Preface
Foreword
Contents
Editors, Contributors and Reviewers
Introduction
Part 1. General Background
Chapter 1. Biological and Health Effects of Ionizing Radiation – The Radiological Protection System
1.1. Biological and health effects of ionizing radiation
1.1.1. Biological processes
1.1.2. Review of units of measure
1.1.3. Natural radioactivity
1.1.4. Health effects
1.1.4.1. Deterministic effects, tissue reactions
1.1.4.2. Stochastic or random effects
1.1.4.3. Induction of diseases other than cancer
1.1.5. Example of the limitations of epidemiology
1.2. Radiological protection system
1.2.1. Types of exposure situations
1.2.2. Exposure categories
1.2.3. Justification principle
1.2.4. Optimization (ALARA) principle
1.2.5. Principle of application of dose limits
Chapter 2. Organization of Nuclear Safety Control and Regulation for Nuclear Facilities and Activities in France
2.1. From the founding of CEA to the TSN Act
2.2. A few definitions
2.3. The different contributors to nuclear safety and their missions
2.4. A few basic principles and notions in the field of nuclear safety
2.5. Statutory and quasi-statutory frameworks applicable to basic nuclear installations
Chapter 3. The International Dimensionand the Social Dimension
3.1. International dimension
3.1.1. Introduction
3.1.2. IAEA standards
3.1.3. International Reporting System for Operating Experience (IRS)
3.1.4. Services developed by the IAEA
3.1.4.1. OSART reviews
3.1.4.2. IRRS reviews
3.1.4.3. Other services and study frameworks set up by the IAEA
3.1.5. WANO
3.1.6. NEA
3.1.7. Organizations dedicated to radiation protection and health
3.1.8. From bilateral Franco-German cooperation to European structures for the exchange and capitalization of knowledge and practices, training and assessment services
3.1.9. Nuclear regulator associations
3.2. The social dimension
3.2.1. Introduction – the context in France
3.2.2. Examples of initiatives and issues raised concerning reactor safety in the French nuclear power plant fleet
Chapter 4. Nuclear Reactors: Complex Sociotechnical Systems – the Importance of Human and Organizational Factors
4.1. The introduction of human and organizational factors in the field of nuclear power reactors and lessons learned from the Three Mile Island nuclear power plant accident
4.2. The accident at the Chernobyl nuclear power plantand the concept of ‘safety culture’
4.3. The Fukushima Daiichi nuclear power plant accident: the social dimension and the conceptof organization ‘resilience’
4.4. Changes in the perception of the role of people in achieving a high level of reliability in complex sociotechnical systems
4.5. Main topics studied in the development of resources and skills pertaining to human and organizational factors
4.5.1. Resources and skills
4.5.2. Main topics studied
4.6. Human and organizational factors in French regulations
Part 2. Safety by Design
Chapter 5. The Development of Nuclear Power Using Uranium-235 Fission – A Few Notions of Physics Used in Pressurized Water Reactors
5.1. Important milestones in the development of nuclear power using fission of the uranium-235 isotope
5.2. Fission and important concepts in reactor kinetics
5.3. Removing power from the core during operation
5.4. Decay heat
5.5. Main features of pressurized water reactor cores
5.6. Control and monitoring of pressurized water reactor cores
5.7. Using uranium and plutonium mixed oxide (MOX) fuel
Chapter 6. General Objectives, Principles and Basic Concepts of the Safety Approach
6.1. General approach to risks – General objectives
6.2. Fundamental safety functions
6.3. Confinement barriers
6.4. Defence in depth
6.4.1. Levels of defence in depth
6.4.2. Elements common to the different levels of defence in depth
6.5. Events considered: terminology adopted for nuclear power reactors
6.6. WENRA reference levels
6.7. Deterministic safety analysis and probabilistic safety assessments
6.8. Lessons learned from the accident at the Fukushima Daiichi nuclear power plant on the concept of defence in depth and deterministic analysis
6.9. Safety culture – Quality control
Chapter 7. Safety Options and Considerations at the Design Phase
7.1. Different types of design provisions associated with safety considerations
7.2. Single-failure criterion
7.3. The specific nature of computer-based systems (based on instrumentation and control software)
7.4. Equipment safety classification
7.4.1. Importance of equipment for safety and safety classification
7.4.2. Generic requirements associated with the different safety classes
7.4.3. Qualification of equipment for accident conditions
7.5. Information on designing nuclear pressure equipment
7.6. General considerations on provisions for hazards in facility design
7.7. Anticipating decommissioning in the design stage
Chapter 8. Study of Operating Conditions in the Deterministic Safety Analysis
8.1. Categories of operating conditions
8.2. Choice of operating conditions
8.2.1. Concept of ‘bounding’ incident or accident
8.2.2. Accident exclusion
8.3. List and breakdown of operating conditions
8.4. Methods for studying operating conditions
8.4.1. Choice of initial conditions, conservatism
8.4.2. Consideration of an aggravating event in the study on operating conditions – ‘Passive’ failures
8.4.3. Conventional combinations
8.4.4. Preventing accident aggravation
8.4.5. Operator response time
8.4.6. Using qualified simulation software
8.4.7. Main criteria to be met for fuel in the reactor core
8.5. Concept of ‘design-basis situations’ for equipment
8.6. Situations to be taken into account in application of pressure equipment regulations
8.7. Assessing the radiological consequences of incidents, accidents and hazards
8.7.1. Assessing radioactive substances released from the facility
8.7.2. Assessing radiological consequences of radioactive release from the facility
8.7.3. Assessing radiological consequences
Chapter 9. Loss-of-Coolant Accident
9.1. Short- and medium-term aspects of a LOCA
9.1.1. Mechanical effects on vessel internals and fuel assembly structures
9.1.2. Thermal-hydraulic aspects and behaviour of fuel rods
9.1.2.1. Large-break LOCA
9.1.2.2. Intermediate-break LOCA
9.1.3. Effects on reactor containment and internals
9.1.4. Long-term aspect
9.2. Safety demonstration
9.2.1. General information and background
9.2.2. Fuel assemblies and fuel rods, vessel internals, reactor coolant system components
9.2.2.1. Mechanical strength of vessel internals, fuel assembly structures and reactor coolant system components
9.2.2.2. Fuel behaviour
9.2.3. Reactor containment and equipment located inside
Chapter 10. A Special Issue: Steam Generator Tubes
10.1. Steam generator tube rupture as a Category 3 event
10.2. Preventing an SGTR accident, risk of multiple ruptures
10.3. Steam generator tube rupture(s) studied as a Category 4 event
10.3.1. 900 MWe and 1300 MWe reactors
10.3.2. 1450 MWe reactors and EPR (Flamanville 3)
10.4. Provisions to mitigate the radiological consequences of SGTR accidents
Chapter 11. Providing for Hazards: General Considerations and Internal Hazards
11.1. General considerations on providing for hazards
11.2. Potential projectiles inside the containment
11.3. Effects of pipe breaks
11.4. Projectiles generated by a turbine rotor failure
11.5. Protection against load drops
11.5.1. Risks related to spent fuel transport packaging
11.5.2. Other handling risks
11.6. Fire protection
11.7. Explosion protection
11.8. Internal flooding
Chapter 12. Providing for External Hazards
12.1. General considerations on providingfor external hazards
12.2. ‘Climate watch’ implemented by EDF
12.3. Earthquakes
12.4. External floods
12.5. Extreme temperatures
12.5.1. Extreme cold
12.5.2. Extreme heat
12.6. Possible heat sink hazards
12.7. Other naturally-occurring external hazards
12.8. Accidental aeroplane crashes (excluding malicious acts)
12.9. Risks related to the industrial environment (excluding malicious acts)
Chapter 13. Complementary Domain of Events
13.1. The origin of studies belonging to the complementary domain
13.2. Background of the complementary domain
13.3. Analysis of complementary domain events
13.4. ‘New complementary domain’
13.5. Case of the Flamanville 3 EPR
Chapter 14. Development and Use of Probabilistic Safety Assessments
14.1. History and regulatory context
14.1.1. International situation
14.1.2. Situation in France
14.2. Level 1 PSA
14.2.1. Scope
14.2.2. Method for carrying out a Level 1 PSA
