REGDOC-2.4.4, Safety Analysis for Class IB Nuclear Facilities - Public Consultation

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REGDOC-2.4.4, Safety Analysis for Class IB Nuclear Facilities sets out requirements and guidance for applicants and licensees to demonstrate the safety of a Class IB nuclear facility, including:

  • a safety analysis program (the managed process that governs conduct of a safety analysis)
  • conduct of a safety analysis (a systematic evaluation of the potential hazards)
  • safety analysis documents, records and reporting

This document is the first version of REGDOC‑2.4.4, Safety Analysis for Class IB Nuclear Facilities.

For additional information on safety analysis for the post-closure phase of a disposal facility, see REGDOC‑2.11.1, Waste Management, Volume III: Safety Case for Disposal of Radioactive Waste.

Consultation on REGDOC-2.4.4, Safety Analysis for Class IB Nuclear Facilities is now closed. Thank you to everyone who submitted comments.

REGDOC-2.4.4, Safety Analysis for Class IB Nuclear Facilities sets out requirements and guidance for applicants and licensees to demonstrate the safety of a Class IB nuclear facility, including:

  • a safety analysis program (the managed process that governs conduct of a safety analysis)
  • conduct of a safety analysis (a systematic evaluation of the potential hazards)
  • safety analysis documents, records and reporting

This document is the first version of REGDOC‑2.4.4, Safety Analysis for Class IB Nuclear Facilities.

For additional information on safety analysis for the post-closure phase of a disposal facility, see REGDOC‑2.11.1, Waste Management, Volume III: Safety Case for Disposal of Radioactive Waste.

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    This appendix provides a sample structure for an SAR. The applicant or licensee is under no obligation to follow this format; however, as described in sections 2 through 5 of this regulatory document, the report shall include all information as applicable. 

    Table of contents

    Chapter 1: Introduction

    Chapter 2: General facility description

    • applicable regulations, codes and standards
    • basic technical characteristics
    • facility layout
    • operating modes
    • additional referenced analyses
    • a summary of significant modifications or changes to the site or facility since the previous safety analysis report, including modifications to any facility buildings, processes, equipment, procedures, programs or organizational structure

    Chapter 3: Management of safety

    • organizational structure
    • operational management philosophy
    • safety culture
    • quality assurance
    • monitoring and review of safety performance

    Chapter 4: Site evaluation

    • site reference data (area under the control of the licensee and the surrounding area)
    • hydrology
    • hydrogeological characteristics
    • meteorology
    • seismology
    • present and projected surrounding population distribution
    • present and projected surrounding land use
    • evaluation of site specific hazards
    • proximity of industrial, transport and military facilities
    • activities at the facility site that may influence the facility’s safety
    • radiological conditions due to external sources
    • site related issues in emergency planning and accident management
    • monitoring of site related parameters

    Chapter 5: General design aspects

    • safety objectives, design principles and criteria
    • conformance with the design principles and criteria
    • classification of structures, systems and components
    • civil engineering aspects of facility design
    • equipment qualification and environmental factors
    • human performance program
    • protection against internal and external hazards

    Chapter 6: Description of facility systems and components

    • nuclear systems and components
    • non-nuclear systems and components
    • instrumentation and control
    • electrical systems
    • auxiliary systems
    • fire protection systems
    • radioactive waste treatment system
    • other safety relevant systems

    Chapter 7: Safety analyses

    • safety objectives and acceptance criteria
    • identification and classification of PIEs
    • human actions
    • deterministic approach
    • probabilistic approach
    • summary of results of the safety analyses

    Chapter 8: Commissioning (for new facilities)

    Chapter 9: Operational aspects

    • organization
    • administrative procedures
    • operating procedures
    • emergency operating procedures
    • guidelines for accident management
    • maintenance, surveillance, inspection and testing
    • management of aging
    • control of modifications
    • qualification and training of personnel
    • human factors
    • feedback of operational experience
    • documents and records

    Chapter 10: Operational limits and conditions

    Chapter 11: Radiation protection

    • application of the ALARA principle
    • sources of radiation
    • design features for radiation protection
    • radiation monitoring
    • radiation protection program

    Chapter 12: Emergency preparedness

    • emergency management
    • emergency response facilities
    • fire protection program

    Chapter 13: Environmental aspects

    • radiological effects
    • non-radiological effects

    Chapter 14: Radioactive waste management

    • control of waste
    • handling of radioactive waste
    • minimizing the accumulation of waste
    • conditioning of waste
    • storage of waste
    • disposal of waste

    Chapter 15: Decommissioning and end of life aspects

    • decommissioning plan
    • financial guarantee

    Chapter 16: Public information program

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    This appendix provides some examples of limiting conditions for safe operation of the facility, applying requirements for:

