Safety analysis
Safety analysis is a systematic evaluation of the potential hazards associated with the conduct of a proposed activity or facility and considers the effectiveness of preventative measures and strategies in reducing the effects of such hazards. There are two basic types of safety analysis: probabilistic safety analysis and deterministic safety analysis.
Probabilistic safety analysis focuses on evaluating the risk arising from various events to confirm that safety goals are met. This analysis is performed using best-estimate data and assumptions, and considers all existing plant systems to provide a realistic prediction. It provides insights into plant design and operation, including the identification of dominant risk contributors and safety improvement opportunities, and the comparison of options for reducing risk.
Deterministic safety analysis focuses on evaluating the consequences of various events to confirm that the dose acceptance criteria and safety goals are met with a high degree of confidence for all accidents within the design basis. This includes demonstrating the efficiency of defence in depth and the integrity of protective barriers.
In this category you will find research and technical information about deterministic and probabilistic safety analysis, hazard analysis, criticality safety, severe accident analysis, and management of safety issues.
Research and support program
- RSP-760.1 A Study for the Canadian Nuclear Safety Commission on Artificial Intelligence Applications and Implications for the Nuclear Industry
- RSP-723.2 Technical Basis for Flood Protection and Flood Hazard Assessment for Canadian Nuclear Facilities
- RSP-593.2 Implementation of 4D/4P tool for all NPPs
- RSP-762.1, Review of the Canadian Nuclear Safety Commission’s Regulatory Framework for Readiness to Regulate Fusion Technologies
- RSP-723.1, Assessment of the potential impacts of climate change on probable maximum precipitations applicable to nuclear facilities in Canada
- RSP-682.1, An Experimental Study of the Effects of Flat Bar Supports on Streamwise Fluidelastic Instability in Nuclear Steam Generators
- RSP-674.1, Investigation of two-phase flow phenomena in reactor headers
- RSP-671.1, Studies of Molten Metal Solidification in Internal Pipe Flows
- RSP-669.1, Hydrogen/CO Combustion and Passive Autocatalytic Recombiner Behaviour
- RSP-665.1, Corrosion of Steel H-Piles at Nuclear Generating Stations
- RSP-660.1, Radioactive Material Transport Probabilistic Risk Assessment: Large Truck Accidents on Canadian Roadways
- R656.2 - Licensing of safety critical software for nuclear reactors - Common position of international nuclear regulators and authorised technical support organisations
- RSP-646.1, Assessment of RELAP5 for Natural Circulation
- RSP-614.1, Flood hazard assessment for nuclear power plants in Canada
- RSP-613.6, Safety Assessment Code Development and Application
- RSP-612.1, Application of Bayes method in evaluation of ROP/NOP trip setpoint
- RSP-602.3, Develop irradiated fuel bay Severe Accident Analysis computer code (PKPIRT) - Phase 3 (BAYLOCA)
- RSP-602.2, Phenomena Identification and Ranking Table (PKPIRT) for a Severe Accident in a CANDU Irradiated Fuel Bay
- RSP-598.2, Integrated Framework for Propagation of Uncertainties in Nuclear Cross-Sections in CANDU Steady-State and Transient Reactor Physics Simulations
- RSP-598.1, Feasibility study of an integrated framework for characterization of uncertainties with application to CANDU steady state and transient reactor physics simulation
- RSP-586.2, IAEA technical document on fuel safety criteria for pressurized heavy-water fuel
- RSP-557.1, Assessing regulatory requirements and guidelines for the single failure criterion
- RSP-444.3, Effect of non-linear soil behaviour on the seismic response of soil-structure systems
- RSP-0304, Incorporating aging effects into probabilistic safety assessments
- RSP-0298, International CFD Benchmark Problem
- RSP-0296, Statistical analysis of common-cause failure data to support safety and reliability analysis of nuclear plant systems
