareness of CFSI issues and to prevent CFSI material from entering into NPP systems. These include:
  • enhanced training of receipt inspection staff
  • visual tools to assist in receipt inspection (including pictures of actual items for comparision with the received material)
  • enhanced training of supplier auditors in CFSI
  • contractual requirements for suppliers and sub-suppliers to report CFSI issues
  • implementation of CSA Group standard N299 "Quality Assurance Program Requirements for Supply of Items and Services for Nuclear Power Plants" by suppliers
Germany Article 16 p. 143 The National Report states that within Canada’s constitutional framework, emergency management is a shared responsibility among the three levels of government (municipal, provincial and federal), operators, and non-government organizations in a bottom-up approach. Could Canada please clarify which non-governmental organizations are involved and what is their role? In the Provinces, volunteer organizations like the Red Cross, Saint Johns Ambulance, and the Salvation Army are used to provide social services to members of the public affected by an emergency. For example, in the case of a nuclear emergency, volunteer organizations can be deployed at reception and evacuation centres to provide food, beds, blankets and first aid supplies. These reception centres and evacuation centres are at pre-determined locations outside of the detailed planning zones. Additionally, in an emergency, Ham Radio networks can also be used for communications. 
Germany Article 16 p. 146 The National Report states that Canada has specific measures in place for the post-accident recovery phase. Could Canada please explain in more detail what these specific measures are? Current arrangements for recovery are described in the provincial nuclear emergency plans of Ontario and New Brunswick, and include provisions for a range of actions including environmental monitoring, health surveillance, communications, and financial assistance.  Canada has drafted a Framework for Recovery Following a Nuclear Emergency, which was released for public consultation. The document is currently being finalized to address to the results of the  consultation and is planned to be published by fall 2020 by Health Canada with the support of the CNSC, Natural Resources Canada, and Public Safety. It will provide guidance to stakeholders, including provincial emergency management authorities. 

The document will consider all of the aspects required to recover from a nuclear emergency. This includes background based on up-to-date international guidance, transition to recovery, and best practices for key recovery elements. The key recovery elements considered are: implementation of protection strategies, environment and food chain, exposure pathways and dose assessments, health monitoring, epidemiological studies, management of contamination, remediation, waste management, protection of workers, and public communications. Future publications will focus on the roles and responsibilities of organizations within Canada in the recovery from a nuclear emergency.

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India Article 7 Page 38 It is stated “The CNSC conducts cyclical reviews of regulatory documents. Documents are reviewed to determine which ones should be withdrawn and archived, retained “as is” for continued use or scheduled for revision.”

Can Canada clarify whether any periodicity is specified for taking up the review for revision of regulatory documents?

The CNSC is committed to continuous improvement and aims to ensure CNSC's regulations and regulatory documents contine to be clear and effective, allowing CNSC to remain an effective and efficient regulator. The CNSC's Regulatory Framework Plan sets out the regulations and regulatory documents that the CNSC intends to develop or amend in the coming years. The plan aligns with the CNSC's corporate priorities, considering the current nuclear environment and is managed under the guidance of the CNSC Regulatory Framework Steering Committee and Management Committee. The plan is reviewed annually and adjusted when new developments in the nuclear industry arise.
India Article 7 Page 53 It is stated that:

“Licence renewal is a mechanism to implement new requirements from recently published CNSC REGDOCs or standards, thus contributing to the continuous safety improvement of NPPs.”

It is understood that PSR, which is supposedly another mechanism for implementation of new requirements, would be coinciding with licence renewal once in 10 years (page 55).

Can Canada clarify whether any other regulatory mechanism exists for implementation of new requirements?

There are other mechanisms for implementing new requirements.  One possibility is the voluntary adoption of new requirements by the licensees.  This is fairly common, since licensees and other industry members regularly participate in the public consultations that CNSC executes when researching and preparing REGDOCs.  The same groups are also participants in the writing and revision of CSA Group standards.  Thus, the final published versions of the new requirements sometimes reflect adjustments to make them more readily implementable without compromising safety.  Licensees may propose to implement new requirements prior to the next PSR/licence renewal.  If CNSC staff finds the proposal acceptable, it includes the committment in the licence conditions handbook (LCH) for the licensee (e.g., the LCH will note when CNSC staff will start using the new REGDOC or CSA Group standard as a source of compliance verification criteria in its inspections or compliance assessments.

If CNSC staff need to accelerate compliance with new requirements, rather than waiting for voluntary compliance or the next PSR/licence renewal, other options are available.  For example, CNSC can issue a regulatory request for information about potential implementation, as allowed by subsection 12(2) of the General Nuclear Safety and Control Regulations.  If a more direct route is needed, the CNSC can issue an order to comply with the new requirements; another option is for the Commission to amend the licence to impose the new requirements.  These options are descrribed in Section 7.2(v) of the Canadian report.

India Article 7 Page 43 It is mentioned that license is required from CNSC for all the 5 types of activities in the lifecycle of Class-IA nuclear facilities (e.g. NPPs). It is noted that, in addition to ‘license to decommission’ there is requirement for ‘license to abandon’.

Can Canada elaborate on the criteria/ regulatory requirements  considered for  ‘license to abandon’ ?

Release from CNSC regulatory control occurs when the licensee has successfully decommissioned the facility and restored the site to a state in which it can be released for future use (e.g., green field or brown field [industrial]). The CNSC requires a submission for a licence to abandon or exemption from licensing. This submission must be supported by reports on the results of the decommissioning and site restoration activities, as well as the results of the radiological and environmental monitoring, to demonstrate that the site no longer needs to be licensed under the NSCA.