    • nuclear substances (type, chemical and physical form, maximum capacity in the facility, isotopic composition)
    • hazardous substances (such as chemicals) inside the facility and its equipment
    • minimum availability requirements for:
      • SSCs important to safety
      • requirements on testing values of the SSCs

                      Note: in some cases, OLCs relating to the availability of SSCs may include requirements for their testing, including:
                              - initial and periodic tests
                              - type of tests
                              - verification
                              - calibration or inspection
                              - required intervals for inspections
                              - time between two successive tests

    • means of confinement:
      • air flows (and where appropriate, temperatures and humidity) within the facility and its processes
      • target pressure drops across filters
      • pressures within the facility buildings (rooms, cells or boxes as appropriate) relative to the atmosphere (under normal and emergency conditions)
      • isolation of means of confinement and starting of emergency ventilation
      • operations that require confinement
      • configuration and minimum equipment for ventilation system
      • leak rate from the means of confinement
      • efficiency of filters
    • radiation protection and management of radioactive waste:
      • alarm setting for criticality alarm systems and for radiation detection and monitoring instrumentation and equipment
      • limits on the airborne concentration of nuclear substances
      • radiation exposure control levels for operation, including radiation dose action levels
      • limits for controlling surface contamination
      • storage capacity for liquid and solid nuclear waste
    • material handling, including requirements for:
      • movements of nuclear and hazardous substances, including onsite and offsite transportation
      • the material handling tools and equipment including cranes (maximum allowable loads and testing requirements)
      • storage containers
    • electrical systems, including requirements for:
      • emergency power supply
      • testing frequency
      • availability and reliability of uninterruptable power supply and diesel generators
    • other systems; some examples are:
      • fire protection systems
      • process auxiliaries
      • communications systems
      • emergency lighting systems
    • monitoring system and associated alarm settings:
      • values of the settings for instrumentation in the facility
      • values of the settings for process equipment necessary for safety
    • administrative requirements:
      • staffing (for example, minimum staffing and hours of work)
      • prerequisites for activities important to safety (such as transport of radioactive or fissile material (both onsite and offsite))
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    This appendix describes the types of PIEs and the ranges of conditions to be considered for applicability at Class IB nuclear facilities.

    C.1 Selected postulated initiating events

    Some examples of PIEs are:

    1. incorrect specification of incoming and transferred material
    2. loss of services
      • loss of electrical power
      • loss of compressed air
      • loss of inert atmosphere
      • loss of coolant
      • loss of ultimate heat sink

    3. loss of criticality safety controls
      • drop of fuel during handling
      • loss of geometry
      • flooding
      • loss of neutron poison
      • excess reflection or moderation
      • unintentional change of phase
      • failure or collapse of structural components
      • maintenance error
      • control system error
      • over (double) batching

    4. processing errors
      • incorrect facility configuration
      • insufficient reagent or coolant, added too fast or too early
      • incorrect pressure or gas flow, rupture
      • incorrect or extreme temperature
      • unexpected phase changes leading to criticality or loss of confinement
      • function required for safety not applied
      • safety function applied too late

    5. facility and equipment failures
      • failure of confinement or leak
      • inadequate isolation of process fluids
      • blockage or bypass of filter or column
      • spurious actuation of item important to safety
      • structural failures

    6. handling errors
      • hazardous load dropped
      • heavy load dropped on item(s) important to safety
      • safety interlocks failure on demand
      • brakes, overspeed or overload protection inadequate
      • obstructed pathway leading to collision
      • failure of lifting component (for example, hook, beam, or cable)
      • load fixed to floor

    7. special internal events
      • internal fires or explosions
      • internal flooding (for example, from sprinkler systems or other water pipes)
      • malfunction in experiment
      • improper access by persons to restricted areas
      • criticality event
      • fluid jets, pipe whip, internal missiles
      • exothermic chemical reaction
      • ignition of accumulated combustible gases (for example, hydrogen)
      • failure due to corrosion
      • loss of neutron absorption
      • accidents on transport routes (including collisions into the facility building)

    8. external events
      • external fires or explosions
      • earthquakes (including seismically induced faulting and landslides)
      • flooding (including failure of an upstream/downstream dam; blockage of a river; or damage due to storm surges or high waves)
      • tornados and tornado missiles
      • extreme meteorological phenomena (including precipitation, sandstorms, hurricanes, storms and lightning
      • aircraft crashes
      • toxic spills
      • effects from adjacent facilities (for example, nuclear facilities, chemical facilities and waste management facilities)
      • biological hazards such as microbial corrosion, structural damage or damage to equipment by rodents or insects
      • power or voltage surges on the external supply line

    9. human errors
      • operator error or omission
      • maintenance error or omission

    C.2 Range of selected events to be considered for applicability

    The following classification and ranges of internal events are to be considered for applicability:

    • anticipated operational occurrence (AOO): an event with a likelihood of occurrence that is greater than 10‑2 per year
    • design basis accident (DBA): an event with a likelihood of occurrence that is less than 10‑2 per year and greater than 10‑5 per year
    • design extension conditions (DEC): an event with a likelihood of occurrence that is less than 10‑5 per year and greater than 10‑6 per year

    The following ranges of selected external events are to be considered for applicability.