14.2.2.1. General information
14.2.2.2. Specific point: probabilistic human reliability analysis
14.2.3. Level 1 PSA results and lessons learned
14.3. Level 2 PSA
14.3.1. Scope
14.3.2. Method for carrying out a Level 2 PSA
14.3.2.1. General information
14.3.2.2. Probabilistic human reliability analysis for Level 2 PSAs
14.3.3. Examples of lessons learned from Level 2 PSAs
14.3.3.1. Steam explosion risk assessment
14.3.3.3. Isolating penetrations in the containment
14.3.3.4. Modifying the pressure relief system of the reactor coolant system
14.3.3.5. Improvement of operating procedures to reduce risk of core melt under pressure
14.3.3.6. Contribution of Level 2 PSAs to emergency response measures
14.4. Expanding the scope of PSA coverage
14.5. Using probabilistic safety assessments
14.5.1. Using PSAs in the design phase
14.5.1.1. Usefulness and particularities of PSAs in the design phase
14.5.1.2. PSAs conducted to support the Flamanville 3 EPR design
14.5.2. Using PSAs in periodic reviews
14.5.2.1. Level 1 PSA
14.5.2.2. Level 2 PSA
14.5.3. Using PSAs for reactor operation
14.5.3.1. Using PSAs to analyse event severity
14.5.3.2. Using PSAs to analyse operational limits and conditions and temporary changes
14.5.3.3. Using PSAs to analyse operating procedures
Chapter 15. Aspects Specific to PWR Spent Fuel Storage Pools
15.1. Spent fuel pool design
15.1.1. Confinement barriers
15.1.2. Initiating events defined at the design stage
15.2. Experience feedback
15.2.1. Loss of cooling
15.2.1.1. Loss of heat sink
15.2.1.2. Risks related to maintenance during unit outages
15.2.1.3. Suction of foreign matter into the cooling system
15.2.1.4. Exceeding the decay heat defined in facility design
15.2.2. Water losses
15.2.2.1. Gate or sluice gate failures
15.2.2.2. Line-up errors
15.2.2.3. Failure of a reactor coolant system pipe nozzle dam
15.2.2.4. Rupture of a pipe connected to the spent fuel pool
15.3. Safety reassessments
15.4. Experience feedback from the accident that affected the Unit 4 pool at the Fukushima Daiichi nuclear power plant
15.4.1. Events
15.4.2. Complementary safety assessments conducted in France
15.5. Measures adopted for the EPR
15.6. Recommendations for new reactor designs
15.7. New systems for storing spent fuel
Chapter 16. Taking into Account Human and Organizational Factors in Facility Design
16.1. Taking into account human and organizational factors in nuclear power reactor design
16.1.1. Importance of considering human and organizational factors at the design stage
16.1.2. Approach at the design stage
16.1.2.1. Prior to the design phase: analysis of ‘existing elements’
16.1.2.2. Design objectives
16.1.2.3. Definition of detailed design provisions
16.1.2.4. Validation of design provisions
16.1.2.5. Assessments conducted during reactor startup and after commissioning
16.1.3. Project management and human and organizational factors engineering programme
16.2. Considering human and organizational aspects when designing changes to nuclear power plants
16.2.1. Importance of human and organizational factors in designing modifications
16.2.2. ‘Human, social and organizational approach’ implemented by EDF
16.2.3. Changes, a subject that always deserves special attention from a human and organizational factors perspective
16.3. Human and organizational factors for future nuclear power reactor projects
Chapter 17. Studying Core-Melt Accidents to Enhance Safety
17.1. Core degradation and vessel failure
17.1.1. Core uncovery
17.1.2. Fuel degradation
17.1.3. Failure of the reactor coolant system
17.1.4. Phenomena that can cause early containment failure
17.1.5. Phenomena that can ultimately lead to containment failure
17.2. Containment failure modes
17.3. Classification of releases associated with core-melt accidents − ‘source terms’
17.4. Improving knowledge
17.5. Studies in France on containment failure modes
17.5.1. Introduction
17.5.2. Initial containment leakage
17.5.3. Direct heating of gases in the containment
17.5.4. Hydrogen explosion in the containment
17.5.5. Steam explosion in the vessel or reactor pit
17.5.6. Gradual pressure increase in the containment
17.5.7. Penetration of the concrete basemat of the containment by corium