If unrestricted release is yet to be achieved, perpetual licensing from the CNSC may be required. Or, oversight by another regulatory or governmental body may allow the Commission to exempt the site indefinitely from CNSC licensing (determined on a case-by-case basis) if the risks are determined to be acceptable by the Commission. In order to be granted an exemption, the licensee must present a safety case that demonstrates long-term safety.

Section 4 of the General Nuclear Safety and Control Regulations (GNSCR) outlines the information that must be submitted to obtain a licence to abandon a nuclear substance, nuclear facility, prescribed equipment or prescribed information. Additional information for an application for a licence to abandon is provided under section 8 of the Class I Nuclear Facilities Regulations (CINFR) and under section 8 of the Uranium Mines and Mills Regulations (UMMR). Section 5 of the Class II Nuclear Facilities and Prescribed Equipment Regulations (CIINFPER) outlines the information that is required in an application for a licence to decommission a Class II nuclear facility. All licensees must submit policies and procedures regarding decommissioning as part of their application.

In reviewing a submission for a licence to abandon, the CNSC must be satisfied that the abandonment of the site, nuclear substance, prescribed equipment or information does not pose an unreasonable risk to the environment, the health and safety of persons, or national security. The abandonment must also not result in a failure to comply with Canada’s international obligations.

India Article 7 Page 42 Para 7.2(i)(b)Small modular reactors (SMR)

“Small modular reactors are not legally defined in Canada because they fall under the Class I Facilities Regulations. However, they are generally understood to include reactors for the generation of power or heat for industrial purposes.”

If SMRs fall under Class I Facilities and requirements for such facilities are already in place following a graded approach, could Canada clarify whether there is any difficulty due to SMRs not being legally defined?

CNSC has not yet encountered any difficulties with not having the term "SMR" defined.  In fact, having such a definition would actually introduce unnecessary complexities to regulatory decision making by focusing on definitions instead of actual risk-informed considerations.  For example, experience has shown that the term "Module" varies from design to design and the implications of the use of modularity may depend on how a module has been defined and implmented. Canada has a long history of regulating reactor projects of various types and sizes under the existing framework.  Based on this past experience, the existing Canadian framework is sufficiently flexible to address the licensing of activities involving new reactor technologies, recognizing that new technologies will require a degree of judgement supported by sufficient technical evidence that demonstrates the safety case. CNSC expectations are further articulated in regulatory documents which are subject to continuous improvement. CNSC incorporates lessons learned from pre-licensing activities (e.g., vendor design reviews) as well as from other licensing activities to better clarify existing regulatory expectations, in particular, related to the application of a graded approach.  The existing regulatory framework also allows the CNSC to leverage experiences from other countries in current decision making.
India Article 14 Page 130 It is stated that, “CNSC licences to operate the existing NPPs contain conditions governing the licensee’s verification of safety through various fitness-for-service programs. The licensees’ programs include testing (see subsection 14(ii)(a)) and various aging management programs to address specific critical systems and aging mechanisms” It has been elaborated that aging management program like the fuel channel, feeder pipe lifecycle management plan, FAC program are available for managing the degradation of components.

Can Canada elaborate its regulatory practices/ requirements with respect to ageing or residual life assessment of the inaccessible critical components and its verification as part of license renewal?

CNSC requires licensees to incorporate approaches to estimate the condition of inaccessible critical components in their aging management programs. Consideration should be given to appropriate historical operating experience and may require additional research and support activities to generate performance data under simulated operating conditions or advanced modelling techniques considering expected operating conditions.

For critical components, CNSC requires the licensee to model aging using the best available understanding of the degradation mechanisms, to understand the consequences of process failures, and to account for uncertainty in the models. For example, for inaccessible feeder dissimilar metal welds, Canadian licensees have performed leak-before-break assessments to demonstrate that any feeder leakage from a postulated through-wall crack would be detected, and the plant shutdown, well in advance of such a growing crack reaching its critical crack length, to avoid catastrophic piping failures. For assessments of inaccessible components, uncertainty of the input parameters of the assessments must be carefully considered by the licensees. The uncertainty of these parameters are often accounted for in probabilistic assessments.

In summary, for inaccessible components, CNSC provides ongoignoversight of the types of assessment methodologies used to manage aging, carefully consider licensee assumptions, and ensure that the uncertainties of the models are properly considered.

At the time of licence renewal, CNSC staff confirms that the applicant complies with the appropriate regulayory requirements, including CNSC REGDOC-2.6.3, Aging Management and various applicable CSA Group standards. CNSC staff confirm that the applicant has programs to ensure that the aging of inaccessible critical components is managed safely. If necessary, the CNSC uses regulatory hold points (established by the Commission in the licence) to limit operation based on certain age-related parameters.

India Article 14 Page 250 It is stated “Bruce Power has developed an approach for deterministic analyses in support of seismic events, fire and floods, drawing from post-Fukushima assessments and probabilistic safety assessments”.

Could Canada explain the considerations for margin assessment for beyond design basis earthquake?