- RSP-0293, OPG/BP 2010 EVS methodology for calculation of NOP trip setpoint: Independent verification and benchmarking of statistical method and mathematical framework
- RSP-0292, Nuclear research trends post Fukushima
- RSP-0288, Technical report UOME-RBB-2012-04 (Grid spacer simulations)
- RSP-0274, Uncertainties in calculation of kinetics parameters of CANDU cores
- RSP-0270, Comments on existing AECL documents used in the seismic evaluation of the NRU facility and recommended acceptance criteria for a current evaluation of the seismic adequacy of the NRU facility
- RSP-0260, Bruce fire modeling review
- RSP-0259, Industrial fire brigade staffing
- RSP-0258, Fire safe shutdown analysis review
- RSP-0257, Devenir environnemental du tritium dans le sol et la végétation
- RSP-0248, International Common Cause Data Exchange (ICDE) Project: Preparation of data to be submitted to the clearing house
- RSP-0246, An evaluation of severe accident computer codes for CANDU nuclear power plants
- RSP-0242, Verification of the PROMETHEUS core analysis code and its advanced pption
- RSP-0240, Technical basis for G-144 trip parameter acceptance criteria for the safety analysis of CANDU nuclear power plants
- RSP-0238, Applicability of multidimensional methods to prediction of flow and void distributions in CANDU headers
- RSP-0233, Pickering B Unit 6 probabilistic assessment
- RSP-0231, Numerical model of the thermal and mechanical behaviour of a CANDU 37-element bundle
- RSP-0230 Modeling of molten-fuel-moderator interactions
- RSP-0229, Review of the analysis basis for Pickering B LBLOCA BEAU analysis
- RSP-0225, International Common Cause Data Exchange (ICDE) Project: Preparation of data to be submitted to the clearing house
- RSP-0220, Scaling of RD-14M large LOCA experiments GAI 00G01: A review of COG-04-2023: A scaling assessment of RD-14M for coolant voiding during the power pulse phase of a postulated large-break LOCA scenario: application to a 20% reactor inlet header break in the Point Lepreau reactor
- RSP-0216, Review comments on NRU severe accident studies-related reports
- RSP-0211, International activities in development and application of best estimate analysis methods
- RSP-0208, Fuel failure mechanisms under accident conditions
- RSP-0200, Review of the 5-kg corium commissioning test completed in the MFMI Facility
- RSP-0195, Applicability review of fuel and fuel-channel models in thermalhydraulics computer codes for CANFLEX 43-element fuel – Phase 1
- RSP-0192, Guidelines for safety assessment of application using best-estimate and uncertainty analysis methods for CANDU nuclear power plants (Revision 1)
- RSP-0186, Safety analysis review guide for non-power reactors
- RSP-0181, Survey of CANDU fuel bundle experiments under high temperature conditions
- RSP-0179A, Technologies for mitigating tritium releases to the environment
- RSP-0179B, Good work practice for effective tritium management
- RSP-0177, Data collection for the ICDE (International Common-Cause Data Exchange) Project
- RSP-0172, International regulatory practices in fuel design qualification
- RSP-0168, Review of the coverage limit in the Canadian Nuclear Liability Act – Phase II
- RSP-0167, Proposed plan for scaling analysis
- RSP-0163, Independence expert panel review of reactor physics uncertainties
- RSP-0160, Independent expert peer review: Canadian industry best estimate analysis and uncertainty methodology
- RSP-0156, Guidelines for the assessment of electromagnetic interference in the CANDU plant – Phase 2
- RSP-0152, International Common Cause Data Exchange (ICDE) Project: Preparation of data to be submitted to the clearing house
- RSP-0150, Assessing the uncertainty in the pressure tube ballooning predictions using safety code
- RSP-0148, Safety report review guide for research and radioisotope production reactor facilities – Phase 2
- RSP-0143, PHT system acoustics: Validation of code assumptions
- RSP-0138, Review of the coverage limit in the Canadian Nuclear Liability Act – Phase 1: Development of the methodology