17.5.8. ‘U4’ provisions
17.5.9. Bypass of containment by outgoing pipes (the V mode)
17.5.10. Fast reactivity insertion accidents
17.6. Severe accident operating guidelines
17.7. Radiological consequences associated with the S3 source term and emergency response plans implemented by public authorities
17.8. Ultimate emergency operating procedures
17.9. On-site emergency plan
17.10. Approach adopted for the EPR
17.10.1. General safety objectives
17.10.2. ‘Practical elimination’ of core-melt conditions that could lead to significant early releases
17.10.3. Provisions for low-pressure core melt
Chapter 18. New-Generation Reactors
18.1. Organization and framework of Franco-German discussions
18.2. Progression of safety objectives and design options for the EPR project
18.2.1. General safety objectives
18.2.2. Events to be taken into account at the design stage and in deterministic and probabilistic analyses
18.2.3. Main provisions for preventing incidents and accidents
18.2.4. Functional redundancy, independence between systems, system reliability
18.2.5. Confinement preservation
18.2.6. Radiological protection
18.2.7. Incorporating lessons learned from the Fukushima Daiichi nuclear power plant accident
18.3. International context: general safety objectives for new-generation reactors
18.4. Concepts highlighted in new reactor designs
18.4.1. AP1000: gravity systems
18.4.2. VVER: SPOT system
18.4.3. NM EPR: ‘multi-group’ technology, diversified heat sink
18.4.4. ATMEA 1: safety injection accumulatorsin the reactor coolant system
18.4.5. NuScale: common pool for modular reactors
Part 3. Safety in Operation
Chapter 19. Startup Testsfor Pressurized Water Reactors
19.1. Introduction
19.2. Commissioning
19.2.1. Defining startup tests
19.2.2. Phasing of startup tests
19.2.2.1. Preliminary and pre-operational tests
19.2.2.2. Operational tests
19.2.2.3. General principles for test sequencing and execution
19.2.3. Documentation for startup tests
19.2.3.1. Integrated system test procedures and startup test procedures
19.2.3.2. Test programmes, test procedures, standard test guidelines
19.2.3.3. Completeness analysis, adequacy analysis
19.2.3.4. Acceptance criteria
19.3. Objectives and general rules to take into account for startup tests
19.4. Key lessons learned from startup tests on nuclear power reactors in France
19.4.1. Qualification tests and on-site tests
19.4.2. Long-term on-site testing
19.4.3. Test configurations and completeness, transpositions
19.4.4. Safety measures that cannot be verified by testing
19.4.5. Criteria
19.4.6. Cleanness, keeping system lines clean, foreign matter
19.4.7. Piping support structures and displacement
19.4.8. Pump and piping vibrations
19.4.9. Validation of operating procedures and periodic tests
19.4.10. Uncertainty and ‘set points’
19.4.11. Condition of facilities during startup tests
19.4.12. Other aspects
19.5. Examples of findings resulting from startup tests
Chapter 20. General Operating Rules
20.1. General Operating Rules
20.1.1. Content of general operating rules
20.1.2. Limits of general operating rules
20.2. Operational limits and conditions
20.2.1. Content of operational limits and conditions
20.2.1.1. Operating modes and standard states
20.2.1.2. Requirements and unavailability
20.2.1.3. Fallback states and time required to reach them
20.2.1.4. Events and event groups
20.2.1.5. Combined types of unavailability
20.2.1.6. Concepts of ‘boundary condition’ and ‘specific requirement’
20.2.2. Average pressure and temperature range of the reactor coolant system
20.2.3. Changes in operational limits and conditions
20.3. Initial and periodic tests
20.4. Incident and accident operating procedures
Chapter 21. Operating Experience Feedback from Events: Rules and Practices
21.1. Background
21.2. Objectives of an operating experience feedback system
21.3. Components of an operating experience feedback system – Regulations
21.4. Operating experience feedback practices adopted for the French nuclear power plant fleet
Chapter 22. Operating Experience from Events Attributable to Shortcomings in Initial Reactor Design or the Quality of Maintenance