This analysis was carried out as part of Bruce Power's safety report improvement project to align the Bruce A and Bruce B safety reports with the requirements of CNSC REGDOC 2.4.1, Deterministic Safety Analysis. Beyond design basis earthquakes are not considered in the deterministic analysis but is dealt with by probablistic safety assessment (PSA). The assessments are based on guidance provided in EPRI NP-6041-SL "A Methodology for Assessment of Nuclear Power Plant Seismic Margin".  A number of assessments feed into Level 1 and Level 2 PSA's to determine severe core damage and large release probabilities; these include seismic hazard characterization, seismic response characterization and seismic fragility.
India Article 16 Page 146 It is mentioned that “In June 2018, following extensive public consultation and incorporation of lessons learned from emergency exercises, Health Canada published Generic Criteria and Operational Intervention Levels for Nuclear Emergency Planning and Response, which contained updated guidelines for public protective measures.”

Could it be clarified if Canada follows the same generic criteria and Operational Intervention Levels for Nuclear Emergency Planning and Response as stated in IAEA safety guide GSG 2 / GSR Part 7? If the criteria is different, then could Canada share the basis of the developed criteria?

Health Canada has published "Generic Criteria and Operational Intervention Levels for Nuclear Emergency Planning and Response" which align with the latest recommendations from the IAEA (including GSR Part 7) and ICRP.  The development of the guidance also took into consideration the types of scenarios associated with CANDU reactors.

In most cases, Health Canada has adopted the IAEA values for generic criteria, but there are two exceptions:

  • Health Canada has adopted a lower dose level for ingestion controls because the lower threshold is considered reasonably achievable in the Canadian context, given the ready availability of alternative sources of food and water in those areas most likely to be impacted by a significant release of radioactivity;
  • Health Canada has maintained the practice of having a separate generic criteria for sheltering, primarily in order to accommodate the current needs of some Canadian organizations. It corresponds to two days at the daily intervention level recommended in the previous (2003) edition of the guidelines.

In most cases, Health Canada has adopted the IAEA values for operational intervention levels (OILs), with the following exceptions:

  • Health Canada has not adopted IAEA’s OIL7. This OIL is calculated so that food and water can be efficiently screened using measurements of marker radionuclides (Cs-137 and I-131). It is based on the estimated ratios of radionuclides in a release from an accident at an NPP. Health Canada has not included it in the recommendations because it is not appropriate for all categories of emergency that are addressed in the Federal Nuclear Emergency Plan. Instead, Health Canada retained and expanded upon the radionuclide-specific activity concentrations presented in Canadian Guidelines for the Restriction of Radioactively Contaminated Food and Water Following a Nuclear Emergency (HC2000). Canadian specific ingestion data was used to inform the OIL calculations (HC 1993; HC 2011; HWC 1976).
  • Thyroid monitoring is an essential activity if radioiodine exposure is suspected. However, Health Canada has not adopted IAEA’s OIL8 and recommends that thyroid monitoring and dosimetry be carried out, when required, by specialists in internal assessment.
India Article 17 Page 165 It is mentioned :

”…it is important to consider population-related factors to meet radiation dose limits set by regulations. ….Other factors include: local weather, seismicity, neighbouring facilities, and air and rail transport corridor activity.

Can Canada clarify how seismicity is factored in for dose assessment?

As part of the overall integrated determination of site suitability considered in review of an application for a licence to prepare site, the CNSC evaluates if a facility can be designed and operated so that dose limits, safety goals and environmental measures can be effectively implemented. This includes the evaluation of natural external events such as seismic hazards to ensure that dose limits are not exceeded.

For the final selected site to be referenced in an application for a licence to prepare site, the applicant must prepare a site-specific seismic hazard assessment.

Additionally, site evaluation information such as information regarding seismic hazards, obtained during site preparation activities, is used and revisited in subsequent lifecycle phases of construction and operation.

India General Annex 18 (i) Page 286 Could Canada share information about severe accident sampling and monitoring system, in particular, through this monitoring which important radionuclides are monitored? Examples of monitoring provisions at OPG are:
  • Automated Near Boundary Gamma Monitoring System measures gross gamma only. It is supplemented by Off-Site Survey Teams that manually sample air and particulates and conduct smear surveys. This information is processed in OPG labs.
  • Automated Source Term Gamma Monitor System takes samples from the Filtered Air Discharge System which is then analyzed in OPG labs. The information is shared with the Emergency Response Organization. The expected main sources of gross gamma radiation are noble gases, Iodines and Tritium.
India Article 19 Page 181 It is mentioned that ‘Integration of Plant procedures (e.g. abnormal incident manuals, emergency operating procedures) with SAMGs and EMEGs is complete’.

It is understood that abnormal incident manuals are used to handle events of much less severity (e.g. transients/anticipated operational occurrences), therefore, integration of such procedures with SAMGs and EMEGs is not very clear. Does this integration mean that failure of lower hierarchy procedure leads to invoking of procedure next in hierarchy?

Could Canada elaborate on the implementation of plant procedures integration?

Yes, this integration is meant to ensure a proper transition to SAMGs and EMEGs should the lower hierarchy procedure fail to reach the desired objectives (typically the reactor shutdown with forced circulation cooling in place). All Canadian NPPs have procedures in place that detail when to transition from abnormal incident manuals/normal emergency operating procedure to SAMGs and EMEGs.

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United States of America

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United States of America General Page 8 In 2019, CNSC received both an IRRS and EPREV mission.
  1. (Considering their safety significance, what  recommendations and suggestions are considered of highest priority?

Preparing for peer review missions is a very resource intensive activity. Hosting two missions almost concurrently is commendable.

