Fitness for service
The fitness for service safety and control area covers activities that evaluate the systems, structures and components at nuclear power plants to confirm that they are maintained as intended by design and perform design functions. The fitness for service evaluation involves various activities including engineering assessments for structural integrity, periodic in-service inspections using non-destructive examinations, maintenance activities and research and developments depending on component conditions and operating experiences. The assessment for fitness for service is to be continuously confirmed for its validity through periodic in-service inspections, surveillances and research activities throughout the period of service life. To manage all these activities in a systematic way, the CNSC requires that licensees develop and implement aging management programs or maintenance programs. CNSC staff closely monitor the effectiveness of those programs through compliance verification activities.
In this category, you will find research and technical information related to the following specific areas: equipment fitness for service/equipment performance, maintenance, structural integrity, aging management, chemistry control, and periodic inspection and testing.
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-739.1, Regulatory Oversight Program for Authorized Inspection Agencies (AIA) at Nuclear Facilities
- RSP-690.1, Best Practices for Probabilistic Fracture Mechanics Evaluations
- RSP-734.1, Development of an Innovative Probabilistic Compliance Reliability Index for Nuclear Power Plant Oversight
- RSP-706.1, Probabilistic Assessments: Principles and Computational Methods
- RSP-682.1, An Experimental Study of the Effects of Flat Bar Supports on Streamwise Fluidelastic Instability in Nuclear Steam Generators
- RSP-665.1 – Corrosion of Steel H-Piles at Nuclear Generating Stations
- RSP-658.1, Technical Presentation by Oak Ridge National Laboratory – Design Overview and Operating Experience for the Molten Salt and Sodium-Cooled Fast Reactors
- R656.2 - Licensing of safety critical software for nuclear reactors - Common position of international nuclear regulators and authorised technical support organisations
- RSP-645.1, Statistical Modelling of Aging Effects in Failure Rates of Piping Components
- RSP-637.1 – Review of the applicability of EPRI recommendations for a flow-accelerated corrosion program to CANDU nuclear piping systems
- RSP-590.1, Regulatory Assessment of Leakage Through Cracks in Piping Components
- RSP-586.2, IAEA technical document on fuel safety criteria for pressurized heavy-water fuel
- RSP-567.1, Review and Standardization of Testing Procedures for Irradiated Zr-2.5Nb Pressure Tube Material
- RSP-523.1, Investigation of Consequences of Concrete Alkali–Aggregate Reaction on Existing Nuclear Structures
- RSP-444.2 – Development of Analytical Tools for Soil-Structure Analysis
- RSP-0306 – Third party review of PRAISE-CANDU probabilistic fracture mechanics code
- RSP-0305 – Loading of steam generator tubes during main steam line breaks
- RSP-0295 – Investigation of the fatigue cracking and leakage rate potential of U-bend tube bundles subjected to flow-induced vibrations
- RSP-0294 – Assessment of Leak Rates through Steam Generator Tubes
- RSP-0286 – Irradiation effects on material properties for 304L stainless steel base metal and welds
- RSP-0284 – Ageing management of cable in nuclear generating stations
- RSP-0275 – OECD piping failure data exchange project (OPDE)
- RSP-0268 – Loading of steam generator tubes during main steam line break
- RSP-0253 – Review of fitness-for-service guidelines for steam generator
- RSP-0252 – Probabilistic assessment of leak rates through steam generator tubes
- RSP-0251 – Effect of inspection uncertainties on the operational assessment of reliability of steam generator tubing
- RSP-0250 – Future directions for using the leak-before-break concept in regulatory assessments
- RSP-0249 – OECD/NEA Pipe Failure Data Exchange (OPDE)
- RSP-0245 – Development of an integrated approach for implementing the Earned Quality Method
- RSP-0244 – Loading of steam generator tubes during main steam line break – Phase 1
- RSP-0236 – Piping failure frequency analysis using OECD/NEA data
- RSP-0232 – Application of the Earned Quality Method to the probabilistic assessment of leak rate through steam generator tubes research project
- RSP-0228 – OECD Piping Failure Data Exchange (OPDE) Project: Results and insight into the first phase
- RSP-0226 – Type 2 inspection checklists development project
- RSP-0224 – Evaluation of a new approach for the assessment and disposition of pressure tube crevice corrosion flaws
- RSP-0217 – Investigation into the role of pipe breaks in the licensing of CANDU reactors with positive void reactivity feedback and the credible application of early detection (leak-before-break)
- RSP-0212 – Technical support for implementation of IAEA periodic safety review in support of life extension of NPPs
- RSP-0198 – Development of regulatory guidelines on the effectiveness of NPP ageing management programs
- RSP-0197 – Assessment of LBB applicability to CANDU primary heat transport piping
- RSP-0196 – A report on performance demonstration of NDR techniques
- RSP-0190 – Independent review of current status of leak-before-break assessment at Darlington NGS
- RSP-0169 – Review of Bruce A steam generator and preheater conditions assessment and life cycle managament plan for research project; Condition assessment and life cycle management of aging steam generator
- RSP-0147 – Review of CANDU steam generator fitness-for-service guidelines and Darlington life cycle management plan
Technical papers and abstracts
- Investigation of the influence of some key parameters in the groundwater flow and solute transport modelling for in-situ decommissioning projects
- CODAP Database Project: 20 Years of International Collaboration on Passive Component Operating Experience and Materials Degradation
- Probabilistic Modelling Approaches for the Flaw Growth Rate Estimation in a Multi-Component System