22.1. Events attributable to design shortcomings: core cooling deficiencies when reactor is shut down with water level in mid-loop operating range of the residual heat removal system (RHRS)
22.2. Recurrent loss of safety function events related to maintenance operations – Lessons learned
22.2.1. Events
22.2.2. General discussion initiated by EDF in the late 1980s on the quality of maintenance operations
22.2.3. Applying the defence-in-depth concept when working on a reactor in service
22.2.4. Problems that may recur
Chapter 23. Operating Experience from Events Related to Maintenance Operations, Electrical Power Sources and Distribution, Internal and External Hazards
23.1. Risks of failure related to equipment or maintenance
23.1.1. Risks of common-mode failure
23.1.1.1. Risks of common-mode failure related to settings
23.1.1.2. Risks of common-mode failure on electrical switchboards
23.1.1.3. Unavailability of two out of three high-head safety injection lines in the cold legs of the reactor coolant system
23.1.1.4. Loss of electrical power supplies
23.1.2. Introduction of non-borated water into the reactor coolant system
23.1.3. Cooling the reactor coolant system after inhibition of automatic actions
23.1.4. A temporary device prevents switching the safety injection system to the water recirculation mode
23.2. Events related to internal hazards
23.2.1. Risk of common-mode failure due to internal flooding
23.2.2. Risk of failure due to fire
23.2.3. Risks associated with the use of hydrogen in 900 MWe reactors
23.3. External hazards: events related to periods of extreme cold
Chapter 24. Enhanced Protectionof Estuary and River Sites: Flooding at the Blayais Nuclear Power Plant and Obstruction of a Water Intake at the Cruas-Meysse Nuclear Power Plant
24.1. Partial loss of engineered safety systems following flooding of the Blayais nuclear power plant
24.2. Total loss of heat sink due to clogging of filter drums by a massive influx of plant matter at the Cruas-Meysse nuclear power plant
Chapter 25. Taking into Account Human and Organizational Factors in Facility Operation
25.1. Skills management
25.1.1. Historical background
25.1.2. Managing training
25.1.3. Strategic workforce planning
25.1.4. Personnel certification
25.2. Safety and risk management
25.2.1. Historical background
25.2.2. Decision-making and safety
25.2.3. Risk analyses applied to work activities
25.2.4. Operating experience feedback
25.2.5. Managing organizational change
25.3. Managing operational activities
25.3.1. Characteristics of operational activities
25.3.2. Monitoring by the operating crew in the control room
25.3.3. Compliance with general operating rules
25.3.4. Line-up
25.3.5. Operation in extreme situations
25.4. Management of maintenance activities
25.4.1. Management of a scheduled reactor outage for refuelling and maintenance
25.4.2. Risks during reactor outages
25.4.3. Preparation for scheduled reactor outages
25.4.4. Managing scheduled reactor outages
25.5. Supervising outsourced activities
25.5.1. Contractor qualification and contracting
25.5.2. Matching workload and resources
25.2.3. Risk analyses applied to work activities
25.2.4. Operating experience feedback
25.2.5. Managing organizational change
Chapter 26. Facility Maintenance
26.1. Maintenance objectives
26.2. Maintenance
26.2.1. Definition
26.2.2. Maintenance strategies
26.3. Optimizing maintenance
26.3.1. Reliability-centred maintenance
26.3.2. Conditional maintenance
26.3.3. Conditional maintenance by sampling – Maintenance based on reference equipment items
26.3.4. ‘AP-913’ method
26.4. Maintenance baselines
26.5. On-site maintenance
26.5.1. The various stages of maintenance operations
26.5.2. Main conditions for successful maintenance
26.5.3. Examples of anomalies or deviations discovered during routine maintenance, explained by an inadequate maintenance baseline with regard to deterioration mechanisms
26.5.4. Examples of events associated with non-quality maintenance
26.5.4.1. Example of an event explained by an incorrect setting on redundant equipment
26.5.4.2. Example of an event explained by an incorrect setting of electrical protection thresholds
26.5.4.3. Examples of events explained by a failure to return equipment to a compliant state after maintenance or an error in performing a work procedure