  1. What preparatory activities lessons and  best practices would you share with other countries?
  2. Would you have done something differently to improve efficiency and minimize the impact on resources?
  1. The results from the EPREV mission to Canada confirmed that Canada has a well-developed and mature nuclear emergency preparedness and response system in place across all levels of government. The EPREV mission identified several good practices, as well as opportunities for improvement to continue to strengthen Canada’s ability to prepare for, and respond to, nuclear emergencies. No recommendations and suggestions received were considered of high safety significance as the mission concluded that Canada is ready to respond to a nuclear emergency. Nevertheless, Canada is committed to continuous improvement and as such will address all the recommendations and suggestions (Canada has drafted a report entitled: Canada’s response to the 2019 report on the International Atomic Energy Agency Emergency Preparedness Review Mission to Canada: Peer Appraisal of the Arrangements in Canada Regarding Preparedness and Response for a Nuclear or Radiological Emergency). When Canada considers it has adequately addressed the findings, it will invite the IAEA for a follow-up mission, typically within four years of the EPREV mission.
  2. Canada was also commended on its arrangements for hosting the EPREV mission. Emergency preparedness and response in Canada is multi-jurisdictional and involves a broad range of stakeholders at various levels of government, and the nuclear operators. In the preparation and conduct of the EPREV mission to Canada, the following lessons were noted:
    1. Scope of mission - ensure that other IAEA missions, taking place within a short time frame, do not influence the scope of the EPREV Mission. For example, Module 10 (Emergency Preparedness and Response) was not included in the IRRS mission to Canada since the EPREV mission had occurred three months earlier. Ensure that all involved stakeholders are in agreement with the final arrangement.
    2. Participants - early during preparation phase of the project, confirm other participating departments and agencies. This will ensure that the right people are engaged to complete the self-assessment and participate in the interviews during the mission.
    3. Project governance - provide clear governance to engage all participants. To ensure good coordination, a project charter was developed to obtain the support of all partners, and an overarching steering committee with senior officials from key stakeholder organizations was created. The committee maintained oversight, provided strategic direction, and enabled timely decision-making and effective conflict resolution.
    4. Project plan – involve all participant organizations in systematic plannning. As part of the project governance, Canada also established a core planning team with representatives from each jurisdiction to serve as coordination points for all of the relevant organizations. The core planning team, which was responsible for authoring the self-assessment, engaging stakeholders, and planning for the mission, carried out planning activities as directed by the steering committee.
    5. Dialogue - discuss planning with the IAEA on ongoing basis
    6. Choice of peer reviewers – ensure reviewers are suitable for context of the mission. It was important to have representation on the EPREV review team that understood the constitutional and legal framework of Canada as well as the reactor technology used in Canada.
  3. No, the EPREV Mission was conducted within the project-planning period and within budget. Overlap with the IRRS mission was minimized by removing duplication (i.e., Module 10 of the IRRS Mission).  Additionally, video conferencing was used to include the DDG of Safeguards and Security and the Chief, IEC to close out the mission.
United States of America Article 7 Page 34 CNSC's most recent long-term regulatory framework plan covers the period from 2019 to 2024 and takes into account CNSC's priorities and project plans. Please discuss some notable items included in this updated plan. The CNSC’s most recent long-term regulatory framework plan covers the period from 2019 to 2024 and outlines the regulations and regulatory documents the CNSC will be developing or amending during that time.  Notable planned reviews of regulations made under the NSCA include:
  • Radiation Protection Regulations,
  • Nuclear Security Regulations
  • Class II Nuclear Facilities and Prescribed Equipment Regulations

Notable key regulatory documents (REGDOCs) the CNSC has published or will be developing or amending in the coming years include:

  • REGDOC-1.1.5, Supplemental Information for Small Modular Reactor Proponents (2019),
  • REGDOC-2.2.4, Fitness for Duty: Managing Alcohol and Drug Use (revision, 2020),
  • REGDOC-2.5.2, Design of Reactor Facilities: Nuclear Power Plants (revision, 2020),
  • REGDOC-2.7.1, Radiation Protection (2020),
  • REGDOC-2.7.2, Dosimetry: Ascertaining Occupational Dose (2020)
  • REGDOC-2.9.2, Environmental Protection: Environmental Principles, Assessments and Protection Measures including Controlling Releases to the Environment (2020),
  • REGDOC-2.11.1, Waste Management (2020),
  • REGDOC-2.11.2, Decommissioning (2020),
  • REGDOC-2.4.1, Deterministic Safety Analysis (revision, 2022),
  • REGDOC-2.4.2, Probabilistic Safety Assessment (revision, 2022)

Activities are aligned with the CNSC's corporate priorities and consider current developments in the nuclear environment. The plan is reviewed regularly and adjusted as necessary.

United States of America Article 10 Page 91-96 In 2018, CNSC published a regulatory document on safety culture to formalize CNSC's commitment to promoting a  healthy safety culture in the nuclear industry and clarify requirements and expectations for licensees regarding safety culture at NPPs. Licensees began to implement this regulatory document during the reporting period. CNSC also performed a regulatory safety culture self-assessment during the reporting period and produced the findings, recommendations, and management action plan based on the results.
  1. Please discuss key findings from both the licensee and CNSC safety culture self-assessments.
  2. What are some of the planned actions to be taken as a result of these assessments?
For clarity, REGDOC 2.1.2, Safety Culture concerns how licensees monitor and assess their own safety culture, thus the primary CNSC review/ inspection findings would be of this process, not of the state of licensee's safety culture.  As such, the findings of the licensees' safety culture assessments are pertinent to the licensee.