    Wind and tornado loading

    For assessment of design basis accidents (DBA):
    The potential for the occurrence of tornadoes in the region of interest shall be assessed on the basis of detailed historical and instrumentally recorded data for the region. For example, wind design for an existing facility if prescribed by an applicable building code would have an annual exceedance probability of greater than or equal to 2 x 10‑2. For more information, see Standard Review Plan for Fuel Cycle Facilities Licence Applications (NUREG‑1520) [17].

    For assessment of design extension conditions (DEC):
    Depending on the geographical location of the facility, the effects of a tornado with an annual exceedance probability of 10‑5 or greater may need to be considered if a potential exists at the facility for offsite consequences of DEC that may lead to offsite emergency mitigation measures.


    Flooding hazards

    For assessment of DBA:
     Existing facilities are generally to be sited above the 100‑year flood plain.

    For assessment of DEC:
     Maximum probable flood plain should be used if a potential exists at the facility for offsite consequences of DEC that may lead to offside emergency mitigation measures.


    Seismic hazards

    Near regional studies should include a geographical area typically not less than 25 km in radius. Site vicinity studies should cover a geographical area typically not less than 5 km in radius. Site area studies should include the entire area covered by the facility. For more information, see IAEA SSG‑9, Seismic Hazards in Site Evaluation for Nuclear Installations [18].

    Information on prehistorical, historical and instrumentally recorded earthquakes in the region should be collected and documented. For more information, see IAEA NS‑R‑3 (Rev. 1), Site Evaluation for Nuclear Installations [19].

    For assessment of DBA:
    Structures at existing nuclear fuel cycle facilities are built to a building code. Guidance in CSA N289.5, Seismic instrumentation requirements for nuclear power plants and nuclear facilities [20] should be used if a potential exists at the facility for offsite consequences of DBA that may lead to offsite emergency mitigation measures.

    For assessment of DEC:
    CSA N289.5, Seismic instrumentation requirements for nuclear power plants and nuclear facilities [20] provides guidance that includes meeting a design-basis earthquake having an exceedance probability of 10‑3 per year to less than 10‑4 per year. CSA N289.5 [20] should be used if a potential exists at the facility for offsite consequences of DEC that may lead to offsite emergency mitigation measures.


    Aircraft crashes

    The potential for aircraft crashes, including impacts, fires and explosions on site, should be considered with account taken of:

    • the foreseeable characteristics of air traffic, the locations and types of airports
    • the characteristics of aircraft, including those with special permission to fly over or near the facility such as firefighting aircraft and helicopters

    For more information, see IAEA NS‑R‑3 (Rev. 1), Site Evaluation for Nuclear Installations [19].

    For assessment of both DBA and DEC:
     
    The potential hazards arising from aircraft crashes are taken into account if:

    • airways or airport approaches pass within 4 km of the site
    • airports are located within 10 km of the site for all but the biggest airports
    • for large airports, if the distance (d) in kilometers to the proposed site is less than 16 km and the number of projected yearly flight operations is greater than 500d2

    Where the distance (d) is greater than 16 km, the hazard is considered if the number of projected yearly flight operations is greater than 1000d2.

    For military installations or air space usage such as practice bombing or firing ranges, which might pose a hazard to the site, the hazard is considered if there are such installations within 30 km of the proposed site.

    For more information, see IAEA NS‑G‑3.1, External Human Induced Events in Site Evaluation for Nuclear Power Plants [21].

  • Glossary

    over 2 years ago

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    For definitions of terms used in this document, see REGDOC‑3.6, Glossary of CNSC Terminology.

    REGDOC‑3.6 includes terms and definitions used in the Nuclear Safety and Control Act (NSCA), the regulations made under the NSCA, and CNSC regulatory documents and other publications. REGDOC‑3.6 is provided for reference and information.