Chapter 27. In-service Monitoring and Inspection of Equipment
27.1. Main internal equipment items on a pressurized water reactor vessel
27.1.1. ‘Core baffle’ around the core
27.1.2. RCCA guide tubes
27.2. Reactor vessel, nozzles and head
27.2.1. Vessel underclad defects
27.2.2. Cracking on vessel head adaptors
27.2.2.1. Condition of other reactors
27.2.2.2. Impact on safety
27.2.2.3. Prevention, monitoring and mitigation
27.2.2.4. Developing inspection tools
27.2.2.5. Repairs
27.2.2.6. Leak detection
27.2.2.7. Anti-ejection devices
27.2.2.8. Current situation
27.2.2.9. Cracks observed on reactor vessel heads in other countries
27.2.2.10. Implementation of special monitoring for ‘Inconel areas’ beginning in 1992
27.2.3. Cracking on vessel lower head penetrations detected in 2011
27.2.4. Monitoring the ‘beltline’ region of the vessel
27.2.5. Defects observed on reactor vessels in Belgium
27.3. Steam generators
27.3.1. The different types of defects
27.3.2. Associated risks
27.3.3. Monitoring during operation and inspection during outages
27.3.3.1. Monitoring during operation
27.3.3.2. Inspection during reactor outages
27.3.4. Steps to be taken when a defect is detected
27.3.4.1. Tube wear due to foreign matter
27.3.4.2. Wear due to contact with anti-vibration bars
27.3.4.3. Cracking in U-bend tubes
27.3.4.4. Tube deformation and cracking
27.3.5. Steam generator replacement
27.3.6. Clogging observed in the 2000s
27.3.7. Conclusion
27.4. Steam lines
27.5. Auxiliary systems: cracks induced by local thermal-hydraulic phenomena
27.5.1. Cracking in non-isolatable sections connected to the reactor coolant loops
27.5.2. RHRS thermal fatigue at Unit 1 of the Civaux nuclear power plant
27.6. Civil works: containment structures
27.6.1. Anticipated degradation phenomena
27.6.2. Devices for direct monitoring of containment building concrete walls
27.6.3. Leak tests and measurements
27.6.4. Main anomalies
Chapter 28. Fuel Management, Monitoring and Developments
28.1. Procedures for monitoring fuel rod integrity
28.1.1. Radiochemical specifications for reactor coolant
28.1.2. Inspections and measurements carried out directly on fuel assemblies
28.1.2.1. Liquid penetrant testing in the refuelling machine mast
28.1.2.2. Liquid penetrant testing in the FB cell
28.1.2.3. Inspections performed on fuel rods
28.2. Operating experience feedback and changes in cladding material
28.3. Anomalies and significant events involving fuel assemblies
28.3.1. Baffle jetting
28.3.2. Fretting
28.3.3. Events encountered during handling operations
28.3.4. Lateral deformation of fuel assemblies interfering with RCCA drop
Chapter 29. Facility Compliance
29.1. Introduction
29.2. Detection and treatment of compliance deviations for pressurized water reactors in the nuclear power plant fleet
29.2.1. Process for handling compliance gaps
29.2.2. Examples of compliance gaps
29.2.2.1. Compliance gap in electrical connection boxes qualified for accident conditions
29.2.2.2. Failure in the seismic resistance of metal floors in electrical and auxiliary buildings of 900 MWe reactors (CPY series)
29.2.2.3. Risk of containment sump screen blockage
29.2.2.4. Anomaly found in engines of emergency and SBO diesel generators for 900 MWe reactors
29.2.2.5. Temperature resistance fault in the high-head safety injection pumps
29.2.2.6. Mixed lubricants in equipment required for accident situations
29.2.2.7. Flow imbalance between safety injection lines of 900 MWe reactors
29.2.2.8. Anomaly in CATHARE software modelling of natural circulation in the upper part of the vessel
29.2.2.9. Vibrations and rotor lift on engineered safety motor-driven pump units
Chapter 30. Periodic Reviews
30.1. Introduction
30.2. History of periodic reviews in France for nuclear power reactors
30.2.1. Reactors other than PWRs in the French nuclear power plant fleet
30.2.2. PWRs in the French nuclear power plant fleet (900 MWe, 1300 MWe and 1450 MWe)
30.3. Periodic review process for PWRs in the French nuclear power plant fleet
30.3.1. Regulations
30.3.2. Outline of a PWR periodic review
30.4. Case of the review associated with the third ten-yearly outage of 900 MWe reactors