Responses to questions on licensees’ self-assessment of safety culture:

  1. Section 10(b) of the 8th Canadian report, from page 91, discusses the key findings from the licensee's most recent nuclear safety culture self-assessments. The response to Sequence 81 describes the main findings (strengths and areas for improvement) from the CNSC’s safety culture self-assessment.
  2. CNSC analysed the areas for improvement from the safety culture self-assessment, identified five recommendations to address them and developed actions to respond to each recommendation. The following briefly summarizes the recommendations.
    1. The CNSC should provide supervisors, managers and executives with ongoing coaching and mentoring in the leadership characteristics necessary to maintain a healthy safety oversight culture.
    2. The CNSC should develop a problem identification, resolution, reporting, and communication tool to further increase transparency around the communication of safety issues.
    3. The CNSC should develop an overarching safety oversight culture vision or policy statement outlining the desired culture that the CNSC is striving to achieve
    4. The CNSC should develop strategies to ensure that critical technical and regulatory knowledge, including knowledge of past experience and decision making, is actively managed as a resource and is readily available to staff.
    5. The CNSC should conduct a follow-up safety oversight culture assessment in three to five years to confirm the effectiveness of the actions resulting from this assessment and to deepen the commitment to continuously strengthen the CNSC’s safety oversight culture.
The presentation to the Commission and resulting discussion are documented in paragraphs 69 to 85 of the minutes of the Commission meeting ( CNSC staff’s response to the recommendations involved 48 distinct actions, many of which have already been completed. Others are scheduled for completion prior to the next scheduled CNSC safety culture self assessment in May 2022.

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Russian Federation

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Russian Federation Article 16.1 Section 16.1 (g) What measures for ensuring availability and response to emergencies were proposed for implementation following the results of the mission in June 2019? What are the mission results? The Emergency Preparedness Review (EPREV) mission provided valuable insight and commended Canada on its well-developed and mature nuclear emergency preparedness and response system in place across all levels of government. The EPREV review team also acknowledged Canada’s successful implementation of the IAEA Safety Standards throughout its emergency preparedness and response program, and for exceeding them in some cases. The EPREV mission identified several good practices, as well as opportunities for improvement to continue to strengthen Canada’s ability to prepare for, and respond to, nuclear emergencies. In the 2019 Report on the Peer Appraisal of the Arrangements in Canada Regarding Preparedness and Response for a Nuclear or Radiological Emergency (2019 EPREV Mission to Canada Report), the IAEA review team highlighted five (5) good practices and provided six (6) suggestions and six (6) recommendations. Both the IAEA’s report and Canada’s response will be publicly available in February, 2020.
Russian Federation General General Are the cybersecurity works being done at NPPs, if yes, what do they include? Yes, all licensees have requirements on cybersecurity embedded in their licences. The majority of the cybersecurity intiatives at Canadian NPPs are focused on protection of the licensee's local area networks to ensure that sensitive and controlled nuclear information is protected. Given that the NPP control computers are not connected to the internet, protection of NPP systems is implemented through cybersecurity programs that require strict checks and cybersecurity requirements on electronic equipment entering the NPP to ensure that no malware is inadvertantly introduced into NPP systems.

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Saudi Arabia

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Saudi Arabia General Safety Culture Knowing that Canada in one of the leading countries in nuclear industry, what is the impact of following IAEA modifed consept ( Culture for Safety) on Canadian regulatory systems, what changes canadian regulatory authority will do? Canada's approach to assessing safety culture is modeled upon the IAEA Safety Report Series 83, Performing  Safety Culture Self Assessments.  The approach "draws on experience from behavioural, social, psychological and organizational sciences to emphasize the importance of describing aspects of an organization’s culture in an impartial manner before making comparisons with international norms and expectations"(SRS 83, p2).  CNSC REGDOC-2.1.2 is aligned with the approach espoused in SRS 83.
Saudi Arabia Article 19.4 Section 19 p.181-183 Is the development of EMEG a requirement and could it be considered as part of SAMG and how could EMEG differed among SAMG and EOP? The CNSC has developed requirements for EMEG in REGDOC-2.3.2 “Accident Management” (Version 2). At the NPPs, EMEG are called up from station abnormal incidents manuals (AIMs) and emergency operating procedures (EOPs), in response to specific beyond design basis event scenarios. Unlike AIMs and EOPs, EMEG provide instructions for the use of EME which is part of equipment and instruments that are not installed as part of the original plant design, but are used as an additional provision to mitigate the consequences of an accident. EMEG are a complementary enhancement to SAMG and could be considered as part of the SAMG strategy. The fundamental objective of EMEG is different from the fundamental objective of SAMG, although they are complementary. The objective of EMEG is to provide diverse water supply to address the loss of heat sinks, and electrical power system enhancements to address total loss of AC power to prevent core degradation, restore power to critical systems and components and re-establish key monitoring function(s) to detect changes in event status, thus preventing the progression of beyond design basis accident to a severe accident. If the situation continues to deteriorate and SAMG entry conditions are met, it is assumed that core degradation has occurred and priority shifts to using SAMG strategies to maintain containment integrity and minimize radionuclide release to the public. However, although the EMEG were designed for losses of heat sink or power, they could be used to prevent or terminate a severe accident as an alternative to the SAMG.