- RSP-0136, Study of the potential implications of adopting single failure/severe accident licensing requirements
Technical papers and abstracts
- Mathematical Modelling of Fault Activation from Water Injection at an Underground Research Facility
- SERPENT-based Few-Group Cross Section Data for NESTLE-C
- Challenging Test Problems for Coupled Transient Simulations of Pressurized Heavy Water Reactors
- Web-based PSA application: Risk Handbook to Support Regulatory Site Inspections of Nuclear Power Plants in Canada
- Recent Insights from the International Common-Cause Failure Data Exchange (ICDE) Project
- Defence in Depth Considerations for Inherent Safety Features and Passive Safety Systems in Advanced Reactor Technology Designs
- Severe Accident Prevention and Mitigation in Pressurized Heavy Water Reactors
- The Use of Risk Insights to Support Inspections for Nuclear Power Plants in Canada
- State of the Art of Molten Salt Reactors (MSR) and Areas of Regulatory Focus
- Canadian Regulator’s Perspective on Severe Accident Management of CANDU Reactors: In-Vessel Retention
- Technical and Scientific Support Organization Forum – Supporting the Development of Technical and Scientific Capacities in Member States
- Status Report on Hydrogen and Combustible Gases SOAR to NEA–WGAMA
- Use of the Commercial Grade Dedication (CGD) Process for Installed Commercial Grade Equipment for Safety Related Application
- Regulating Energetic Beams for Medical Isotope Production in Canada
- Safety Impact of Exceeding Linac Vault Design Workload Limits
- Overview of CNSC EOC Technical Tools Supporting Accident Prognosis
- A Comparison of Fault Trees and the Dynamic Flowgraph Methodology for the Analysis of FPGA-Based Safety Systems Part 2: Theoretical Investigations
- Use of the Phenomena Identification and Ranking Table (PIRT) in the assessment of Source Term (ST) phenomena
- Safety Objectives Proposed for CANDU Fuel in Design Extension Conditions
- Convective Heat Transfer in CANDU Spent Fuel Racks After a Loss of Coolant
- Incorporation of Post-Fukushima Upgrades Into Severe Accident Mitigating Strategies
- Assessment of Risks to Human Health and the Environment From Hazardous Substances in Nuclear Fuel Cycle Facilities in Canada
- Modelling the Short- and Long-Term Hydro-Mechanical Behaviour of Argillaceous Limestone
- Analysis of Fault-Tolerant Design Methods and Architectures for Digital I&C Systems Using the Dynamic Flowgraph Methodology
- The Application of CNSC Fukushima Action Plan in the Design of Small Modular Reactors
- Severe Accident Strategies, Review and Validation of Guidelines at Canadian NPPs
- CNSC Technical EOC Tools for Accident Assessment and Prognosis
- Hydrogen Assessment Tool for the Emergency Operating Centre
- Addressing Challenges in the Application of the Design Safety Requirements for Nuclear Power Plants to Small and Medium Sized Reactors in Pre-Licensing Vendor Design Reviews in Canada
- Regulatory Research on Thermo-Hydro-Mechanical-Chemical Processes
- Effects of Glaciation on the Rock Formations Around a Proposed Nuclear Waste Repository
- Challenges in Application of BEPU for Risk Evaluation
- BEPU and Evaluation of Predictive Capability of Physics Simulations of CANDU Transients
- Overview of the Historical and Regulatory Basis for Exclusion-Zone Sizing in Canada
- Mathematical Modelling of a Fault Slip Induced by Water Injection
- Regulatory Oversight of New Accelerator Technologies
- CNSC Severe Accident Research
- Canadian Approach to Defence in Depth, Design Extension Conditions and Severe Accident Management
- CANDU Fuel Safety Criteria: A CNSC Perspective
- Overview of Canadian Regulatory Practice and Requirements for Nuclear Core Design
- Modelling a heater experiment for radioactive waste disposal
- Comparative modelling approaches of hydro-mechanical processes in sealing experiments at the Tournemire URL
- Geoscientific arguments in support of the safety case for a deep geological repository in Southern Ontario, Canada
- Development of a Production MCNP CANDU 3D Full-Core Model With Practical Remedies to the Issues in Deriving Reliable Tally Results
- What Could Be Learnt From the Development and Qualification of Canadian Reactor-Physics Codes
- Regulatory Review of CANDU Fuel Thermal Hydraulic Analysis and Associated Challenges
- Application of Bayes Method in Evaluation of ROP/NOP Trip Setpoint
- Integrated Framework for Propagation of Uncertainties in Nuclear Cross-Sections in CANDU Steady-State and Transient Reactor Physics Simulations
- The Pseudo-Resonant Isotope Model for Predicting the Resonance-Interference Effect in Self-Shielding Calculation
- On the Functional Failure Concept and Probabilistic Safety Margins: Challenges in Application for Evaluation of Effectiveness of Shutdown Systems
- Assessment of the Effect of the Advanced Drift-Flux Model of ASSERT-PV on CHF, Flow and Void Fraction Distributions
- Modern Application of the Single Failure Criterion in Nuclear Power Plant Design and Its Future Evolution
- Status of Code Application and Maintenance Program (CAMP) Activities in Canada
- Validation of Thermalhydraulic Computer Codes Used in Safety Analysis: Regulatory Perspective
- WGRISK Site-Level PSA Project: Status Update and Preliminary Insights for the Risk Aggregation Focus Area
- Quantification of Plant-level HCLPF Capacity in Probabilistic Safety Assessment-Based Seismic Margin Assessment
- Status of Code Application and Maintenance Program (CAMP) Activities in Canada
- 3-D System-Scale Thermal-Hydraulic Code Applications in Canada
- Application of a Graded Approach to Safety Standards of Research Reactors and Subcritical Assemblies
- Use of a Graded Approach in the Regulation of Research Reactors at the Canadian Nuclear Safety Commission
- Executive Summary: Regulatory Role of Probabilistic Safety Assessment
- CANDU Reactors: Long-Term Operation and Refurbishment, CNSC Perspective
- CNSC Review of the Long-Term Safety Case for a Deep Geological Repository
- Regulatory Challenges for Human Factors in Small Modular Reactors
- Reliability Requirements and Use of Risk Applications for the Reliability Program in Canadian Nuclear Power Plants
- Modification of the SPAR-H method to support HRA for Level 2 PSA
- Modelling & Simulation (M&S) and Uncertainly Qualification (UQ) for Safety Analysis of Nuclear Power Plants
- Highlights of Canadian Nuclear Criticality Safety Standards, Regulation, and Guidance
- Adequacy of bleed condenser/degasser condenser relief valves capacity in the CANDU operating plants to mitigate consequences of a total loss of heat sink accident
- Regulatory considerations in long-lead items for nuclear reactor facilities
- Severe accident progression without operator action
- Current fire protection regulatory approach for nuclear facilities in Canada
- Regulatory evaluation of the research and development activities in support of nuclear safety
- Regulatory review of the CANDU Fuel Modification Program in Canada
- CANDU safety research and development in Canada: Current progress and challenges
- WGAMA/WGFS status report on spent fuel pool (SFP) under loss of cooling accident conditions
- Regulatory framework and insights from fire PSA of Canadian nuclear power plants
- Focus areas for the regulatory review of the probabilistic safety assessment of seismic events
- CNSC regulatory requirements on reliability of nuclear power plants
- Effect of transverse reinforcement for missile impact on reinforced concrete slabs
- Canadian regulatory approach to steam generator life cycle management
- Comparison of methods used for internal flood probabilistic safety analyses
- Evaluation of severe accident mitigation actions through simulation
- A risk-informed perspective on deterministic safety analysis of nuclear power plants
- Blending deterministic and probabilistic arguments in the regulatory decision making: The Canadian approach
- Amplification of seismic input due to 1D, 2D and 3D effects, and their importance for nuclear power plant structures
- IRIS – 2012 benchmark, Parts I and II
- Lack of correlation (incoherence) modeling and effect from realistic 3D inclined, body and surface seismic motions
- IRIS 2010 – Part III: Numerical simulations of Meppen II-4 test and VTT-IRSN-CNSC punching tests
- CANDU heat sinks improvements as a follow up to Fukushima Daiichi Accident "The Regulator Perspective"
- PSA approach for evaluation of external hazards as part of CNSC Fukushima action items
- CNSC expectations for resolution of hydrogen related safety issues
- Fuel performance in aging CANDU reactors – A quick overview of the CNSC regulatory oversight activities of the past 15 years and of the lessons learned
- A stochastic model for piping failure frequency analysis using OPDE data
- Crack growth model for the probabilistic assessment of inspection strategies for steam generator tubes
- Assessment of steam generator tube flaw size and leak rate models
- Canadian regulatory requirements for safety analysis of nuclear power plants
- Accident precursor program – A CNSC perspective
- Testing of statistical procedures for use in optimization of reactor performance under aged conditions
- Assessment of the impact of ageing on the performance of CANDU special safety and safety related systems; Safety analysis perspective
- CNSC approach to the review of reactor core thermalhydraulic design for new nuclear power plants
- Application of probabilistic safety goals to regulation of nuclear power plants in Canada
- Challenges in PSA regulatory expectations for nuclear power plants in Canada
- Tritium releases from Canadian nuclear facilities – synthesis and associated controls
- Pragmatic application of the precautionary principle to deal with unknown safety challenges
- Testing and qualification of confidence in statistical procedures
- A stylized approach for evaluation of incremental change in CANDU ROP/NOP trip functional failure probability under aging conditions
- A stochastic-deterministic approach for evaluation of uncertainty in the predicted maximum fuel bundle enthalpy in a CANDU postulated LBLOCA event
- Evolution of Canadian reliability requirements in a risk-informed environment
- Analysing heat transfer between used nuclear fuel bundles in spent fuel pools using COMSOL multiphysics
- CNSC nuclear power plant accident handbook
- Regulatory assessment of integrated safety reviews for nuclear plants refurbishment
- Development of a simplified generic PRA model for regulatory application
- Regulatory evaluation of the research and development activities in support of nuclear safety
- Fundamental safety principles for future reactors and the role of a technology demonstration program
- Safety margins in deterministic safety analysis
- Safety analysis: its role and current trends
- Numerical prediction of heat transfer and pressure tube/calendria tube deformation during a contact boiling test
- Balanced design of a CANDU 6 NPP: Insights from full and simplified PRA models
- Regulatory oversight – Approach to life extension of nuclear research reactors
- CNSC Fuel Performance Oversight Programme
- A regulatory perspective on the establishment of fuel safety criteria for the large loss of coolant accident
- Risk informed decision making – Specific aspects for risk-informing decisions in a regulatory environment
- Current fire protection regulatory approach for nuclear facilities in Canada
- Regulatory approach for the assessment of fire protection programs at nuclear power plants in Canada
- Synthetic earthquake ground motions at closely spaced distances with SYNACC
- Why an effective national regulatory infrastructure is essential for a country's radiation protection system
Publications
Health studies
Videos
- Post-Fukushima improvements to nuclear power plants
- Nuclear power plant safety systems
- Understanding nuclear power plants: total station blackout
- Nuclear power plant safety systems – Part 1: Introduction
- Nuclear power plant safety systems – Part 2: Controlling the reactor
- Nuclear power plant safety systems – Part 3: Cooling the fuel
- Nuclear power plant safety systems – Part 4: Containing radiation
- Nuclear power plant safety systems – Part 5: Canada's nuclear regulator
- Void coefficient of reactivity and CANDU reactors
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