30.4.1. Plant unit compliance reviews, the complementary investigation and ageing management
30.4.1.1. Plant unit compliance reviews
30.4.1.2. Complementary investigation programme
30.4.1.3. Ageing management
30.4.2. Compliance studies on the design of civil works systems and structures
30.4.3. Studies to reassess system design
30.4.4. Reassessment of reactor resistance to internal and external hazards
30.4.5. Accident studies
30.4.6. Taking into account lessons learned during the review associated with the VD3 900 outage for subsequent reviews
30.5. Fourth ten-yearly outage of 900 MWe reactors: integrating the extension of the operating lifetime of nuclear power reactors in France
30.5.1. Background
30.5.2. Periodic Review Strategic Plan – Setting objectives
30.5.3. A few significant issues identified in reviews conducted by safety organizations
30.6. Overview of international practices – IAEA Guides
30.6.1. International practices
30.6.2. IAEA Guides
30.7. Multilateral practices
Chapter 31. Optimizing Radiation Protection and Limiting Doses Received by Workers During Operations in a Nuclear Power Plant
31.1. Sources of ionizing radiationin a nuclear power reactor
31.2. Examples of optimization of worker radiation protection
31.3. Arrangements for ‘Major Refit’ operations
31.4. Approach and objectives adopted for the EPR
Part 4. The Accidents at Three Mile Island, Chernobyl and Fukushima Daiichi Nuclear Power Plants, Lessons Learned and Emergency Response Management
Chapter 32. The Three Mile Island Nuclear Power Plant Accident
32.1. Accident sequence – Reconstitution through simulation
32.2. Accident consequences
32.3. Analysis of the accident causes
32.3.1. Error in identifying the position of the relief valve
32.3.2. Understanding the behaviour of the pressurizer
32.3.3. Stopping safety injection
32.3.4. Human-machine interface
32.3.5. Isolating the reactor containment
32.3.6. Confinement inside the auxiliary building
32.3.7. Emergency feedwater supply to the steam generators
32.4. Lessons learned from the Three Mile Island accident
32.4.1. The human factor in facility operation
32.4.2. Importance of precursor events
32.4.3. Study of complex situations and core-melt accidents, handling emergency situations
32.5. Conclusions
Chapter 33. Incident and Accident Operation: from the Event-Oriented Approach to the State-Oriented Approach
33.1. Limits of the event-oriented approach
33.2. The State-Oriented Approach concept
33.3. First application of the state-oriented approach
33.4. Widespread application of the state-oriented approach
33.5. ‘Stabilized’ state-oriented approach
33.6. State-oriented approach adopted for the EPR
Chapter 34. The Chernobyl Nuclear Power Plant Accident
34.1. The Chernobyl nuclear power plant and RBMK reactors
34.2. The accident sequence
34.3. Analysis of the accident causes and changes made to RBMK units soon after the accident
34.4. The other units at the facility
34.5. Radioactive release and protection of the population
34.5.1. Radioactive release kinetics
34.5.2. Protection of the population
34.6. Consequences on human health and the environment
34.6.1. Direct effects of radiation
34.6.2. Thyroid cancer in children
34.6.3. Long-term contamination in the Dnieper Basin
34.7. Radioactive fallout in France and its consequences
34.7.1. Doses attributable to the plume
34.7.2. External doses due to soil deposition
34.7.3. Doses due to ingestion of contaminated foodstuffs
34.7.4. Overall levels
34.7.5. Thyroid cancer
34.7.5.1. Monitoring thyroid cancer in France
34.7.5.2. Assessment of the number of cancer cases induced in France by the Chernobyl accident
34.8. Lessons learned by the international community from a general viewpoint and with regard to RBMK reactors
34.9. Lessons learned in France
34.10. Keeping the public informed
34.11. After the Chernobyl accident
Chapter 35. Options and Control of Reactivity Insertion in Pressurized Water Reactors
35.1. Research and study of event sequences
35.1.1. Cooling accidents
35.1.2. Incidents and accidents related to rod cluster control assemblies
35.1.3. Boron dilution accidents
35.1.4. Inserting a cold water plug in the core
35.2. Changes in criteria
35.3. The case of outage states
35.4. Regulations
Chapter 36. The Reactor Accident at the Fukushima Daiichi Nuclear Power Plant and Lessons Learned in France
36.1. Reactor units at the Fukushima Daiichi nuclear power plant
36.1.1. General operation of a boiling water reactor
36.1.2. Containment
36.1.3. Emergency cooling systems
36.2. Sequence of events during the accident
36.3. Radioactive release
36.3.1. Airborne radioactive release, residual caesium deposits and contamination of foodstuffs
36.3.1.1. Airborne radioactive release
36.3.1.2. Persistent caesium deposits
36.3.1.3. Contamination of foodstuffs
36.3.2. Release of radioactive substances in the Pacific Ocean
36.3.3. Long-distance atmospheric dispersion of the radioactive plume
36.4. Action taken to control facilities and released contaminated water
36.5. Socioeconomic and health impact in numbers
36.5.1. Socioeconomic impact
36.5.2. Health impact
36.6. Lessons learned from the accident
36.6.1. Complementary safety assessments carried out in Europe and France following the Fukushima Daiichi nuclear power plant accident
36.6.2. Complementary safety assessments carried out in France
36.6.3. Procedure for complementary safety assessments carried out in France
36.6.4. Conclusions of the complementary safety assessments carried out in France
36.6.5. The ‘hardened safety core’
36.6.5.1. Purpose
36.6.5.2. Principles
36.6.5.3. Illustrations
36.6.6. Nuclear Rapid Response Force (FARN)
36.6.7. Deployment of post-Fukushima measures in French nuclear power plants
36.7. Other lessons learned in France from the Fukushima Daiichi nuclear power plant accident
Chapter 37. Lessons Learned from the Fukushima Daiichi Nuclear Power Plant Accident: Work Conducted by the IAEA and WENRA, Action Taken in Countries Other than France
37.1. Work conducted by the IAEA
37.2. Work conducted by WENRA
37.3. Japan
37.4. Belgium
37.4.1. Nuclear power plants in Belgium
37.4.2. General details on the design of Belgian nuclear power plants
37.4.3. Stress tests and main lessons learned
37.4.3.1. Improving protection of facilities against external hazards
37.4.3.2. Improving protection of facilities against loss of electrical power supplies or loss of heat sink
37.4.3.3. Improving on-site emergency plans
37.4.3.4. Improving management of core-melt accidents
37.5. USA
Chapter 38. Emergency Preparedness and Response
38.1. Defining a radiological emergency and ‘response’ objectives
38.2. General organization of radiological emergency management
38.2.1. Organization and entities concerned
38.2.2. Major Nuclear or Radiological Accident National Response Plan and emergency plans
38.2.2.1. Major Nuclear or Radiological Accident National Response Plan
38.2.2.2. Emergency plans
38.2.2.3. Provisions for protecting the public in the event of an accidental release of radioactivity
38.3. Management by the operator
38.4. Prefectural authorities and mayors
38.5. ASN, the Nuclear Safety Authority
38.6. IRSN
38.7. Assessment approach in the event of an accident affecting a reactor in the nuclear power plant fleet
38.7.1. ‘3D/3P’ method
38.7.2. The ‘aggravated prognosis’ approach
38.7.3. Extending the 3D/3P method to severe accidents (the ‘D/P AG’ method)
38.8. Emergency preparedness
38.8.1. Emergency response exercises
38.8.2. Operating experience feedback
Part 5. PWR Safety Studies, R&D and Simulation Software
Chapter 39. PWR Safety Studies and R&D
39.1. Contribution of studies to the improvement of pressurized water reactor safety
39.2. Purpose and overview of R&D work, dedicated programmes and organizations involved, and research facilities in France
39.2.1. Purpose and overview of R&D
39.2.2. Dedicated frameworks and organizations involved
39.2.3. Facilities in France used for research and development
Chapter 40. Examples of Simulation Software Developed for Safety Analysis of Pressurized Water Reactors
40.1. Simulation software for neutronics
40.2. Simulation software for thermal hydraulics (and mechanics)
40.3. Simulation software for thermal mechanics
40.4. Software for simulating core-melt situations
40.5. Simulation software for mechanics
40.6. Fire simulation software
List of Acronyms
Acronyms for institutions, bodies and groups
Technical acronyms and abbreviations
Technical glossary