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Denmark Article 7 Article 11,

p. 98

It dismantling is not undergoing throughout the "storage and surveillance period (approximately 30 years) please explain the rationale behind this long cooling off period, as this could likely preclude the use of personnel, familiar with the units when they were in operation, from participating in the decommissioning. In the National Report it is stated, that “The preliminary decommissioning plan provides the long-term vision for the storage and surveillance period (approximately 30 years) prior to demolition and site restoration. In the preliminary decommissioning plan, the estimated costs associated with decommissioning are presented form the basis for the decommissioning financial guarantees.” The CNSC’s performance-based regulatory philosophy, with respect to decommissioning strategies, is not prescriptive. Proponents must propose their preferred strategy as part of their decommissioning plan and must support it with a science-based safety assessment. Any proposed decommissioning strategy will be assessed by the CNSC against regulatory requirements to ensure the protection of health and safety of the public and the environment.  CNSC regulatory guide G-219, Decommissioning Planning for Licensed Activities (under revision as REGDOC 2.11., does however outline the following decommissioning strategies:
  • immediate (prompt) decommissioning
  • deferred decommissioning
  • in-situ decommissioning

When determining the appropriate decommissioning strategy, the following are to be considered and prioritized:

  • forms and characteristics of radioactive and conventional contaminants
  • integrity of containment and other structures over time
  • availability of decontamination and disassembly technologies
  • potential for recycle or reuse of equipment and materials
  • availability of knowledgeable staff
  • potential environmental impacts
  • potential worker and public doses
  • end-state objectives and site redevelopment pressures
  • potential revenues, costs and available funding
  • availability of waste management and disposal capacity
  • regulatory requirements
  • public input

For deferred decommissioning, the CNSC requires a decommissioning plan that details the programs and activities that will be maintained during the storage phase including with surveillance provisions. The decommissioning plan includes:

  • end-state objectives
  • major decontamination, disassembly and remediation step
  • approximate quantities and types of waste generated
  • overview of the principal hazards and protection strategies
  • estimate of cost
  • method(s) of guaranteeing financing for the decommissioning activities

This decommissioning plan would then be later revised to detail the dismantling work to be completed. The financial guarantee, covering the period of storage with surveillance, eventual dismantlement of the facility, and any long-term monitoring, would be reviewed and updated periodically (typically every five years) as per licence requirements. This is documented in annex C of REGDOC-3.5.1, Licensing Process for Class I Nuclear Facilities and Uranium Mines and Mills.

During the storage with surveillance phase, licensee activities, while reduced, are focused on inspection, servicing and maintenance in order to confirm that the structures, systems and components needed to maintain safe storage are functioning as required. Programs and physical systems need for storage with surveillance must be maintained. During storage with surveillance, there will be opportunities for activities to reduce risks at the facility. These include:

  • reduction or removal of combustibles
  • removal and recycling of non-contaminated or slightly-contaminated equipment (e.g., turbines, pumps, heat exchangers)
  • reduction or isolation of asbestos
  • demolition of non-nuclear facilities
  • removal of radioactive waste to an off-site licensed storage facility
  • reduction or removal of hazardous wastes

In accordance with CSA N294.0, Decommissioning of facilities containing nuclear substances, during storage with surveillance, a plan is required for surveillance, monitoring, physical protection, and maintenance of the facility during such periods shall be developed and implemented to:

  • maintain the facility in a safe state
  • control the release of materials to the environment
  • prevent access by unauthorized persons

As noted in previous questions, licensees of nuclear facilities must provide guarantees that adequate financial resources are available for the decommissioning of these facilities and management of the resulting radioactive wastes.

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Romania Article 18.1 Summary of other safety improvements during the reporting period

page 23

page 286

Please provide more information on the design and operation requirements / technical specifications for the portable filtered air supply systems procured for the secondary control areas. The secondary control area (SCA) and SCA tunnel was an area for which modifications were identified in order to improve the habitability and meet radiation dose targets in the unlikely event of a severe accident. The overall function of portable filtration is to provide a safe habitable atmosphere for the SCA operator, in the event of a partial release (e.g., emergency filtered vent release after a loss-of-coolant accident).

A portable filter unit is now available  in the event of a partial release to minimize radionuclide concentrations in the SCA during the event. It would be deployed from its storage location and connected to the existing SCA ventilation system through a dedicated (new) emergency connection at the SCA as required to mitigate accident consequences. It would be powered by the emergency power supply with redundancy from a portable diesel generator (emergency mitigating equipment).

Romania Article 7.2.1 CSA Group standards

Pages 39-41

How are the new editions and new standards used by the existing NPPs? What are the mechanisms through which these new standards can lead to practicable safety improvements? Does the CNSC require the licensees to use the latest CSA standards? See the response to Sequence 109 for information on how the CNSC imposes new requirements as they are developed.  Some requirements (e.g., prescriptive ones) automatically drive safety improvements once the licensee implements them.  In other cases, the resulting safety improvements are executed through the PSR and the resultant integrated implementation plan (IIP), which becomes part of the licensing basis when considered by the Commission at the time of licence renewal.  As part of the periodic safey review (PSR), CNSC staff reviews and accepts the licensee's proposed PSR basis document, which proposes the specific versions of REGDOCs and CSA Group standards that will be used in the PSR to identify gaps. 
Romania Article 7.2.1 7.2 (i) (b) Regulatory framework documents General description of CNSC regulatory documents Page 36 Is the minimum staff complement, addressed by REGDOC-2.2.5, required to cope also with design extension conditions?

CNSC REGDOC-2.2.5, Minimum Staff Complement, sets out the key factors that CNSC staff will take into account when assessing whether the licensee has made or the applicant will make adequate provision for ensuring the presence of a sufficient number of qualified staff as required by Section 12 (1) (a) of the General Nuclear Safety and Control Regulations.

One aspect of ensuring the presence of a sufficient number of qualified workers is defining the minimum number of workers with specific qualifications who will be available to the nuclear facility at all times, known as the minimum staff complement. The number and qualifications of workers in the minimum staff complement should be adequate to successfully respond to the most resource-intensive conditions for any credible facility state, including normal operations, anticipated operational occurrences, design-basis accidents, and emergencies.

Validation scenarios for the minimum staff complement are also validated using a variety of representative conditions that should include the most resource-intensive credible events. For multi-unit NPPs, validation scenarios should include the most resource intensive events that could affect more than one unit, such as, but not limited to, seismic events, loss of off-site power and steam line or feed water line breaks.

The robust expectations for the systematic analysis and validation of the minimum staff complement provides confidence that the facility will have a sufficient number of qualified staff available at all times to respond immediately to the above mentioned conditions. As for design extension conditions, REGDOC 2.2.5 does not specifically identify that these conditions should be taken into consideration during the analysis and validation phases. However, the minimum staff complement will be available for the initial response to any BDBA prior to arrival of any additional staff that is called-in for assistance. REGDOC 2.2.5 is scheduled for review and update as part of the CNSC regulatory document process. Additional considerations for the update will include design extension conditions, beyond-design-basis accidents (BDBAs), workload assessment and provision of additional staff for response to BDBAs, among others.

Romania Article 7.2.2 Licence conditions handbooks Pag 48 We consider that CNSC's use of Licence conditions handbooks represents a good practice that can be adopted by other countries, as it promotes consistency and objectivity of regulatory oversight. The Licence conditions handbook gathers in a single document all the regulatory details, explanations, expectations and associated processes for definitions, interpretations and administrative control of the licence conditions. The Licence conditions handbook is read in conjunction with the licence. The Licence condition handbook associates each licence condition with compliance verification criteria that are used by CNSC staff to confirm the licensee’s compliance with the licence condition. Canada acknowledges the comment
Romania Article 12 12 (d) Human actions in safety analysis Page 111 How does CNSC regulate and review time-critical operator actions?  “Time-critical operator actions” are not directly addressed in current CNSC requirements for deterministic safety analysis (DSA) and probabilistic safety assessment (PSA), but they are taken into consideration. In DSA, the operator is given a certain amount of time to act in response to a clear signal for error. For example, per CNSC REGDOC-2.4.1, Deterministic Safety Analysis, operator error has to be taken into consideration in developing accident scenarios for safety analysis:
  • Section 3.1 states: “Deterministic safety analysis methods can be applied to a wide range of plant operating modes and events, including normal operation and abnormal operation resulting from equipment failure, operator errors and challenges arising from events like fires, floods or earthquakes”
  • Section 4.2.2 states “The list of events identified for the safety analysis shall include all credible 1) component and system failures or malfunctions, 2) operator errors”
  • Section states “However, the generic implications of human errors as initiating events should be considered to identify any further potential system failures. As such, if a specific operator error could result in a unique initiating event, it should be included in the list of PIEs for the deterministic safety analysis.”

Also, REGDOC-2.4.1 stipulates that the licensee performs DSA to assist in developing a strategy for the operator to follow. For example, Section 3.2 item 4 states that deterministic safety analyses are performed to:

  • “assist in confirming or validating the strategies that have been selected to recover the plant from an AOO or DBA”
  • “assist in developing a strategy for the operator to follow, should the automatic actions and emergency operating procedures fail to prevent a severe accident”
Romania Article 14.1 Development of probabilistic safety assessment and implementation of REGDOC-2.4.2

Use of probabilistic safety assessment

Pages 124-125

It is mentioned in the report that recent PSA updates have included estimates of the multi-unit PSA results. What is the impact of multi-unit PSA estimates for the risk monitors (e.g. EOOS)? In Canada, PSAs for the multi-unit NPPs are developed based on a representative unit, taking into consideration the multi-unit impact (e.g., impact of site-level initiating events, impact of initiating events occurred from adjacent units, and other multi-unit interactions (shared systems/components and human interactions between units). The PSA results (e.g., core damage frequency (CDF) or large release frequency (LRF)) are expressed in per unit basis and the risk monitor models (e.g., for EOOS) are also the per-unit models.

The recent PSA updates provided estimates of site level risk (CDF and LRF per site), but they were not used in the risk monitors. The risk monitors, such as EOOS, focus on the risk of an individual unit, taking into consideration impacts from common/shared equipment.

Romania Article 19.2 19 (ii) (b) Safe operating envelope

Page 178

How have the OP&Ps and Impairment Manuals for Canadian NPPs been supplemented based on the implementation of CSA N290.15-10 and thorougher definition of the safe operation envelope? The OP&Ps, impairment manuals, and SOE serve different purposes. The SOE defines safe operating limits comprehensively for NPPs, while the OP&P defines the high-level operating policies and principles, and only includes a subset of the relevant operating limits. With the implementation of the SOE program at each station, a subset of the relevant operating limits is no longer needed in the OP&P, although the licensee may choose to keep them. To date, no licensee has put more operating limits into the OP&P based on the SOE. Regarding the impairment manuals, they have to be consistent with SOE limits.
Romania Article 19.2 19 (ii) (b) Safe operating envelope

Page 178

Have the Canadian NPPs defined operational limits and conditions for the equipment and components installed to support accident management for design extension conditions, including severe accidents? No operational limits and conditions are defined for the equipment and components installed to support accident management for design extension conditions (DEC). The reason for this is the large degree of uncertainty associated with the DEC analysis, and also to encourage utilities to include wide range of equipment for this purpose with little imposed conditions.
Romania Article 14.1 14 (i) (d) Reviews by the World Association of Nuclear Operators and IAEA

Page 127

Please provide more details on the good practices identified by the OSART missions related to severe accident management simulator application for supporting multi-unit guideline development, respectively to obsolescence management that considers long term aging management assessments Obsolescence management takes into consideration long-term aging management assessments and the requirements associated with transitioning to decommissioning. The plant’s obsolescence management program supports long-term operation plans and the transition to decommissioning. The obsolescence team consists of individuals dedicated to identification and resolution of obsolescence issues. These individuals implement obsolescence solutions by providing both procurement and design engineering support. This arrangement allows the plant to have a more focused and consistent approach and reduce the number of hand-offs between various engineering work groups. This arrangement allows the team to look at wider application strategically to select the most cost-effective solution. The co-operation between these groups with different mandates enables the preparation of condition assessments that consider full component life span needs. This approach leads to improved safety when approaching the transition to final decommissioning.
Romania Article 19.6 19 (vi) Reporting incidents significant to safety

Pages 184-185

Does the CNSC perform its own independent analyses and investigations for events reported by the licensees? The CNSC has a low reporting threshold of events at NPPs. Nevertheless, CNSC staff review details of all events reported, cross-informing other specialists and licensing staff as needed and following up through regulatory activities with the aid of site staff to assure underlying issues are adequately addressed.  Follow-up may include reviews of the licensee's assessment, independent analysis, review of station records, observation of existing conditions, oversight of licensee follow-up, reactive inspection, etc, as well as conduct of enforcement action if needed.  CNSC also has the capacity and authority to investigate events on its own.

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Bulgaria Article 8.1 page 80. Targeted consultation with Indigenous communities On page 80 is written that CNSC is in the process of implementing a long-term engagement strategy with Indigenous communities with the purpose to formalize relationships and continue engagement activities outside of the CNSC’s regulatory processes. Please could you give more details about this strategy and clarify what is meant by relationships and activities outside the CNSC’s regulatory processes. As a life-cycle regulator and an Agent of the Crown the CNSC is committed to building long-term relationships with Indigenous groups with an interest in CNSC regulated facilities and activities. The approach to formalizing the relationship  with interested Indigenous communities can include the signing of a Terms of Reference with an associated work plan that lays out the agreed upon frequency of meetings for regular information sharing and collaborative activities (such as environmental monitoring).

The CNSC is committed to conducting these engagement activities throughout the life-cycle of CNSC regulated facilities and activities of interest, within and outside of a formal regulatory process (i.e. a licence application or environmental assessment). This means that the CNSC will continue to engage and collaborate with Indigenous communities during a facility's licence term and not just during a formal regulatory process before the Commission for decision (which may trigger formal consultation requirements).

Bulgaria Article 10 page 96. Priority to safety at CSNC On page 96 it is written that CSNC has conducted SC self-assessment in 2018. It will be useful to learn more details regarding the methodology that has been followed for the self-assessment, i.e. the methods used to collect SC data, as well as the framework for assessing regulatory SC. The assessment methodology was developed to align with the International Atomic Energy Agency’s (IAEA) Safety Reports Series No. 83: Performing Safety Culture Self-assessments, to the extent practicable.

The assessment used a multi-method approach to gather and collect data from document reviews, focus group discussions and the administration of a questionnaire on safety oversight culture perceptions. The CNSC also utilized the Nuclear Energy Agency's (NEA’s) The Safety Culture of an Effective Nuclear Regulatory Body.

Bulgaria Article 12 page 109. Human performance programs On page 109 is written that the requirement for a licensee to have a human performance program is a license condition in NPP operating licenses. Please give more details about the regulatory requirements to the content of the programs, as well as the main components of such a program. The licence condition states “The licensee shall implement and maintain a human performance program”. The related Regulatory document REGDOC-2.2.1, Human Performance is being updated to formalize the content and components of human performance programs. It should be noted that a human performance program is not necessarily a stand-alone program and is generally embedded within the licensee’s management system, and a “road-map” approach can be used to describe the human performance program. NPP licensees describe the program in their licence applications to renew their licences.

Guidance in the current licence condition handbooks includes the following:

The human performance program should address and integrate the range of human factors that influence human perofrmance, which include, but may not be limited to the following:

  • the provision of qualified staff (certification and training, staffing, minimum shift complement, hours of work, fatigue management)
  • the reduction of human error (human factors in design, procedures, turnover, work protection, pre- and post-job briefings, safe work strategies/practices)
  • organizational support for safe work activities (human actions in safety analysis, organizational performance and safety culture)
  • continuous improvement of human performance

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Hungary General p.19 What is the overall state of the regulatory framework development  process with respect to the (licencing of) SMRs? The CNSC regulatory framework as it stands is technology-neutral and fit for licensing projects that reference SMRs and advanced reactors.  One of the key strengths of the framework is that requirements are derived from internationally-accepted and proven fundamental safety objectives.  An applicant has the ability to propose and demonstrate how it will meet these objectives using proven engineering practices and supported by relevant and credible information.  Any proposal will also need to demonstrate how uncertainties are being addressed in the proposed safety and control measures.    As the industry and the CNSC continues to develop operational experience from R&D as well as construction and operation, lessons learned will be analyzed to determine how to further clarify CNSC expectations as well as industry standards.

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