- Advancements in Probabilistic Approaches for CANDU Components and Systems in Canada
- Simplified FEA of Missile Impact on Reinforced Concrete Structures with Attached Equipment
- A Regulatory Perspective on Canadian Practices on the Use of Indirect Methods in Seismic Qualification and Their Evolution
- From NURETH-2013 to NURETH-2019: Non-Critical Summary and Brief Contribution to the History of Nuclear Reactor Thermal Hydraulics
- Shrinkage Mitigation Of An Ultra-High Performance Concrete Submitted To Various Mixing And Curing Conditions
- Applicability of Sub-modelling Technique for Dynamic Analysis of Concrete Structures with Attached Equipment Under Missile Impact
- Proposal for New Design Provisions for Impactive and Impulsive Loading in ACI349 and ACI359/ASME Section III Division 2
- Proposal to Introduce Design Extension Loading Cases and Corresponding Acceptance Criteria in the ASME Code
- Recent Insights from the International Common-Cause Failure Data Exchange (ICDE) Project
- The Use of Risk Insights to Support Inspections for Nuclear Power Plants in Canada
- The Thermal Hydraulics of High-Pressure Molten Fuel Ejection
- Effects of Shrinkage-Reducing Admixtures, Mixing and Curing Conditions on the Shrinkage of an Ultra-High-Performance Fibre-Reinforced Concrete
- Impact of Mixing And Curing Temperatures on Fresh and Hardened States Properties of UHPC
- Probabilistic Assessments: Principles and Computational Methods
- OECD/NEA/CSNI Project ASCET on Numerical Simulations of Squat Shear Walls with Alkali-Silica Reaction
- FE Analysis of Reinforced Concrete Structures Under Missile Impact Using Sub-modelling Technique
- Battery Aging: Battery Degradations in Safety Electrical Power Systems for Nuclear Power Plants
- Assessing Cyber Security in Small Modular Reactors
- Regulatory Perspective on Chemistry Control at NPPs
- Specimen Curvature and Size Effects on Crack Growth Resistance from Compact Tension Specimens of CANDU Pressure Tubes
- Technical and Scientific Support Organization Forum – Supporting the Development of Technical and Scientific Capacities in Member States
- Applications of Probabilistic Fracture Mechanics for Pressure Tubes
- Finite Element Analysis of Walls with Alkali–Silica Reaction Subject to Quasi Static Cyclic Loading
- Regulatory Perspective on Fitness-for-Service Requirements for the Pressure Tube to Calandria Tube Contact in CANDU Reactors
- Regulatory Considerations for the Adoption of Probabilistic Assessment Methodologies for Pressure Boundary Component Aging Evaluations
- Overview of Classification Requirements in Canadian Codes and Standards for Nuclear Power Plants
- Finite Element Analysis of Walls With Alkali–Silica Reaction (ASR) Subjected to Constant Axial and Cyclic Lateral Loadings
- Proposed Review Framework for Design of Pressure Retaining Components in Small Modular Reactor (SMR)
- Effects of Non-Normal Input Distributions and Sampling Region on Monte Carlo Results
- Regulatory Perspective on the Fitness-for-Service Requirements for the Pressure Tube to Calandria Tube Contact in CANDU Reactors
- Introduction to steam generators – from Heron of Alexandria to nuclear power plants: Brief history and literature survey
- Assessment of Critical Subcooled Flow Through Cracks in Large and Small Pipes Using TRACE and RELAP5
- Applications of Probabilistic Fracture Mechanics for Pressure Tubes With Flaws
- Regulatory Perspectives on the Condition Assessment and Monitoring of Safety Significant Major Reactor Components for Long-Term Operation
- Regulatory Perspective on Fitness for Service Assessments of CANDU Pressure Boundary Components
- Problems With Pigs – Radioisotope Handling Outside the Hot Cell
- Considerations for the Use of Probabilistic Assessments in Regulatory Decision Making Related to Pressure Boundary Component Aging
- CNSC research related to impactive and impulsive loading
- OECD/NEA/CSNI Benchmark ASCET Phase II: Summary, Conclusions and Recommendations
- CANDU Fuel Safety Criteria: A CNSC Perspective
- Managing Structural Integrity of Key Components for Long-Term Operation of Nuclear Power Plants – Regulatory staff's perspectives
- Regulatory Perspective for the Definition of Probabilistic Acceptance Criteria for the CANDU Pressure Tubes
- Canadian Regulatory Perspectives on Probabilistic Fracture Mechanics
- CANDU Reactors: Long-Term Operation and Refurbishment, CNSC Perspective
- Reliability Requirements and Use of Risk Applications for the Reliability Program in Canadian Nuclear Power Plants
- Status of Canadian NPPs’ LTO preparation
- Lack of correlation (incoherence) modeling and effect from realistic 3D inclined, body and surface seismic motions
- Synthetic earthquake ground motions at closely spaced distances with SYNACC
- Regulatory oversight – Approach to life extension of nuclear research reactors
- Evolution of Canadian reliability requirements in a risk-informed environment
- Material surveillance program regulatory guidance for steam generator tubes extracted from Canadian CANDU reactors
- Canadian regulatory approach to steam generator life cycle management
- Assessment of the impact of ageing on the performance of CANDU special safety and safety related systems; Safety analysis perspective
- Regulatory assessment of integrated safety reviews for nuclear plants refurbishment
- Effect of environmental and material conditions on stress corrosion cracking (SCC) of carbon steels in high temperature water
- Stress re-distribution induced by creep relaxation around a notch loaded in tension – measurement and modeling
- Update on Canadian regulatory oversight of ageing management for nuclear power plants
- Regulatory perspective on CANDU feeder pipe degradation due to flow accelerated corrosion (FAC) and intergranular stress corrosion cracking (IGSCC)
- 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
- Application of the leak before break concept to CANDU feeder piping with service induced cracking
- Activities of OECD/NEA in the field of integrity and ageing of components and structures
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
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