    The following terms are being defined in this draft for public consultation. Following public consultation, the final versions of the terms and definitions will be included in the next version of REGDOC‑3.6.

    credible abnormal event (événement anormal crédible)

    As defined in the CSA Group publication CSA N292.1, Wet storage or irradiated fuel and other radioactive materials [14], a naturally occurring or human-generated event or event sequence that has a frequency of occurrence equal to or greater than 10‑6 per year.

    control location (lieu de commande)

    A location that is permanently staffed during periods when the event in question may occur; for example, a control room.

    safety analysis program (programme d'analyse de la sûreté)

    Activities to plan, execute, verify and document safety analyses; to identify and act upon research and experience; to train analysts; and to preserve knowledge. The safety analysis program includes interfaces with other programs to ensure that safety analysis is initiated when needed and that the results of the safety analysis are used appropriately.

    To be added to “Appendix A: Acronyms and abbreviations” in REGDOC‑3.6:

    ELAP (PPACA)                 extended loss of AC power

    FINAS (FINAS)                fuel incident notification and analysis system

    SAR (RAS)                    safety analysis report

  • References

    over 2 years ago

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    The CNSC may include references to information on best practices and standards such as those published by CSA Group. With permission of the publisher, CSA Group, all nuclear-related CSA standards may be viewed at no cost through the CNSC Web page “How to gain free access to all nuclear-related CSA standards”.

    1. Canadian Nuclear Safety Commission (CNSC), REGDOC‑2.11.1, Waste Management, Volume III: Safety Case for Disposal of Radioactive Waste, Ottawa, Canada, 2019
    2. CNSC, REGDOC‑3.5.3, Regulatory Fundamentals, Ottawa, Canada, 2018
    3. International Atomic Energy Agency (IAEA), SSR‑4, Safety of Nuclear Fuel Cycle Facilities, Vienna, Austria, 2017 
    4. CNSC, REGDOC‑3.6, Glossary of CNSC Terminology, Ottawa, Canada
    5. CSA Group, CSA N286‑12, Management system requirements for nuclear facilities, reaffirmed in 2017
    6. CSA Group, CSA N292.1, Wet storage of irradiated fuel and other radioactive materials, 2016
    7. CNSC, REGDOC‑2.5.2, Design of Reactor Facilities: Nuclear Power Plants, Ottawa, Canada, 2014
    8. IAEA, Safety Guide SSG-5, Safety of Conversion Facilities and Uranium Enrichment Facilities, Vienna, Austria, 2010 
    9. IAEA, Safety Guide SSG-6, Safety of Uranium Fuel Fabrication Facilities, Vienna, Austria, 2010 
    10. IAEA, TECDOC No. 1267, Procedures for Conducting Probabilistic Safety Assessment for Non‑Reactor Nuclear Facilities, Vienna, Austria, 2002
    11. CNSC, REGDOC‑2.2.5, Minimum Staff Complement, Ottawa, Canada, 2019 
    12. CNSC, REGDOC‑2.4.3, Nuclear Criticality Safety, Ottawa, Canada, 2018 
    13. Health Canada, H46‑2/03‑326E, Canadian Guidelines for Intervention During a Nuclear Emergency, Ottawa, Canada, 2003
    14. IAEA, General Safety Requirements No. GSR Part 7, Preparedness and Response for a Nuclear or Radiological Emergency, Vienna, Austria, 2015
    15. CSA Group, CSA standard N292.0, General principles for the management of radioactive waste and irradiated fuel, 2019
    16. CSA Group, CSA standard N292.2, Interim dry storage of irradiated fuel, 2013 (reaffirmed in 2018)
    17. United States Nuclear Regulatory Commission (NUREG), Standard Review Plan for Fuel Cycle Facilities License Applications (NUREG‑1520), Revision 2, 2015
    18. IAEA, Specific Safety Guide SSG‑9, Seismic Hazards in Site Evaluation for Nuclear Installations, Vienna, Austria, 2010
    19. IAEA, Safety Standard No. NS‑R‑3 (Rev. 1), Site Evaluation for Nuclear Installations, Vienna, Austria, 2016
    20. CSA Group, CSA standard N289.5, Seismic instrumentation requirements for nuclear power plants and nuclear facilities, reaffirmed in 2017
    21. IAEA, Safety Guide NS‑G‑3.1, External Human Induced Events in Site Evaluation for Nuclear Power Plants, Vienna, Austria, 2002
  • Additional Information

    over 2 years ago

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    The CNSC may recommend additional information on best practices and standards such as those published by CSA Group. With permission of the publisher, CSA Group, all nuclear-related CSA standards may be viewed at no cost through the CNSC webpage “How to gain free access to all nuclear-related CSA standards”.

    The following documents provide additional information that may be relevant and useful for understanding the requirements and guidance provided in this regulatory document: