EDP Sciences 1775555484 20260407 fre

laboutique.edpsciences.fr-002108 03 01 01 002108 03 9782759827220 15 9782759827220 00 BA 02 160 mm 01 240 mm 10 01 02 Institut de Radioprotection et de Sûreté Nucléaire 01 01 ELEMENTS OF NUCLEAR SAFETY PRESSURIZED WATER REACTORS 1 B15 01 A1684 Jean Couturier Couturier, Jean Jean Couturier 01 eng 00 1170 03 01 24 Izibook:Subject Radioprotection 24 Izibook:SubjectAndCategoryAndTags |Radioprotection| 20 pressurized water reactor safety;nuclear power;nuclear power plant;safety;reactor;radiological emergency;nuclear emergency;safety analysis method;nuclear safety;nuclear risk control;radiological risk control 10 2011 BISACSCI058000 01 620 06 06 05 03 00 Everything that is important to know about pressurized water reactor safety in nuclear power plants is compiled in this work of reference, from safety fundamentals and reactor design to provisions for managing a radiological or nuclear emergency. The book narrates the determined and continuous quest to enhance nuclear safety. Jean Couturier, coordinator and senior editor, covers the forty-year history of evolution in the safety objectives, approaches, analysis methods and assessment criteria that have defined pressurized water reactor safety, mainly within the French nuclear power plant fleet, from the 1970s up to today’s Flamanville 3 EPR. Everyone, including current and future generations of engineers, researchers and, more broadly, any citizen interested in nuclear safety issues will find that the book reinforces their knowledge on this important subject, providing an understanding of the fundamentals, to the benefit of nuclear and radiological risk control. 02 00 Everything that is important to know about pressurized water reactorsafety in nuclear power plants is compiled in this work of reference, fromsafety fundamentals and reactor design to provisions for managing aradiological or nuclear emergency. 04 00 ContentsPreface............................................................................................................................................. IIIForeword......................................................................................................................................... VIIEditors, Contributors and Reviewers................................................................................... XXXVIIIntroduction................................................................................................................................... XLIIIPart 1General BackgroundChapter 1Biological and Health Effects of Ionizing Radiation – The Radiological Protection System1.1. Biological and health effects of ionizing radiation............................................. 41.1.1. Biological processes................................................................................. 41.1.2. Review of units of measure................................................................... 61.1.3. Natural radioactivity............................................................................... 71.1.4. Health effects............................................................................................ 81.1.4.1. Deterministic effects, tissue reactions......................... 81.1.4.2. Stochastic or random effects.......................................... 91.1.4.3. Induction of diseases other than cancer..................... 111.1.5. Example of the limitations of epidemiology................................... 111.2. Radiological protection system................................................................................ 131.2.1. Types of exposure situations................................................................ 14 1.2.2. Exposure categories................................................................................. 141.2.3. Justification principle............................................................................... 151.2.4. Optimization (ALARA) principle ......................................................... 161.2.5. Principle of application of dose limits............................................... 21Chapter 2Organization of Nuclear Safety Control and Regulation for Nuclear Facilities and Activities in France2.1. From the founding of CEA to the TSN Act........................................................... 232.2. A few definitions........................................................................................................... 282.3. The different contributors to nuclear safety and their missions.................. 302.4. A few basic principles and notions in the field of nuclear safety................. 462.5. Statutory and quasi-statutory frameworks applicable to basic nuclear installations.............................................. 48Chapter 3The International Dimension and the Social Dimension3.1. International dimension.............................................................................................. 773.1.1. Introduction................................................................................................ 773.1.2. IAEA standards........................................................................................... 803.1.3. International Reporting System for Operating Experience (IRS).......................................................... 823.1.4. Services developed by the IAEA........................................................... 843.1.4.1. OSART reviews.................................................................... 843.1.4.2. IRRS reviews......................................................................... 873.1.4.3. Other services and study frameworks set up by the IAEA.............................................................. 883.1.5. WANO.......................................................................................................... 903.1.6. NEA............................................................................................................... 913.1.7. Organizations dedicated to radiation protection and health........................................................................ 933.1.8. From bilateral Franco-German cooperation to European structures for the exchange and capitalization of knowledge and practices, training and assessment services........................................................ 943.1.9. Nuclear regulator associations............................................................. 1003.2. The social dimension................................................................................................... 1023.2.1. Introduction – the context in France................................................. 1023.2.2. Examples of initiatives and issues raised concerning reactor safety in the French nuclear power plant fleet........... 102Chapter 4Nuclear Reactors: Complex Sociotechnical Systems – the Importance of Human and Organizational Factors4.1. 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.............................. 1084.2. The accident at the Chernobyl nuclear power plant and the concept of ‘safety culture’................................. 1094.3. The Fukushima Daiichi nuclear power plant accident: the social dimension and the concept of organization ‘resilience’... 1134.4. Changes in the perception of the role of people in achieving a high level of reliability in complex sociotechnical systems...... 1134.5. Main topics studied in the development of resources and skills pertaining to human and organizational factors............ 1164.5.1. Resources and skills................................................................................. 1164.5.2. Main topics studied.................................................................................. 1184.6. Human and organizational factors in French regulations................................ 119 Part 2Safety by DesignChapter 5The Development of Nuclear Power Using Uranium-235 Fission – A Few Notions of Physics Used in Pressurized Water Reactors5.1. Important milestones in the development of nuclear power using fission of the uranium-235 isotope.......................... 1235.2. Fission and important concepts in reactor kinetics........................................... 1275.3. Removing power from the core during operation............................................. 1345.4. Decay heat...................................................................................................................... 1355.5. Main features of pressurized water reactor cores.............................................. 1365.6. Control and monitoring of pressurized water reactor cores.......................... 1375.7. Using uranium and plutonium mixed oxide (MOX) fuel................................. 147Chapter 6General Objectives, Principles and Basic Concepts of the Safety Approach6.1. General approach to risks – General objectives................................................. 1526.2. Fundamental safety functions.................................................................................. 1566.3. Confinement barriers................................................................................................... 1576.4. Defence in depth........................................................................................................... 1616.4.1. Levels of defence in depth..................................................................... 1626.4.2. Elements common to the different levels of defence in depth............................................................................. 1696.5. Events considered: terminology adopted for nuclear power reactors......... 1696.6. WENRA reference levels............................................................................................. 1716.7. Deterministic safety analysis and probabilistic safety assessments............ 1716.8. Lessons learned from the accident at the Fukushima Daiichi nuclear power plant on the concept of defence in depth and deterministic analysis......................................................................................... 1736.9. Safety culture – Quality control.............................................................................. 174Chapter 7Safety Options and Considerations at the Design Phase7.1. Different types of design provisions associated with safety considerations................................................. 1887.2. Single-failure criterion................................................................................................. 1897.3. The specific nature of computer-based systems (based on instrumentation and control software)............... 1937.4. Equipment safety classification................................................................................ 1957.4.1. Importance of equipment for safety and safety classification...................................................................... 1957.4.2. Generic requirements associated with the different safety classes.................................................................. 1997.4.3. Qualification of equipment for accident conditions..................... 2027.5. Information on designing nuclear pressure equipment.................................... 2077.6. General considerations on provisions for hazards in facility design............ 2107.7. Anticipating decommissioning in the design stage........................................... 212Chapter 8Study of Operating Conditions in the Deterministic Safety Analysis8.1. Categories of operating conditions......................................................................... 2188.2. Choice of operating conditions................................................................................ 2228.2.1. Concept of ‘bounding’ incident or accident..................................... 2238.2.2. Accident exclusion.................................................................................... 2248.3. List and breakdown of operating conditions....................................................... 2258.4. Methods for studying operating conditions......................................................... 2278.4.1. Choice of initial conditions, conservatism....................................... 2288.4.2. Consideration of an aggravating event in the study on operating conditions – ‘Passive’ failures............ 2298.4.3. Conventional combinations.................................................................. 2318.4.4. Preventing accident aggravation......................................................... 2328.4.5. Operator response time......................................................................... 2328.4.6. Using qualified simulation software................................................... 2338.4.7. Main criteria to be met for fuel in the reactor core..................... 2348.5. Concept of ‘design-basis situations’ for equipment.......................................... 2368.6. Situations to be taken into account in application of pressure equipment regulations................................................. 2378.7. Assessing the radiological consequences of incidents, accidents and hazards.............................................. 2388.7.1. Assessing radioactive substances released from the facility...... 2408.7.2. Assessing radiological consequences of radioactive release from the facility................................................ 2428.7.3. Assessing radiological consequences................................................. 243Chapter 9Loss-of-Coolant Accident9.1. Short- and medium-term aspects of a LOCA...................................................... 2529.1.1. Mechanical effects on vessel internals and fuel assembly structures................................................................ 2539.1.2. Thermal-hydraulic aspects and behaviour of fuel rods................ 2559.1.2.1. Large-break LOCA............................................................... 2559.1.2.2. Intermediate-break LOCA................................................ 2579.1.3. Effects on reactor containment and internals................................ 2589.1.4. Long-term aspect...................................................................................... 2599.2. Safety demonstration.................................................................................................. 2619.2.1. General information and background................................................ 2619.2.2. Fuel assemblies and fuel rods, vessel internals, reactor coolant system components................................................. 2639.2.2.1. Mechanical strength of vessel internals, fuel assembly structures and reactor coolant system components.............. 2639.2.2.2. Fuel behaviour...................................................................... 2659.2.3. Reactor containment and equipment located inside.................... 266Chapter 10A Special Issue: Steam Generator Tubes10.1. Steam generator tube rupture as a Category 3 event...................................... 27010.2. Preventing an SGTR accident, risk of multiple ruptures................................... 27210.3. Steam generator tube rupture(s) studied as a Category 4 event................. 27410.3.1. 900 MWe and 1300 MWe reactors.................................................... 27410.3.2. 1450 MWe reactors and EPR (Flamanville 3).................................. 27410.4. Provisions to mitigate the radiological consequences of SGTR accidents. 276Chapter 11Providing for Hazards: General Considerations and Internal Hazards11.1. General considerations on providing for hazards............................................... 27911.2. Potential projectiles inside the containment....................................................... 28211.3. Effects of pipe breaks.................................................................................................. 28411.4. Projectiles generated by a turbine rotor failure.................................................. 28511.5. Protection against load drops................................................................................... 28811.5.1. Risks related to spent fuel transport packaging............................. 28811.5.2. Other handling risks................................................................................ 29111.6. Fire protection............................................................................................................... 29211.7. Explosion protection.................................................................................................... 29711.8. Internal flooding............................................................................................................ 300Chapter 12Providing for External Hazards12.1. General considerations on providing for external hazards.............................. 30512.2. ‘Climate watch’ implemented by EDF.................................................................... 30712.3. Earthquakes..................................................................................................................... 30712.4. External floods............................................................................................................... 32212.5. Extreme temperatures................................................................................................. 33012.5.1. Extreme cold.............................................................................................. 33012.5.2. Extreme heat.............................................................................................. 33112.6. Possible heat sink hazards.......................................................................................... 33212.7. Other naturally-occurring external hazards......................................................... 33612.8. Accidental aeroplane crashes (excluding malicious acts)................................ 33612.9. Risks related to the industrial environment (excluding malicious acts)..... 340Chapter 13Complementary Domain of Events13.1. The origin of studies belonging to the complementary domain................... 34413.2. Background of the complementary domain........................................................ 34413.3. Analysis of complementary domain events......................................................... 35113.4. ‘New complementary domain’................................................................................. 35113.5. Case of the Flamanville 3 EPR.................................................................................. 354Chapter 14Development and Use of Probabilistic Safety Assessments14.1. History and regulatory context................................................................................ 35714.1.1. International situation............................................................................ 35714.1.2. Situation in France................................................................................... 35914.2. Level 1 PSA...................................................................................................................... 36114.2.1. Scope............................................................................................................ 36114.2.2. Method for carrying out a Level 1 PSA............................................. 36214.2.2.1. General information........................................................... 36214.2.2.2. Specific point: probabilistic human reliability analysis................................................................................... 36414.2.3. Level 1 PSA results and lessons learned............................................ 36914.3. Level 2 PSA...................................................................................................................... 37314.3.1. Scope............................................................................................................ 37314.3.2. Method for carrying out a Level 2 PSA............................................. 37414.3.2.1. General information........................................................... 37414.3.2.2. Probabilistic human reliability analysis for Level 2 PSAs................................................................... 37914.3.3. Examples of lessons learned from Level 2 PSAs............................. 38214.3.3.1. Steam explosion risk assessment.................................. 38214.3.3.2. Mechanical integrity of the 900 MWe reactor containments........................................................ 38314.3.3.3. Isolating penetrations in the containment................. 38314.3.3.4. Modifying the pressure relief system of the reactor coolant system........................................ 38414.3.3.5. Improvement of operating procedures to reduce risk of core melt under pressure................ 38414.3.3.6. Contribution of Level 2 PSAs to emergency response measures............................................................. 38514.4. Expanding the scope of PSA coverage................................................................... 38514.5. Using probabilistic safety assessments.................................................................. 38614.5.1. Using PSAs in the design phase........................................................... 38614.5.1.1. Usefulness and particularities of PSAs in the design phase............................................................. 38614.5.1.2. PSAs conducted to support the Flamanville 3 EPR design............................................................................. 38714.5.2. Using PSAs in periodic reviews............................................................ 38914.5.2.1. Level 1 PSA............................................................................ 39014.5.2.2. Level 2 PSA............................................................................ 39114.5.3. Using PSAs for reactor operation........................................................ 39214.5.3.1. Using PSAs to analyse event severity.......................... 39214.5.3.2. Using PSAs to analyse operational limits and conditions and temporary changes...................... 39414.5.3.3. Using PSAs to analyse operating procedures............. 396Chapter 15Aspects Specific to PWR Spent Fuel Storage Pools15.1. Spent fuel pool design................................................................................................. 39915.1.1. Confinement barriers.............................................................................. 39915.1.2. Initiating events defined at the design stage.................................. 40015.2. Experience feedback..................................................................................................... 40115.2.1. Loss of cooling........................................................................................... 40115.2.1.1. Loss of heat sink.................................................................. 40115.2.1.2. Risks related to maintenance during unit outages................................................................................... 40215.2.1.3. Suction of foreign matter into the cooling system.................................................................................... 40315.2.1.4. Exceeding the decay heat defined in facility design...................................................................................... 40315.2.2. Water losses............................................................................................... 404 15.2.2.1. Gate or sluice gate failures.............................................. 40415.2.2.2. Line-up errors....................................................................... 40515.2.2.3. Failure of a reactor coolant system pipe nozzle dam.......................................................................................... 40915.2.2.4. Rupture of a pipe connected to the spent fuel pool................................................................................. 41115.3. Safety reassessments................................................................................................... 41115.4. Experience feedback from the accident that affected the Unit 4 pool at the Fukushima Daiichi nuclear power plant..... 41415.4.1. Events........................................................................................................... 41415.4.2. Complementary safety assessments conducted in France......... 41715.5. Measures adopted for the EPR.................................................................................. 41915.6. Recommendations for new reactor designs......................................................... 42015.7. New systems for storing spent fuel........................................................................ 423Chapter 16Taking into Account Human and Organizational Factors in Facility Design16.1. Taking into account human and organizational factors in nuclear power reactor design....................................... 42516.1.1. Importance of considering human and organizational factors at the design stage....................................... 42516.1.2. Approach at the design stage............................................................... 43116.1.2.1. Prior to the design phase: analysis of ‘existing elements’........................................ 43216.1.2.2. Design objectives................................................................ 43516.1.2.3. Definition of detailed design provisions...................... 43616.1.2.4. Validation of design provisions...................................... 43916.1.2.5. Assessments conducted during reactor startup and after commissioning.................................................. 44216.1.3. Project management and human and organizational factors engineering programme..................................... 44316.2. Considering human and organizational aspects when designing changes to nuclear power plants..................... 44316.2.1. Importance of human and organizational factors in designing modifications..................................................................... 44416.2.2. ‘Human, social and organizational approach’ implemented by EDF........................................................................... 44516.2.3. Changes, a subject that always deserves special attention from a human and organizational factors perspective.... 44716.3. Human and organizational factors for future nuclear power reactor projects.................................................. 448Chapter 17Studying Core-Melt Accidents to Enhance Safety17.1. Core degradation and vessel failure....................................................................... 45217.1.1. Core uncovery............................................................................................ 45317.1.2. Fuel degradation....................................................................................... 45417.1.3. Failure of the reactor coolant system............................................... 45517.1.4. Phenomena that can cause early containment failure................. 45517.1.5. Phenomena that can ultimately lead to containment failure....................................................... 45717.2. Containment failure modes....................................................................................... 45717.3. Classification of releases associated with core-melt accidents − ‘source terms’................................................. 46017.4. Improving knowledge.................................................................................................. 46217.5. Studies in France on containment failure modes............................................... 46217.5.1. Introduction................................................................................................ 46217.5.2. Initial containment leakage................................................................... 46317.5.3. Direct heating of gases in the containment.................................... 46317.5.4. Hydrogen explosion in the containment.......................................... 46417.5.5. Steam explosion in the vessel or reactor pit................................... 46517.5.6. Gradual pressure increase in the containment............................... 46617.5.7. Penetration of the concrete basemat of the containment by corium............................................ 46717.5.8. ‘U4’ provisions........................................................................................... 46817.5.9. Bypass of containment by outgoing pipes (the V mode).............................................................................................. 46817.5.10. Fast reactivity insertion accidents...................................................... 46917.6. Severe accident operating guidelines..................................................................... 46917.7. Radiological consequences associated with the S3 source term and emergency response plans implemented by public authorities............ 47017.8. Ultimate emergency operating procedures.......................................................... 47317.9. On-site emergency plan............................................................................................. 47317.10. Approach adopted for the EPR................................................................................. 47617.10.1. General safety objectives....................................................................... 47617.10.2. ‘Practical elimination’ of core-melt conditions that could lead to significant early releases............................. 47617.10.3. Provisions for low-pressure core melt............................................... 480Chapter 18New-Generation Reactors18.1. Organization and framework of Franco-German discussions........................ 48518.2. Progression of safety objectives and design options for the EPR project.......................................................... 48618.2.1. General safety objectives....................................................................... 48618.2.2. Events to be taken into account at the design stage and in deterministic and probabilistic analyses.............................. 48818.2.3. Main provisions for preventing incidents and accidents............. 49018.2.4. Functional redundancy, independence between systems, system reliability................................................ 49518.2.5. Confinement preservation..................................................................... 49618.2.6. Radiological protection........................................................................... 49718.2.7. Incorporating lessons learned from the Fukushima Daiichi nuclear power plant accident......................... 49718.3. International context: general safety objectives for new-generation reactors............................................ 49918.4. Concepts highlighted in new reactor designs...................................................... 50118.4.1. AP1000: gravity systems........................................................................ 50118.4.2. VVER: SPOT system................................................................................. 50318.4.3. NM EPR: ‘multi-group’ technology, diversified heat sink............ 504 18.4.4. ATMEA 1: safety injection accumulators in the reactor coolant system.............................................................. 50518.4.5. NuScale: common pool for modular reactors................................. 505 Part 3Safety in OperationChapter 19Startup Tests for Pressurized Water Reactors19.1. Introduction.................................................................................................................... 51119.2. Commissioning.............................................................................................................. 51419.2.1. Defining startup tests.............................................................................. 51419.2.2. Phasing of startup tests.......................................................................... 51519.2.2.1. Preliminary and pre-operational tests......................... 51519.2.2.2. Operational tests................................................................ 51719.2.2.3. General principles for test sequencing and execution....................................................................... 51719.2.3. Documentation for startup tests......................................................... 51819.2.3.1. Integrated system test procedures and startup test procedures............................................ 51819.2.3.2. Test programmes, test procedures, standard test guidelines.................................................... 51819.2.3.3. Completeness analysis, adequacy analysis................. 51819.2.3.4. Acceptance criteria............................................................. 51919.3. Objectives and general rules to take into account for startup tests........... 52019.4. Key lessons learned from startup tests on nuclear power reactors in France............................................................ 52119.4.1. Qualification tests and on-site tests.................................................. 52219.4.2. Long-term on-site testing...................................................................... 52419.4.3. Test configurations and completeness, transpositions................ 525 19.4.4. Safety measures that cannot be verified by testing..................... 52719.4.5. Criteria......................................................................................................... 52719.4.6. Cleanness, keeping system lines clean, foreign matter................ 52819.4.7. Piping support structures and displacement................................... 53219.4.8. Pump and piping vibrations................................................................... 53219.4.9. Validation of operating procedures and periodic tests................ 53419.4.10. Uncertainty and ‘set points’.................................................................. 53519.4.11. Condition of facilities during startup tests...................................... 53619.4.12. Other aspects............................................................................................. 53719.5. Examples of findings resulting from startup tests............................................. 538Chapter 20General Operating Rules20.1. General Operating Rules............................................................................................ 55220.1.1. Content of general operating rules..................................................... 55320.1.2. Limits of general operating rules......................................................... 55420.2. Operational limits and conditions........................................................................... 55420.2.1. Content of operational limits and conditions................................. 55520.2.1.1. Operating modes and standard states......................... 55620.2.1.2. Requirements and unavailability................................... 55820.2.1.3. Fallback states and time required to reach them.... 55820.2.1.4. Events and event groups.................................................. 55920.2.1.5. Combined types of unavailability.................................. 56020.2.1.6. Concepts of ‘boundary condition’ and ‘specific requirement’................................................ 56020.2.2. Average pressure and temperature range of the reactor coolant system.............................................................. 56120.2.3. Changes in operational limits and conditions................................. 56320.3. Initial and periodic tests............................................................................................. 56420.4. Incident and accident operating procedures........................................................ 566Chapter 21 Operating Experience Feedback from Events: Rules and Practices21.1. Background...................................................................................................................... 56921.2. Objectives of an operating experience feedback system................................ 57121.3. Components of an operating experience feedback system – Regulations............................................................................... 57221.4. Operating experience feedback practices adopted for the French nuclear power plant fleet................................. 576Chapter 22Operating Experience from Events Attributable to Shortcomings in Initial Reactor Design or the Quality of Maintenance22.1. 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).......................................... 59022.2. Recurrent loss of safety function events related to maintenance operations – Lessons learned........................... 59522.2.1. Events........................................................................................................... 59522.2.2. General discussion initiated by EDF in the late 1980s on the quality of maintenance operations....................................... 60522.2.3. Applying the defence-in-depth concept when working on a reactor in service.................................. 60822.2.4. Problems that may recur....................................................................... 609Chapter 23Operating Experience from Events Related to Maintenance Operations, Electrical Power Sources and Distribution, Internal and External Hazards23.1. Risks of failure related to equipment or maintenance..................................... 61223.1.1. Risks of common-mode failure............................................................ 61223.1.1.1. Risks of common-mode failure related to settings............................................................................. 61223.1.1.2. Risks of common-mode failure on electrical switchboards.................................................61323.1.1.3. Unavailability of two out of three high-head safety injection lines in the cold legs of the reactor coolant system.............. 61423.1.1.4. Loss of electrical power supplies................................... 61723.1.2. Introduction of non-borated water into the reactor coolant system.......................................................... 62023.1.3. Cooling the reactor coolant system after inhibition of automatic actions............................................................................... 62223.1.4. A temporary device prevents switching the safety injection system to the water recirculation mode................... 62423.2. Events related to internal hazards........................................................................... 62523.2.1. Risk of common-mode failure due to internal flooding.............. 62523.2.2. Risk of failure due to fire....................................................................... 62823.2.3. Risks associated with the use of hydrogen in 900 MWe reactors............................................................................... 63123.3. External hazards: events related to periods of extreme cold......................... 635Chapter 24Enhanced Protection of Estuary and River Sites: Flooding at the Blayais Nuclear Power Plant and Obstruction of a Water Intakeat the Cruas-Meysse Nuclear Power Plant24.1. Partial loss of engineered safety systems following flooding of the Blayais nuclear power plant............................... 64224.2. 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...................................................... 648Chapter 25Taking into Account Human and Organizational Factors in Facility Operation25.1. Skills management....................................................................................................... 65325.1.1. Historical background............................................................................. 65425.1.2. Managing training.................................................................................... 65525.1.3. Strategic workforce planning................................................................ 65625.1.4. Personnel certification............................................................................ 65725.2. Safety and risk management.................................................................................... 65825.2.1. Historical background............................................................................. 65825.2.2. Decision-making and safety.................................................................. 66025.2.3. Risk analyses applied to work activities............................................ 66225.2.4. Operating experience feedback........................................................... 66325.2.5. Managing organizational change......................................................... 66425.3. Managing operational activities............................................................................... 66625.3.1. Characteristics of operational activities............................................ 66625.3.2. Monitoring by the operating crew in the control room.............. 66925.3.3. Compliance with general operating rules......................................... 67025.3.4. Line-up......................................................................................................... 67125.3.5. Operation in extreme situations......................................................... 67225.4. Management of maintenance activities................................................................ 67225.4.1. Management of a scheduled reactor outage for refuelling and maintenance...................................................... 67225.4.2. Risks during reactor outages................................................................. 67325.4.3. Preparation for scheduled reactor outages...................................... 67425.4.4. Managing scheduled reactor outages................................................ 67525.5. Supervising outsourced activities............................................................................ 67525.5.1. Contractor qualification and contracting......................................... 67625.5.2. Matching workload and resources...................................................... 67625.5.3. Carrying out work.................................................................................... 67725.5.4. Surveillance of outsourced activities................................................. 67825.5.5. Operating experience feedback and assessing outsourced activities...................... 678Chapter 26Facility Maintenance26.1. Maintenance objectives.............................................................................................. 68126.2. Maintenance................................................................................................................... 68226.2.1. Definition.................................................................................................... 68226.2.2. Maintenance strategies........................................................................... 68326.3. Optimizing maintenance............................................................................................ 68426.3.1. Reliability-centred maintenance.......................................................... 68426.3.2. Conditional maintenance....................................................................... 68626.3.3. Conditional maintenance by sampling – Maintenance based on reference equipment items...................... 68826.3.4. ‘AP-913’ method....................................................................................... 68926.4. Maintenance baselines................................................................................................ 69126.5. On-site maintenance................................................................................................... 69426.5.1. The various stages of maintenance operations.............................. 69426.5.2. Main conditions for successful maintenance.................................. 69626.5.3. Examples of anomalies or deviations discovered during routine maintenance, explained by an inadequate maintenancebaseline with regard to deterioration mechanisms....................... 70526.5.4. Examples of events associated with non-quality maintenance.............................................. 70826.5.4.1. Example of an event explained by an incorrect setting on redundant equipment................................... 70826.5.4.2. Example of an event explained by an incorrect setting of electrical protection thresholds................. 70826.5.4.3. Examples of events explained by a failure to return equipment to a compliant state after maintenance or an error in performing a work procedure................................................................ 709Chapter 27In-service Monitoring and Inspection of Equipment27.1. Main internal equipment items on a pressurized water reactor vessel...... 71927.1.1. ‘Core baffle’ around the core................................................................ 72027.1.2. RCCA guide tubes..................................................................................... 72227.2. Reactor vessel, nozzles and head............................................................................. 72227.2.1. Vessel underclad defects........................................................................ 72427.2.2. Cracking on vessel head adaptors....................................................... 72527.2.2.1. Condition of other reactors............................................. 72727.2.2.2. Impact on safety................................................................. 72827.2.2.3. Prevention, monitoring and mitigation....................... 72927.2.2.4. Developing inspection tools............................................ 72927.2.2.5. Repairs.................................................................................... 73027.2.2.6. Leak detection...................................................................... 73027.2.2.7. Anti-ejection devices......................................................... 73127.2.2.8. Current situation................................................................. 73127.2.2.9. Cracks observed on reactor vessel heads in other countries............................................................... 73127.2.2.10. Implementation of special monitoring for ‘Inconel areas’ beginning in 1992........................... 73227.2.3. Cracking on vessel lower head penetrations detected in 2011.......................................................... 73327.2.4. Monitoring the ‘beltline’ region of the vessel................................. 73527.2.5. Defects observed on reactor vessels in Belgium............................ 73627.3. Steam generators.......................................................................................................... 73827.3.1. The different types of defects.............................................................. 73827.3.2. Associated risks......................................................................................... 74027.3.3. Monitoring during operation and inspection during outages.... 74127.3.3.1. Monitoring during operation........................................... 74127.3.3.2. Inspection during reactor outages................................. 74227.3.4. Steps to be taken when a defect is detected.................................. 74227.3.4.1. Tube wear due to foreign matter.................................. 74327.3.4.2. Wear due to contact with anti-vibration bars.......... 74327.3.4.3. Cracking in U-bend tubes................................................. 74427.3.4.4. Tube deformation and cracking..................................... 74427.3.5. Steam generator replacement.............................................................. 74527.3.6. Clogging observed in the 2000s.......................................................... 74627.3.7. Conclusion.................................................................................................. 74827.4. Steam lines...................................................................................................................... 74827.5. Auxiliary systems: cracks induced by local thermal-hydraulic phenomena................................................................................. 75027.5.1. Cracking in non-isolatable sections connected to the reactor coolant loops................................................................. 75027.5.2. RHRS thermal fatigue at Unit 1 of the Civaux nuclear power plant......................................... 75127.6. Civil works: containment structures....................................................................... 75527.6.1. Anticipated degradation phenomena................................................. 75627.6.2. Devices for direct monitoring of containment building concrete walls......................................... 75727.6.3. Leak tests and measurements.............................................................. 75927.6.4. Main anomalies......................................................................................... 759Chapter 28Fuel Management, Monitoring and Developments28.1. Procedures for monitoring fuel rod integrity....................................................... 76828.1.1. Radiochemical specifications for reactor coolant.......................... 76828.1.2. Inspections and measurements carried out directly on fuel assemblies.............................................. 77728.1.2.1. Liquid penetrant testing in the refuelling machine mast....................................................................... 77828.1.2.2. Liquid penetrant testing in the FB cell......................... 77928.1.2.3. Inspections performed on fuel rods.............................. 78028.2. Operating experience feedback and changes in cladding material.............. 78228.3. Anomalies and significant events involving fuel assemblies.......................... 78628.3.1. Baffle jetting............................................................................................... 78628.3.2. Fretting........................................................................................................ 78728.3.3. Events encountered during handling operations............................ 78928.3.4. Lateral deformation of fuel assemblies interfering with RCCA drop.......................... 791Chapter 29Facility Compliance29.1. Introduction.................................................................................................................... 79529.2. Detection and treatment of compliance deviations for pressurized water reactors in the nuclear power plant fleet............... 79629.2.1. Process for handling compliance gaps............................................... 79629.2.2. Examples of compliance gaps............................................................... 79829.2.2.1. Compliance gap in electrical connection boxes qualified for accident conditions................................... 79829.2.2.2. Failure in the seismic resistance of metal floors in electrical and auxiliary buildings of 900 MWe reactors (CPY series). 79929.2.2.3. Risk of containment sump screen blockage............... 79929.2.2.4. Anomaly found in engines of emergency and SBO diesel generators for 900 MWe reactors.................... 80029.2.2.5. Temperature resistance fault in the high-head safety injection pumps...................................................... 80229.2.2.6. Mixed lubricants in equipment required for accident situations....................................................... 80229.2.2.7. Flow imbalance between safety injection lines of 900 MWe reactors......................................................... 80429.2.2.8. Anomaly in CATHARE software modelling of natural circulation in the upper part of the vessel...................................... 80529.2.2.9. Vibrations and rotor lift on engineered safety motor-driven pump units.................................... 805Chapter 30Periodic Reviews30.1. Introduction.................................................................................................................... 80730.2. History of periodic reviews in France for nuclear power reactors................ 80930.2.1. Reactors other than PWRs in the French nuclear power plant fleet......................................................... 80930.2.2. PWRs in the French nuclear power plant fleet (900 MWe, 1300 MWe and 1450 MWe).......................................... 81130.3. Periodic review process for PWRs in the French nuclear power plant fleet..................................................................... 81530.3.1. Regulations................................................................................................. 81530.3.2. Outline of a PWR periodic review....................................................... 81730.4. Case of the review associated with the third ten-yearly outage of 900 MWe reactors..................................................... 82230.4.1. Plant unit compliance reviews, the complementary investigation and ageing management............................................. 82430.4.1.1. Plant unit compliance reviews........................................ 82430.4.1.2. Complementary investigation programme................ 82530.4.1.3. Ageing management.......................................................... 826 30.4.2. Compliance studies on the design of civil works systems and structures............................................. 82630.4.3. Studies to reassess system design...................................................... 82730.4.4. Reassessment of reactor resistance to internal and external hazards................................................................................ 83030.4.5. Accident studies........................................................................................ 83230.4.6. Taking into account lessons learned during the review associated with the VD3 900 outage for subsequent reviews....83730.5. Fourth ten-yearly outage of 900 MWe reactors: integrating the extension of the operating lifetime of nuclear power reactorsin France.......................................................................................................................... 83830.5.1. Background................................................................................................. 83830.5.2. Periodic Review Strategic Plan – Setting objectives..................... 83930.5.3. A few significant issues identified in reviews conducted by safety organizations.................. 84230.6. Overview of international practices – IAEA Guides........................................... 84530.6.1. International practices............................................................................ 84630.6.2. IAEA Guides................................................................................................ 84730.7. Multilateral practices................................................................................................... 848Chapter 31Optimizing Radiation Protection and Limiting Doses Received by Workers During Operations in a Nuclear Power Plant31.1. Sources of ionizing radiation in a nuclear power reactor................................ 85231.2. Examples of optimization of worker radiation protection.............................. 85331.3. Arrangements for ‘Major Refit’ operations........................................................... 85531.4. Approach and objectives adopted for the EPR.................................................... 858 Part 4The Accidents at Three Mile Island, Chernobyl and Fukushima Daiichi Nuclear Power Plants, Lessons Learned and Emergency Response ManagementChapter 32The Three Mile Island Nuclear Power Plant Accident32.1. Accident sequence – Reconstitution through simulation................................ 86432.2. Accident consequences............................................................................................... 87232.3. Analysis of the accident causes................................................................................ 87432.3.1. Error in identifying the position of the relief valve....................... 87432.3.2. Understanding the behaviour of the pressurizer............................ 87532.3.3. Stopping safety injection....................................................................... 87632.3.4. Human-machine interface..................................................................... 87632.3.5. Isolating the reactor containment...................................................... 87732.3.6. Confinement inside the auxiliary building....................................... 87732.3.7. Emergency feedwater supply to the steam generators............... 87732.4. Lessons learned from the Three Mile Island accident....................................... 87732.4.1. The human factor in facility operation............................................. 87832.4.2. Importance of precursor events........................................................... 88132.4.3. Study of complex situations and core-melt accidents, handling emergency situations....................................... 88232.5. Conclusions..................................................................................................................... 883Chapter 33Incident and Accident Operation: from the Event-Oriented Approach to the State-Oriented Approach33.1. Limits of the event-oriented approach.................................................................. 88533.2. The State-Oriented Approach concept.................................................................. 88633.3. First application of the state-oriented approach............................................... 88833.4. Widespread application of the state-oriented approach................................. 89133.5. ‘Stabilized’ state-oriented approach....................................................................... 89233.6. State-oriented approach adopted for the EPR.................................................... 894Chapter 34The Chernobyl Nuclear Power Plant Accident34.1. The Chernobyl nuclear power plant and RBMK reactors................................. 89734.2. The accident sequence................................................................................................ 90134.3. Analysis of the accident causes and changes made to RBMK units soon after the accident.................................. 90634.4. The other units at the facility................................................................................... 90834.5. Radioactive release and protection of the population..................................... 90834.5.1. Radioactive release kinetics.................................................................. 90834.5.2. Protection of the population................................................................ 91134.6. Consequences on human health and the environment................................... 91534.6.1. Direct effects of radiation...................................................................... 91534.6.2. Thyroid cancer in children..................................................................... 91734.6.3. Long-term contamination in the Dnieper Basin............................. 91934.7. Radioactive fallout in France and its consequences.......................................... 92134.7.1. Doses attributable to the plume......................................................... 92134.7.2. External doses due to soil deposition................................................ 92134.7.3. Doses due to ingestion of contaminated foodstuffs.................... 92234.7.4. Overall levels.............................................................................................. 92334.7.5. Thyroid cancer........................................................................................... 92434.7.5.1. Monitoring thyroid cancer in France............................ 92534.7.5.2. Assessment of the number of cancer cases induced in France by the Chernobyl accident........... 92634.8. Lessons learned by the international community from a general viewpoint and with regard to RBMK reactors....................... 92734.9. Lessons learned in France........................................................................................... 92834.10. Keeping the public informed..................................................................................... 93134.11. After the Chernobyl accident.................................................................................... 933Chapter 35Options and Control of Reactivity Insertion in Pressurized Water Reactors35.1. Research and study of event sequences................................................................ 93735.1.1. Cooling accidents..................................................................................... 938 35.1.2. Incidents and accidents related to rod cluster control assemblies.................................................................................... 94035.1.3. Boron dilution accidents........................................................................ 94335.1.4. Inserting a cold water plug in the core............................................. 95035.2. Changes in criteria........................................................................................................ 95135.3. The case of outage states.......................................................................................... 95335.4. Regulations...................................................................................................................... 957Chapter 36The Reactor Accident at the Fukushima Daiichi Nuclear Power Plant and Lessons Learned in France36.1. Reactor units at the Fukushima Daiichi nuclear power plant........................ 96236.1.1. General operation of a boiling water reactor.................................. 96236.1.2. Containment.............................................................................................. 96236.1.3. Emergency cooling systems.................................................................. 96436.2. Sequence of events during the accident............................................................... 96636.3. Radioactive release....................................................................................................... 97236.3.1. Airborne radioactive release, residual caesium deposits and contamination of foodstuffs............................... 97236.3.1.1. Airborne radioactive release............................................ 97236.3.1.2. Persistent caesium deposits............................................ 97236.3.1.3. Contamination of foodstuffs.......................................... 97336.3.2. Release of radioactive substances in the Pacific Ocean.............. 97536.3.3. Long-distance atmospheric dispersion of the radioactive plume........................................................................ 97636.4. Action taken to control facilities and released contaminated water.......... 97836.5. Socioeconomic and health impact in numbers................................................... 98136.5.1. Socioeconomic impact............................................................................ 98136.5.2. Health impact............................................................................................ 98236.6. Lessons learned from the accident.......................................................................... 98536.6.1. Complementary safety assessments carried out in Europe and France following the Fukushima Daiichi nuclear power plant accident............................................................................................ 98636.6.2. Complementary safety assessments carried out in France........ 98836.6.3. Procedure for complementary safety assessments carried out in France...................................... 98936.6.4. Conclusions of the complementary safety assessments carried out in France............................................. 99036.6.5. The ‘hardened safety core’.................................................................... 99136.6.5.1. Purpose................................................................................... 99136.6.5.2. Principles................................................................................ 99136.6.5.3. Illustrations........................................................................... 99336.6.6. Nuclear Rapid Response Force (FARN).............................................. 99536.6.7. Deployment of post-Fukushima measures in French nuclear power plants............................................................ 99736.7. Other lessons learned in France from the Fukushima Daiichi nuclear power plant accident...................................... 998Chapter 37Lessons Learned from the Fukushima Daiichi Nuclear PowerPlant Accident: Work Conducted by the IAEA and WENRA, Action Taken in Countries Other than France37.1. Work conducted by the IAEA.................................................................................... 100237.2. Work conducted by WENRA..................................................................................... 100337.3. Japan................................................................................................................................. 100437.4. Belgium............................................................................................................................ 100637.4.1. Nuclear power plants in Belgium........................................................ 100637.4.2. General details on the design of Belgian nuclear power plants................................................................ 100637.4.3. Stress tests and main lessons learned............................................... 100737.4.3.1. Improving protection of facilities against external hazards.................................................................. 100737.4.3.2. Improving protection of facilities against loss of electrical power supplies or loss of heat sink....... 101037.4.3.3. Improving on-site emergency plans............................. 1011 37.4.3.4. Improving management of core-melt accidents...... 101237.5. USA.................................................................................................................................... 1013Chapter 38Emergency Preparedness and Response38.1. Defining a radiological emergency and ‘response’ objectives........................ 102138.2. General organization of radiological emergency management..................... 102338.2.1. Organization and entities concerned................................................. 102338.2.2. Major Nuclear or Radiological Accident National Response Plan and emergency plans........................................ 102538.2.2.1. Major Nuclear or Radiological Accident National Response Plan...................................................................... 102538.2.2.2. Emergency plans................................................................. 102638.2.2.3. Provisions for protecting the public in the event of an accidental release of radioactivity..................... 102838.3. Management by the operator................................................................................... 102938.4. Prefectural authorities and mayors......................................................................... 103138.5. ASN, the Nuclear Safety Authority......................................................................... 103238.6. IRSN................................................................................................................................... 103338.7. Assessment approach in the event of an accident affecting a reactor in the nuclear power plant fleet.................. 103838.7.1. ‘3D/3P’ method......................................................................................... 103938.7.2. The ‘aggravated prognosis’ approach................................................ 104138.7.3. Extending the 3D/3P method to severe accidents (the ‘D/P AG’ method)...................................... 104138.8. Emergency preparedness............................................................................................ 104238.8.1. Emergency response exercises............................................................. 104338.8.2. Operating experience feedback........................................................... 1045 Part 5PWR Safety Studies, R&D and Simulation SoftwareChapter 39PWR Safety Studies and R&D39.1. Contribution of studies to the improvement of pressurized water reactor safety................................................................ 105039.2. Purpose and overview of R&D work, dedicated programmes and organizations involved, and research facilities in France... 105239.2.1. Purpose and overview of R&D............................................................. 105239.2.2. Dedicated frameworks and organizations involved....................... 106339.2.3. Facilities in France used for research and development.............. 1065Chapter 40Examples of Simulation Software Developed for Safety Analysis of Pressurized Water Reactors40.1. Simulation software for neutronics........................................................................ 107640.2. Simulation software for thermal hydraulics (and mechanics)....................... 107940.3. Simulation software for thermal mechanics........................................................ 108440.4. Software for simulating core-melt situations..................................................... 108540.5. Simulation software for mechanics........................................................................ 108840.6. Fire simulation software............................................................................................. 1089List of Acronyms.......................................................................................................................... 1093Acronyms for institutions, bodies and groups..................................................................... 1093Technical acronyms and abbreviations.................................................................................. 1103Technical glossary......................................................................................................................... 1121  ContentsPreface............................................................................................................................................. IIIForeword......................................................................................................................................... VIIEditors, Contributors and Reviewers................................................................................... XXXVIIIntroduction................................................................................................................................... XLIIIPart 1General BackgroundChapter 1Biological and Health Effects of Ionizing Radiation – The Radiological Protection System1.1. Biological and health effects of ionizing radiation............................................. 41.1.1. Biological processes................................................................................. 41.1.2. Review of units of measure................................................................... 61.1.3. Natural radioactivity............................................................................... 71.1.4. Health effects............................................................................................ 81.1.4.1. Deterministic effects, tissue reactions......................... 81.1.4.2. Stochastic or random effects.......................................... 91.1.4.3. Induction of diseases other than cancer..................... 111.1.5. Example of the limitations of epidemiology................................... 111.2. Radiological protection system................................................................................ 131.2.1. Types of exposure situations................................................................ 14 1.2.2. Exposure categories................................................................................. 141.2.3. Justification principle............................................................................... 151.2.4. Optimization (ALARA) principle ......................................................... 161.2.5. Principle of application of dose limits............................................... 21Chapter 2Organization of Nuclear Safety Control and Regulation for Nuclear Facilities and Activities in France2.1. From the founding of CEA to the TSN Act........................................................... 232.2. A few definitions........................................................................................................... 282.3. The different contributors to nuclear safety and their missions.................. 302.4. A few basic principles and notions in the field of nuclear safety................. 462.5. Statutory and quasi-statutory frameworks applicable to basic nuclear installations.............................................. 48Chapter 3The International Dimension and the Social Dimension3.1. International dimension.............................................................................................. 773.1.1. Introduction................................................................................................ 773.1.2. IAEA standards........................................................................................... 803.1.3. International Reporting System for Operating Experience (IRS).......................................................... 823.1.4. Services developed by the IAEA........................................................... 843.1.4.1. OSART reviews.................................................................... 843.1.4.2. IRRS reviews......................................................................... 873.1.4.3. Other services and study frameworks set up by the IAEA.............................................................. 883.1.5. WANO.......................................................................................................... 903.1.6. NEA............................................................................................................... 913.1.7. Organizations dedicated to radiation protection and health........................................................................ 933.1.8. From bilateral Franco-German cooperation to European structures for the exchange and capitalization of knowledge and practices, training and assessment services........................................................ 943.1.9. Nuclear regulator associations............................................................. 1003.2. The social dimension................................................................................................... 1023.2.1. Introduction – the context in France................................................. 1023.2.2. Examples of initiatives and issues raised concerning reactor safety in the French nuclear power plant fleet........... 102Chapter 4Nuclear Reactors: Complex Sociotechnical Systems – the Importance of Human and Organizational Factors4.1. 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.............................. 1084.2. The accident at the Chernobyl nuclear power plant and the concept of ‘safety culture’................................. 1094.3. The Fukushima Daiichi nuclear power plant accident: the social dimension and the concept of organization ‘resilience’... 1134.4. Changes in the perception of the role of people in achieving a high level of reliability in complex sociotechnical systems...... 1134.5. Main topics studied in the development of resources and skills pertaining to human and organizational factors............ 1164.5.1. Resources and skills................................................................................. 1164.5.2. Main topics studied.................................................................................. 1184.6. Human and organizational factors in French regulations................................ 119Part 2Safety by DesignChapter 5The Development of Nuclear Power Using Uranium-235 Fission – A Few Notions of Physics Used in Pressurized Water Reactors5.1. Important milestones in the development of nuclear power using fission of the uranium-235 isotope.......................... 1235.2. Fission and important concepts in reactor kinetics........................................... 1275.3. Removing power from the core during operation............................................. 1345.4. Decay heat...................................................................................................................... 1355.5. Main features of pressurized water reactor cores.............................................. 1365.6. Control and monitoring of pressurized water reactor cores.......................... 1375.7. Using uranium and plutonium mixed oxide (MOX) fuel................................. 147Chapter 6General Objectives, Principles and Basic Concepts of the Safety Approach6.1. General approach to risks – General objectives................................................. 1526.2. Fundamental safety functions.................................................................................. 1566.3. Confinement barriers................................................................................................... 1576.4. Defence in depth........................................................................................................... 1616.4. Defence in depth........................................................................................................... 1616.4.1. Levels of defence in depth..................................................................... 1626.4.2. Elements common to the different levels of defence in depth............................................................................. 1696.5. Events considered: terminology adopted for nuclear power reactors......... 1696.6. WENRA reference levels............................................................................................. 1716.7. Deterministic safety analysis and probabilistic safety assessments............ 1716.8. Lessons learned from the accident at the Fukushima Daiichi nuclear power plant on the concept of defence in depth and deterministic analysis......................................................................................... 1736.9. Safety culture – Quality control.............................................................................. 174Chapter 7Safety Options and Considerations at the Design Phase7.1. Different types of design provisions associated with safety considerations................................................. 1887.2. Single-failure criterion................................................................................................. 1897.3. The specific nature of computer-based systems (based on instrumentation and control software)............... 1937.4. Equipment safety classification................................................................................ 1957.4.1. Importance of equipment for safety and safety classification...................................................................... 1957.4.2. Generic requirements associated with the different safety classes.................................................................. 1997.4.3. Qualification of equipment for accident conditions..................... 2027.5. Information on designing nuclear pressure equipment.................................... 2077.6. General considerations on provisions for hazards in facility design............ 2107.7. Anticipating decommissioning in the design stage........................................... 212Chapter 8Study of Operating Conditions in the Deterministic Safety Analysis8.1. Categories of operating conditions......................................................................... 2188.2. Choice of operating conditions................................................................................ 2228.2.1. Concept of ‘bounding’ incident or accident..................................... 2238.2.2. Accident exclusion.................................................................................... 2248.3. List and breakdown of operating conditions....................................................... 2258.4. Methods for studying operating conditions......................................................... 2278.4.1. Choice of initial conditions, conservatism....................................... 2288.4.2. Consideration of an aggravating event in the study on operating conditions – ‘Passive’ failures............ 2298.4.3. Conventional combinations.................................................................. 2318.4.4. Preventing accident aggravation......................................................... 2328.4.5. Operator response time......................................................................... 2328.4.6. Using qualified simulation software................................................... 2338.4.7. Main criteria to be met for fuel in the reactor core..................... 2348.5. Concept of ‘design-basis situations’ for equipment.......................................... 2368.6. Situations to be taken into account in application of pressure equipment regulations................................................. 2378.7. Assessing the radiological consequences of incidents, accidents and hazards.............................................. 2388.7.1. Assessing radioactive substances released from the facility...... 2408.7.2. Assessing radiological consequences of radioactive release from the facility................................................ 2428.7.3. Assessing radiological consequences................................................. 243Chapter 9Loss-of-Coolant Accident9.1. Short- and medium-term aspects of a LOCA...................................................... 2529.1.1. Mechanical effects on vessel internals and fuel assembly structures................................................................ 2539.1.2. Thermal-hydraulic aspects and behaviour of fuel rods................ 2559.1.2.1. Large-break LOCA............................................................... 2559.1.2.2. Intermediate-break LOCA................................................ 2579.1.3. Effects on reactor containment and internals................................ 2589.1.4. Long-term aspect...................................................................................... 2599.2. Safety demonstration.................................................................................................. 2619.2.1. General information and background................................................ 2619.2.2. Fuel assemblies and fuel rods, vessel internals, reactor coolant system components................................................. 2639.2.2.1. Mechanical strength of vessel internals, fuel assembly structures and reactor coolant system components.............. 2639.2.2.2. Fuel behaviour...................................................................... 2659.2.3. Reactor containment and equipment located inside.................... 266Chapter 10A Special Issue: Steam Generator Tubes10.1. Steam generator tube rupture as a Category 3 event...................................... 27010.2. Preventing an SGTR accident, risk of multiple ruptures................................... 27210.3. Steam generator tube rupture(s) studied as a Category 4 event................. 27410.3.1. 900 MWe and 1300 MWe reactors.................................................... 27410.3.2. 1450 MWe reactors and EPR (Flamanville 3).................................. 27410.4. Provisions to mitigate the radiological consequences of SGTR accidents. 276Chapter 11Providing for Hazards: General Considerations and Internal Hazards11.1. General considerations on providing for hazards............................................... 27911.2. Potential projectiles inside the containment....................................................... 28211.3. Effects of pipe breaks.................................................................................................. 28411.4. Projectiles generated by a turbine rotor failure.................................................. 28511.5. Protection against load drops................................................................................... 28811.5.1. Risks related to spent fuel transport packaging............................. 28811.5.2. Other handling risks................................................................................ 29111.6. Fire protection............................................................................................................... 29211.7. Explosion protection.................................................................................................... 29711.8. Internal flooding............................................................................................................ 300Chapter 12Providing for External Hazards12.1. General considerations on providing for external hazards.............................. 30512.2. ‘Climate watch’ implemented by EDF.................................................................... 30712.3. Earthquakes..................................................................................................................... 30712.4. External floods............................................................................................................... 32212.5. Extreme temperatures................................................................................................. 33012.5.1. Extreme cold.............................................................................................. 33012.5.2. Extreme heat.............................................................................................. 33112.6. Possible heat sink hazards.......................................................................................... 33212.7. Other naturally-occurring external hazards......................................................... 33612.8. Accidental aeroplane crashes (excluding malicious acts)................................ 33612.9. Risks related to the industrial environment (excluding malicious acts)..... 340Chapter 13Complementary Domain of Events13.1. The origin of studies belonging to the complementary domain................... 34413.2. Background of the complementary domain........................................................ 34413.3. Analysis of complementary domain events......................................................... 35113.4. ‘New complementary domain’................................................................................. 35113.5. Case of the Flamanville 3 EPR.................................................................................. 354Chapter 14Development and Use of Probabilistic Safety Assessments14.1. History and regulatory context................................................................................ 35714.1.1. International situation............................................................................ 35714.1.2. Situation in France................................................................................... 35914.2. Level 1 PSA...................................................................................................................... 36114.2.1. Scope............................................................................................................ 36114.2.2. Method for carrying out a Level 1 PSA............................................. 36214.2.2.1. General information........................................................... 36214.2.2.2. Specific point: probabilistic human reliability analysis................................................................................... 36414.2.3. Level 1 PSA results and lessons learned............................................ 36914.3. Level 2 PSA...................................................................................................................... 37314.3.1. Scope............................................................................................................ 37314.3.2. Method for carrying out a Level 2 PSA............................................. 37414.3.2.1. General information........................................................... 37414.3.2.2. Probabilistic human reliability analysis for Level 2 PSAs................................................................... 37914.3.3. Examples of lessons learned from Level 2 PSAs............................. 38214.3.3.1. Steam explosion risk assessment.................................. 38214.3.3.2. Mechanical integrity of the 900 MWe reactor containments........................................................ 38314.3.3.3. Isolating penetrations in the containment................. 38314.3.3.4. Modifying the pressure relief system of the reactor coolant system........................................ 38414.3.3.5. Improvement of operating procedures to reduce risk of core melt under pressure................ 38414.3.3.6. Contribution of Level 2 PSAs to emergency response measures............................................................. 38514.4. Expanding the scope of PSA coverage................................................................... 38514.5. Using probabilistic safety assessments.................................................................. 38614.5.1. Using PSAs in the design phase........................................................... 38614.5.1.1. Usefulness and particularities of PSAs in the design phase............................................................. 38614.5.1.2. PSAs conducted to support the Flamanville 3 EPR design............................................................................. 38714.5.2. Using PSAs in periodic reviews............................................................ 38914.5.2.1. Level 1 PSA............................................................................ 39014.5.2.2. Level 2 PSA............................................................................ 39114.5.3. Using PSAs for reactor operation........................................................ 39214.5.3.1. Using PSAs to analyse event severity.......................... 39214.5.3.2. Using PSAs to analyse operational limits and conditions and temporary changes...................... 39414.5.3.3. Using PSAs to analyse operating procedures............. 396Chapter 15Aspects Specific to PWR Spent Fuel Storage Pools15.1. Spent fuel pool design................................................................................................. 39915.1.1. Confinement barriers.............................................................................. 39915.1.2. Initiating events defined at the design stage.................................. 40015.2. Experience feedback..................................................................................................... 40115.2.1. Loss of cooling........................................................................................... 40115.2.1.1. Loss of heat sink.................................................................. 40115.2.1.2. Risks related to maintenance during unit outages................................................................................... 40215.2.1.3. Suction of foreign matter into the cooling system.................................................................................... 40315.2.1.4. Exceeding the decay heat defined in facility design...................................................................................... 40315.2.2. Water losses............................................................................................... 404 15.2.2.1. Gate or sluice gate failures.............................................. 40415.2.2.2. Line-up errors....................................................................... 40515.2.2.3. Failure of a reactor coolant system pipe nozzle dam.......................................................................................... 40915.2.2.4. Rupture of a pipe connected to the spent fuel pool................................................................................. 41115.3. Safety reassessments................................................................................................... 41115.4. Experience feedback from the accident that affected the Unit 4 pool at the Fukushima Daiichi nuclear power plant..... 41415.4.1. Events........................................................................................................... 41415.4.2. Complementary safety assessments conducted in France......... 41715.5. Measures adopted for the EPR.................................................................................. 41915.6. Recommendations for new reactor designs......................................................... 42015.7. New systems for storing spent fuel........................................................................ 423Chapter 16Taking into Account Human and Organizational Factors in Facility Design16.1. Taking into account human and organizational factors in nuclear power reactor design....................................... 42516.1.1. Importance of considering human and organizational factors at the design stage....................................... 42516.1.2. Approach at the design stage............................................................... 43116.1.2.1. Prior to the design phase: analysis of ‘existing elements’........................................ 43216.1.2.2. Design objectives................................................................ 43516.1.2.3. Definition of detailed design provisions...................... 43616.1.2.4. Validation of design provisions...................................... 43916.1.2.5. Assessments conducted during reactor startup and after commissioning.................................................. 44216.1.3. Project management and human and organizational factors engineering programme..................................... 44316.2. Considering human and organizational aspects when designing changes to nuclear power plants..................... 44316.2.1. Importance of human and organizational factors in designing modifications..................................................................... 44416.2.2. ‘Human, social and organizational approach’ implemented by EDF........................................................................... 44516.2.3. Changes, a subject that always deserves special attention from a human and organizational factors perspective.... 44716.3. Human and organizational factors for future nuclear power reactor projects.................................................. 448Chapter 17Studying Core-Melt Accidents to Enhance Safety17.1. Core degradation and vessel failure....................................................................... 45217.1.1. Core uncovery............................................................................................ 45317.1.2. Fuel degradation....................................................................................... 45417.1.3. Failure of the reactor coolant system............................................... 45517.1.4. Phenomena that can cause early containment failure................. 45517.1.5. Phenomena that can ultimately lead to containment failure....................................................... 45717.2. Containment failure modes....................................................................................... 45717.3. Classification of releases associated with core-melt accidents − ‘source terms’................................................. 46017.4. Improving knowledge.................................................................................................. 46217.5. Studies in France on containment failure modes............................................... 46217.5.1. Introduction................................................................................................ 46217.5.2. Initial containment leakage................................................................... 46317.5.3. Direct heating of gases in the containment.................................... 46317.5.4. Hydrogen explosion in the containment.......................................... 46417.5.5. Steam explosion in the vessel or reactor pit................................... 46517.5.6. Gradual pressure increase in the containment............................... 46617.5.7. Penetration of the concrete basemat of the containment by corium............................................ 46717.5.8. ‘U4’ provisions........................................................................................... 46817.5.9. Bypass of containment by outgoing pipes (the V mode).............................................................................................. 46817.5.10. Fast reactivity insertion accidents...................................................... 46917.6. Severe accident operating guidelines..................................................................... 46917.7. Radiological consequences associated with the S3 source term and emergency response plans implemented by public authorities............ 47017.8. Ultimate emergency operating procedures.......................................................... 47317.9. On-site emergency plan............................................................................................. 47317.10. Approach adopted for the EPR................................................................................. 47617.10.1. General safety objectives....................................................................... 47617.10.2. ‘Practical elimination’ of core-melt conditions that could lead to significant early releases............................. 47617.10.3. Provisions for low-pressure core melt............................................... 480Chapter 18New-Generation Reactors18.1. Organization and framework of Franco-German discussions........................ 48518.2. Progression of safety objectives and design options for the EPR project.......................................................... 48618.2.1. General safety objectives....................................................................... 48618.2.2. Events to be taken into account at the design stage and in deterministic and probabilistic analyses.............................. 48818.2.3. Main provisions for preventing incidents and accidents............. 49018.2.4. Functional redundancy, independence between systems, system reliability................................................ 49518.2.5. Confinement preservation..................................................................... 49618.2.6. Radiological protection........................................................................... 49718.2.7. Incorporating lessons learned from the Fukushima Daiichi nuclear power plant accident......................... 49718.3. International context: general safety objectives for new-generation reactors............................................ 49918.4. Concepts highlighted in new reactor designs...................................................... 50118.4.1. AP1000: gravity systems........................................................................ 50118.4.2. VVER: SPOT system................................................................................. 50318.4.3. NM EPR: ‘multi-group’ technology, diversified heat sink............ 504 18.4.4. ATMEA 1: safety injection accumulators in the reactor coolant system.............................................................. 50518.4.5. NuScale: common pool for modular reactors................................. 505 Part 3Safety in OperationChapter 19Startup Tests for Pressurized Water Reactors19.1. Introduction.................................................................................................................... 51119.2. Commissioning.............................................................................................................. 51419.2.1. Defining startup tests.............................................................................. 51419.2.2. Phasing of startup tests.......................................................................... 51519.2.2.1. Preliminary and pre-operational tests......................... 51519.2.2.2. Operational tests................................................................ 51719.2.2.3. General principles for test sequencing and execution....................................................................... 51719.2.3. Documentation for startup tests......................................................... 51819.2.3.1. Integrated system test procedures and startup test procedures............................................ 51819.2.3.2. Test programmes, test procedures, standard test guidelines.................................................... 51819.2.3.3. Completeness analysis, adequacy analysis................. 51819.2.3.4. Acceptance criteria............................................................. 51919.3. Objectives and general rules to take into account for startup tests........... 52019.4. Key lessons learned from startup tests on nuclear power reactors in France............................................................ 52119.4.1. Qualification tests and on-site tests.................................................. 52219.4.2. Long-term on-site testing...................................................................... 52419.4.3. Test configurations and completeness, transpositions................ 525 19.4.4. Safety measures that cannot be verified by testing..................... 52719.4.5. Criteria......................................................................................................... 52719.4.6. Cleanness, keeping system lines clean, foreign matter................ 52819.4.7. Piping support structures and displacement................................... 53219.4.8. Pump and piping vibrations................................................................... 53219.4.9. Validation of operating procedures and periodic tests................ 53419.4.10. Uncertainty and ‘set points’.................................................................. 53519.4.11. Condition of facilities during startup tests...................................... 53619.4.12. Other aspects............................................................................................. 53719.5. Examples of findings resulting from startup tests............................................. 538Chapter 20General Operating Rules20.1. General Operating Rules............................................................................................ 55220.1.1. Content of general operating rules..................................................... 55320.1.2. Limits of general operating rules......................................................... 55420.2. Operational limits and conditions........................................................................... 55420.2.1. Content of operational limits and conditions................................. 55520.2.1.1. Operating modes and standard states......................... 55620.2.1.2. Requirements and unavailability................................... 55820.2.1.3. Fallback states and time required to reach them.... 55820.2.1.4. Events and event groups.................................................. 55920.2.1.5. Combined types of unavailability.................................. 56020.2.1.6. Concepts of ‘boundary condition’ and ‘specific requirement’................................................ 56020.2.2. Average pressure and temperature range of the reactor coolant system.............................................................. 56120.2.3. Changes in operational limits and conditions................................. 56320.3. Initial and periodic tests............................................................................................. 56420.4. Incident and accident operating procedures........................................................ 566Chapter 21 Operating Experience Feedback from Events: Rules and Practices21.1. Background...................................................................................................................... 56921.2. Objectives of an operating experience feedback system................................ 57121.3. Components of an operating experience feedback system – Regulations............................................................................... 57221.4. Operating experience feedback practices adopted for the French nuclear power plant fleet................................. 576Chapter 22Operating Experience from Events Attributable to Shortcomings in Initial Reactor Design or the Quality of Maintenance22.1. 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).......................................... 59022.2. Recurrent loss of safety function events related to maintenance operations – Lessons learned........................... 59522.2.1. Events........................................................................................................... 59522.2.2. General discussion initiated by EDF in the late 1980s on the quality of maintenance operations....................................... 60522.2.3. Applying the defence-in-depth concept when working on a reactor in service.................................. 60822.2.4. Problems that may recur....................................................................... 609Chapter 23Operating Experience from Events Related to Maintenance Operations, Electrical Power Sources and Distribution, Internal and External Hazards23.1. Risks of failure related to equipment or maintenance..................................... 61223.1.1. Risks of common-mode failure............................................................ 61223.1.1.1. Risks of common-mode failure related to settings............................................................................. 61223.1.1.2. Risks of common-mode failure on electrical switchboards.................................................61323.1.1.3. Unavailability of two out of three high-head safety injection lines in the cold legs of the reactor coolant system.............. 61423.1.1.4. Loss of electrical power supplies................................... 61723.1.2. Introduction of non-borated water into the reactor coolant system.......................................................... 62023.1.3. Cooling the reactor coolant system after inhibition of automatic actions............................................................................... 62223.1.4. A temporary device prevents switching the safety injection system to the water recirculation mode................... 62423.2. Events related to internal hazards........................................................................... 62523.2.1. Risk of common-mode failure due to internal flooding.............. 62523.2.2. Risk of failure due to fire....................................................................... 62823.2.3. Risks associated with the use of hydrogen in 900 MWe reactors............................................................................... 63123.3. External hazards: events related to periods of extreme cold......................... 635Chapter 24Enhanced Protection of Estuary and River Sites: Flooding at the Blayais Nuclear Power Plant and Obstruction of a Water Intakeat the Cruas-Meysse Nuclear Power Plant24.1. Partial loss of engineered safety systems following flooding of the Blayais nuclear power plant............................... 64224.2. 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...................................................... 648Chapter 25Taking into Account Human and Organizational Factors in Facility Operation25.1. Skills management....................................................................................................... 65325.1.1. Historical background............................................................................. 65425.1.2. Managing training.................................................................................... 65525.1.3. Strategic workforce planning................................................................ 65625.1.4. Personnel certification............................................................................ 65725.2. Safety and risk management.................................................................................... 65825.2.1. Historical background............................................................................. 65825.2.2. Decision-making and safety.................................................................. 66025.2.3. Risk analyses applied to work activities............................................ 66225.2.4. Operating experience feedback........................................................... 66325.2.5. Managing organizational change......................................................... 66425.3. Managing operational activities............................................................................... 66625.3.1. Characteristics of operational activities............................................ 66625.3.2. Monitoring by the operating crew in the control room.............. 66925.3.3. Compliance with general operating rules......................................... 67025.3.4. Line-up......................................................................................................... 67125.3.5. Operation in extreme situations......................................................... 67225.4. Management of maintenance activities................................................................ 67225.4.1. Management of a scheduled reactor outage for refuellingand maintenance...................................................................................... 67225.4.2. Risks during reactor outages................................................................. 67325.4.3. Preparation for scheduled reactor outages...................................... 67425.4.4. Managing scheduled reactor outages................................................ 67525.5. Supervising outsourced activities............................................................................ 67525.5.1. Contractor qualification and contracting......................................... 67625.5.2. Matching workload and resources...................................................... 67625.5.3. Carrying out work.................................................................................... 67725.5.4. Surveillance of outsourced activities................................................. 67825.5.5. Operating experience feedback and assessingoutsourced activities............................................................................... 678Chapter 26Facility Maintenance26.1. Maintenance objectives.............................................................................................. 68126.2. Maintenance................................................................................................................... 68226.2.1. Definition.................................................................................................... 68226.2.2. Maintenance strategies........................................................................... 68326.3. Optimizing maintenance............................................................................................ 68426.3.1. Reliability-centred maintenance.......................................................... 68426.3.2. Conditional maintenance....................................................................... 686XXVI Elements of nuclear safety – Pressurized water reactors26.3.3. Conditional maintenance by sampling –Maintenance based on reference equipment items...................... 68826.3.4. ‘AP-913’ method....................................................................................... 68926.4. Maintenance baselines................................................................................................ 69126.5. On-site maintenance................................................................................................... 69426.5.1. The various stages of maintenance operations.............................. 69426.5.2. Main conditions for successful maintenance.................................. 69626.5.3. Examples of anomalies or deviations discovered during routinemaintenance, explained by an inadequate maintenancebaseline with regard to deterioration mechanisms....................... 70526.5.4. Examples of events associated with non-qualitymaintenance............................................................................................... 70826.5.4.1. Example of an event explained by an incorrectsetting on redundant equipment................................... 70826.5.4.2. Example of an event explained by an incorrectsetting of electrical protection thresholds................. 70826.5.4.3. Examples of events explained by a failureto return equipment to a compliant stateafter maintenance or an error in performinga work procedure................................................................ 709Chapter 27In-service Monitoring and Inspection of Equipment27.1. Main internal equipment items on a pressurized water reactor vessel...... 71927.1.1. ‘Core baffle’ around the core................................................................ 72027.1.2. RCCA guide tubes..................................................................................... 72227.2. Reactor vessel, nozzles and head............................................................................. 72227.2.1. Vessel underclad defects........................................................................ 72427.2.2. Cracking on vessel head adaptors....................................................... 72527.2.2.1. Condition of other reactors............................................. 72727.2.2.2. Impact on safety................................................................. 72827.2.2.3. Prevention, monitoring and mitigation....................... 72927.2.2.4. Developing inspection tools............................................ 72927.2.2.5. Repairs.................................................................................... 730Contents XXVII27.2.2.6. Leak detection...................................................................... 73027.2.2.7. Anti-ejection devices......................................................... 73127.2.2.8. Current situation................................................................. 73127.2.2.9. Cracks observed on reactor vessel headsin other countries............................................................... 73127.2.2.10. Implementation of special monitoringfor ‘Inconel areas’ beginning in 1992........................... 73227.2.3. Cracking on vessel lower head penetrations detectedin 2011......................................................................................................... 73327.2.4. Monitoring the ‘beltline’ region of the vessel................................. 73527.2.5. Defects observed on reactor vessels in Belgium............................ 73627.3. Steam generators.......................................................................................................... 73827.3.1. The different types of defects.............................................................. 73827.3.2. Associated risks......................................................................................... 74027.3.3. Monitoring during operation and inspection during outages.... 74127.3.3.1. Monitoring during operation........................................... 74127.3.3.2. Inspection during reactor outages................................. 74227.3.4. Steps to be taken when a defect is detected.................................. 74227.3.4.1. Tube wear due to foreign matter.................................. 74327.3.4.2. Wear due to contact with anti-vibration bars.......... 74327.3.4.3. Cracking in U-bend tubes................................................. 74427.3.4.4. Tube deformation and cracking..................................... 74427.3.5. Steam generator replacement.............................................................. 74527.3.6. Clogging observed in the 2000s.......................................................... 74627.3.7. Conclusion.................................................................................................. 74827.4. Steam lines...................................................................................................................... 74827.5. Auxiliary systems: cracks induced by local thermal-hydraulicphenomena..................................................................................................................... 75027.5.1. Cracking in non-isolatable sections connectedto the reactor coolant loops................................................................. 75027.5.2. RHRS thermal fatigue at Unit 1 of the Civaux nuclearpower plant................................................................................................ 75127.6. Civil works: containment structures....................................................................... 75527.6.1. Anticipated degradation phenomena................................................. 75627.6.2. Devices for direct monitoring of containment buildingconcrete walls............................................................................................ 75727.6.3. Leak tests and measurements.............................................................. 75927.6.4. Main anomalies......................................................................................... 759Chapter 28Fuel Management, Monitoring and Developments28.1. Procedures for monitoring fuel rod integrity....................................................... 76828.1.1. Radiochemical specifications for reactor coolant.......................... 76828.1.2. Inspections and measurements carried out directlyon fuel assemblies.................................................................................... 77728.1.2.1. Liquid penetrant testing in the refuellingmachine mast....................................................................... 77828.1.2.2. Liquid penetrant testing in the FB cell......................... 77928.1.2.3. Inspections performed on fuel rods.............................. 78028.2. Operating experience feedback and changes in cladding material.............. 78228.3. Anomalies and significant events involving fuel assemblies.......................... 78628.3.1. Baffle jetting............................................................................................... 78628.3.2. Fretting........................................................................................................ 78728.3.3. Events encountered during handling operations............................ 78928.3.4. Lateral deformation of fuel assemblies interferingwith RCCA drop......................................................................................... 791Chapter 29Facility Compliance29.1. Introduction.................................................................................................................... 79529.2. Detection and treatment of compliance deviations for pressurizedwater reactors in the nuclear power plant fleet................................................. 79629.2.1. Process for handling compliance gaps............................................... 79629.2.2. Examples of compliance gaps............................................................... 79829.2.2.1. Compliance gap in electrical connection boxesqualified for accident conditions................................... 79829.2.2.2. Failure in the seismic resistance of metal floorsin electrical and auxiliary buildings of 900 MWereactors (CPY series).......................................................... 79929.2.2.3. Risk of containment sump screen blockage............... 79929.2.2.4. Anomaly found in engines of emergency and SBOdiesel generators for 900 MWe reactors.................... 80029.2.2.5. Temperature resistance fault in the high-headsafety injection pumps...................................................... 80229.2.2.6. Mixed lubricants in equipment requiredfor accident situations....................................................... 80229.2.2.7. Flow imbalance between safety injection linesof 900 MWe reactors......................................................... 80429.2.2.8. Anomaly in CATHARE software modellingof natural circulation in the upper partof the vessel.......................................................................... 80529.2.2.9. Vibrations and rotor lift on engineeredsafety motor-driven pump units.................................... 805Chapter 30Periodic Reviews30.1. Introduction.................................................................................................................... 80730.2. History of periodic reviews in France for nuclear power reactors................ 80930.2.1. Reactors other than PWRs in the French nuclear powerplant fleet.................................................................................................... 80930.2.2. PWRs in the French nuclear power plant fleet(900 MWe, 1300 MWe and 1450 MWe).......................................... 81130.3. Periodic review process for PWRs in the French nuclear powerplant fleet........................................................................................................................ 81530.3.1. Regulations................................................................................................. 81530.3.2. Outline of a PWR periodic review....................................................... 81730.4. Case of the review associated with the third ten-yearly outageof 900 MWe reactors................................................................................................... 82230.4.1. Plant unit compliance reviews, the complementaryinvestigation and ageing management............................................. 82430.4.1.1. Plant unit compliance reviews........................................ 82430.4.1.2. Complementary investigation programme................ 82530.4.1.3. Ageing management.......................................................... 82630.4.2. Compliance studies on the design of civil works systemsand structures............................................................................................ 82630.4.3. Studies to reassess system design...................................................... 82730.4.4. Reassessment of reactor resistance to internaland external hazards................................................................................ 83030.4.5. Accident studies........................................................................................ 83230.4.6. Taking into account lessons learned during the reviewassociated with the VD3 900 outagefor subsequent reviews........................................................................... 83730.5. Fourth ten-yearly outage of 900 MWe reactors: integratingthe extension of the operating lifetime of nuclear power reactorsin France.......................................................................................................................... 83830.5.1. Background................................................................................................. 83830.5.2. Periodic Review Strategic Plan – Setting objectives..................... 83930.5.3. A few significant issues identified in reviews conductedby safety organizations........................................................................... 84230.6. Overview of international practices – IAEA Guides........................................... 84530.6.1. International practices............................................................................ 84630.6.2. IAEA Guides................................................................................................ 84730.7. Multilateral practices................................................................................................... 848Chapter 31Optimizing Radiation Protection and Limiting Doses Receivedby Workers During Operations in a Nuclear Power Plant31.1. Sources of ionizing radiation in a nuclear power reactor................................ 85231.2. Examples of optimization of worker radiation protection.............................. 85331.3. Arrangements for ‘Major Refit’ operations........................................................... 85531.4. Approach and objectives adopted for the EPR.................................................... 858 Part 4The Accidents at Three Mile Island,Chernobyl and Fukushima Daiichi Nuclear Power Plants,Lessons Learned and Emergency Response ManagementChapter 32The Three Mile Island Nuclear Power Plant Accident32.1. Accident sequence – Reconstitution through simulation................................ 86432.2. Accident consequences............................................................................................... 87232.3. Analysis of the accident causes................................................................................ 87432.3.1. Error in identifying the position of the relief valve....................... 87432.3.2. Understanding the behaviour of the pressurizer............................ 87532.3.3. Stopping safety injection....................................................................... 87632.3.4. Human-machine interface..................................................................... 87632.3.5. Isolating the reactor containment...................................................... 87732.3.6. Confinement inside the auxiliary building....................................... 87732.3.7. Emergency feedwater supply to the steam generators............... 87732.4. Lessons learned from the Three Mile Island accident....................................... 87732.4.1. The human factor in facility operation............................................. 87832.4.2. Importance of precursor events........................................................... 88132.4.3. Study of complex situations and core-melt accidents,handling emergency situations............................................................ 88232.5. Conclusions..................................................................................................................... 883Chapter 33Incident and Accident Operation:from the Event-Oriented Approach to the State-Oriented Approach33.1. Limits of the event-oriented approach.................................................................. 88533.2. The State-Oriented Approach concept.................................................................. 88633.3. First application of the state-oriented approach............................................... 88833.4. Widespread application of the state-oriented approach................................. 89133.5. ‘Stabilized’ state-oriented approach....................................................................... 89233.6. State-oriented approach adopted for the EPR.................................................... 894Chapter 34The Chernobyl Nuclear Power Plant Accident34.1. The Chernobyl nuclear power plant and RBMK reactors................................. 89734.2. The accident sequence................................................................................................ 90134.3. Analysis of the accident causes and changes made to RBMK unitssoon after the accident............................................................................................... 90634.4. The other units at the facility................................................................................... 90834.5. Radioactive release and protection of the population..................................... 90834.5.1. Radioactive release kinetics.................................................................. 90834.5.2. Protection of the population................................................................ 91134.6. Consequences on human health and the environment................................... 91534.6.1. Direct effects of radiation...................................................................... 91534.6.2. Thyroid cancer in children..................................................................... 91734.6.3. Long-term contamination in the Dnieper Basin............................. 91934.7. Radioactive fallout in France and its consequences.......................................... 92134.7.1. Doses attributable to the plume......................................................... 92134.7.2. External doses due to soil deposition................................................ 92134.7.3. Doses due to ingestion of contaminated foodstuffs.................... 92234.7.4. Overall levels.............................................................................................. 92334.7.5. Thyroid cancer........................................................................................... 92434.7.5.1. Monitoring thyroid cancer in France............................ 92534.7.5.2. Assessment of the number of cancer casesinduced in France by the Chernobyl accident........... 92634.8. Lessons learned by the international communityfrom a general viewpoint and with regard to RBMK reactors....................... 92734.9. Lessons learned in France........................................................................................... 92834.10. Keeping the public informed..................................................................................... 93134.11. After the Chernobyl accident.................................................................................... 933Chapter 35Options and Control of Reactivity Insertionin Pressurized Water Reactors35.1. Research and study of event sequences................................................................ 93735.1.1. Cooling accidents..................................................................................... 93835.1.2. Incidents and accidents related to rod clustercontrol assemblies.................................................................................... 94035.1.3. Boron dilution accidents........................................................................ 94335.1.4. Inserting a cold water plug in the core............................................. 95035.2. Changes in criteria........................................................................................................ 95135.3. The case of outage states.......................................................................................... 95335.4. Regulations...................................................................................................................... 957Chapter 36The Reactor Accident at the Fukushima Daiichi Nuclear Power Plantand Lessons Learned in France36.1. Reactor units at the Fukushima Daiichi nuclear power plant........................ 96236.1.1. General operation of a boiling water reactor.................................. 96236.1.2. Containment.............................................................................................. 96236.1.3. Emergency cooling systems.................................................................. 96436.2. Sequence of events during the accident............................................................... 96636.3. Radioactive release....................................................................................................... 97236.3.1. Airborne radioactive release, residual caesium depositsand contamination of foodstuffs........................................................ 97236.3.1.1. Airborne radioactive release............................................ 97236.3.1.2. Persistent caesium deposits............................................ 97236.3.1.3. Contamination of foodstuffs.......................................... 97336.3.2. Release of radioactive substances in the Pacific Ocean.............. 97536.3.3. Long-distance atmospheric dispersionof the radioactive plume........................................................................ 97636.4. Action taken to control facilities and released contaminated water.......... 97836.5. Socioeconomic and health impact in numbers................................................... 98136.5.1. Socioeconomic impact............................................................................ 98136.5.2. Health impact............................................................................................ 98236.6. Lessons learned from the accident.......................................................................... 98536.6.1. Complementary safety assessments carried out in Europeand France following the Fukushima Daiichi nuclear powerplant accident............................................................................................ 98636.6.2. Complementary safety assessments carried out in France........ 98836.6.3. Procedure for complementary safety assessmentscarried out in France................................................................................ 98936.6.4. Conclusions of the complementary safety assessmentscarried out in France................................................................................ 99036.6.5. The ‘hardened safety core’.................................................................... 99136.6.5.1. Purpose................................................................................... 99136.6.5.2. Principles................................................................................ 99136.6.5.3. Illustrations........................................................................... 99336.6.6. Nuclear Rapid Response Force (FARN).............................................. 99536.6.7. Deployment of post-Fukushima measuresin French nuclear power plants............................................................ 99736.7. Other lessons learned in France from the Fukushima Daiichi nuclearpower plant accident................................................................................................... 998Chapter 37Lessons Learned from the Fukushima Daiichi Nuclear PowerPlant Accident: Work Conducted by the IAEA and WENRA,Action Taken in Countries Other than France37.1. Work conducted by the IAEA.................................................................................... 100237.2. Work conducted by WENRA..................................................................................... 100337.3. Japan................................................................................................................................. 100437.4. Belgium............................................................................................................................ 100637.4.1. Nuclear power plants in Belgium........................................................ 100637.4.2. General details on the design of Belgian nuclear powerplants............................................................................................................ 100637.4.3. Stress tests and main lessons learned............................................... 100737.4.3.1. Improving protection of facilities againstexternal hazards.................................................................. 100737.4.3.2. Improving protection of facilities against lossof electrical power supplies or loss of heat sink....... 1010Contents XXXV37.4.3.3. Improving on-site emergency plans............................. 101137.4.3.4. Improving management of core-melt accidents...... 101237.5. USA.................................................................................................................................... 1013Chapter 38Emergency Preparedness and Response38.1. Defining a radiological emergency and ‘response’ objectives........................ 102138.2. General organization of radiological emergency management..................... 102338.2.1. Organization and entities concerned................................................. 102338.2.2. Major Nuclear or Radiological Accident National ResponsePlan and emergency plans..................................................................... 102538.2.2.1. Major Nuclear or Radiological Accident NationalResponse Plan...................................................................... 102538.2.2.2. Emergency plans................................................................. 102638.2.2.3. Provisions for protecting the public in the eventof an accidental release of radioactivity..................... 102838.3. Management by the operator................................................................................... 102938.4. Prefectural authorities and mayors......................................................................... 103138.5. ASN, the Nuclear Safety Authority......................................................................... 103238.6. IRSN................................................................................................................................... 103338.7. Assessment approach in the event of an accident affecting a reactorin the nuclear power plant fleet.............................................................................. 103838.7.1. ‘3D/3P’ method......................................................................................... 103938.7.2. The ‘aggravated prognosis’ approach................................................ 104138.7.3. Extending the 3D/3P method to severe accidents(the ‘D/P AG’ method)............................................................................ 104138.8. Emergency preparedness............................................................................................ 104238.8.1. Emergency response exercises............................................................. 104338.8.2. Operating experience feedback........................................................... 1045 Part 5PWR Safety Studies, R&D and Simulation SoftwareChapter 39PWR Safety Studies and R&D39.1. Contribution of studies to the improvement of pressurized waterreactor safety................................................................................................................. 105039.2. Purpose and overview of R&D work, dedicated programmesand organizations involved, and research facilities in France........................ 105239.2.1. Purpose and overview of R&D............................................................. 105239.2.2. Dedicated frameworks and organizations involved....................... 106339.2.3. Facilities in France used for research and development.............. 1065Chapter 40Examples of Simulation Software Developed for Safety Analysisof Pressurized Water Reactors40.1. Simulation software for neutronics........................................................................ 107640.2. Simulation software for thermal hydraulics (and mechanics)....................... 107940.3. Simulation software for thermal mechanics........................................................ 108440.4. Software for simulating core-melt situations..................................................... 108540.5. Simulation software for mechanics........................................................................ 108840.6. Fire simulation software............................................................................................. 1089List of Acronyms.......................................................................................................................... 1093Acronyms for institutions, bodies and groups..................................................................... 1093Technical acronyms and abbreviations.................................................................................. 1103Technical glossary......................................................................................................................... 1121 21 00 06 01 https://laboutique.edpsciences.fr/produit/1280/9782759827237/elements-of-nuclear-safety 01 00 03 02 https://laboutique.edpsciences.fr/system/product_pictures/data/009/982/690/original/9782759827220-PressurizedWaterReactors_couv-sofedis.jpg 17 14 20240913T180957+0200 01 P1 EDP Sciences 01 01 P1 EDP Sciences 17 avenue du Hoggar - PA de Courtaboeuf - 91944 Les Ulis cedex A http://publications.edpsciences.org/ 04 11 00 20220707 01 00 20220825 19 00 20220825 2026 06 3052868830012 01 WORLD 27 03 9782759827237 15 9782759827237 27 03 9782759827244 15 9782759827244 WORLD 08 EDP Sciences 04 01 00 20220825 03 01 D1 EDP Sciences 21 04 05 01 02 1 00 300.00 01 5.50 15.64 EUR laboutique.edpsciences.fr-R001719 03 01 01 R001719 00 ED E107 00 01 01 ELEMENTS OF NUCLEAR SAFETY PRESSURIZED WATER REACTORS 01 eng 08 1173 03 22 89009587 17 01 P1 EDP Sciences 01 01 P1 EDP Sciences 17 avenue du Hoggar - PA de Courtaboeuf - 91944 Les Ulis cedex A http://publications.edpsciences.org/ 11 00 20220707 02 01 002109 03 9782759827237 15 9782759827237 03 01 D1 EDP Sciences 29 https://laboutique.edpsciences.fr/open_access_download/2109/1719 45 03 laboutique.edpsciences.fr-002109 03 01 01 002109 03 9782759827237 15 9782759827237 10 EA E107 10 00 01 R001719 ED E107 1 10 01 02 Institut de Radioprotection et de Sûreté Nucléaire 01 01 ELEMENTS OF NUCLEAR SAFETY PRESSURIZED WATER REACTORS 1 B15 01 A1684 Jean Couturier Couturier, Jean Jean Couturier 01 eng 08 1168 03 01 24 Izibook:Subject Radioprotection 24 Izibook:SubjectAndCategoryAndTags |Radioprotection| 20 pressurized water reactor safety;nuclear power;nuclear power plant;safety;reactor;radiological emergency;nuclear emergency;safety analysis method;nuclear safety;nuclear risk control;radiological risk control 10 2011 BISACSCI058000 01 620 06 06 05 03 00 Everything that is important to know about pressurized water reactor safety in nuclear power plants is compiled in this work of reference, from safety fundamentals and reactor design to provisions for managing a radiological or nuclear emergency. The book narrates the determined and continuous quest to enhance nuclear safety. Jean Couturier, coordinator and senior editor, covers the forty-year history of evolution in the safety objectives, approaches, analysis methods and assessment criteria that have defined pressurized water reactor safety, mainly within the French nuclear power plant fleet, from the 1970s up to today’s Flamanville 3 EPR. Everyone, including current and future generations of engineers, researchers and, more broadly, any citizen interested in nuclear safety issues will find that the book reinforces their knowledge on this important subject, providing an understanding of the fundamentals, to the benefit of nuclear and radiological risk control. 02 00 Everything that is important to know about pressurized water reactorsafety in nuclear power plants is compiled in this work of reference, fromsafety fundamentals and reactor design to provisions for managing aradiological or nuclear emergency. 04 00 ContentsPreface............................................................................................................................................. IIIForeword......................................................................................................................................... VIIEditors, Contributors and Reviewers................................................................................... XXXVIIIntroduction................................................................................................................................... XLIIIPart 1General BackgroundChapter 1Biological and Health Effects of Ionizing Radiation – The Radiological Protection System1.1. Biological and health effects of ionizing radiation............................................. 41.1.1. Biological processes................................................................................. 41.1.2. Review of units of measure................................................................... 61.1.3. Natural radioactivity............................................................................... 71.1.4. Health effects............................................................................................ 81.1.4.1. Deterministic effects, tissue reactions......................... 81.1.4.2. Stochastic or random effects.......................................... 91.1.4.3. Induction of diseases other than cancer..................... 111.1.5. Example of the limitations of epidemiology................................... 111.2. Radiological protection system................................................................................ 131.2.1. Types of exposure situations................................................................ 14 1.2.2. Exposure categories................................................................................. 141.2.3. Justification principle............................................................................... 151.2.4. Optimization (ALARA) principle ......................................................... 161.2.5. Principle of application of dose limits............................................... 21Chapter 2Organization of Nuclear Safety Control and Regulation for Nuclear Facilities and Activities in France2.1. From the founding of CEA to the TSN Act........................................................... 232.2. A few definitions........................................................................................................... 282.3. The different contributors to nuclear safety and their missions.................. 302.4. A few basic principles and notions in the field of nuclear safety................. 462.5. Statutory and quasi-statutory frameworks applicable to basic nuclear installations.............................................. 48Chapter 3The International Dimension and the Social Dimension3.1. International dimension.............................................................................................. 773.1.1. Introduction................................................................................................ 773.1.2. IAEA standards........................................................................................... 803.1.3. International Reporting System for Operating Experience (IRS).......................................................... 823.1.4. Services developed by the IAEA........................................................... 843.1.4.1. OSART reviews.................................................................... 843.1.4.2. IRRS reviews......................................................................... 873.1.4.3. Other services and study frameworks set up by the IAEA.............................................................. 883.1.5. WANO.......................................................................................................... 903.1.6. NEA............................................................................................................... 913.1.7. Organizations dedicated to radiation protection and health........................................................................ 933.1.8. From bilateral Franco-German cooperation to European structures for the exchange and capitalization of knowledge and practices, training and assessment services........................................................ 943.1.9. Nuclear regulator associations............................................................. 1003.2. The social dimension................................................................................................... 1023.2.1. Introduction – the context in France................................................. 1023.2.2. Examples of initiatives and issues raised concerning reactor safety in the French nuclear power plant fleet........... 102Chapter 4Nuclear Reactors: Complex Sociotechnical Systems – the Importance of Human and Organizational Factors4.1. 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.............................. 1084.2. The accident at the Chernobyl nuclear power plant and the concept of ‘safety culture’................................. 1094.3. The Fukushima Daiichi nuclear power plant accident: the social dimension and the concept of organization ‘resilience’... 1134.4. Changes in the perception of the role of people in achieving a high level of reliability in complex sociotechnical systems...... 1134.5. Main topics studied in the development of resources and skills pertaining to human and organizational factors............ 1164.5.1. Resources and skills................................................................................. 1164.5.2. Main topics studied.................................................................................. 1184.6. Human and organizational factors in French regulations................................ 119 Part 2Safety by DesignChapter 5The Development of Nuclear Power Using Uranium-235 Fission – A Few Notions of Physics Used in Pressurized Water Reactors5.1. Important milestones in the development of nuclear power using fission of the uranium-235 isotope.......................... 1235.2. Fission and important concepts in reactor kinetics........................................... 1275.3. Removing power from the core during operation............................................. 1345.4. Decay heat...................................................................................................................... 1355.5. Main features of pressurized water reactor cores.............................................. 1365.6. Control and monitoring of pressurized water reactor cores.......................... 1375.7. Using uranium and plutonium mixed oxide (MOX) fuel................................. 147Chapter 6General Objectives, Principles and Basic Concepts of the Safety Approach6.1. General approach to risks – General objectives................................................. 1526.2. Fundamental safety functions.................................................................................. 1566.3. Confinement barriers................................................................................................... 1576.4. Defence in depth........................................................................................................... 1616.4.1. Levels of defence in depth..................................................................... 1626.4.2. Elements common to the different levels of defence in depth............................................................................. 1696.5. Events considered: terminology adopted for nuclear power reactors......... 1696.6. WENRA reference levels............................................................................................. 1716.7. Deterministic safety analysis and probabilistic safety assessments............ 1716.8. Lessons learned from the accident at the Fukushima Daiichi nuclear power plant on the concept of defence in depth and deterministic analysis......................................................................................... 1736.9. Safety culture – Quality control.............................................................................. 174Chapter 7Safety Options and Considerations at the Design Phase7.1. Different types of design provisions associated with safety considerations................................................. 1887.2. Single-failure criterion................................................................................................. 1897.3. The specific nature of computer-based systems (based on instrumentation and control software)............... 1937.4. Equipment safety classification................................................................................ 1957.4.1. Importance of equipment for safety and safety classification...................................................................... 1957.4.2. Generic requirements associated with the different safety classes.................................................................. 1997.4.3. Qualification of equipment for accident conditions..................... 2027.5. Information on designing nuclear pressure equipment.................................... 2077.6. General considerations on provisions for hazards in facility design............ 2107.7. Anticipating decommissioning in the design stage........................................... 212Chapter 8Study of Operating Conditions in the Deterministic Safety Analysis8.1. Categories of operating conditions......................................................................... 2188.2. Choice of operating conditions................................................................................ 2228.2.1. Concept of ‘bounding’ incident or accident..................................... 2238.2.2. Accident exclusion.................................................................................... 2248.3. List and breakdown of operating conditions....................................................... 2258.4. Methods for studying operating conditions......................................................... 2278.4.1. Choice of initial conditions, conservatism....................................... 2288.4.2. Consideration of an aggravating event in the study on operating conditions – ‘Passive’ failures............ 2298.4.3. Conventional combinations.................................................................. 2318.4.4. Preventing accident aggravation......................................................... 2328.4.5. Operator response time......................................................................... 2328.4.6. Using qualified simulation software................................................... 2338.4.7. Main criteria to be met for fuel in the reactor core..................... 2348.5. Concept of ‘design-basis situations’ for equipment.......................................... 2368.6. Situations to be taken into account in application of pressure equipment regulations................................................. 2378.7. Assessing the radiological consequences of incidents, accidents and hazards.............................................. 2388.7.1. Assessing radioactive substances released from the facility...... 2408.7.2. Assessing radiological consequences of radioactive release from the facility................................................ 2428.7.3. Assessing radiological consequences................................................. 243Chapter 9Loss-of-Coolant Accident9.1. Short- and medium-term aspects of a LOCA...................................................... 2529.1.1. Mechanical effects on vessel internals and fuel assembly structures................................................................ 2539.1.2. Thermal-hydraulic aspects and behaviour of fuel rods................ 2559.1.2.1. Large-break LOCA............................................................... 2559.1.2.2. Intermediate-break LOCA................................................ 2579.1.3. Effects on reactor containment and internals................................ 2589.1.4. Long-term aspect...................................................................................... 2599.2. Safety demonstration.................................................................................................. 2619.2.1. General information and background................................................ 2619.2.2. Fuel assemblies and fuel rods, vessel internals, reactor coolant system components................................................. 2639.2.2.1. Mechanical strength of vessel internals, fuel assembly structures and reactor coolant system components.............. 2639.2.2.2. Fuel behaviour...................................................................... 2659.2.3. Reactor containment and equipment located inside.................... 266Chapter 10A Special Issue: Steam Generator Tubes10.1. Steam generator tube rupture as a Category 3 event...................................... 27010.2. Preventing an SGTR accident, risk of multiple ruptures................................... 27210.3. Steam generator tube rupture(s) studied as a Category 4 event................. 27410.3.1. 900 MWe and 1300 MWe reactors.................................................... 27410.3.2. 1450 MWe reactors and EPR (Flamanville 3).................................. 27410.4. Provisions to mitigate the radiological consequences of SGTR accidents. 276Chapter 11Providing for Hazards: General Considerations and Internal Hazards11.1. General considerations on providing for hazards............................................... 27911.2. Potential projectiles inside the containment....................................................... 28211.3. Effects of pipe breaks.................................................................................................. 28411.4. Projectiles generated by a turbine rotor failure.................................................. 28511.5. Protection against load drops................................................................................... 28811.5.1. Risks related to spent fuel transport packaging............................. 28811.5.2. Other handling risks................................................................................ 29111.6. Fire protection............................................................................................................... 29211.7. Explosion protection.................................................................................................... 29711.8. Internal flooding............................................................................................................ 300Chapter 12Providing for External Hazards12.1. General considerations on providing for external hazards.............................. 30512.2. ‘Climate watch’ implemented by EDF.................................................................... 30712.3. Earthquakes..................................................................................................................... 30712.4. External floods............................................................................................................... 32212.5. Extreme temperatures................................................................................................. 33012.5.1. Extreme cold.............................................................................................. 33012.5.2. Extreme heat.............................................................................................. 33112.6. Possible heat sink hazards.......................................................................................... 33212.7. Other naturally-occurring external hazards......................................................... 33612.8. Accidental aeroplane crashes (excluding malicious acts)................................ 33612.9. Risks related to the industrial environment (excluding malicious acts)..... 340Chapter 13Complementary Domain of Events13.1. The origin of studies belonging to the complementary domain................... 34413.2. Background of the complementary domain........................................................ 34413.3. Analysis of complementary domain events......................................................... 35113.4. ‘New complementary domain’................................................................................. 35113.5. Case of the Flamanville 3 EPR.................................................................................. 354Chapter 14Development and Use of Probabilistic Safety Assessments14.1. History and regulatory context................................................................................ 35714.1.1. International situation............................................................................ 35714.1.2. Situation in France................................................................................... 35914.2. Level 1 PSA...................................................................................................................... 36114.2.1. Scope............................................................................................................ 36114.2.2. Method for carrying out a Level 1 PSA............................................. 36214.2.2.1. General information........................................................... 36214.2.2.2. Specific point: probabilistic human reliability analysis................................................................................... 36414.2.3. Level 1 PSA results and lessons learned............................................ 36914.3. Level 2 PSA...................................................................................................................... 37314.3.1. Scope............................................................................................................ 37314.3.2. Method for carrying out a Level 2 PSA............................................. 37414.3.2.1. General information........................................................... 37414.3.2.2. Probabilistic human reliability analysis for Level 2 PSAs................................................................... 37914.3.3. Examples of lessons learned from Level 2 PSAs............................. 38214.3.3.1. Steam explosion risk assessment.................................. 38214.3.3.2. Mechanical integrity of the 900 MWe reactor containments........................................................ 38314.3.3.3. Isolating penetrations in the containment................. 38314.3.3.4. Modifying the pressure relief system of the reactor coolant system........................................ 38414.3.3.5. Improvement of operating procedures to reduce risk of core melt under pressure................ 38414.3.3.6. Contribution of Level 2 PSAs to emergency response measures............................................................. 38514.4. Expanding the scope of PSA coverage................................................................... 38514.5. Using probabilistic safety assessments.................................................................. 38614.5.1. Using PSAs in the design phase........................................................... 38614.5.1.1. Usefulness and particularities of PSAs in the design phase............................................................. 38614.5.1.2. PSAs conducted to support the Flamanville 3 EPR design............................................................................. 38714.5.2. Using PSAs in periodic reviews............................................................ 38914.5.2.1. Level 1 PSA............................................................................ 39014.5.2.2. Level 2 PSA............................................................................ 39114.5.3. Using PSAs for reactor operation........................................................ 39214.5.3.1. Using PSAs to analyse event severity.......................... 39214.5.3.2. Using PSAs to analyse operational limits and conditions and temporary changes...................... 39414.5.3.3. Using PSAs to analyse operating procedures............. 396Chapter 15Aspects Specific to PWR Spent Fuel Storage Pools15.1. Spent fuel pool design................................................................................................. 39915.1.1. Confinement barriers.............................................................................. 39915.1.2. Initiating events defined at the design stage.................................. 40015.2. Experience feedback..................................................................................................... 40115.2.1. Loss of cooling........................................................................................... 40115.2.1.1. Loss of heat sink.................................................................. 40115.2.1.2. Risks related to maintenance during unit outages................................................................................... 40215.2.1.3. Suction of foreign matter into the cooling system.................................................................................... 40315.2.1.4. Exceeding the decay heat defined in facility design...................................................................................... 40315.2.2. Water losses............................................................................................... 404 15.2.2.1. Gate or sluice gate failures.............................................. 40415.2.2.2. Line-up errors....................................................................... 40515.2.2.3. Failure of a reactor coolant system pipe nozzle dam.......................................................................................... 40915.2.2.4. Rupture of a pipe connected to the spent fuel pool................................................................................. 41115.3. Safety reassessments................................................................................................... 41115.4. Experience feedback from the accident that affected the Unit 4 pool at the Fukushima Daiichi nuclear power plant..... 41415.4.1. Events........................................................................................................... 41415.4.2. Complementary safety assessments conducted in France......... 41715.5. Measures adopted for the EPR.................................................................................. 41915.6. Recommendations for new reactor designs......................................................... 42015.7. New systems for storing spent fuel........................................................................ 423Chapter 16Taking into Account Human and Organizational Factors in Facility Design16.1. Taking into account human and organizational factors in nuclear power reactor design....................................... 42516.1.1. Importance of considering human and organizational factors at the design stage....................................... 42516.1.2. Approach at the design stage............................................................... 43116.1.2.1. Prior to the design phase: analysis of ‘existing elements’........................................ 43216.1.2.2. Design objectives................................................................ 43516.1.2.3. Definition of detailed design provisions...................... 43616.1.2.4. Validation of design provisions...................................... 43916.1.2.5. Assessments conducted during reactor startup and after commissioning.................................................. 44216.1.3. Project management and human and organizational factors engineering programme..................................... 44316.2. Considering human and organizational aspects when designing changes to nuclear power plants..................... 44316.2.1. Importance of human and organizational factors in designing modifications..................................................................... 44416.2.2. ‘Human, social and organizational approach’ implemented by EDF........................................................................... 44516.2.3. Changes, a subject that always deserves special attention from a human and organizational factors perspective.... 44716.3. Human and organizational factors for future nuclear power reactor projects.................................................. 448Chapter 17Studying Core-Melt Accidents to Enhance Safety17.1. Core degradation and vessel failure....................................................................... 45217.1.1. Core uncovery............................................................................................ 45317.1.2. Fuel degradation....................................................................................... 45417.1.3. Failure of the reactor coolant system............................................... 45517.1.4. Phenomena that can cause early containment failure................. 45517.1.5. Phenomena that can ultimately lead to containment failure....................................................... 45717.2. Containment failure modes....................................................................................... 45717.3. Classification of releases associated with core-melt accidents − ‘source terms’................................................. 46017.4. Improving knowledge.................................................................................................. 46217.5. Studies in France on containment failure modes............................................... 46217.5.1. Introduction................................................................................................ 46217.5.2. Initial containment leakage................................................................... 46317.5.3. Direct heating of gases in the containment.................................... 46317.5.4. Hydrogen explosion in the containment.......................................... 46417.5.5. Steam explosion in the vessel or reactor pit................................... 46517.5.6. Gradual pressure increase in the containment............................... 46617.5.7. Penetration of the concrete basemat of the containment by corium............................................ 46717.5.8. ‘U4’ provisions........................................................................................... 46817.5.9. Bypass of containment by outgoing pipes (the V mode).............................................................................................. 46817.5.10. Fast reactivity insertion accidents...................................................... 46917.6. Severe accident operating guidelines..................................................................... 46917.7. Radiological consequences associated with the S3 source term and emergency response plans implemented by public authorities............ 47017.8. Ultimate emergency operating procedures.......................................................... 47317.9. On-site emergency plan............................................................................................. 47317.10. Approach adopted for the EPR................................................................................. 47617.10.1. General safety objectives....................................................................... 47617.10.2. ‘Practical elimination’ of core-melt conditions that could lead to significant early releases............................. 47617.10.3. Provisions for low-pressure core melt............................................... 480Chapter 18New-Generation Reactors18.1. Organization and framework of Franco-German discussions........................ 48518.2. Progression of safety objectives and design options for the EPR project.......................................................... 48618.2.1. General safety objectives....................................................................... 48618.2.2. Events to be taken into account at the design stage and in deterministic and probabilistic analyses.............................. 48818.2.3. Main provisions for preventing incidents and accidents............. 49018.2.4. Functional redundancy, independence between systems, system reliability................................................ 49518.2.5. Confinement preservation..................................................................... 49618.2.6. Radiological protection........................................................................... 49718.2.7. Incorporating lessons learned from the Fukushima Daiichi nuclear power plant accident......................... 49718.3. International context: general safety objectives for new-generation reactors............................................ 49918.4. Concepts highlighted in new reactor designs...................................................... 50118.4.1. AP1000: gravity systems........................................................................ 50118.4.2. VVER: SPOT system................................................................................. 50318.4.3. NM EPR: ‘multi-group’ technology, diversified heat sink............ 504 18.4.4. ATMEA 1: safety injection accumulators in the reactor coolant system.............................................................. 50518.4.5. NuScale: common pool for modular reactors................................. 505 Part 3Safety in OperationChapter 19Startup Tests for Pressurized Water Reactors19.1. Introduction.................................................................................................................... 51119.2. Commissioning.............................................................................................................. 51419.2.1. Defining startup tests.............................................................................. 51419.2.2. Phasing of startup tests.......................................................................... 51519.2.2.1. Preliminary and pre-operational tests......................... 51519.2.2.2. Operational tests................................................................ 51719.2.2.3. General principles for test sequencing and execution....................................................................... 51719.2.3. Documentation for startup tests......................................................... 51819.2.3.1. Integrated system test procedures and startup test procedures............................................ 51819.2.3.2. Test programmes, test procedures, standard test guidelines.................................................... 51819.2.3.3. Completeness analysis, adequacy analysis................. 51819.2.3.4. Acceptance criteria............................................................. 51919.3. Objectives and general rules to take into account for startup tests........... 52019.4. Key lessons learned from startup tests on nuclear power reactors in France............................................................ 52119.4.1. Qualification tests and on-site tests.................................................. 52219.4.2. Long-term on-site testing...................................................................... 52419.4.3. Test configurations and completeness, transpositions................ 525 19.4.4. Safety measures that cannot be verified by testing..................... 52719.4.5. Criteria......................................................................................................... 52719.4.6. Cleanness, keeping system lines clean, foreign matter................ 52819.4.7. Piping support structures and displacement................................... 53219.4.8. Pump and piping vibrations................................................................... 53219.4.9. Validation of operating procedures and periodic tests................ 53419.4.10. Uncertainty and ‘set points’.................................................................. 53519.4.11. Condition of facilities during startup tests...................................... 53619.4.12. Other aspects............................................................................................. 53719.5. Examples of findings resulting from startup tests............................................. 538Chapter 20General Operating Rules20.1. General Operating Rules............................................................................................ 55220.1.1. Content of general operating rules..................................................... 55320.1.2. Limits of general operating rules......................................................... 55420.2. Operational limits and conditions........................................................................... 55420.2.1. Content of operational limits and conditions................................. 55520.2.1.1. Operating modes and standard states......................... 55620.2.1.2. Requirements and unavailability................................... 55820.2.1.3. Fallback states and time required to reach them.... 55820.2.1.4. Events and event groups.................................................. 55920.2.1.5. Combined types of unavailability.................................. 56020.2.1.6. Concepts of ‘boundary condition’ and ‘specific requirement’................................................ 56020.2.2. Average pressure and temperature range of the reactor coolant system.............................................................. 56120.2.3. Changes in operational limits and conditions................................. 56320.3. Initial and periodic tests............................................................................................. 56420.4. Incident and accident operating procedures........................................................ 566Chapter 21 Operating Experience Feedback from Events: Rules and Practices21.1. Background...................................................................................................................... 56921.2. Objectives of an operating experience feedback system................................ 57121.3. Components of an operating experience feedback system – Regulations............................................................................... 57221.4. Operating experience feedback practices adopted for the French nuclear power plant fleet................................. 576Chapter 22Operating Experience from Events Attributable to Shortcomings in Initial Reactor Design or the Quality of Maintenance22.1. 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).......................................... 59022.2. Recurrent loss of safety function events related to maintenance operations – Lessons learned........................... 59522.2.1. Events........................................................................................................... 59522.2.2. General discussion initiated by EDF in the late 1980s on the quality of maintenance operations....................................... 60522.2.3. Applying the defence-in-depth concept when working on a reactor in service.................................. 60822.2.4. Problems that may recur....................................................................... 609Chapter 23Operating Experience from Events Related to Maintenance Operations, Electrical Power Sources and Distribution, Internal and External Hazards23.1. Risks of failure related to equipment or maintenance..................................... 61223.1.1. Risks of common-mode failure............................................................ 61223.1.1.1. Risks of common-mode failure related to settings............................................................................. 61223.1.1.2. Risks of common-mode failure on electrical switchboards.................................................61323.1.1.3. Unavailability of two out of three high-head safety injection lines in the cold legs of the reactor coolant system.............. 61423.1.1.4. Loss of electrical power supplies................................... 61723.1.2. Introduction of non-borated water into the reactor coolant system.......................................................... 62023.1.3. Cooling the reactor coolant system after inhibition of automatic actions............................................................................... 62223.1.4. A temporary device prevents switching the safety injection system to the water recirculation mode................... 62423.2. Events related to internal hazards........................................................................... 62523.2.1. Risk of common-mode failure due to internal flooding.............. 62523.2.2. Risk of failure due to fire....................................................................... 62823.2.3. Risks associated with the use of hydrogen in 900 MWe reactors............................................................................... 63123.3. External hazards: events related to periods of extreme cold......................... 635Chapter 24Enhanced Protection of Estuary and River Sites: Flooding at the Blayais Nuclear Power Plant and Obstruction of a Water Intakeat the Cruas-Meysse Nuclear Power Plant24.1. Partial loss of engineered safety systems following flooding of the Blayais nuclear power plant............................... 64224.2. 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...................................................... 648Chapter 25Taking into Account Human and Organizational Factors in Facility Operation25.1. Skills management....................................................................................................... 65325.1.1. Historical background............................................................................. 65425.1.2. Managing training.................................................................................... 65525.1.3. Strategic workforce planning................................................................ 65625.1.4. Personnel certification............................................................................ 65725.2. Safety and risk management.................................................................................... 65825.2.1. Historical background............................................................................. 65825.2.2. Decision-making and safety.................................................................. 66025.2.3. Risk analyses applied to work activities............................................ 66225.2.4. Operating experience feedback........................................................... 66325.2.5. Managing organizational change......................................................... 66425.3. Managing operational activities............................................................................... 66625.3.1. Characteristics of operational activities............................................ 66625.3.2. Monitoring by the operating crew in the control room.............. 66925.3.3. Compliance with general operating rules......................................... 67025.3.4. Line-up......................................................................................................... 67125.3.5. Operation in extreme situations......................................................... 67225.4. Management of maintenance activities................................................................ 67225.4.1. Management of a scheduled reactor outage for refuelling and maintenance...................................................... 67225.4.2. Risks during reactor outages................................................................. 67325.4.3. Preparation for scheduled reactor outages...................................... 67425.4.4. Managing scheduled reactor outages................................................ 67525.5. Supervising outsourced activities............................................................................ 67525.5.1. Contractor qualification and contracting......................................... 67625.5.2. Matching workload and resources...................................................... 67625.5.3. Carrying out work.................................................................................... 67725.5.4. Surveillance of outsourced activities................................................. 67825.5.5. Operating experience feedback and assessing outsourced activities...................... 678Chapter 26Facility Maintenance26.1. Maintenance objectives.............................................................................................. 68126.2. Maintenance................................................................................................................... 68226.2.1. Definition.................................................................................................... 68226.2.2. Maintenance strategies........................................................................... 68326.3. Optimizing maintenance............................................................................................ 68426.3.1. Reliability-centred maintenance.......................................................... 68426.3.2. Conditional maintenance....................................................................... 68626.3.3. Conditional maintenance by sampling – Maintenance based on reference equipment items...................... 68826.3.4. ‘AP-913’ method....................................................................................... 68926.4. Maintenance baselines................................................................................................ 69126.5. On-site maintenance................................................................................................... 69426.5.1. The various stages of maintenance operations.............................. 69426.5.2. Main conditions for successful maintenance.................................. 69626.5.3. Examples of anomalies or deviations discovered during routine maintenance, explained by an inadequate maintenancebaseline with regard to deterioration mechanisms....................... 70526.5.4. Examples of events associated with non-quality maintenance.............................................. 70826.5.4.1. Example of an event explained by an incorrect setting on redundant equipment................................... 70826.5.4.2. Example of an event explained by an incorrect setting of electrical protection thresholds................. 70826.5.4.3. Examples of events explained by a failure to return equipment to a compliant state after maintenance or an error in performing a work procedure................................................................ 709Chapter 27In-service Monitoring and Inspection of Equipment27.1. Main internal equipment items on a pressurized water reactor vessel...... 71927.1.1. ‘Core baffle’ around the core................................................................ 72027.1.2. RCCA guide tubes..................................................................................... 72227.2. Reactor vessel, nozzles and head............................................................................. 72227.2.1. Vessel underclad defects........................................................................ 72427.2.2. Cracking on vessel head adaptors....................................................... 72527.2.2.1. Condition of other reactors............................................. 72727.2.2.2. Impact on safety................................................................. 72827.2.2.3. Prevention, monitoring and mitigation....................... 72927.2.2.4. Developing inspection tools............................................ 72927.2.2.5. Repairs.................................................................................... 73027.2.2.6. Leak detection...................................................................... 73027.2.2.7. Anti-ejection devices......................................................... 73127.2.2.8. Current situation................................................................. 73127.2.2.9. Cracks observed on reactor vessel heads in other countries............................................................... 73127.2.2.10. Implementation of special monitoring for ‘Inconel areas’ beginning in 1992........................... 73227.2.3. Cracking on vessel lower head penetrations detected in 2011.......................................................... 73327.2.4. Monitoring the ‘beltline’ region of the vessel................................. 73527.2.5. Defects observed on reactor vessels in Belgium............................ 73627.3. Steam generators.......................................................................................................... 73827.3.1. The different types of defects.............................................................. 73827.3.2. Associated risks......................................................................................... 74027.3.3. Monitoring during operation and inspection during outages.... 74127.3.3.1. Monitoring during operation........................................... 74127.3.3.2. Inspection during reactor outages................................. 74227.3.4. Steps to be taken when a defect is detected.................................. 74227.3.4.1. Tube wear due to foreign matter.................................. 74327.3.4.2. Wear due to contact with anti-vibration bars.......... 74327.3.4.3. Cracking in U-bend tubes................................................. 74427.3.4.4. Tube deformation and cracking..................................... 74427.3.5. Steam generator replacement.............................................................. 74527.3.6. Clogging observed in the 2000s.......................................................... 74627.3.7. Conclusion.................................................................................................. 74827.4. Steam lines...................................................................................................................... 74827.5. Auxiliary systems: cracks induced by local thermal-hydraulic phenomena................................................................................. 75027.5.1. Cracking in non-isolatable sections connected to the reactor coolant loops................................................................. 75027.5.2. RHRS thermal fatigue at Unit 1 of the Civaux nuclear power plant......................................... 75127.6. Civil works: containment structures....................................................................... 75527.6.1. Anticipated degradation phenomena................................................. 75627.6.2. Devices for direct monitoring of containment building concrete walls......................................... 75727.6.3. Leak tests and measurements.............................................................. 75927.6.4. Main anomalies......................................................................................... 759Chapter 28Fuel Management, Monitoring and Developments28.1. Procedures for monitoring fuel rod integrity....................................................... 76828.1.1. Radiochemical specifications for reactor coolant.......................... 76828.1.2. Inspections and measurements carried out directly on fuel assemblies.............................................. 77728.1.2.1. Liquid penetrant testing in the refuelling machine mast....................................................................... 77828.1.2.2. Liquid penetrant testing in the FB cell......................... 77928.1.2.3. Inspections performed on fuel rods.............................. 78028.2. Operating experience feedback and changes in cladding material.............. 78228.3. Anomalies and significant events involving fuel assemblies.......................... 78628.3.1. Baffle jetting............................................................................................... 78628.3.2. Fretting........................................................................................................ 78728.3.3. Events encountered during handling operations............................ 78928.3.4. Lateral deformation of fuel assemblies interfering with RCCA drop.......................... 791Chapter 29Facility Compliance29.1. Introduction.................................................................................................................... 79529.2. Detection and treatment of compliance deviations for pressurized water reactors in the nuclear power plant fleet............... 79629.2.1. Process for handling compliance gaps............................................... 79629.2.2. Examples of compliance gaps............................................................... 79829.2.2.1. Compliance gap in electrical connection boxes qualified for accident conditions................................... 79829.2.2.2. Failure in the seismic resistance of metal floors in electrical and auxiliary buildings of 900 MWe reactors (CPY series). 79929.2.2.3. Risk of containment sump screen blockage............... 79929.2.2.4. Anomaly found in engines of emergency and SBO diesel generators for 900 MWe reactors.................... 80029.2.2.5. Temperature resistance fault in the high-head safety injection pumps...................................................... 80229.2.2.6. Mixed lubricants in equipment required for accident situations....................................................... 80229.2.2.7. Flow imbalance between safety injection lines of 900 MWe reactors......................................................... 80429.2.2.8. Anomaly in CATHARE software modelling of natural circulation in the upper part of the vessel...................................... 80529.2.2.9. Vibrations and rotor lift on engineered safety motor-driven pump units.................................... 805Chapter 30Periodic Reviews30.1. Introduction.................................................................................................................... 80730.2. History of periodic reviews in France for nuclear power reactors................ 80930.2.1. Reactors other than PWRs in the French nuclear power plant fleet......................................................... 80930.2.2. PWRs in the French nuclear power plant fleet (900 MWe, 1300 MWe and 1450 MWe).......................................... 81130.3. Periodic review process for PWRs in the French nuclear power plant fleet..................................................................... 81530.3.1. Regulations................................................................................................. 81530.3.2. Outline of a PWR periodic review....................................................... 81730.4. Case of the review associated with the third ten-yearly outage of 900 MWe reactors..................................................... 82230.4.1. Plant unit compliance reviews, the complementary investigation and ageing management............................................. 82430.4.1.1. Plant unit compliance reviews........................................ 82430.4.1.2. Complementary investigation programme................ 82530.4.1.3. Ageing management.......................................................... 826 30.4.2. Compliance studies on the design of civil works systems and structures............................................. 82630.4.3. Studies to reassess system design...................................................... 82730.4.4. Reassessment of reactor resistance to internal and external hazards................................................................................ 83030.4.5. Accident studies........................................................................................ 83230.4.6. Taking into account lessons learned during the review associated with the VD3 900 outage for subsequent reviews....83730.5. Fourth ten-yearly outage of 900 MWe reactors: integrating the extension of the operating lifetime of nuclear power reactorsin France.......................................................................................................................... 83830.5.1. Background................................................................................................. 83830.5.2. Periodic Review Strategic Plan – Setting objectives..................... 83930.5.3. A few significant issues identified in reviews conducted by safety organizations.................. 84230.6. Overview of international practices – IAEA Guides........................................... 84530.6.1. International practices............................................................................ 84630.6.2. IAEA Guides................................................................................................ 84730.7. Multilateral practices................................................................................................... 848Chapter 31Optimizing Radiation Protection and Limiting Doses Received by Workers During Operations in a Nuclear Power Plant31.1. Sources of ionizing radiation in a nuclear power reactor................................ 85231.2. Examples of optimization of worker radiation protection.............................. 85331.3. Arrangements for ‘Major Refit’ operations........................................................... 85531.4. Approach and objectives adopted for the EPR.................................................... 858 Part 4The Accidents at Three Mile Island, Chernobyl and Fukushima Daiichi Nuclear Power Plants, Lessons Learned and Emergency Response ManagementChapter 32The Three Mile Island Nuclear Power Plant Accident32.1. Accident sequence – Reconstitution through simulation................................ 86432.2. Accident consequences............................................................................................... 87232.3. Analysis of the accident causes................................................................................ 87432.3.1. Error in identifying the position of the relief valve....................... 87432.3.2. Understanding the behaviour of the pressurizer............................ 87532.3.3. Stopping safety injection....................................................................... 87632.3.4. Human-machine interface..................................................................... 87632.3.5. Isolating the reactor containment...................................................... 87732.3.6. Confinement inside the auxiliary building....................................... 87732.3.7. Emergency feedwater supply to the steam generators............... 87732.4. Lessons learned from the Three Mile Island accident....................................... 87732.4.1. The human factor in facility operation............................................. 87832.4.2. Importance of precursor events........................................................... 88132.4.3. Study of complex situations and core-melt accidents, handling emergency situations....................................... 88232.5. Conclusions..................................................................................................................... 883Chapter 33Incident and Accident Operation: from the Event-Oriented Approach to the State-Oriented Approach33.1. Limits of the event-oriented approach.................................................................. 88533.2. The State-Oriented Approach concept.................................................................. 88633.3. First application of the state-oriented approach............................................... 88833.4. Widespread application of the state-oriented approach................................. 89133.5. ‘Stabilized’ state-oriented approach....................................................................... 89233.6. State-oriented approach adopted for the EPR.................................................... 894Chapter 34The Chernobyl Nuclear Power Plant Accident34.1. The Chernobyl nuclear power plant and RBMK reactors................................. 89734.2. The accident sequence................................................................................................ 90134.3. Analysis of the accident causes and changes made to RBMK units soon after the accident.................................. 90634.4. The other units at the facility................................................................................... 90834.5. Radioactive release and protection of the population..................................... 90834.5.1. Radioactive release kinetics.................................................................. 90834.5.2. Protection of the population................................................................ 91134.6. Consequences on human health and the environment................................... 91534.6.1. Direct effects of radiation...................................................................... 91534.6.2. Thyroid cancer in children..................................................................... 91734.6.3. Long-term contamination in the Dnieper Basin............................. 91934.7. Radioactive fallout in France and its consequences.......................................... 92134.7.1. Doses attributable to the plume......................................................... 92134.7.2. External doses due to soil deposition................................................ 92134.7.3. Doses due to ingestion of contaminated foodstuffs.................... 92234.7.4. Overall levels.............................................................................................. 92334.7.5. Thyroid cancer........................................................................................... 92434.7.5.1. Monitoring thyroid cancer in France............................ 92534.7.5.2. Assessment of the number of cancer cases induced in France by the Chernobyl accident........... 92634.8. Lessons learned by the international community from a general viewpoint and with regard to RBMK reactors....................... 92734.9. Lessons learned in France........................................................................................... 92834.10. Keeping the public informed..................................................................................... 93134.11. After the Chernobyl accident.................................................................................... 933Chapter 35Options and Control of Reactivity Insertion in Pressurized Water Reactors35.1. Research and study of event sequences................................................................ 93735.1.1. Cooling accidents..................................................................................... 938 35.1.2. Incidents and accidents related to rod cluster control assemblies.................................................................................... 94035.1.3. Boron dilution accidents........................................................................ 94335.1.4. Inserting a cold water plug in the core............................................. 95035.2. Changes in criteria........................................................................................................ 95135.3. The case of outage states.......................................................................................... 95335.4. Regulations...................................................................................................................... 957Chapter 36The Reactor Accident at the Fukushima Daiichi Nuclear Power Plant and Lessons Learned in France36.1. Reactor units at the Fukushima Daiichi nuclear power plant........................ 96236.1.1. General operation of a boiling water reactor.................................. 96236.1.2. Containment.............................................................................................. 96236.1.3. Emergency cooling systems.................................................................. 96436.2. Sequence of events during the accident............................................................... 96636.3. Radioactive release....................................................................................................... 97236.3.1. Airborne radioactive release, residual caesium deposits and contamination of foodstuffs............................... 97236.3.1.1. Airborne radioactive release............................................ 97236.3.1.2. Persistent caesium deposits............................................ 97236.3.1.3. Contamination of foodstuffs.......................................... 97336.3.2. Release of radioactive substances in the Pacific Ocean.............. 97536.3.3. Long-distance atmospheric dispersion of the radioactive plume........................................................................ 97636.4. Action taken to control facilities and released contaminated water.......... 97836.5. Socioeconomic and health impact in numbers................................................... 98136.5.1. Socioeconomic impact............................................................................ 98136.5.2. Health impact............................................................................................ 98236.6. Lessons learned from the accident.......................................................................... 98536.6.1. Complementary safety assessments carried out in Europe and France following the Fukushima Daiichi nuclear power plant accident............................................................................................ 98636.6.2. Complementary safety assessments carried out in France........ 98836.6.3. Procedure for complementary safety assessments carried out in France...................................... 98936.6.4. Conclusions of the complementary safety assessments carried out in France............................................. 99036.6.5. The ‘hardened safety core’.................................................................... 99136.6.5.1. Purpose................................................................................... 99136.6.5.2. Principles................................................................................ 99136.6.5.3. Illustrations........................................................................... 99336.6.6. Nuclear Rapid Response Force (FARN).............................................. 99536.6.7. Deployment of post-Fukushima measures in French nuclear power plants............................................................ 99736.7. Other lessons learned in France from the Fukushima Daiichi nuclear power plant accident...................................... 998Chapter 37Lessons Learned from the Fukushima Daiichi Nuclear PowerPlant Accident: Work Conducted by the IAEA and WENRA, Action Taken in Countries Other than France37.1. Work conducted by the IAEA.................................................................................... 100237.2. Work conducted by WENRA..................................................................................... 100337.3. Japan................................................................................................................................. 100437.4. Belgium............................................................................................................................ 100637.4.1. Nuclear power plants in Belgium........................................................ 100637.4.2. General details on the design of Belgian nuclear power plants................................................................ 100637.4.3. Stress tests and main lessons learned............................................... 100737.4.3.1. Improving protection of facilities against external hazards.................................................................. 100737.4.3.2. Improving protection of facilities against loss of electrical power supplies or loss of heat sink....... 101037.4.3.3. Improving on-site emergency plans............................. 1011 37.4.3.4. Improving management of core-melt accidents...... 101237.5. USA.................................................................................................................................... 1013Chapter 38Emergency Preparedness and Response38.1. Defining a radiological emergency and ‘response’ objectives........................ 102138.2. General organization of radiological emergency management..................... 102338.2.1. Organization and entities concerned................................................. 102338.2.2. Major Nuclear or Radiological Accident National Response Plan and emergency plans........................................ 102538.2.2.1. Major Nuclear or Radiological Accident National Response Plan...................................................................... 102538.2.2.2. Emergency plans................................................................. 102638.2.2.3. Provisions for protecting the public in the event of an accidental release of radioactivity..................... 102838.3. Management by the operator................................................................................... 102938.4. Prefectural authorities and mayors......................................................................... 103138.5. ASN, the Nuclear Safety Authority......................................................................... 103238.6. IRSN................................................................................................................................... 103338.7. Assessment approach in the event of an accident affecting a reactor in the nuclear power plant fleet.................. 103838.7.1. ‘3D/3P’ method......................................................................................... 103938.7.2. The ‘aggravated prognosis’ approach................................................ 104138.7.3. Extending the 3D/3P method to severe accidents (the ‘D/P AG’ method)...................................... 104138.8. Emergency preparedness............................................................................................ 104238.8.1. Emergency response exercises............................................................. 104338.8.2. Operating experience feedback........................................................... 1045 Part 5PWR Safety Studies, R&D and Simulation SoftwareChapter 39PWR Safety Studies and R&D39.1. Contribution of studies to the improvement of pressurized water reactor safety................................................................ 105039.2. Purpose and overview of R&D work, dedicated programmes and organizations involved, and research facilities in France... 105239.2.1. Purpose and overview of R&D............................................................. 105239.2.2. Dedicated frameworks and organizations involved....................... 106339.2.3. Facilities in France used for research and development.............. 1065Chapter 40Examples of Simulation Software Developed for Safety Analysis of Pressurized Water Reactors40.1. Simulation software for neutronics........................................................................ 107640.2. Simulation software for thermal hydraulics (and mechanics)....................... 107940.3. Simulation software for thermal mechanics........................................................ 108440.4. Software for simulating core-melt situations..................................................... 108540.5. Simulation software for mechanics........................................................................ 108840.6. Fire simulation software............................................................................................. 1089List of Acronyms.......................................................................................................................... 1093Acronyms for institutions, bodies and groups..................................................................... 1093Technical acronyms and abbreviations.................................................................................. 1103Technical glossary......................................................................................................................... 1121  ContentsPreface............................................................................................................................................. IIIForeword......................................................................................................................................... VIIEditors, Contributors and Reviewers................................................................................... XXXVIIIntroduction................................................................................................................................... XLIIIPart 1General BackgroundChapter 1Biological and Health Effects of Ionizing Radiation – The Radiological Protection System1.1. Biological and health effects of ionizing radiation............................................. 41.1.1. Biological processes................................................................................. 41.1.2. Review of units of measure................................................................... 61.1.3. Natural radioactivity............................................................................... 71.1.4. Health effects............................................................................................ 81.1.4.1. Deterministic effects, tissue reactions......................... 81.1.4.2. Stochastic or random effects.......................................... 91.1.4.3. Induction of diseases other than cancer..................... 111.1.5. Example of the limitations of epidemiology................................... 111.2. Radiological protection system................................................................................ 131.2.1. Types of exposure situations................................................................ 14 1.2.2. Exposure categories................................................................................. 141.2.3. Justification principle............................................................................... 151.2.4. Optimization (ALARA) principle ......................................................... 161.2.5. Principle of application of dose limits............................................... 21Chapter 2Organization of Nuclear Safety Control and Regulation for Nuclear Facilities and Activities in France2.1. From the founding of CEA to the TSN Act........................................................... 232.2. A few definitions........................................................................................................... 282.3. The different contributors to nuclear safety and their missions.................. 302.4. A few basic principles and notions in the field of nuclear safety................. 462.5. Statutory and quasi-statutory frameworks applicable to basic nuclear installations.............................................. 48Chapter 3The International Dimension and the Social Dimension3.1. International dimension.............................................................................................. 773.1.1. Introduction................................................................................................ 773.1.2. IAEA standards........................................................................................... 803.1.3. International Reporting System for Operating Experience (IRS).......................................................... 823.1.4. Services developed by the IAEA........................................................... 843.1.4.1. OSART reviews.................................................................... 843.1.4.2. IRRS reviews......................................................................... 873.1.4.3. Other services and study frameworks set up by the IAEA.............................................................. 883.1.5. WANO.......................................................................................................... 903.1.6. NEA............................................................................................................... 913.1.7. Organizations dedicated to radiation protection and health........................................................................ 933.1.8. From bilateral Franco-German cooperation to European structures for the exchange and capitalization of knowledge and practices, training and assessment services........................................................ 943.1.9. Nuclear regulator associations............................................................. 1003.2. The social dimension................................................................................................... 1023.2.1. Introduction – the context in France................................................. 1023.2.2. Examples of initiatives and issues raised concerning reactor safety in the French nuclear power plant fleet........... 102Chapter 4Nuclear Reactors: Complex Sociotechnical Systems – the Importance of Human and Organizational Factors4.1. 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.............................. 1084.2. The accident at the Chernobyl nuclear power plant and the concept of ‘safety culture’................................. 1094.3. The Fukushima Daiichi nuclear power plant accident: the social dimension and the concept of organization ‘resilience’... 1134.4. Changes in the perception of the role of people in achieving a high level of reliability in complex sociotechnical systems...... 1134.5. Main topics studied in the development of resources and skills pertaining to human and organizational factors............ 1164.5.1. Resources and skills................................................................................. 1164.5.2. Main topics studied.................................................................................. 1184.6. Human and organizational factors in French regulations................................ 119Part 2Safety by DesignChapter 5The Development of Nuclear Power Using Uranium-235 Fission – A Few Notions of Physics Used in Pressurized Water Reactors5.1. Important milestones in the development of nuclear power using fission of the uranium-235 isotope.......................... 1235.2. Fission and important concepts in reactor kinetics........................................... 1275.3. Removing power from the core during operation............................................. 1345.4. Decay heat...................................................................................................................... 1355.5. Main features of pressurized water reactor cores.............................................. 1365.6. Control and monitoring of pressurized water reactor cores.......................... 1375.7. Using uranium and plutonium mixed oxide (MOX) fuel................................. 147Chapter 6General Objectives, Principles and Basic Concepts of the Safety Approach6.1. General approach to risks – General objectives................................................. 1526.2. Fundamental safety functions.................................................................................. 1566.3. Confinement barriers................................................................................................... 1576.4. Defence in depth........................................................................................................... 1616.4. Defence in depth........................................................................................................... 1616.4.1. Levels of defence in depth..................................................................... 1626.4.2. Elements common to the different levels of defence in depth............................................................................. 1696.5. Events considered: terminology adopted for nuclear power reactors......... 1696.6. WENRA reference levels............................................................................................. 1716.7. Deterministic safety analysis and probabilistic safety assessments............ 1716.8. Lessons learned from the accident at the Fukushima Daiichi nuclear power plant on the concept of defence in depth and deterministic analysis......................................................................................... 1736.9. Safety culture – Quality control.............................................................................. 174Chapter 7Safety Options and Considerations at the Design Phase7.1. Different types of design provisions associated with safety considerations................................................. 1887.2. Single-failure criterion................................................................................................. 1897.3. The specific nature of computer-based systems (based on instrumentation and control software)............... 1937.4. Equipment safety classification................................................................................ 1957.4.1. Importance of equipment for safety and safety classification...................................................................... 1957.4.2. Generic requirements associated with the different safety classes.................................................................. 1997.4.3. Qualification of equipment for accident conditions..................... 2027.5. Information on designing nuclear pressure equipment.................................... 2077.6. General considerations on provisions for hazards in facility design............ 2107.7. Anticipating decommissioning in the design stage........................................... 212Chapter 8Study of Operating Conditions in the Deterministic Safety Analysis8.1. Categories of operating conditions......................................................................... 2188.2. Choice of operating conditions................................................................................ 2228.2.1. Concept of ‘bounding’ incident or accident..................................... 2238.2.2. Accident exclusion.................................................................................... 2248.3. List and breakdown of operating conditions....................................................... 2258.4. Methods for studying operating conditions......................................................... 2278.4.1. Choice of initial conditions, conservatism....................................... 2288.4.2. Consideration of an aggravating event in the study on operating conditions – ‘Passive’ failures............ 2298.4.3. Conventional combinations.................................................................. 2318.4.4. Preventing accident aggravation......................................................... 2328.4.5. Operator response time......................................................................... 2328.4.6. Using qualified simulation software................................................... 2338.4.7. Main criteria to be met for fuel in the reactor core..................... 2348.5. Concept of ‘design-basis situations’ for equipment.......................................... 2368.6. Situations to be taken into account in application of pressure equipment regulations................................................. 2378.7. Assessing the radiological consequences of incidents, accidents and hazards.............................................. 2388.7.1. Assessing radioactive substances released from the facility...... 2408.7.2. Assessing radiological consequences of radioactive release from the facility................................................ 2428.7.3. Assessing radiological consequences................................................. 243Chapter 9Loss-of-Coolant Accident9.1. Short- and medium-term aspects of a LOCA...................................................... 2529.1.1. Mechanical effects on vessel internals and fuel assembly structures................................................................ 2539.1.2. Thermal-hydraulic aspects and behaviour of fuel rods................ 2559.1.2.1. Large-break LOCA............................................................... 2559.1.2.2. Intermediate-break LOCA................................................ 2579.1.3. Effects on reactor containment and internals................................ 2589.1.4. Long-term aspect...................................................................................... 2599.2. Safety demonstration.................................................................................................. 2619.2.1. General information and background................................................ 2619.2.2. Fuel assemblies and fuel rods, vessel internals, reactor coolant system components................................................. 2639.2.2.1. Mechanical strength of vessel internals, fuel assembly structures and reactor coolant system components.............. 2639.2.2.2. Fuel behaviour...................................................................... 2659.2.3. Reactor containment and equipment located inside.................... 266Chapter 10A Special Issue: Steam Generator Tubes10.1. Steam generator tube rupture as a Category 3 event...................................... 27010.2. Preventing an SGTR accident, risk of multiple ruptures................................... 27210.3. Steam generator tube rupture(s) studied as a Category 4 event................. 27410.3.1. 900 MWe and 1300 MWe reactors.................................................... 27410.3.2. 1450 MWe reactors and EPR (Flamanville 3).................................. 27410.4. Provisions to mitigate the radiological consequences of SGTR accidents. 276Chapter 11Providing for Hazards: General Considerations and Internal Hazards11.1. General considerations on providing for hazards............................................... 27911.2. Potential projectiles inside the containment....................................................... 28211.3. Effects of pipe breaks.................................................................................................. 28411.4. Projectiles generated by a turbine rotor failure.................................................. 28511.5. Protection against load drops................................................................................... 28811.5.1. Risks related to spent fuel transport packaging............................. 28811.5.2. Other handling risks................................................................................ 29111.6. Fire protection............................................................................................................... 29211.7. Explosion protection.................................................................................................... 29711.8. Internal flooding............................................................................................................ 300Chapter 12Providing for External Hazards12.1. General considerations on providing for external hazards.............................. 30512.2. ‘Climate watch’ implemented by EDF.................................................................... 30712.3. Earthquakes..................................................................................................................... 30712.4. External floods............................................................................................................... 32212.5. Extreme temperatures................................................................................................. 33012.5.1. Extreme cold.............................................................................................. 33012.5.2. Extreme heat.............................................................................................. 33112.6. Possible heat sink hazards.......................................................................................... 33212.7. Other naturally-occurring external hazards......................................................... 33612.8. Accidental aeroplane crashes (excluding malicious acts)................................ 33612.9. Risks related to the industrial environment (excluding malicious acts)..... 340Chapter 13Complementary Domain of Events13.1. The origin of studies belonging to the complementary domain................... 34413.2. Background of the complementary domain........................................................ 34413.3. Analysis of complementary domain events......................................................... 35113.4. ‘New complementary domain’................................................................................. 35113.5. Case of the Flamanville 3 EPR.................................................................................. 354Chapter 14Development and Use of Probabilistic Safety Assessments14.1. History and regulatory context................................................................................ 35714.1.1. International situation............................................................................ 35714.1.2. Situation in France................................................................................... 35914.2. Level 1 PSA...................................................................................................................... 36114.2.1. Scope............................................................................................................ 36114.2.2. Method for carrying out a Level 1 PSA............................................. 36214.2.2.1. General information........................................................... 36214.2.2.2. Specific point: probabilistic human reliability analysis................................................................................... 36414.2.3. Level 1 PSA results and lessons learned............................................ 36914.3. Level 2 PSA...................................................................................................................... 37314.3.1. Scope............................................................................................................ 37314.3.2. Method for carrying out a Level 2 PSA............................................. 37414.3.2.1. General information........................................................... 37414.3.2.2. Probabilistic human reliability analysis for Level 2 PSAs................................................................... 37914.3.3. Examples of lessons learned from Level 2 PSAs............................. 38214.3.3.1. Steam explosion risk assessment.................................. 38214.3.3.2. Mechanical integrity of the 900 MWe reactor containments........................................................ 38314.3.3.3. Isolating penetrations in the containment................. 38314.3.3.4. Modifying the pressure relief system of the reactor coolant system........................................ 38414.3.3.5. Improvement of operating procedures to reduce risk of core melt under pressure................ 38414.3.3.6. Contribution of Level 2 PSAs to emergency response measures............................................................. 38514.4. Expanding the scope of PSA coverage................................................................... 38514.5. Using probabilistic safety assessments.................................................................. 38614.5.1. Using PSAs in the design phase........................................................... 38614.5.1.1. Usefulness and particularities of PSAs in the design phase............................................................. 38614.5.1.2. PSAs conducted to support the Flamanville 3 EPR design............................................................................. 38714.5.2. Using PSAs in periodic reviews............................................................ 38914.5.2.1. Level 1 PSA............................................................................ 39014.5.2.2. Level 2 PSA............................................................................ 39114.5.3. Using PSAs for reactor operation........................................................ 39214.5.3.1. Using PSAs to analyse event severity.......................... 39214.5.3.2. Using PSAs to analyse operational limits and conditions and temporary changes...................... 39414.5.3.3. Using PSAs to analyse operating procedures............. 396Chapter 15Aspects Specific to PWR Spent Fuel Storage Pools15.1. Spent fuel pool design................................................................................................. 39915.1.1. Confinement barriers.............................................................................. 39915.1.2. Initiating events defined at the design stage.................................. 40015.2. Experience feedback..................................................................................................... 40115.2.1. Loss of cooling........................................................................................... 40115.2.1.1. Loss of heat sink.................................................................. 40115.2.1.2. Risks related to maintenance during unit outages................................................................................... 40215.2.1.3. Suction of foreign matter into the cooling system.................................................................................... 40315.2.1.4. Exceeding the decay heat defined in facility design...................................................................................... 40315.2.2. Water losses............................................................................................... 404 15.2.2.1. Gate or sluice gate failures.............................................. 40415.2.2.2. Line-up errors....................................................................... 40515.2.2.3. Failure of a reactor coolant system pipe nozzle dam.......................................................................................... 40915.2.2.4. Rupture of a pipe connected to the spent fuel pool................................................................................. 41115.3. Safety reassessments................................................................................................... 41115.4. Experience feedback from the accident that affected the Unit 4 pool at the Fukushima Daiichi nuclear power plant..... 41415.4.1. Events........................................................................................................... 41415.4.2. Complementary safety assessments conducted in France......... 41715.5. Measures adopted for the EPR.................................................................................. 41915.6. Recommendations for new reactor designs......................................................... 42015.7. New systems for storing spent fuel........................................................................ 423Chapter 16Taking into Account Human and Organizational Factors in Facility Design16.1. Taking into account human and organizational factors in nuclear power reactor design....................................... 42516.1.1. Importance of considering human and organizational factors at the design stage....................................... 42516.1.2. Approach at the design stage............................................................... 43116.1.2.1. Prior to the design phase: analysis of ‘existing elements’........................................ 43216.1.2.2. Design objectives................................................................ 43516.1.2.3. Definition of detailed design provisions...................... 43616.1.2.4. Validation of design provisions...................................... 43916.1.2.5. Assessments conducted during reactor startup and after commissioning.................................................. 44216.1.3. Project management and human and organizational factors engineering programme..................................... 44316.2. Considering human and organizational aspects when designing changes to nuclear power plants..................... 44316.2.1. Importance of human and organizational factors in designing modifications..................................................................... 44416.2.2. ‘Human, social and organizational approach’ implemented by EDF........................................................................... 44516.2.3. Changes, a subject that always deserves special attention from a human and organizational factors perspective.... 44716.3. Human and organizational factors for future nuclear power reactor projects.................................................. 448Chapter 17Studying Core-Melt Accidents to Enhance Safety17.1. Core degradation and vessel failure....................................................................... 45217.1.1. Core uncovery............................................................................................ 45317.1.2. Fuel degradation....................................................................................... 45417.1.3. Failure of the reactor coolant system............................................... 45517.1.4. Phenomena that can cause early containment failure................. 45517.1.5. Phenomena that can ultimately lead to containment failure....................................................... 45717.2. Containment failure modes....................................................................................... 45717.3. Classification of releases associated with core-melt accidents − ‘source terms’................................................. 46017.4. Improving knowledge.................................................................................................. 46217.5. Studies in France on containment failure modes............................................... 46217.5.1. Introduction................................................................................................ 46217.5.2. Initial containment leakage................................................................... 46317.5.3. Direct heating of gases in the containment.................................... 46317.5.4. Hydrogen explosion in the containment.......................................... 46417.5.5. Steam explosion in the vessel or reactor pit................................... 46517.5.6. Gradual pressure increase in the containment............................... 46617.5.7. Penetration of the concrete basemat of the containment by corium............................................ 46717.5.8. ‘U4’ provisions........................................................................................... 46817.5.9. Bypass of containment by outgoing pipes (the V mode).............................................................................................. 46817.5.10. Fast reactivity insertion accidents...................................................... 46917.6. Severe accident operating guidelines..................................................................... 46917.7. Radiological consequences associated with the S3 source term and emergency response plans implemented by public authorities............ 47017.8. Ultimate emergency operating procedures.......................................................... 47317.9. On-site emergency plan............................................................................................. 47317.10. Approach adopted for the EPR................................................................................. 47617.10.1. General safety objectives....................................................................... 47617.10.2. ‘Practical elimination’ of core-melt conditions that could lead to significant early releases............................. 47617.10.3. Provisions for low-pressure core melt............................................... 480Chapter 18New-Generation Reactors18.1. Organization and framework of Franco-German discussions........................ 48518.2. Progression of safety objectives and design options for the EPR project.......................................................... 48618.2.1. General safety objectives....................................................................... 48618.2.2. Events to be taken into account at the design stage and in deterministic and probabilistic analyses.............................. 48818.2.3. Main provisions for preventing incidents and accidents............. 49018.2.4. Functional redundancy, independence between systems, system reliability................................................ 49518.2.5. Confinement preservation..................................................................... 49618.2.6. Radiological protection........................................................................... 49718.2.7. Incorporating lessons learned from the Fukushima Daiichi nuclear power plant accident......................... 49718.3. International context: general safety objectives for new-generation reactors............................................ 49918.4. Concepts highlighted in new reactor designs...................................................... 50118.4.1. AP1000: gravity systems........................................................................ 50118.4.2. VVER: SPOT system................................................................................. 50318.4.3. NM EPR: ‘multi-group’ technology, diversified heat sink............ 504 18.4.4. ATMEA 1: safety injection accumulators in the reactor coolant system.............................................................. 50518.4.5. NuScale: common pool for modular reactors................................. 505 Part 3Safety in OperationChapter 19Startup Tests for Pressurized Water Reactors19.1. Introduction.................................................................................................................... 51119.2. Commissioning.............................................................................................................. 51419.2.1. Defining startup tests.............................................................................. 51419.2.2. Phasing of startup tests.......................................................................... 51519.2.2.1. Preliminary and pre-operational tests......................... 51519.2.2.2. Operational tests................................................................ 51719.2.2.3. General principles for test sequencing and execution....................................................................... 51719.2.3. Documentation for startup tests......................................................... 51819.2.3.1. Integrated system test procedures and startup test procedures............................................ 51819.2.3.2. Test programmes, test procedures, standard test guidelines.................................................... 51819.2.3.3. Completeness analysis, adequacy analysis................. 51819.2.3.4. Acceptance criteria............................................................. 51919.3. Objectives and general rules to take into account for startup tests........... 52019.4. Key lessons learned from startup tests on nuclear power reactors in France............................................................ 52119.4.1. Qualification tests and on-site tests.................................................. 52219.4.2. Long-term on-site testing...................................................................... 52419.4.3. Test configurations and completeness, transpositions................ 525 19.4.4. Safety measures that cannot be verified by testing..................... 52719.4.5. Criteria......................................................................................................... 52719.4.6. Cleanness, keeping system lines clean, foreign matter................ 52819.4.7. Piping support structures and displacement................................... 53219.4.8. Pump and piping vibrations................................................................... 53219.4.9. Validation of operating procedures and periodic tests................ 53419.4.10. Uncertainty and ‘set points’.................................................................. 53519.4.11. Condition of facilities during startup tests...................................... 53619.4.12. Other aspects............................................................................................. 53719.5. Examples of findings resulting from startup tests............................................. 538Chapter 20General Operating Rules20.1. General Operating Rules............................................................................................ 55220.1.1. Content of general operating rules..................................................... 55320.1.2. Limits of general operating rules......................................................... 55420.2. Operational limits and conditions........................................................................... 55420.2.1. Content of operational limits and conditions................................. 55520.2.1.1. Operating modes and standard states......................... 55620.2.1.2. Requirements and unavailability................................... 55820.2.1.3. Fallback states and time required to reach them.... 55820.2.1.4. Events and event groups.................................................. 55920.2.1.5. Combined types of unavailability.................................. 56020.2.1.6. Concepts of ‘boundary condition’ and ‘specific requirement’................................................ 56020.2.2. Average pressure and temperature range of the reactor coolant system.............................................................. 56120.2.3. Changes in operational limits and conditions................................. 56320.3. Initial and periodic tests............................................................................................. 56420.4. Incident and accident operating procedures........................................................ 566Chapter 21 Operating Experience Feedback from Events: Rules and Practices21.1. Background...................................................................................................................... 56921.2. Objectives of an operating experience feedback system................................ 57121.3. Components of an operating experience feedback system – Regulations............................................................................... 57221.4. Operating experience feedback practices adopted for the French nuclear power plant fleet................................. 576Chapter 22Operating Experience from Events Attributable to Shortcomings in Initial Reactor Design or the Quality of Maintenance22.1. 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).......................................... 59022.2. Recurrent loss of safety function events related to maintenance operations – Lessons learned........................... 59522.2.1. Events........................................................................................................... 59522.2.2. General discussion initiated by EDF in the late 1980s on the quality of maintenance operations....................................... 60522.2.3. Applying the defence-in-depth concept when working on a reactor in service.................................. 60822.2.4. Problems that may recur....................................................................... 609Chapter 23Operating Experience from Events Related to Maintenance Operations, Electrical Power Sources and Distribution, Internal and External Hazards23.1. Risks of failure related to equipment or maintenance..................................... 61223.1.1. Risks of common-mode failure............................................................ 61223.1.1.1. Risks of common-mode failure related to settings............................................................................. 61223.1.1.2. Risks of common-mode failure on electrical switchboards.................................................61323.1.1.3. Unavailability of two out of three high-head safety injection lines in the cold legs of the reactor coolant system.............. 61423.1.1.4. Loss of electrical power supplies................................... 61723.1.2. Introduction of non-borated water into the reactor coolant system.......................................................... 62023.1.3. Cooling the reactor coolant system after inhibition of automatic actions............................................................................... 62223.1.4. A temporary device prevents switching the safety injection system to the water recirculation mode................... 62423.2. Events related to internal hazards........................................................................... 62523.2.1. Risk of common-mode failure due to internal flooding.............. 62523.2.2. Risk of failure due to fire....................................................................... 62823.2.3. Risks associated with the use of hydrogen in 900 MWe reactors............................................................................... 63123.3. External hazards: events related to periods of extreme cold......................... 635Chapter 24Enhanced Protection of Estuary and River Sites: Flooding at the Blayais Nuclear Power Plant and Obstruction of a Water Intakeat the Cruas-Meysse Nuclear Power Plant24.1. Partial loss of engineered safety systems following flooding of the Blayais nuclear power plant............................... 64224.2. 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...................................................... 648Chapter 25Taking into Account Human and Organizational Factors in Facility Operation25.1. Skills management....................................................................................................... 65325.1.1. Historical background............................................................................. 65425.1.2. Managing training.................................................................................... 65525.1.3. Strategic workforce planning................................................................ 65625.1.4. Personnel certification............................................................................ 65725.2. Safety and risk management.................................................................................... 65825.2.1. Historical background............................................................................. 65825.2.2. Decision-making and safety.................................................................. 66025.2.3. Risk analyses applied to work activities............................................ 66225.2.4. Operating experience feedback........................................................... 66325.2.5. Managing organizational change......................................................... 66425.3. Managing operational activities............................................................................... 66625.3.1. Characteristics of operational activities............................................ 66625.3.2. Monitoring by the operating crew in the control room.............. 66925.3.3. Compliance with general operating rules......................................... 67025.3.4. Line-up......................................................................................................... 67125.3.5. Operation in extreme situations......................................................... 67225.4. Management of maintenance activities................................................................ 67225.4.1. Management of a scheduled reactor outage for refuellingand maintenance...................................................................................... 67225.4.2. Risks during reactor outages................................................................. 67325.4.3. Preparation for scheduled reactor outages...................................... 67425.4.4. Managing scheduled reactor outages................................................ 67525.5. Supervising outsourced activities............................................................................ 67525.5.1. Contractor qualification and contracting......................................... 67625.5.2. Matching workload and resources...................................................... 67625.5.3. Carrying out work.................................................................................... 67725.5.4. Surveillance of outsourced activities................................................. 67825.5.5. Operating experience feedback and assessingoutsourced activities............................................................................... 678Chapter 26Facility Maintenance26.1. Maintenance objectives.............................................................................................. 68126.2. Maintenance................................................................................................................... 68226.2.1. Definition.................................................................................................... 68226.2.2. Maintenance strategies........................................................................... 68326.3. Optimizing maintenance............................................................................................ 68426.3.1. Reliability-centred maintenance.......................................................... 68426.3.2. Conditional maintenance....................................................................... 686XXVI Elements of nuclear safety – Pressurized water reactors26.3.3. Conditional maintenance by sampling –Maintenance based on reference equipment items...................... 68826.3.4. ‘AP-913’ method....................................................................................... 68926.4. Maintenance baselines................................................................................................ 69126.5. On-site maintenance................................................................................................... 69426.5.1. The various stages of maintenance operations.............................. 69426.5.2. Main conditions for successful maintenance.................................. 69626.5.3. Examples of anomalies or deviations discovered during routinemaintenance, explained by an inadequate maintenancebaseline with regard to deterioration mechanisms....................... 70526.5.4. Examples of events associated with non-qualitymaintenance............................................................................................... 70826.5.4.1. Example of an event explained by an incorrectsetting on redundant equipment................................... 70826.5.4.2. Example of an event explained by an incorrectsetting of electrical protection thresholds................. 70826.5.4.3. Examples of events explained by a failureto return equipment to a compliant stateafter maintenance or an error in performinga work procedure................................................................ 709Chapter 27In-service Monitoring and Inspection of Equipment27.1. Main internal equipment items on a pressurized water reactor vessel...... 71927.1.1. ‘Core baffle’ around the core................................................................ 72027.1.2. RCCA guide tubes..................................................................................... 72227.2. Reactor vessel, nozzles and head............................................................................. 72227.2.1. Vessel underclad defects........................................................................ 72427.2.2. Cracking on vessel head adaptors....................................................... 72527.2.2.1. Condition of other reactors............................................. 72727.2.2.2. Impact on safety................................................................. 72827.2.2.3. Prevention, monitoring and mitigation....................... 72927.2.2.4. Developing inspection tools............................................ 72927.2.2.5. Repairs.................................................................................... 730Contents XXVII27.2.2.6. Leak detection...................................................................... 73027.2.2.7. Anti-ejection devices......................................................... 73127.2.2.8. Current situation................................................................. 73127.2.2.9. Cracks observed on reactor vessel headsin other countries............................................................... 73127.2.2.10. Implementation of special monitoringfor ‘Inconel areas’ beginning in 1992........................... 73227.2.3. Cracking on vessel lower head penetrations detectedin 2011......................................................................................................... 73327.2.4. Monitoring the ‘beltline’ region of the vessel................................. 73527.2.5. Defects observed on reactor vessels in Belgium............................ 73627.3. Steam generators.......................................................................................................... 73827.3.1. The different types of defects.............................................................. 73827.3.2. Associated risks......................................................................................... 74027.3.3. Monitoring during operation and inspection during outages.... 74127.3.3.1. Monitoring during operation........................................... 74127.3.3.2. Inspection during reactor outages................................. 74227.3.4. Steps to be taken when a defect is detected.................................. 74227.3.4.1. Tube wear due to foreign matter.................................. 74327.3.4.2. Wear due to contact with anti-vibration bars.......... 74327.3.4.3. Cracking in U-bend tubes................................................. 74427.3.4.4. Tube deformation and cracking..................................... 74427.3.5. Steam generator replacement.............................................................. 74527.3.6. Clogging observed in the 2000s.......................................................... 74627.3.7. Conclusion.................................................................................................. 74827.4. Steam lines...................................................................................................................... 74827.5. Auxiliary systems: cracks induced by local thermal-hydraulicphenomena..................................................................................................................... 75027.5.1. Cracking in non-isolatable sections connectedto the reactor coolant loops................................................................. 75027.5.2. RHRS thermal fatigue at Unit 1 of the Civaux nuclearpower plant................................................................................................ 75127.6. Civil works: containment structures....................................................................... 75527.6.1. Anticipated degradation phenomena................................................. 75627.6.2. Devices for direct monitoring of containment buildingconcrete walls............................................................................................ 75727.6.3. Leak tests and measurements.............................................................. 75927.6.4. Main anomalies......................................................................................... 759Chapter 28Fuel Management, Monitoring and Developments28.1. Procedures for monitoring fuel rod integrity....................................................... 76828.1.1. Radiochemical specifications for reactor coolant.......................... 76828.1.2. Inspections and measurements carried out directlyon fuel assemblies.................................................................................... 77728.1.2.1. Liquid penetrant testing in the refuellingmachine mast....................................................................... 77828.1.2.2. Liquid penetrant testing in the FB cell......................... 77928.1.2.3. Inspections performed on fuel rods.............................. 78028.2. Operating experience feedback and changes in cladding material.............. 78228.3. Anomalies and significant events involving fuel assemblies.......................... 78628.3.1. Baffle jetting............................................................................................... 78628.3.2. Fretting........................................................................................................ 78728.3.3. Events encountered during handling operations............................ 78928.3.4. Lateral deformation of fuel assemblies interferingwith RCCA drop......................................................................................... 791Chapter 29Facility Compliance29.1. Introduction.................................................................................................................... 79529.2. Detection and treatment of compliance deviations for pressurizedwater reactors in the nuclear power plant fleet................................................. 79629.2.1. Process for handling compliance gaps............................................... 79629.2.2. Examples of compliance gaps............................................................... 79829.2.2.1. Compliance gap in electrical connection boxesqualified for accident conditions................................... 79829.2.2.2. Failure in the seismic resistance of metal floorsin electrical and auxiliary buildings of 900 MWereactors (CPY series).......................................................... 79929.2.2.3. Risk of containment sump screen blockage............... 79929.2.2.4. Anomaly found in engines of emergency and SBOdiesel generators for 900 MWe reactors.................... 80029.2.2.5. Temperature resistance fault in the high-headsafety injection pumps...................................................... 80229.2.2.6. Mixed lubricants in equipment requiredfor accident situations....................................................... 80229.2.2.7. Flow imbalance between safety injection linesof 900 MWe reactors......................................................... 80429.2.2.8. Anomaly in CATHARE software modellingof natural circulation in the upper partof the vessel.......................................................................... 80529.2.2.9. Vibrations and rotor lift on engineeredsafety motor-driven pump units.................................... 805Chapter 30Periodic Reviews30.1. Introduction.................................................................................................................... 80730.2. History of periodic reviews in France for nuclear power reactors................ 80930.2.1. Reactors other than PWRs in the French nuclear powerplant fleet.................................................................................................... 80930.2.2. PWRs in the French nuclear power plant fleet(900 MWe, 1300 MWe and 1450 MWe).......................................... 81130.3. Periodic review process for PWRs in the French nuclear powerplant fleet........................................................................................................................ 81530.3.1. Regulations................................................................................................. 81530.3.2. Outline of a PWR periodic review....................................................... 81730.4. Case of the review associated with the third ten-yearly outageof 900 MWe reactors................................................................................................... 82230.4.1. Plant unit compliance reviews, the complementaryinvestigation and ageing management............................................. 82430.4.1.1. Plant unit compliance reviews........................................ 82430.4.1.2. Complementary investigation programme................ 82530.4.1.3. Ageing management.......................................................... 82630.4.2. Compliance studies on the design of civil works systemsand structures............................................................................................ 82630.4.3. Studies to reassess system design...................................................... 82730.4.4. Reassessment of reactor resistance to internaland external hazards................................................................................ 83030.4.5. Accident studies........................................................................................ 83230.4.6. Taking into account lessons learned during the reviewassociated with the VD3 900 outagefor subsequent reviews........................................................................... 83730.5. Fourth ten-yearly outage of 900 MWe reactors: integratingthe extension of the operating lifetime of nuclear power reactorsin France.......................................................................................................................... 83830.5.1. Background................................................................................................. 83830.5.2. Periodic Review Strategic Plan – Setting objectives..................... 83930.5.3. A few significant issues identified in reviews conductedby safety organizations........................................................................... 84230.6. Overview of international practices – IAEA Guides........................................... 84530.6.1. International practices............................................................................ 84630.6.2. IAEA Guides................................................................................................ 84730.7. Multilateral practices................................................................................................... 848Chapter 31Optimizing Radiation Protection and Limiting Doses Receivedby Workers During Operations in a Nuclear Power Plant31.1. Sources of ionizing radiation in a nuclear power reactor................................ 85231.2. Examples of optimization of worker radiation protection.............................. 85331.3. Arrangements for ‘Major Refit’ operations........................................................... 85531.4. Approach and objectives adopted for the EPR.................................................... 858 Part 4The Accidents at Three Mile Island,Chernobyl and Fukushima Daiichi Nuclear Power Plants,Lessons Learned and Emergency Response ManagementChapter 32The Three Mile Island Nuclear Power Plant Accident32.1. Accident sequence – Reconstitution through simulation................................ 86432.2. Accident consequences............................................................................................... 87232.3. Analysis of the accident causes................................................................................ 87432.3.1. Error in identifying the position of the relief valve....................... 87432.3.2. Understanding the behaviour of the pressurizer............................ 87532.3.3. Stopping safety injection....................................................................... 87632.3.4. Human-machine interface..................................................................... 87632.3.5. Isolating the reactor containment...................................................... 87732.3.6. Confinement inside the auxiliary building....................................... 87732.3.7. Emergency feedwater supply to the steam generators............... 87732.4. Lessons learned from the Three Mile Island accident....................................... 87732.4.1. The human factor in facility operation............................................. 87832.4.2. Importance of precursor events........................................................... 88132.4.3. Study of complex situations and core-melt accidents,handling emergency situations............................................................ 88232.5. Conclusions..................................................................................................................... 883Chapter 33Incident and Accident Operation:from the Event-Oriented Approach to the State-Oriented Approach33.1. Limits of the event-oriented approach.................................................................. 88533.2. The State-Oriented Approach concept.................................................................. 88633.3. First application of the state-oriented approach............................................... 88833.4. Widespread application of the state-oriented approach................................. 89133.5. ‘Stabilized’ state-oriented approach....................................................................... 89233.6. State-oriented approach adopted for the EPR.................................................... 894Chapter 34The Chernobyl Nuclear Power Plant Accident34.1. The Chernobyl nuclear power plant and RBMK reactors................................. 89734.2. The accident sequence................................................................................................ 90134.3. Analysis of the accident causes and changes made to RBMK unitssoon after the accident............................................................................................... 90634.4. The other units at the facility................................................................................... 90834.5. Radioactive release and protection of the population..................................... 90834.5.1. Radioactive release kinetics.................................................................. 90834.5.2. Protection of the population................................................................ 91134.6. Consequences on human health and the environment................................... 91534.6.1. Direct effects of radiation...................................................................... 91534.6.2. Thyroid cancer in children..................................................................... 91734.6.3. Long-term contamination in the Dnieper Basin............................. 91934.7. Radioactive fallout in France and its consequences.......................................... 92134.7.1. Doses attributable to the plume......................................................... 92134.7.2. External doses due to soil deposition................................................ 92134.7.3. Doses due to ingestion of contaminated foodstuffs.................... 92234.7.4. Overall levels.............................................................................................. 92334.7.5. Thyroid cancer........................................................................................... 92434.7.5.1. Monitoring thyroid cancer in France............................ 92534.7.5.2. Assessment of the number of cancer casesinduced in France by the Chernobyl accident........... 92634.8. Lessons learned by the international communityfrom a general viewpoint and with regard to RBMK reactors....................... 92734.9. Lessons learned in France........................................................................................... 92834.10. Keeping the public informed..................................................................................... 93134.11. After the Chernobyl accident.................................................................................... 933Chapter 35Options and Control of Reactivity Insertionin Pressurized Water Reactors35.1. Research and study of event sequences................................................................ 93735.1.1. Cooling accidents..................................................................................... 93835.1.2. Incidents and accidents related to rod clustercontrol assemblies.................................................................................... 94035.1.3. Boron dilution accidents........................................................................ 94335.1.4. Inserting a cold water plug in the core............................................. 95035.2. Changes in criteria........................................................................................................ 95135.3. The case of outage states.......................................................................................... 95335.4. Regulations...................................................................................................................... 957Chapter 36The Reactor Accident at the Fukushima Daiichi Nuclear Power Plantand Lessons Learned in France36.1. Reactor units at the Fukushima Daiichi nuclear power plant........................ 96236.1.1. General operation of a boiling water reactor.................................. 96236.1.2. Containment.............................................................................................. 96236.1.3. Emergency cooling systems.................................................................. 96436.2. Sequence of events during the accident............................................................... 96636.3. Radioactive release....................................................................................................... 97236.3.1. Airborne radioactive release, residual caesium depositsand contamination of foodstuffs........................................................ 97236.3.1.1. Airborne radioactive release............................................ 97236.3.1.2. Persistent caesium deposits............................................ 97236.3.1.3. Contamination of foodstuffs.......................................... 97336.3.2. Release of radioactive substances in the Pacific Ocean.............. 97536.3.3. Long-distance atmospheric dispersionof the radioactive plume........................................................................ 97636.4. Action taken to control facilities and released contaminated water.......... 97836.5. Socioeconomic and health impact in numbers................................................... 98136.5.1. Socioeconomic impact............................................................................ 98136.5.2. Health impact............................................................................................ 98236.6. Lessons learned from the accident.......................................................................... 98536.6.1. Complementary safety assessments carried out in Europeand France following the Fukushima Daiichi nuclear powerplant accident............................................................................................ 98636.6.2. Complementary safety assessments carried out in France........ 98836.6.3. Procedure for complementary safety assessmentscarried out in France................................................................................ 98936.6.4. Conclusions of the complementary safety assessmentscarried out in France................................................................................ 99036.6.5. The ‘hardened safety core’.................................................................... 99136.6.5.1. Purpose................................................................................... 99136.6.5.2. Principles................................................................................ 99136.6.5.3. Illustrations........................................................................... 99336.6.6. Nuclear Rapid Response Force (FARN).............................................. 99536.6.7. Deployment of post-Fukushima measuresin French nuclear power plants............................................................ 99736.7. Other lessons learned in France from the Fukushima Daiichi nuclearpower plant accident................................................................................................... 998Chapter 37Lessons Learned from the Fukushima Daiichi Nuclear PowerPlant Accident: Work Conducted by the IAEA and WENRA,Action Taken in Countries Other than France37.1. Work conducted by the IAEA.................................................................................... 100237.2. Work conducted by WENRA..................................................................................... 100337.3. Japan................................................................................................................................. 100437.4. Belgium............................................................................................................................ 100637.4.1. Nuclear power plants in Belgium........................................................ 100637.4.2. General details on the design of Belgian nuclear powerplants............................................................................................................ 100637.4.3. Stress tests and main lessons learned............................................... 100737.4.3.1. Improving protection of facilities againstexternal hazards.................................................................. 100737.4.3.2. Improving protection of facilities against lossof electrical power supplies or loss of heat sink....... 1010Contents XXXV37.4.3.3. Improving on-site emergency plans............................. 101137.4.3.4. Improving management of core-melt accidents...... 101237.5. USA.................................................................................................................................... 1013Chapter 38Emergency Preparedness and Response38.1. Defining a radiological emergency and ‘response’ objectives........................ 102138.2. General organization of radiological emergency management..................... 102338.2.1. Organization and entities concerned................................................. 102338.2.2. Major Nuclear or Radiological Accident National ResponsePlan and emergency plans..................................................................... 102538.2.2.1. Major Nuclear or Radiological Accident NationalResponse Plan...................................................................... 102538.2.2.2. Emergency plans................................................................. 102638.2.2.3. Provisions for protecting the public in the eventof an accidental release of radioactivity..................... 102838.3. Management by the operator................................................................................... 102938.4. Prefectural authorities and mayors......................................................................... 103138.5. ASN, the Nuclear Safety Authority......................................................................... 103238.6. IRSN................................................................................................................................... 103338.7. Assessment approach in the event of an accident affecting a reactorin the nuclear power plant fleet.............................................................................. 103838.7.1. ‘3D/3P’ method......................................................................................... 103938.7.2. The ‘aggravated prognosis’ approach................................................ 104138.7.3. Extending the 3D/3P method to severe accidents(the ‘D/P AG’ method)............................................................................ 104138.8. Emergency preparedness............................................................................................ 104238.8.1. Emergency response exercises............................................................. 104338.8.2. Operating experience feedback........................................................... 1045 Part 5PWR Safety Studies, R&D and Simulation SoftwareChapter 39PWR Safety Studies and R&D39.1. Contribution of studies to the improvement of pressurized waterreactor safety................................................................................................................. 105039.2. Purpose and overview of R&D work, dedicated programmesand organizations involved, and research facilities in France........................ 105239.2.1. Purpose and overview of R&D............................................................. 105239.2.2. Dedicated frameworks and organizations involved....................... 106339.2.3. Facilities in France used for research and development.............. 1065Chapter 40Examples of Simulation Software Developed for Safety Analysisof Pressurized Water Reactors40.1. Simulation software for neutronics........................................................................ 107640.2. Simulation software for thermal hydraulics (and mechanics)....................... 107940.3. Simulation software for thermal mechanics........................................................ 108440.4. Software for simulating core-melt situations..................................................... 108540.5. Simulation software for mechanics........................................................................ 108840.6. Fire simulation software............................................................................................. 1089List of Acronyms.......................................................................................................................... 1093Acronyms for institutions, bodies and groups..................................................................... 1093Technical acronyms and abbreviations.................................................................................. 1103Technical glossary......................................................................................................................... 1121 21 00 06 01 https://laboutique.edpsciences.fr/produit/1280/9782759827237/elements-of-nuclear-safety 01 00 03 02 https://laboutique.edpsciences.fr/system/product_pictures/data/009/982/690/original/9782759827220-PressurizedWaterReactors_couv-sofedis.jpg 17 14 20240913T180957+0200 01 P1 EDP Sciences 01 01 P1 EDP Sciences 17 avenue du Hoggar - PA de Courtaboeuf - 91944 Les Ulis cedex A http://publications.edpsciences.org/ 04 11 00 20220707 01 00 20220707 19 00 20220825 2026 06 3052868830012 01 WORLD 13 03 9782759827220 15 9782759827220 06 03 9782759827244 15 9782759827244 WORLD 08 EDP Sciences 04 01 00 20220707 03 01 D1 EDP Sciences 20 04 05 01 02 1 00 01 EUR laboutique.edpsciences.fr-R001723 03 01 01 R001723 00 ED E101 00 01 01 ELEMENTS OF NUCLEAR SAFETY PRESSURIZED WATER REACTORS 01 eng 22 30909937 17 01 P1 EDP Sciences 01 01 P1 EDP Sciences 17 avenue du Hoggar - PA de Courtaboeuf - 91944 Les Ulis cedex A http://publications.edpsciences.org/ 11 00 20220707 02 01 002116 03 9782759827244 15 9782759827244 03 01 D1 EDP Sciences 29 https://laboutique.edpsciences.fr/open_access_download/2116/1723 45 03 laboutique.edpsciences.fr-002116 03 01 01 002116 03 9782759827244 15 9782759827244 10 EA E101 10 00 01 R001723 ED E101 1 10 01 02 Institut de Radioprotection et de Sûreté Nucléaire 01 01 ELEMENTS OF NUCLEAR SAFETY PRESSURIZED WATER REACTORS 1 B15 01 A1684 Jean Couturier Couturier, Jean Jean Couturier 01 eng 08 1170 03 01 24 Izibook:Subject Radioprotection 24 Izibook:SubjectAndCategoryAndTags |Radioprotection| 20 pressurized water reactor safety;nuclear power;nuclear power plant;safety;reactor;radiological emergency;nuclear emergency;safety analysis method;nuclear safety;nuclear risk control;radiological risk control 10 2011 BISACSCI058000 01 620 06 06 05 03 00 Everything that is important to know about pressurized water reactor safety in nuclear power plants is compiled in this work of reference, from safety fundamentals and reactor design to provisions for managing a radiological or nuclear emergency. The book narrates the determined and continuous quest to enhance nuclear safety. Jean Couturier, coordinator and senior editor, covers the forty-year history of evolution in the safety objectives, approaches, analysis methods and assessment criteria that have defined pressurized water reactor safety, mainly within the French nuclear power plant fleet, from the 1970s up to today’s Flamanville 3 EPR. Everyone, including current and future generations of engineers, researchers and, more broadly, any citizen interested in nuclear safety issues will find that the book reinforces their knowledge on this important subject, providing an understanding of the fundamentals, to the benefit of nuclear and radiological risk control. 02 00 Everything that is important to know about pressurized water reactorsafety in nuclear power plants is compiled in this work of reference, fromsafety fundamentals and reactor design to provisions for managing aradiological or nuclear emergency. 04 00 ContentsPreface............................................................................................................................................. IIIForeword......................................................................................................................................... VIIEditors, Contributors and Reviewers................................................................................... XXXVIIIntroduction................................................................................................................................... XLIIIPart 1General BackgroundChapter 1Biological and Health Effects of Ionizing Radiation – The Radiological Protection System1.1. Biological and health effects of ionizing radiation............................................. 41.1.1. Biological processes................................................................................. 41.1.2. Review of units of measure................................................................... 61.1.3. Natural radioactivity............................................................................... 71.1.4. Health effects............................................................................................ 81.1.4.1. Deterministic effects, tissue reactions......................... 81.1.4.2. Stochastic or random effects.......................................... 91.1.4.3. Induction of diseases other than cancer..................... 111.1.5. Example of the limitations of epidemiology................................... 111.2. Radiological protection system................................................................................ 131.2.1. Types of exposure situations................................................................ 14 1.2.2. Exposure categories................................................................................. 141.2.3. Justification principle............................................................................... 151.2.4. Optimization (ALARA) principle ......................................................... 161.2.5. Principle of application of dose limits............................................... 21Chapter 2Organization of Nuclear Safety Control and Regulation for Nuclear Facilities and Activities in France2.1. From the founding of CEA to the TSN Act........................................................... 232.2. A few definitions........................................................................................................... 282.3. The different contributors to nuclear safety and their missions.................. 302.4. A few basic principles and notions in the field of nuclear safety................. 462.5. Statutory and quasi-statutory frameworks applicable to basic nuclear installations.............................................. 48Chapter 3The International Dimension and the Social Dimension3.1. International dimension.............................................................................................. 773.1.1. Introduction................................................................................................ 773.1.2. IAEA standards........................................................................................... 803.1.3. International Reporting System for Operating Experience (IRS).......................................................... 823.1.4. Services developed by the IAEA........................................................... 843.1.4.1. OSART reviews.................................................................... 843.1.4.2. IRRS reviews......................................................................... 873.1.4.3. Other services and study frameworks set up by the IAEA.............................................................. 883.1.5. WANO.......................................................................................................... 903.1.6. NEA............................................................................................................... 913.1.7. Organizations dedicated to radiation protection and health........................................................................ 933.1.8. From bilateral Franco-German cooperation to European structures for the exchange and capitalization of knowledge and practices, training and assessment services........................................................ 943.1.9. Nuclear regulator associations............................................................. 1003.2. The social dimension................................................................................................... 1023.2.1. Introduction – the context in France................................................. 1023.2.2. Examples of initiatives and issues raised concerning reactor safety in the French nuclear power plant fleet........... 102Chapter 4Nuclear Reactors: Complex Sociotechnical Systems – the Importance of Human and Organizational Factors4.1. 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.............................. 1084.2. The accident at the Chernobyl nuclear power plant and the concept of ‘safety culture’................................. 1094.3. The Fukushima Daiichi nuclear power plant accident: the social dimension and the concept of organization ‘resilience’... 1134.4. Changes in the perception of the role of people in achieving a high level of reliability in complex sociotechnical systems...... 1134.5. Main topics studied in the development of resources and skills pertaining to human and organizational factors............ 1164.5.1. Resources and skills................................................................................. 1164.5.2. Main topics studied.................................................................................. 1184.6. Human and organizational factors in French regulations................................ 119 Part 2Safety by DesignChapter 5The Development of Nuclear Power Using Uranium-235 Fission – A Few Notions of Physics Used in Pressurized Water Reactors5.1. Important milestones in the development of nuclear power using fission of the uranium-235 isotope.......................... 1235.2. Fission and important concepts in reactor kinetics........................................... 1275.3. Removing power from the core during operation............................................. 1345.4. Decay heat...................................................................................................................... 1355.5. Main features of pressurized water reactor cores.............................................. 1365.6. Control and monitoring of pressurized water reactor cores.......................... 1375.7. Using uranium and plutonium mixed oxide (MOX) fuel................................. 147Chapter 6General Objectives, Principles and Basic Concepts of the Safety Approach6.1. General approach to risks – General objectives................................................. 1526.2. Fundamental safety functions.................................................................................. 1566.3. Confinement barriers................................................................................................... 1576.4. Defence in depth........................................................................................................... 1616.4.1. Levels of defence in depth..................................................................... 1626.4.2. Elements common to the different levels of defence in depth............................................................................. 1696.5. Events considered: terminology adopted for nuclear power reactors......... 1696.6. WENRA reference levels............................................................................................. 1716.7. Deterministic safety analysis and probabilistic safety assessments............ 1716.8. Lessons learned from the accident at the Fukushima Daiichi nuclear power plant on the concept of defence in depth and deterministic analysis......................................................................................... 1736.9. Safety culture – Quality control.............................................................................. 174Chapter 7Safety Options and Considerations at the Design Phase7.1. Different types of design provisions associated with safety considerations................................................. 1887.2. Single-failure criterion................................................................................................. 1897.3. The specific nature of computer-based systems (based on instrumentation and control software)............... 1937.4. Equipment safety classification................................................................................ 1957.4.1. Importance of equipment for safety and safety classification...................................................................... 1957.4.2. Generic requirements associated with the different safety classes.................................................................. 1997.4.3. Qualification of equipment for accident conditions..................... 2027.5. Information on designing nuclear pressure equipment.................................... 2077.6. General considerations on provisions for hazards in facility design............ 2107.7. Anticipating decommissioning in the design stage........................................... 212Chapter 8Study of Operating Conditions in the Deterministic Safety Analysis8.1. Categories of operating conditions......................................................................... 2188.2. Choice of operating conditions................................................................................ 2228.2.1. Concept of ‘bounding’ incident or accident..................................... 2238.2.2. Accident exclusion.................................................................................... 2248.3. List and breakdown of operating conditions....................................................... 2258.4. Methods for studying operating conditions......................................................... 2278.4.1. Choice of initial conditions, conservatism....................................... 2288.4.2. Consideration of an aggravating event in the study on operating conditions – ‘Passive’ failures............ 2298.4.3. Conventional combinations.................................................................. 2318.4.4. Preventing accident aggravation......................................................... 2328.4.5. Operator response time......................................................................... 2328.4.6. Using qualified simulation software................................................... 2338.4.7. Main criteria to be met for fuel in the reactor core..................... 2348.5. Concept of ‘design-basis situations’ for equipment.......................................... 2368.6. Situations to be taken into account in application of pressure equipment regulations................................................. 2378.7. Assessing the radiological consequences of incidents, accidents and hazards.............................................. 2388.7.1. Assessing radioactive substances released from the facility...... 2408.7.2. Assessing radiological consequences of radioactive release from the facility................................................ 2428.7.3. Assessing radiological consequences................................................. 243Chapter 9Loss-of-Coolant Accident9.1. Short- and medium-term aspects of a LOCA...................................................... 2529.1.1. Mechanical effects on vessel internals and fuel assembly structures................................................................ 2539.1.2. Thermal-hydraulic aspects and behaviour of fuel rods................ 2559.1.2.1. Large-break LOCA............................................................... 2559.1.2.2. Intermediate-break LOCA................................................ 2579.1.3. Effects on reactor containment and internals................................ 2589.1.4. Long-term aspect...................................................................................... 2599.2. Safety demonstration.................................................................................................. 2619.2.1. General information and background................................................ 2619.2.2. Fuel assemblies and fuel rods, vessel internals, reactor coolant system components................................................. 2639.2.2.1. Mechanical strength of vessel internals, fuel assembly structures and reactor coolant system components.............. 2639.2.2.2. Fuel behaviour...................................................................... 2659.2.3. Reactor containment and equipment located inside.................... 266Chapter 10A Special Issue: Steam Generator Tubes10.1. Steam generator tube rupture as a Category 3 event...................................... 27010.2. Preventing an SGTR accident, risk of multiple ruptures................................... 27210.3. Steam generator tube rupture(s) studied as a Category 4 event................. 27410.3.1. 900 MWe and 1300 MWe reactors.................................................... 27410.3.2. 1450 MWe reactors and EPR (Flamanville 3).................................. 27410.4. Provisions to mitigate the radiological consequences of SGTR accidents. 276Chapter 11Providing for Hazards: General Considerations and Internal Hazards11.1. General considerations on providing for hazards............................................... 27911.2. Potential projectiles inside the containment....................................................... 28211.3. Effects of pipe breaks.................................................................................................. 28411.4. Projectiles generated by a turbine rotor failure.................................................. 28511.5. Protection against load drops................................................................................... 28811.5.1. Risks related to spent fuel transport packaging............................. 28811.5.2. Other handling risks................................................................................ 29111.6. Fire protection............................................................................................................... 29211.7. Explosion protection.................................................................................................... 29711.8. Internal flooding............................................................................................................ 300Chapter 12Providing for External Hazards12.1. General considerations on providing for external hazards.............................. 30512.2. ‘Climate watch’ implemented by EDF.................................................................... 30712.3. Earthquakes..................................................................................................................... 30712.4. External floods............................................................................................................... 32212.5. Extreme temperatures................................................................................................. 33012.5.1. Extreme cold.............................................................................................. 33012.5.2. Extreme heat.............................................................................................. 33112.6. Possible heat sink hazards.......................................................................................... 33212.7. Other naturally-occurring external hazards......................................................... 33612.8. Accidental aeroplane crashes (excluding malicious acts)................................ 33612.9. Risks related to the industrial environment (excluding malicious acts)..... 340Chapter 13Complementary Domain of Events13.1. The origin of studies belonging to the complementary domain................... 34413.2. Background of the complementary domain........................................................ 34413.3. Analysis of complementary domain events......................................................... 35113.4. ‘New complementary domain’................................................................................. 35113.5. Case of the Flamanville 3 EPR.................................................................................. 354Chapter 14Development and Use of Probabilistic Safety Assessments14.1. History and regulatory context................................................................................ 35714.1.1. International situation............................................................................ 35714.1.2. Situation in France................................................................................... 35914.2. Level 1 PSA...................................................................................................................... 36114.2.1. Scope............................................................................................................ 36114.2.2. Method for carrying out a Level 1 PSA............................................. 36214.2.2.1. General information........................................................... 36214.2.2.2. Specific point: probabilistic human reliability analysis................................................................................... 36414.2.3. Level 1 PSA results and lessons learned............................................ 36914.3. Level 2 PSA...................................................................................................................... 37314.3.1. Scope............................................................................................................ 37314.3.2. Method for carrying out a Level 2 PSA............................................. 37414.3.2.1. General information........................................................... 37414.3.2.2. Probabilistic human reliability analysis for Level 2 PSAs................................................................... 37914.3.3. Examples of lessons learned from Level 2 PSAs............................. 38214.3.3.1. Steam explosion risk assessment.................................. 38214.3.3.2. Mechanical integrity of the 900 MWe reactor containments........................................................ 38314.3.3.3. Isolating penetrations in the containment................. 38314.3.3.4. Modifying the pressure relief system of the reactor coolant system........................................ 38414.3.3.5. Improvement of operating procedures to reduce risk of core melt under pressure................ 38414.3.3.6. Contribution of Level 2 PSAs to emergency response measures............................................................. 38514.4. Expanding the scope of PSA coverage................................................................... 38514.5. Using probabilistic safety assessments.................................................................. 38614.5.1. Using PSAs in the design phase........................................................... 38614.5.1.1. Usefulness and particularities of PSAs in the design phase............................................................. 38614.5.1.2. PSAs conducted to support the Flamanville 3 EPR design............................................................................. 38714.5.2. Using PSAs in periodic reviews............................................................ 38914.5.2.1. Level 1 PSA............................................................................ 39014.5.2.2. Level 2 PSA............................................................................ 39114.5.3. Using PSAs for reactor operation........................................................ 39214.5.3.1. Using PSAs to analyse event severity.......................... 39214.5.3.2. Using PSAs to analyse operational limits and conditions and temporary changes...................... 39414.5.3.3. Using PSAs to analyse operating procedures............. 396Chapter 15Aspects Specific to PWR Spent Fuel Storage Pools15.1. Spent fuel pool design................................................................................................. 39915.1.1. Confinement barriers.............................................................................. 39915.1.2. Initiating events defined at the design stage.................................. 40015.2. Experience feedback..................................................................................................... 40115.2.1. Loss of cooling........................................................................................... 40115.2.1.1. Loss of heat sink.................................................................. 40115.2.1.2. Risks related to maintenance during unit outages................................................................................... 40215.2.1.3. Suction of foreign matter into the cooling system.................................................................................... 40315.2.1.4. Exceeding the decay heat defined in facility design...................................................................................... 40315.2.2. Water losses............................................................................................... 404 15.2.2.1. Gate or sluice gate failures.............................................. 40415.2.2.2. Line-up errors....................................................................... 40515.2.2.3. Failure of a reactor coolant system pipe nozzle dam.......................................................................................... 40915.2.2.4. Rupture of a pipe connected to the spent fuel pool................................................................................. 41115.3. Safety reassessments................................................................................................... 41115.4. Experience feedback from the accident that affected the Unit 4 pool at the Fukushima Daiichi nuclear power plant..... 41415.4.1. Events........................................................................................................... 41415.4.2. Complementary safety assessments conducted in France......... 41715.5. Measures adopted for the EPR.................................................................................. 41915.6. Recommendations for new reactor designs......................................................... 42015.7. New systems for storing spent fuel........................................................................ 423Chapter 16Taking into Account Human and Organizational Factors in Facility Design16.1. Taking into account human and organizational factors in nuclear power reactor design....................................... 42516.1.1. Importance of considering human and organizational factors at the design stage....................................... 42516.1.2. Approach at the design stage............................................................... 43116.1.2.1. Prior to the design phase: analysis of ‘existing elements’........................................ 43216.1.2.2. Design objectives................................................................ 43516.1.2.3. Definition of detailed design provisions...................... 43616.1.2.4. Validation of design provisions...................................... 43916.1.2.5. Assessments conducted during reactor startup and after commissioning.................................................. 44216.1.3. Project management and human and organizational factors engineering programme..................................... 44316.2. Considering human and organizational aspects when designing changes to nuclear power plants..................... 44316.2.1. Importance of human and organizational factors in designing modifications..................................................................... 44416.2.2. ‘Human, social and organizational approach’ implemented by EDF........................................................................... 44516.2.3. Changes, a subject that always deserves special attention from a human and organizational factors perspective.... 44716.3. Human and organizational factors for future nuclear power reactor projects.................................................. 448Chapter 17Studying Core-Melt Accidents to Enhance Safety17.1. Core degradation and vessel failure....................................................................... 45217.1.1. Core uncovery............................................................................................ 45317.1.2. Fuel degradation....................................................................................... 45417.1.3. Failure of the reactor coolant system............................................... 45517.1.4. Phenomena that can cause early containment failure................. 45517.1.5. Phenomena that can ultimately lead to containment failure....................................................... 45717.2. Containment failure modes....................................................................................... 45717.3. Classification of releases associated with core-melt accidents − ‘source terms’................................................. 46017.4. Improving knowledge.................................................................................................. 46217.5. Studies in France on containment failure modes............................................... 46217.5.1. Introduction................................................................................................ 46217.5.2. Initial containment leakage................................................................... 46317.5.3. Direct heating of gases in the containment.................................... 46317.5.4. Hydrogen explosion in the containment.......................................... 46417.5.5. Steam explosion in the vessel or reactor pit................................... 46517.5.6. Gradual pressure increase in the containment............................... 46617.5.7. Penetration of the concrete basemat of the containment by corium............................................ 46717.5.8. ‘U4’ provisions........................................................................................... 46817.5.9. Bypass of containment by outgoing pipes (the V mode).............................................................................................. 46817.5.10. Fast reactivity insertion accidents...................................................... 46917.6. Severe accident operating guidelines..................................................................... 46917.7. Radiological consequences associated with the S3 source term and emergency response plans implemented by public authorities............ 47017.8. Ultimate emergency operating procedures.......................................................... 47317.9. On-site emergency plan............................................................................................. 47317.10. Approach adopted for the EPR................................................................................. 47617.10.1. General safety objectives....................................................................... 47617.10.2. ‘Practical elimination’ of core-melt conditions that could lead to significant early releases............................. 47617.10.3. Provisions for low-pressure core melt............................................... 480Chapter 18New-Generation Reactors18.1. Organization and framework of Franco-German discussions........................ 48518.2. Progression of safety objectives and design options for the EPR project.......................................................... 48618.2.1. General safety objectives....................................................................... 48618.2.2. Events to be taken into account at the design stage and in deterministic and probabilistic analyses.............................. 48818.2.3. Main provisions for preventing incidents and accidents............. 49018.2.4. Functional redundancy, independence between systems, system reliability................................................ 49518.2.5. Confinement preservation..................................................................... 49618.2.6. Radiological protection........................................................................... 49718.2.7. Incorporating lessons learned from the Fukushima Daiichi nuclear power plant accident......................... 49718.3. International context: general safety objectives for new-generation reactors............................................ 49918.4. Concepts highlighted in new reactor designs...................................................... 50118.4.1. AP1000: gravity systems........................................................................ 50118.4.2. VVER: SPOT system................................................................................. 50318.4.3. NM EPR: ‘multi-group’ technology, diversified heat sink............ 504 18.4.4. ATMEA 1: safety injection accumulators in the reactor coolant system.............................................................. 50518.4.5. NuScale: common pool for modular reactors................................. 505 Part 3Safety in OperationChapter 19Startup Tests for Pressurized Water Reactors19.1. Introduction.................................................................................................................... 51119.2. Commissioning.............................................................................................................. 51419.2.1. Defining startup tests.............................................................................. 51419.2.2. Phasing of startup tests.......................................................................... 51519.2.2.1. Preliminary and pre-operational tests......................... 51519.2.2.2. Operational tests................................................................ 51719.2.2.3. General principles for test sequencing and execution....................................................................... 51719.2.3. Documentation for startup tests......................................................... 51819.2.3.1. Integrated system test procedures and startup test procedures............................................ 51819.2.3.2. Test programmes, test procedures, standard test guidelines.................................................... 51819.2.3.3. Completeness analysis, adequacy analysis................. 51819.2.3.4. Acceptance criteria............................................................. 51919.3. Objectives and general rules to take into account for startup tests........... 52019.4. Key lessons learned from startup tests on nuclear power reactors in France............................................................ 52119.4.1. Qualification tests and on-site tests.................................................. 52219.4.2. Long-term on-site testing...................................................................... 52419.4.3. Test configurations and completeness, transpositions................ 525 19.4.4. Safety measures that cannot be verified by testing..................... 52719.4.5. Criteria......................................................................................................... 52719.4.6. Cleanness, keeping system lines clean, foreign matter................ 52819.4.7. Piping support structures and displacement................................... 53219.4.8. Pump and piping vibrations................................................................... 53219.4.9. Validation of operating procedures and periodic tests................ 53419.4.10. Uncertainty and ‘set points’.................................................................. 53519.4.11. Condition of facilities during startup tests...................................... 53619.4.12. Other aspects............................................................................................. 53719.5. Examples of findings resulting from startup tests............................................. 538Chapter 20General Operating Rules20.1. General Operating Rules............................................................................................ 55220.1.1. Content of general operating rules..................................................... 55320.1.2. Limits of general operating rules......................................................... 55420.2. Operational limits and conditions........................................................................... 55420.2.1. Content of operational limits and conditions................................. 55520.2.1.1. Operating modes and standard states......................... 55620.2.1.2. Requirements and unavailability................................... 55820.2.1.3. Fallback states and time required to reach them.... 55820.2.1.4. Events and event groups.................................................. 55920.2.1.5. Combined types of unavailability.................................. 56020.2.1.6. Concepts of ‘boundary condition’ and ‘specific requirement’................................................ 56020.2.2. Average pressure and temperature range of the reactor coolant system.............................................................. 56120.2.3. Changes in operational limits and conditions................................. 56320.3. Initial and periodic tests............................................................................................. 56420.4. Incident and accident operating procedures........................................................ 566Chapter 21 Operating Experience Feedback from Events: Rules and Practices21.1. Background...................................................................................................................... 56921.2. Objectives of an operating experience feedback system................................ 57121.3. Components of an operating experience feedback system – Regulations............................................................................... 57221.4. Operating experience feedback practices adopted for the French nuclear power plant fleet................................. 576Chapter 22Operating Experience from Events Attributable to Shortcomings in Initial Reactor Design or the Quality of Maintenance22.1. 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).......................................... 59022.2. Recurrent loss of safety function events related to maintenance operations – Lessons learned........................... 59522.2.1. Events........................................................................................................... 59522.2.2. General discussion initiated by EDF in the late 1980s on the quality of maintenance operations....................................... 60522.2.3. Applying the defence-in-depth concept when working on a reactor in service.................................. 60822.2.4. Problems that may recur....................................................................... 609Chapter 23Operating Experience from Events Related to Maintenance Operations, Electrical Power Sources and Distribution, Internal and External Hazards23.1. Risks of failure related to equipment or maintenance..................................... 61223.1.1. Risks of common-mode failure............................................................ 61223.1.1.1. Risks of common-mode failure related to settings............................................................................. 61223.1.1.2. Risks of common-mode failure on electrical switchboards.................................................61323.1.1.3. Unavailability of two out of three high-head safety injection lines in the cold legs of the reactor coolant system.............. 61423.1.1.4. Loss of electrical power supplies................................... 61723.1.2. Introduction of non-borated water into the reactor coolant system.......................................................... 62023.1.3. Cooling the reactor coolant system after inhibition of automatic actions............................................................................... 62223.1.4. A temporary device prevents switching the safety injection system to the water recirculation mode................... 62423.2. Events related to internal hazards........................................................................... 62523.2.1. Risk of common-mode failure due to internal flooding.............. 62523.2.2. Risk of failure due to fire....................................................................... 62823.2.3. Risks associated with the use of hydrogen in 900 MWe reactors............................................................................... 63123.3. External hazards: events related to periods of extreme cold......................... 635Chapter 24Enhanced Protection of Estuary and River Sites: Flooding at the Blayais Nuclear Power Plant and Obstruction of a Water Intakeat the Cruas-Meysse Nuclear Power Plant24.1. Partial loss of engineered safety systems following flooding of the Blayais nuclear power plant............................... 64224.2. 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...................................................... 648Chapter 25Taking into Account Human and Organizational Factors in Facility Operation25.1. Skills management....................................................................................................... 65325.1.1. Historical background............................................................................. 65425.1.2. Managing training.................................................................................... 65525.1.3. Strategic workforce planning................................................................ 65625.1.4. Personnel certification............................................................................ 65725.2. Safety and risk management.................................................................................... 65825.2.1. Historical background............................................................................. 65825.2.2. Decision-making and safety.................................................................. 66025.2.3. Risk analyses applied to work activities............................................ 66225.2.4. Operating experience feedback........................................................... 66325.2.5. Managing organizational change......................................................... 66425.3. Managing operational activities............................................................................... 66625.3.1. Characteristics of operational activities............................................ 66625.3.2. Monitoring by the operating crew in the control room.............. 66925.3.3. Compliance with general operating rules......................................... 67025.3.4. Line-up......................................................................................................... 67125.3.5. Operation in extreme situations......................................................... 67225.4. Management of maintenance activities................................................................ 67225.4.1. Management of a scheduled reactor outage for refuelling and maintenance...................................................... 67225.4.2. Risks during reactor outages................................................................. 67325.4.3. Preparation for scheduled reactor outages...................................... 67425.4.4. Managing scheduled reactor outages................................................ 67525.5. Supervising outsourced activities............................................................................ 67525.5.1. Contractor qualification and contracting......................................... 67625.5.2. Matching workload and resources...................................................... 67625.5.3. Carrying out work.................................................................................... 67725.5.4. Surveillance of outsourced activities................................................. 67825.5.5. Operating experience feedback and assessing outsourced activities...................... 678Chapter 26Facility Maintenance26.1. Maintenance objectives.............................................................................................. 68126.2. Maintenance................................................................................................................... 68226.2.1. Definition.................................................................................................... 68226.2.2. Maintenance strategies........................................................................... 68326.3. Optimizing maintenance............................................................................................ 68426.3.1. Reliability-centred maintenance.......................................................... 68426.3.2. Conditional maintenance....................................................................... 68626.3.3. Conditional maintenance by sampling – Maintenance based on reference equipment items...................... 68826.3.4. ‘AP-913’ method....................................................................................... 68926.4. Maintenance baselines................................................................................................ 69126.5. On-site maintenance................................................................................................... 69426.5.1. The various stages of maintenance operations.............................. 69426.5.2. Main conditions for successful maintenance.................................. 69626.5.3. Examples of anomalies or deviations discovered during routine maintenance, explained by an inadequate maintenancebaseline with regard to deterioration mechanisms....................... 70526.5.4. Examples of events associated with non-quality maintenance.............................................. 70826.5.4.1. Example of an event explained by an incorrect setting on redundant equipment................................... 70826.5.4.2. Example of an event explained by an incorrect setting of electrical protection thresholds................. 70826.5.4.3. Examples of events explained by a failure to return equipment to a compliant state after maintenance or an error in performing a work procedure................................................................ 709Chapter 27In-service Monitoring and Inspection of Equipment27.1. Main internal equipment items on a pressurized water reactor vessel...... 71927.1.1. ‘Core baffle’ around the core................................................................ 72027.1.2. RCCA guide tubes..................................................................................... 72227.2. Reactor vessel, nozzles and head............................................................................. 72227.2.1. Vessel underclad defects........................................................................ 72427.2.2. Cracking on vessel head adaptors....................................................... 72527.2.2.1. Condition of other reactors............................................. 72727.2.2.2. Impact on safety................................................................. 72827.2.2.3. Prevention, monitoring and mitigation....................... 72927.2.2.4. Developing inspection tools............................................ 72927.2.2.5. Repairs.................................................................................... 73027.2.2.6. Leak detection...................................................................... 73027.2.2.7. Anti-ejection devices......................................................... 73127.2.2.8. Current situation................................................................. 73127.2.2.9. Cracks observed on reactor vessel heads in other countries............................................................... 73127.2.2.10. Implementation of special monitoring for ‘Inconel areas’ beginning in 1992........................... 73227.2.3. Cracking on vessel lower head penetrations detected in 2011.......................................................... 73327.2.4. Monitoring the ‘beltline’ region of the vessel................................. 73527.2.5. Defects observed on reactor vessels in Belgium............................ 73627.3. Steam generators.......................................................................................................... 73827.3.1. The different types of defects.............................................................. 73827.3.2. Associated risks......................................................................................... 74027.3.3. Monitoring during operation and inspection during outages.... 74127.3.3.1. Monitoring during operation........................................... 74127.3.3.2. Inspection during reactor outages................................. 74227.3.4. Steps to be taken when a defect is detected.................................. 74227.3.4.1. Tube wear due to foreign matter.................................. 74327.3.4.2. Wear due to contact with anti-vibration bars.......... 74327.3.4.3. Cracking in U-bend tubes................................................. 74427.3.4.4. Tube deformation and cracking..................................... 74427.3.5. Steam generator replacement.............................................................. 74527.3.6. Clogging observed in the 2000s.......................................................... 74627.3.7. Conclusion.................................................................................................. 74827.4. Steam lines...................................................................................................................... 74827.5. Auxiliary systems: cracks induced by local thermal-hydraulic phenomena................................................................................. 75027.5.1. Cracking in non-isolatable sections connected to the reactor coolant loops................................................................. 75027.5.2. RHRS thermal fatigue at Unit 1 of the Civaux nuclear power plant......................................... 75127.6. Civil works: containment structures....................................................................... 75527.6.1. Anticipated degradation phenomena................................................. 75627.6.2. Devices for direct monitoring of containment building concrete walls......................................... 75727.6.3. Leak tests and measurements.............................................................. 75927.6.4. Main anomalies......................................................................................... 759Chapter 28Fuel Management, Monitoring and Developments28.1. Procedures for monitoring fuel rod integrity....................................................... 76828.1.1. Radiochemical specifications for reactor coolant.......................... 76828.1.2. Inspections and measurements carried out directly on fuel assemblies.............................................. 77728.1.2.1. Liquid penetrant testing in the refuelling machine mast....................................................................... 77828.1.2.2. Liquid penetrant testing in the FB cell......................... 77928.1.2.3. Inspections performed on fuel rods.............................. 78028.2. Operating experience feedback and changes in cladding material.............. 78228.3. Anomalies and significant events involving fuel assemblies.......................... 78628.3.1. Baffle jetting............................................................................................... 78628.3.2. Fretting........................................................................................................ 78728.3.3. Events encountered during handling operations............................ 78928.3.4. Lateral deformation of fuel assemblies interfering with RCCA drop.......................... 791Chapter 29Facility Compliance29.1. Introduction.................................................................................................................... 79529.2. Detection and treatment of compliance deviations for pressurized water reactors in the nuclear power plant fleet............... 79629.2.1. Process for handling compliance gaps............................................... 79629.2.2. Examples of compliance gaps............................................................... 79829.2.2.1. Compliance gap in electrical connection boxes qualified for accident conditions................................... 79829.2.2.2. Failure in the seismic resistance of metal floors in electrical and auxiliary buildings of 900 MWe reactors (CPY series). 79929.2.2.3. Risk of containment sump screen blockage............... 79929.2.2.4. Anomaly found in engines of emergency and SBO diesel generators for 900 MWe reactors.................... 80029.2.2.5. Temperature resistance fault in the high-head safety injection pumps...................................................... 80229.2.2.6. Mixed lubricants in equipment required for accident situations....................................................... 80229.2.2.7. Flow imbalance between safety injection lines of 900 MWe reactors......................................................... 80429.2.2.8. Anomaly in CATHARE software modelling of natural circulation in the upper part of the vessel...................................... 80529.2.2.9. Vibrations and rotor lift on engineered safety motor-driven pump units.................................... 805Chapter 30Periodic Reviews30.1. Introduction.................................................................................................................... 80730.2. History of periodic reviews in France for nuclear power reactors................ 80930.2.1. Reactors other than PWRs in the French nuclear power plant fleet......................................................... 80930.2.2. PWRs in the French nuclear power plant fleet (900 MWe, 1300 MWe and 1450 MWe).......................................... 81130.3. Periodic review process for PWRs in the French nuclear power plant fleet..................................................................... 81530.3.1. Regulations................................................................................................. 81530.3.2. Outline of a PWR periodic review....................................................... 81730.4. Case of the review associated with the third ten-yearly outage of 900 MWe reactors..................................................... 82230.4.1. Plant unit compliance reviews, the complementary investigation and ageing management............................................. 82430.4.1.1. Plant unit compliance reviews........................................ 82430.4.1.2. Complementary investigation programme................ 82530.4.1.3. Ageing management.......................................................... 826 30.4.2. Compliance studies on the design of civil works systems and structures............................................. 82630.4.3. Studies to reassess system design...................................................... 82730.4.4. Reassessment of reactor resistance to internal and external hazards................................................................................ 83030.4.5. Accident studies........................................................................................ 83230.4.6. Taking into account lessons learned during the review associated with the VD3 900 outage for subsequent reviews....83730.5. Fourth ten-yearly outage of 900 MWe reactors: integrating the extension of the operating lifetime of nuclear power reactorsin France.......................................................................................................................... 83830.5.1. Background................................................................................................. 83830.5.2. Periodic Review Strategic Plan – Setting objectives..................... 83930.5.3. A few significant issues identified in reviews conducted by safety organizations.................. 84230.6. Overview of international practices – IAEA Guides........................................... 84530.6.1. International practices............................................................................ 84630.6.2. IAEA Guides................................................................................................ 84730.7. Multilateral practices................................................................................................... 848Chapter 31Optimizing Radiation Protection and Limiting Doses Received by Workers During Operations in a Nuclear Power Plant31.1. Sources of ionizing radiation in a nuclear power reactor................................ 85231.2. Examples of optimization of worker radiation protection.............................. 85331.3. Arrangements for ‘Major Refit’ operations........................................................... 85531.4. Approach and objectives adopted for the EPR.................................................... 858 Part 4The Accidents at Three Mile Island, Chernobyl and Fukushima Daiichi Nuclear Power Plants, Lessons Learned and Emergency Response ManagementChapter 32The Three Mile Island Nuclear Power Plant Accident32.1. Accident sequence – Reconstitution through simulation................................ 86432.2. Accident consequences............................................................................................... 87232.3. Analysis of the accident causes................................................................................ 87432.3.1. Error in identifying the position of the relief valve....................... 87432.3.2. Understanding the behaviour of the pressurizer............................ 87532.3.3. Stopping safety injection....................................................................... 87632.3.4. Human-machine interface..................................................................... 87632.3.5. Isolating the reactor containment...................................................... 87732.3.6. Confinement inside the auxiliary building....................................... 87732.3.7. Emergency feedwater supply to the steam generators............... 87732.4. Lessons learned from the Three Mile Island accident....................................... 87732.4.1. The human factor in facility operation............................................. 87832.4.2. Importance of precursor events........................................................... 88132.4.3. Study of complex situations and core-melt accidents, handling emergency situations....................................... 88232.5. Conclusions..................................................................................................................... 883Chapter 33Incident and Accident Operation: from the Event-Oriented Approach to the State-Oriented Approach33.1. Limits of the event-oriented approach.................................................................. 88533.2. The State-Oriented Approach concept.................................................................. 88633.3. First application of the state-oriented approach............................................... 88833.4. Widespread application of the state-oriented approach................................. 89133.5. ‘Stabilized’ state-oriented approach....................................................................... 89233.6. State-oriented approach adopted for the EPR.................................................... 894Chapter 34The Chernobyl Nuclear Power Plant Accident34.1. The Chernobyl nuclear power plant and RBMK reactors................................. 89734.2. The accident sequence................................................................................................ 90134.3. Analysis of the accident causes and changes made to RBMK units soon after the accident.................................. 90634.4. The other units at the facility................................................................................... 90834.5. Radioactive release and protection of the population..................................... 90834.5.1. Radioactive release kinetics.................................................................. 90834.5.2. Protection of the population................................................................ 91134.6. Consequences on human health and the environment................................... 91534.6.1. Direct effects of radiation...................................................................... 91534.6.2. Thyroid cancer in children..................................................................... 91734.6.3. Long-term contamination in the Dnieper Basin............................. 91934.7. Radioactive fallout in France and its consequences.......................................... 92134.7.1. Doses attributable to the plume......................................................... 92134.7.2. External doses due to soil deposition................................................ 92134.7.3. Doses due to ingestion of contaminated foodstuffs.................... 92234.7.4. Overall levels.............................................................................................. 92334.7.5. Thyroid cancer........................................................................................... 92434.7.5.1. Monitoring thyroid cancer in France............................ 92534.7.5.2. Assessment of the number of cancer cases induced in France by the Chernobyl accident........... 92634.8. Lessons learned by the international community from a general viewpoint and with regard to RBMK reactors....................... 92734.9. Lessons learned in France........................................................................................... 92834.10. Keeping the public informed..................................................................................... 93134.11. After the Chernobyl accident.................................................................................... 933Chapter 35Options and Control of Reactivity Insertion in Pressurized Water Reactors35.1. Research and study of event sequences................................................................ 93735.1.1. Cooling accidents..................................................................................... 938 35.1.2. Incidents and accidents related to rod cluster control assemblies.................................................................................... 94035.1.3. Boron dilution accidents........................................................................ 94335.1.4. Inserting a cold water plug in the core............................................. 95035.2. Changes in criteria........................................................................................................ 95135.3. The case of outage states.......................................................................................... 95335.4. Regulations...................................................................................................................... 957Chapter 36The Reactor Accident at the Fukushima Daiichi Nuclear Power Plant and Lessons Learned in France36.1. Reactor units at the Fukushima Daiichi nuclear power plant........................ 96236.1.1. General operation of a boiling water reactor.................................. 96236.1.2. Containment.............................................................................................. 96236.1.3. Emergency cooling systems.................................................................. 96436.2. Sequence of events during the accident............................................................... 96636.3. Radioactive release....................................................................................................... 97236.3.1. Airborne radioactive release, residual caesium deposits and contamination of foodstuffs............................... 97236.3.1.1. Airborne radioactive release............................................ 97236.3.1.2. Persistent caesium deposits............................................ 97236.3.1.3. Contamination of foodstuffs.......................................... 97336.3.2. Release of radioactive substances in the Pacific Ocean.............. 97536.3.3. Long-distance atmospheric dispersion of the radioactive plume........................................................................ 97636.4. Action taken to control facilities and released contaminated water.......... 97836.5. Socioeconomic and health impact in numbers................................................... 98136.5.1. Socioeconomic impact............................................................................ 98136.5.2. Health impact............................................................................................ 98236.6. Lessons learned from the accident.......................................................................... 98536.6.1. Complementary safety assessments carried out in Europe and France following the Fukushima Daiichi nuclear power plant accident............................................................................................ 98636.6.2. Complementary safety assessments carried out in France........ 98836.6.3. Procedure for complementary safety assessments carried out in France...................................... 98936.6.4. Conclusions of the complementary safety assessments carried out in France............................................. 99036.6.5. The ‘hardened safety core’.................................................................... 99136.6.5.1. Purpose................................................................................... 99136.6.5.2. Principles................................................................................ 99136.6.5.3. Illustrations........................................................................... 99336.6.6. Nuclear Rapid Response Force (FARN).............................................. 99536.6.7. Deployment of post-Fukushima measures in French nuclear power plants............................................................ 99736.7. Other lessons learned in France from the Fukushima Daiichi nuclear power plant accident...................................... 998Chapter 37Lessons Learned from the Fukushima Daiichi Nuclear PowerPlant Accident: Work Conducted by the IAEA and WENRA, Action Taken in Countries Other than France37.1. Work conducted by the IAEA.................................................................................... 100237.2. Work conducted by WENRA..................................................................................... 100337.3. Japan................................................................................................................................. 100437.4. Belgium............................................................................................................................ 100637.4.1. Nuclear power plants in Belgium........................................................ 100637.4.2. General details on the design of Belgian nuclear power plants................................................................ 100637.4.3. Stress tests and main lessons learned............................................... 100737.4.3.1. Improving protection of facilities against external hazards.................................................................. 100737.4.3.2. Improving protection of facilities against loss of electrical power supplies or loss of heat sink....... 101037.4.3.3. Improving on-site emergency plans............................. 1011 37.4.3.4. Improving management of core-melt accidents...... 101237.5. USA.................................................................................................................................... 1013Chapter 38Emergency Preparedness and Response38.1. Defining a radiological emergency and ‘response’ objectives........................ 102138.2. General organization of radiological emergency management..................... 102338.2.1. Organization and entities concerned................................................. 102338.2.2. Major Nuclear or Radiological Accident National Response Plan and emergency plans........................................ 102538.2.2.1. Major Nuclear or Radiological Accident National Response Plan...................................................................... 102538.2.2.2. Emergency plans................................................................. 102638.2.2.3. Provisions for protecting the public in the event of an accidental release of radioactivity..................... 102838.3. Management by the operator................................................................................... 102938.4. Prefectural authorities and mayors......................................................................... 103138.5. ASN, the Nuclear Safety Authority......................................................................... 103238.6. IRSN................................................................................................................................... 103338.7. Assessment approach in the event of an accident affecting a reactor in the nuclear power plant fleet.................. 103838.7.1. ‘3D/3P’ method......................................................................................... 103938.7.2. The ‘aggravated prognosis’ approach................................................ 104138.7.3. Extending the 3D/3P method to severe accidents (the ‘D/P AG’ method)...................................... 104138.8. Emergency preparedness............................................................................................ 104238.8.1. Emergency response exercises............................................................. 104338.8.2. Operating experience feedback........................................................... 1045 Part 5PWR Safety Studies, R&D and Simulation SoftwareChapter 39PWR Safety Studies and R&D39.1. Contribution of studies to the improvement of pressurized water reactor safety................................................................ 105039.2. Purpose and overview of R&D work, dedicated programmes and organizations involved, and research facilities in France... 105239.2.1. Purpose and overview of R&D............................................................. 105239.2.2. Dedicated frameworks and organizations involved....................... 106339.2.3. Facilities in France used for research and development.............. 1065Chapter 40Examples of Simulation Software Developed for Safety Analysis of Pressurized Water Reactors40.1. Simulation software for neutronics........................................................................ 107640.2. Simulation software for thermal hydraulics (and mechanics)....................... 107940.3. Simulation software for thermal mechanics........................................................ 108440.4. Software for simulating core-melt situations..................................................... 108540.5. Simulation software for mechanics........................................................................ 108840.6. Fire simulation software............................................................................................. 1089List of Acronyms.......................................................................................................................... 1093Acronyms for institutions, bodies and groups..................................................................... 1093Technical acronyms and abbreviations.................................................................................. 1103Technical glossary......................................................................................................................... 1121  ContentsPreface............................................................................................................................................. IIIForeword......................................................................................................................................... VIIEditors, Contributors and Reviewers................................................................................... XXXVIIIntroduction................................................................................................................................... XLIIIPart 1General BackgroundChapter 1Biological and Health Effects of Ionizing Radiation – The Radiological Protection System1.1. Biological and health effects of ionizing radiation............................................. 41.1.1. Biological processes................................................................................. 41.1.2. Review of units of measure................................................................... 61.1.3. Natural radioactivity............................................................................... 71.1.4. Health effects............................................................................................ 81.1.4.1. Deterministic effects, tissue reactions......................... 81.1.4.2. Stochastic or random effects.......................................... 91.1.4.3. Induction of diseases other than cancer..................... 111.1.5. Example of the limitations of epidemiology................................... 111.2. Radiological protection system................................................................................ 131.2.1. Types of exposure situations................................................................ 14 1.2.2. Exposure categories................................................................................. 141.2.3. Justification principle............................................................................... 151.2.4. Optimization (ALARA) principle ......................................................... 161.2.5. Principle of application of dose limits............................................... 21Chapter 2Organization of Nuclear Safety Control and Regulation for Nuclear Facilities and Activities in France2.1. From the founding of CEA to the TSN Act........................................................... 232.2. A few definitions........................................................................................................... 282.3. The different contributors to nuclear safety and their missions.................. 302.4. A few basic principles and notions in the field of nuclear safety................. 462.5. Statutory and quasi-statutory frameworks applicable to basic nuclear installations.............................................. 48Chapter 3The International Dimension and the Social Dimension3.1. International dimension.............................................................................................. 773.1.1. Introduction................................................................................................ 773.1.2. IAEA standards........................................................................................... 803.1.3. International Reporting System for Operating Experience (IRS).......................................................... 823.1.4. Services developed by the IAEA........................................................... 843.1.4.1. OSART reviews.................................................................... 843.1.4.2. IRRS reviews......................................................................... 873.1.4.3. Other services and study frameworks set up by the IAEA.............................................................. 883.1.5. WANO.......................................................................................................... 903.1.6. NEA............................................................................................................... 913.1.7. Organizations dedicated to radiation protection and health........................................................................ 933.1.8. From bilateral Franco-German cooperation to European structures for the exchange and capitalization of knowledge and practices, training and assessment services........................................................ 943.1.9. Nuclear regulator associations............................................................. 1003.2. The social dimension................................................................................................... 1023.2.1. Introduction – the context in France................................................. 1023.2.2. Examples of initiatives and issues raised concerning reactor safety in the French nuclear power plant fleet........... 102Chapter 4Nuclear Reactors: Complex Sociotechnical Systems – the Importance of Human and Organizational Factors4.1. 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.............................. 1084.2. The accident at the Chernobyl nuclear power plant and the concept of ‘safety culture’................................. 1094.3. The Fukushima Daiichi nuclear power plant accident: the social dimension and the concept of organization ‘resilience’... 1134.4. Changes in the perception of the role of people in achieving a high level of reliability in complex sociotechnical systems...... 1134.5. Main topics studied in the development of resources and skills pertaining to human and organizational factors............ 1164.5.1. Resources and skills................................................................................. 1164.5.2. Main topics studied.................................................................................. 1184.6. Human and organizational factors in French regulations................................ 119Part 2Safety by DesignChapter 5The Development of Nuclear Power Using Uranium-235 Fission – A Few Notions of Physics Used in Pressurized Water Reactors5.1. Important milestones in the development of nuclear power using fission of the uranium-235 isotope.......................... 1235.2. Fission and important concepts in reactor kinetics........................................... 1275.3. Removing power from the core during operation............................................. 1345.4. Decay heat...................................................................................................................... 1355.5. Main features of pressurized water reactor cores.............................................. 1365.6. Control and monitoring of pressurized water reactor cores.......................... 1375.7. Using uranium and plutonium mixed oxide (MOX) fuel................................. 147Chapter 6General Objectives, Principles and Basic Concepts of the Safety Approach6.1. General approach to risks – General objectives................................................. 1526.2. Fundamental safety functions.................................................................................. 1566.3. Confinement barriers................................................................................................... 1576.4. Defence in depth........................................................................................................... 1616.4. Defence in depth........................................................................................................... 1616.4.1. Levels of defence in depth..................................................................... 1626.4.2. Elements common to the different levels of defence in depth............................................................................. 1696.5. Events considered: terminology adopted for nuclear power reactors......... 1696.6. WENRA reference levels............................................................................................. 1716.7. Deterministic safety analysis and probabilistic safety assessments............ 1716.8. Lessons learned from the accident at the Fukushima Daiichi nuclear power plant on the concept of defence in depth and deterministic analysis......................................................................................... 1736.9. Safety culture – Quality control.............................................................................. 174Chapter 7Safety Options and Considerations at the Design Phase7.1. Different types of design provisions associated with safety considerations................................................. 1887.2. Single-failure criterion................................................................................................. 1897.3. The specific nature of computer-based systems (based on instrumentation and control software)............... 1937.4. Equipment safety classification................................................................................ 1957.4.1. Importance of equipment for safety and safety classification...................................................................... 1957.4.2. Generic requirements associated with the different safety classes.................................................................. 1997.4.3. Qualification of equipment for accident conditions..................... 2027.5. Information on designing nuclear pressure equipment.................................... 2077.6. General considerations on provisions for hazards in facility design............ 2107.7. Anticipating decommissioning in the design stage........................................... 212Chapter 8Study of Operating Conditions in the Deterministic Safety Analysis8.1. Categories of operating conditions......................................................................... 2188.2. Choice of operating conditions................................................................................ 2228.2.1. Concept of ‘bounding’ incident or accident..................................... 2238.2.2. Accident exclusion.................................................................................... 2248.3. List and breakdown of operating conditions....................................................... 2258.4. Methods for studying operating conditions......................................................... 2278.4.1. Choice of initial conditions, conservatism....................................... 2288.4.2. Consideration of an aggravating event in the study on operating conditions – ‘Passive’ failures............ 2298.4.3. Conventional combinations.................................................................. 2318.4.4. Preventing accident aggravation......................................................... 2328.4.5. Operator response time......................................................................... 2328.4.6. Using qualified simulation software................................................... 2338.4.7. Main criteria to be met for fuel in the reactor core..................... 2348.5. Concept of ‘design-basis situations’ for equipment.......................................... 2368.6. Situations to be taken into account in application of pressure equipment regulations................................................. 2378.7. Assessing the radiological consequences of incidents, accidents and hazards.............................................. 2388.7.1. Assessing radioactive substances released from the facility...... 2408.7.2. Assessing radiological consequences of radioactive release from the facility................................................ 2428.7.3. Assessing radiological consequences................................................. 243Chapter 9Loss-of-Coolant Accident9.1. Short- and medium-term aspects of a LOCA...................................................... 2529.1.1. Mechanical effects on vessel internals and fuel assembly structures................................................................ 2539.1.2. Thermal-hydraulic aspects and behaviour of fuel rods................ 2559.1.2.1. Large-break LOCA............................................................... 2559.1.2.2. Intermediate-break LOCA................................................ 2579.1.3. Effects on reactor containment and internals................................ 2589.1.4. Long-term aspect...................................................................................... 2599.2. Safety demonstration.................................................................................................. 2619.2.1. General information and background................................................ 2619.2.2. Fuel assemblies and fuel rods, vessel internals, reactor coolant system components................................................. 2639.2.2.1. Mechanical strength of vessel internals, fuel assembly structures and reactor coolant system components.............. 2639.2.2.2. Fuel behaviour...................................................................... 2659.2.3. Reactor containment and equipment located inside.................... 266Chapter 10A Special Issue: Steam Generator Tubes10.1. Steam generator tube rupture as a Category 3 event...................................... 27010.2. Preventing an SGTR accident, risk of multiple ruptures................................... 27210.3. Steam generator tube rupture(s) studied as a Category 4 event................. 27410.3.1. 900 MWe and 1300 MWe reactors.................................................... 27410.3.2. 1450 MWe reactors and EPR (Flamanville 3).................................. 27410.4. Provisions to mitigate the radiological consequences of SGTR accidents. 276Chapter 11Providing for Hazards: General Considerations and Internal Hazards11.1. General considerations on providing for hazards............................................... 27911.2. Potential projectiles inside the containment....................................................... 28211.3. Effects of pipe breaks.................................................................................................. 28411.4. Projectiles generated by a turbine rotor failure.................................................. 28511.5. Protection against load drops................................................................................... 28811.5.1. Risks related to spent fuel transport packaging............................. 28811.5.2. Other handling risks................................................................................ 29111.6. Fire protection............................................................................................................... 29211.7. Explosion protection.................................................................................................... 29711.8. Internal flooding............................................................................................................ 300Chapter 12Providing for External Hazards12.1. General considerations on providing for external hazards.............................. 30512.2. ‘Climate watch’ implemented by EDF.................................................................... 30712.3. Earthquakes..................................................................................................................... 30712.4. External floods............................................................................................................... 32212.5. Extreme temperatures................................................................................................. 33012.5.1. Extreme cold.............................................................................................. 33012.5.2. Extreme heat.............................................................................................. 33112.6. Possible heat sink hazards.......................................................................................... 33212.7. Other naturally-occurring external hazards......................................................... 33612.8. Accidental aeroplane crashes (excluding malicious acts)................................ 33612.9. Risks related to the industrial environment (excluding malicious acts)..... 340Chapter 13Complementary Domain of Events13.1. The origin of studies belonging to the complementary domain................... 34413.2. Background of the complementary domain........................................................ 34413.3. Analysis of complementary domain events......................................................... 35113.4. ‘New complementary domain’................................................................................. 35113.5. Case of the Flamanville 3 EPR.................................................................................. 354Chapter 14Development and Use of Probabilistic Safety Assessments14.1. History and regulatory context................................................................................ 35714.1.1. International situation............................................................................ 35714.1.2. Situation in France................................................................................... 35914.2. Level 1 PSA...................................................................................................................... 36114.2.1. Scope............................................................................................................ 36114.2.2. Method for carrying out a Level 1 PSA............................................. 36214.2.2.1. General information........................................................... 36214.2.2.2. Specific point: probabilistic human reliability analysis................................................................................... 36414.2.3. Level 1 PSA results and lessons learned............................................ 36914.3. Level 2 PSA...................................................................................................................... 37314.3.1. Scope............................................................................................................ 37314.3.2. Method for carrying out a Level 2 PSA............................................. 37414.3.2.1. General information........................................................... 37414.3.2.2. Probabilistic human reliability analysis for Level 2 PSAs................................................................... 37914.3.3. Examples of lessons learned from Level 2 PSAs............................. 38214.3.3.1. Steam explosion risk assessment.................................. 38214.3.3.2. Mechanical integrity of the 900 MWe reactor containments........................................................ 38314.3.3.3. Isolating penetrations in the containment................. 38314.3.3.4. Modifying the pressure relief system of the reactor coolant system........................................ 38414.3.3.5. Improvement of operating procedures to reduce risk of core melt under pressure................ 38414.3.3.6. Contribution of Level 2 PSAs to emergency response measures............................................................. 38514.4. Expanding the scope of PSA coverage................................................................... 38514.5. Using probabilistic safety assessments.................................................................. 38614.5.1. Using PSAs in the design phase........................................................... 38614.5.1.1. Usefulness and particularities of PSAs in the design phase............................................................. 38614.5.1.2. PSAs conducted to support the Flamanville 3 EPR design............................................................................. 38714.5.2. Using PSAs in periodic reviews............................................................ 38914.5.2.1. Level 1 PSA............................................................................ 39014.5.2.2. Level 2 PSA............................................................................ 39114.5.3. Using PSAs for reactor operation........................................................ 39214.5.3.1. Using PSAs to analyse event severity.......................... 39214.5.3.2. Using PSAs to analyse operational limits and conditions and temporary changes...................... 39414.5.3.3. Using PSAs to analyse operating procedures............. 396Chapter 15Aspects Specific to PWR Spent Fuel Storage Pools15.1. Spent fuel pool design................................................................................................. 39915.1.1. Confinement barriers.............................................................................. 39915.1.2. Initiating events defined at the design stage.................................. 40015.2. Experience feedback..................................................................................................... 40115.2.1. Loss of cooling........................................................................................... 40115.2.1.1. Loss of heat sink.................................................................. 40115.2.1.2. Risks related to maintenance during unit outages................................................................................... 40215.2.1.3. Suction of foreign matter into the cooling system.................................................................................... 40315.2.1.4. Exceeding the decay heat defined in facility design...................................................................................... 40315.2.2. Water losses............................................................................................... 404 15.2.2.1. Gate or sluice gate failures.............................................. 40415.2.2.2. Line-up errors....................................................................... 40515.2.2.3. Failure of a reactor coolant system pipe nozzle dam.......................................................................................... 40915.2.2.4. Rupture of a pipe connected to the spent fuel pool................................................................................. 41115.3. Safety reassessments................................................................................................... 41115.4. Experience feedback from the accident that affected the Unit 4 pool at the Fukushima Daiichi nuclear power plant..... 41415.4.1. Events........................................................................................................... 41415.4.2. Complementary safety assessments conducted in France......... 41715.5. Measures adopted for the EPR.................................................................................. 41915.6. Recommendations for new reactor designs......................................................... 42015.7. New systems for storing spent fuel........................................................................ 423Chapter 16Taking into Account Human and Organizational Factors in Facility Design16.1. Taking into account human and organizational factors in nuclear power reactor design....................................... 42516.1.1. Importance of considering human and organizational factors at the design stage....................................... 42516.1.2. Approach at the design stage............................................................... 43116.1.2.1. Prior to the design phase: analysis of ‘existing elements’........................................ 43216.1.2.2. Design objectives................................................................ 43516.1.2.3. Definition of detailed design provisions...................... 43616.1.2.4. Validation of design provisions...................................... 43916.1.2.5. Assessments conducted during reactor startup and after commissioning.................................................. 44216.1.3. Project management and human and organizational factors engineering programme..................................... 44316.2. Considering human and organizational aspects when designing changes to nuclear power plants..................... 44316.2.1. Importance of human and organizational factors in designing modifications..................................................................... 44416.2.2. ‘Human, social and organizational approach’ implemented by EDF........................................................................... 44516.2.3. Changes, a subject that always deserves special attention from a human and organizational factors perspective.... 44716.3. Human and organizational factors for future nuclear power reactor projects.................................................. 448Chapter 17Studying Core-Melt Accidents to Enhance Safety17.1. Core degradation and vessel failure....................................................................... 45217.1.1. Core uncovery............................................................................................ 45317.1.2. Fuel degradation....................................................................................... 45417.1.3. Failure of the reactor coolant system............................................... 45517.1.4. Phenomena that can cause early containment failure................. 45517.1.5. Phenomena that can ultimately lead to containment failure....................................................... 45717.2. Containment failure modes....................................................................................... 45717.3. Classification of releases associated with core-melt accidents − ‘source terms’................................................. 46017.4. Improving knowledge.................................................................................................. 46217.5. Studies in France on containment failure modes............................................... 46217.5.1. Introduction................................................................................................ 46217.5.2. Initial containment leakage................................................................... 46317.5.3. Direct heating of gases in the containment.................................... 46317.5.4. Hydrogen explosion in the containment.......................................... 46417.5.5. Steam explosion in the vessel or reactor pit................................... 46517.5.6. Gradual pressure increase in the containment............................... 46617.5.7. Penetration of the concrete basemat of the containment by corium............................................ 46717.5.8. ‘U4’ provisions........................................................................................... 46817.5.9. Bypass of containment by outgoing pipes (the V mode).............................................................................................. 46817.5.10. Fast reactivity insertion accidents...................................................... 46917.6. Severe accident operating guidelines..................................................................... 46917.7. Radiological consequences associated with the S3 source term and emergency response plans implemented by public authorities............ 47017.8. Ultimate emergency operating procedures.......................................................... 47317.9. On-site emergency plan............................................................................................. 47317.10. Approach adopted for the EPR................................................................................. 47617.10.1. General safety objectives....................................................................... 47617.10.2. ‘Practical elimination’ of core-melt conditions that could lead to significant early releases............................. 47617.10.3. Provisions for low-pressure core melt............................................... 480Chapter 18New-Generation Reactors18.1. Organization and framework of Franco-German discussions........................ 48518.2. Progression of safety objectives and design options for the EPR project.......................................................... 48618.2.1. General safety objectives....................................................................... 48618.2.2. Events to be taken into account at the design stage and in deterministic and probabilistic analyses.............................. 48818.2.3. Main provisions for preventing incidents and accidents............. 49018.2.4. Functional redundancy, independence between systems, system reliability................................................ 49518.2.5. Confinement preservation..................................................................... 49618.2.6. Radiological protection........................................................................... 49718.2.7. Incorporating lessons learned from the Fukushima Daiichi nuclear power plant accident......................... 49718.3. International context: general safety objectives for new-generation reactors............................................ 49918.4. Concepts highlighted in new reactor designs...................................................... 50118.4.1. AP1000: gravity systems........................................................................ 50118.4.2. VVER: SPOT system................................................................................. 50318.4.3. NM EPR: ‘multi-group’ technology, diversified heat sink............ 504 18.4.4. ATMEA 1: safety injection accumulators in the reactor coolant system.............................................................. 50518.4.5. NuScale: common pool for modular reactors................................. 505 Part 3Safety in OperationChapter 19Startup Tests for Pressurized Water Reactors19.1. Introduction.................................................................................................................... 51119.2. Commissioning.............................................................................................................. 51419.2.1. Defining startup tests.............................................................................. 51419.2.2. Phasing of startup tests.......................................................................... 51519.2.2.1. Preliminary and pre-operational tests......................... 51519.2.2.2. Operational tests................................................................ 51719.2.2.3. General principles for test sequencing and execution....................................................................... 51719.2.3. Documentation for startup tests......................................................... 51819.2.3.1. Integrated system test procedures and startup test procedures............................................ 51819.2.3.2. Test programmes, test procedures, standard test guidelines.................................................... 51819.2.3.3. Completeness analysis, adequacy analysis................. 51819.2.3.4. Acceptance criteria............................................................. 51919.3. Objectives and general rules to take into account for startup tests........... 52019.4. Key lessons learned from startup tests on nuclear power reactors in France............................................................ 52119.4.1. Qualification tests and on-site tests.................................................. 52219.4.2. Long-term on-site testing...................................................................... 52419.4.3. Test configurations and completeness, transpositions................ 525 19.4.4. Safety measures that cannot be verified by testing..................... 52719.4.5. Criteria......................................................................................................... 52719.4.6. Cleanness, keeping system lines clean, foreign matter................ 52819.4.7. Piping support structures and displacement................................... 53219.4.8. Pump and piping vibrations................................................................... 53219.4.9. Validation of operating procedures and periodic tests................ 53419.4.10. Uncertainty and ‘set points’.................................................................. 53519.4.11. Condition of facilities during startup tests...................................... 53619.4.12. Other aspects............................................................................................. 53719.5. Examples of findings resulting from startup tests............................................. 538Chapter 20General Operating Rules20.1. General Operating Rules............................................................................................ 55220.1.1. Content of general operating rules..................................................... 55320.1.2. Limits of general operating rules......................................................... 55420.2. Operational limits and conditions........................................................................... 55420.2.1. Content of operational limits and conditions................................. 55520.2.1.1. Operating modes and standard states......................... 55620.2.1.2. Requirements and unavailability................................... 55820.2.1.3. Fallback states and time required to reach them.... 55820.2.1.4. Events and event groups.................................................. 55920.2.1.5. Combined types of unavailability.................................. 56020.2.1.6. Concepts of ‘boundary condition’ and ‘specific requirement’................................................ 56020.2.2. Average pressure and temperature range of the reactor coolant system.............................................................. 56120.2.3. Changes in operational limits and conditions................................. 56320.3. Initial and periodic tests............................................................................................. 56420.4. Incident and accident operating procedures........................................................ 566Chapter 21 Operating Experience Feedback from Events: Rules and Practices21.1. Background...................................................................................................................... 56921.2. Objectives of an operating experience feedback system................................ 57121.3. Components of an operating experience feedback system – Regulations............................................................................... 57221.4. Operating experience feedback practices adopted for the French nuclear power plant fleet................................. 576Chapter 22Operating Experience from Events Attributable to Shortcomings in Initial Reactor Design or the Quality of Maintenance22.1. 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).......................................... 59022.2. Recurrent loss of safety function events related to maintenance operations – Lessons learned........................... 59522.2.1. Events........................................................................................................... 59522.2.2. General discussion initiated by EDF in the late 1980s on the quality of maintenance operations....................................... 60522.2.3. Applying the defence-in-depth concept when working on a reactor in service.................................. 60822.2.4. Problems that may recur....................................................................... 609Chapter 23Operating Experience from Events Related to Maintenance Operations, Electrical Power Sources and Distribution, Internal and External Hazards23.1. Risks of failure related to equipment or maintenance..................................... 61223.1.1. Risks of common-mode failure............................................................ 61223.1.1.1. Risks of common-mode failure related to settings............................................................................. 61223.1.1.2. Risks of common-mode failure on electrical switchboards.................................................61323.1.1.3. Unavailability of two out of three high-head safety injection lines in the cold legs of the reactor coolant system.............. 61423.1.1.4. Loss of electrical power supplies................................... 61723.1.2. Introduction of non-borated water into the reactor coolant system.......................................................... 62023.1.3. Cooling the reactor coolant system after inhibition of automatic actions............................................................................... 62223.1.4. A temporary device prevents switching the safety injection system to the water recirculation mode................... 62423.2. Events related to internal hazards........................................................................... 62523.2.1. Risk of common-mode failure due to internal flooding.............. 62523.2.2. Risk of failure due to fire....................................................................... 62823.2.3. Risks associated with the use of hydrogen in 900 MWe reactors............................................................................... 63123.3. External hazards: events related to periods of extreme cold......................... 635Chapter 24Enhanced Protection of Estuary and River Sites: Flooding at the Blayais Nuclear Power Plant and Obstruction of a Water Intakeat the Cruas-Meysse Nuclear Power Plant24.1. Partial loss of engineered safety systems following flooding of the Blayais nuclear power plant............................... 64224.2. 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...................................................... 648Chapter 25Taking into Account Human and Organizational Factors in Facility Operation25.1. Skills management....................................................................................................... 65325.1.1. Historical background............................................................................. 65425.1.2. Managing training.................................................................................... 65525.1.3. Strategic workforce planning................................................................ 65625.1.4. Personnel certification............................................................................ 65725.2. Safety and risk management.................................................................................... 65825.2.1. Historical background............................................................................. 65825.2.2. Decision-making and safety.................................................................. 66025.2.3. Risk analyses applied to work activities............................................ 66225.2.4. Operating experience feedback........................................................... 66325.2.5. Managing organizational change......................................................... 66425.3. Managing operational activities............................................................................... 66625.3.1. Characteristics of operational activities............................................ 66625.3.2. Monitoring by the operating crew in the control room.............. 66925.3.3. Compliance with general operating rules......................................... 67025.3.4. Line-up......................................................................................................... 67125.3.5. Operation in extreme situations......................................................... 67225.4. Management of maintenance activities................................................................ 67225.4.1. Management of a scheduled reactor outage for refuellingand maintenance...................................................................................... 67225.4.2. Risks during reactor outages................................................................. 67325.4.3. Preparation for scheduled reactor outages...................................... 67425.4.4. Managing scheduled reactor outages................................................ 67525.5. Supervising outsourced activities............................................................................ 67525.5.1. Contractor qualification and contracting......................................... 67625.5.2. Matching workload and resources...................................................... 67625.5.3. Carrying out work.................................................................................... 67725.5.4. Surveillance of outsourced activities................................................. 67825.5.5. Operating experience feedback and assessingoutsourced activities............................................................................... 678Chapter 26Facility Maintenance26.1. Maintenance objectives.............................................................................................. 68126.2. Maintenance................................................................................................................... 68226.2.1. Definition.................................................................................................... 68226.2.2. Maintenance strategies........................................................................... 68326.3. Optimizing maintenance............................................................................................ 68426.3.1. Reliability-centred maintenance.......................................................... 68426.3.2. Conditional maintenance....................................................................... 686XXVI Elements of nuclear safety – Pressurized water reactors26.3.3. Conditional maintenance by sampling –Maintenance based on reference equipment items...................... 68826.3.4. ‘AP-913’ method....................................................................................... 68926.4. Maintenance baselines................................................................................................ 69126.5. On-site maintenance................................................................................................... 69426.5.1. The various stages of maintenance operations.............................. 69426.5.2. Main conditions for successful maintenance.................................. 69626.5.3. Examples of anomalies or deviations discovered during routinemaintenance, explained by an inadequate maintenancebaseline with regard to deterioration mechanisms....................... 70526.5.4. Examples of events associated with non-qualitymaintenance............................................................................................... 70826.5.4.1. Example of an event explained by an incorrectsetting on redundant equipment................................... 70826.5.4.2. Example of an event explained by an incorrectsetting of electrical protection thresholds................. 70826.5.4.3. Examples of events explained by a failureto return equipment to a compliant stateafter maintenance or an error in performinga work procedure................................................................ 709Chapter 27In-service Monitoring and Inspection of Equipment27.1. Main internal equipment items on a pressurized water reactor vessel...... 71927.1.1. ‘Core baffle’ around the core................................................................ 72027.1.2. RCCA guide tubes..................................................................................... 72227.2. Reactor vessel, nozzles and head............................................................................. 72227.2.1. Vessel underclad defects........................................................................ 72427.2.2. Cracking on vessel head adaptors....................................................... 72527.2.2.1. Condition of other reactors............................................. 72727.2.2.2. Impact on safety................................................................. 72827.2.2.3. Prevention, monitoring and mitigation....................... 72927.2.2.4. Developing inspection tools............................................ 72927.2.2.5. Repairs.................................................................................... 730Contents XXVII27.2.2.6. Leak detection...................................................................... 73027.2.2.7. Anti-ejection devices......................................................... 73127.2.2.8. Current situation................................................................. 73127.2.2.9. Cracks observed on reactor vessel headsin other countries............................................................... 73127.2.2.10. Implementation of special monitoringfor ‘Inconel areas’ beginning in 1992........................... 73227.2.3. Cracking on vessel lower head penetrations detectedin 2011......................................................................................................... 73327.2.4. Monitoring the ‘beltline’ region of the vessel................................. 73527.2.5. Defects observed on reactor vessels in Belgium............................ 73627.3. Steam generators.......................................................................................................... 73827.3.1. The different types of defects.............................................................. 73827.3.2. Associated risks......................................................................................... 74027.3.3. Monitoring during operation and inspection during outages.... 74127.3.3.1. Monitoring during operation........................................... 74127.3.3.2. Inspection during reactor outages................................. 74227.3.4. Steps to be taken when a defect is detected.................................. 74227.3.4.1. Tube wear due to foreign matter.................................. 74327.3.4.2. Wear due to contact with anti-vibration bars.......... 74327.3.4.3. Cracking in U-bend tubes................................................. 74427.3.4.4. Tube deformation and cracking..................................... 74427.3.5. Steam generator replacement.............................................................. 74527.3.6. Clogging observed in the 2000s.......................................................... 74627.3.7. Conclusion.................................................................................................. 74827.4. Steam lines...................................................................................................................... 74827.5. Auxiliary systems: cracks induced by local thermal-hydraulicphenomena..................................................................................................................... 75027.5.1. Cracking in non-isolatable sections connectedto the reactor coolant loops................................................................. 75027.5.2. RHRS thermal fatigue at Unit 1 of the Civaux nuclearpower plant................................................................................................ 75127.6. Civil works: containment structures....................................................................... 75527.6.1. Anticipated degradation phenomena................................................. 75627.6.2. Devices for direct monitoring of containment buildingconcrete walls............................................................................................ 75727.6.3. Leak tests and measurements.............................................................. 75927.6.4. Main anomalies......................................................................................... 759Chapter 28Fuel Management, Monitoring and Developments28.1. Procedures for monitoring fuel rod integrity....................................................... 76828.1.1. Radiochemical specifications for reactor coolant.......................... 76828.1.2. Inspections and measurements carried out directlyon fuel assemblies.................................................................................... 77728.1.2.1. Liquid penetrant testing in the refuellingmachine mast....................................................................... 77828.1.2.2. Liquid penetrant testing in the FB cell......................... 77928.1.2.3. Inspections performed on fuel rods.............................. 78028.2. Operating experience feedback and changes in cladding material.............. 78228.3. Anomalies and significant events involving fuel assemblies.......................... 78628.3.1. Baffle jetting............................................................................................... 78628.3.2. Fretting........................................................................................................ 78728.3.3. Events encountered during handling operations............................ 78928.3.4. Lateral deformation of fuel assemblies interferingwith RCCA drop......................................................................................... 791Chapter 29Facility Compliance29.1. Introduction.................................................................................................................... 79529.2. Detection and treatment of compliance deviations for pressurizedwater reactors in the nuclear power plant fleet................................................. 79629.2.1. Process for handling compliance gaps............................................... 79629.2.2. Examples of compliance gaps............................................................... 79829.2.2.1. Compliance gap in electrical connection boxesqualified for accident conditions................................... 79829.2.2.2. Failure in the seismic resistance of metal floorsin electrical and auxiliary buildings of 900 MWereactors (CPY series).......................................................... 79929.2.2.3. Risk of containment sump screen blockage............... 79929.2.2.4. Anomaly found in engines of emergency and SBOdiesel generators for 900 MWe reactors.................... 80029.2.2.5. Temperature resistance fault in the high-headsafety injection pumps...................................................... 80229.2.2.6. Mixed lubricants in equipment requiredfor accident situations....................................................... 80229.2.2.7. Flow imbalance between safety injection linesof 900 MWe reactors......................................................... 80429.2.2.8. Anomaly in CATHARE software modellingof natural circulation in the upper partof the vessel.......................................................................... 80529.2.2.9. Vibrations and rotor lift on engineeredsafety motor-driven pump units.................................... 805Chapter 30Periodic Reviews30.1. Introduction.................................................................................................................... 80730.2. History of periodic reviews in France for nuclear power reactors................ 80930.2.1. Reactors other than PWRs in the French nuclear powerplant fleet.................................................................................................... 80930.2.2. PWRs in the French nuclear power plant fleet(900 MWe, 1300 MWe and 1450 MWe).......................................... 81130.3. Periodic review process for PWRs in the French nuclear powerplant fleet........................................................................................................................ 81530.3.1. Regulations................................................................................................. 81530.3.2. Outline of a PWR periodic review....................................................... 81730.4. Case of the review associated with the third ten-yearly outageof 900 MWe reactors................................................................................................... 82230.4.1. Plant unit compliance reviews, the complementaryinvestigation and ageing management............................................. 82430.4.1.1. Plant unit compliance reviews........................................ 82430.4.1.2. Complementary investigation programme................ 82530.4.1.3. Ageing management.......................................................... 82630.4.2. Compliance studies on the design of civil works systemsand structures............................................................................................ 82630.4.3. Studies to reassess system design...................................................... 82730.4.4. Reassessment of reactor resistance to internaland external hazards................................................................................ 83030.4.5. Accident studies........................................................................................ 83230.4.6. Taking into account lessons learned during the reviewassociated with the VD3 900 outagefor subsequent reviews........................................................................... 83730.5. Fourth ten-yearly outage of 900 MWe reactors: integratingthe extension of the operating lifetime of nuclear power reactorsin France.......................................................................................................................... 83830.5.1. Background................................................................................................. 83830.5.2. Periodic Review Strategic Plan – Setting objectives..................... 83930.5.3. A few significant issues identified in reviews conductedby safety organizations........................................................................... 84230.6. Overview of international practices – IAEA Guides........................................... 84530.6.1. International practices............................................................................ 84630.6.2. IAEA Guides................................................................................................ 84730.7. Multilateral practices................................................................................................... 848Chapter 31Optimizing Radiation Protection and Limiting Doses Receivedby Workers During Operations in a Nuclear Power Plant31.1. Sources of ionizing radiation in a nuclear power reactor................................ 85231.2. Examples of optimization of worker radiation protection.............................. 85331.3. Arrangements for ‘Major Refit’ operations........................................................... 85531.4. Approach and objectives adopted for the EPR.................................................... 858 Part 4The Accidents at Three Mile Island,Chernobyl and Fukushima Daiichi Nuclear Power Plants,Lessons Learned and Emergency Response ManagementChapter 32The Three Mile Island Nuclear Power Plant Accident32.1. Accident sequence – Reconstitution through simulation................................ 86432.2. Accident consequences............................................................................................... 87232.3. Analysis of the accident causes................................................................................ 87432.3.1. Error in identifying the position of the relief valve....................... 87432.3.2. Understanding the behaviour of the pressurizer............................ 87532.3.3. Stopping safety injection....................................................................... 87632.3.4. Human-machine interface..................................................................... 87632.3.5. Isolating the reactor containment...................................................... 87732.3.6. Confinement inside the auxiliary building....................................... 87732.3.7. Emergency feedwater supply to the steam generators............... 87732.4. Lessons learned from the Three Mile Island accident....................................... 87732.4.1. The human factor in facility operation............................................. 87832.4.2. Importance of precursor events........................................................... 88132.4.3. Study of complex situations and core-melt accidents,handling emergency situations............................................................ 88232.5. Conclusions..................................................................................................................... 883Chapter 33Incident and Accident Operation:from the Event-Oriented Approach to the State-Oriented Approach33.1. Limits of the event-oriented approach.................................................................. 88533.2. The State-Oriented Approach concept.................................................................. 88633.3. First application of the state-oriented approach............................................... 88833.4. Widespread application of the state-oriented approach................................. 89133.5. ‘Stabilized’ state-oriented approach....................................................................... 89233.6. State-oriented approach adopted for the EPR.................................................... 894Chapter 34The Chernobyl Nuclear Power Plant Accident34.1. The Chernobyl nuclear power plant and RBMK reactors................................. 89734.2. The accident sequence................................................................................................ 90134.3. Analysis of the accident causes and changes made to RBMK unitssoon after the accident............................................................................................... 90634.4. The other units at the facility................................................................................... 90834.5. Radioactive release and protection of the population..................................... 90834.5.1. Radioactive release kinetics.................................................................. 90834.5.2. Protection of the population................................................................ 91134.6. Consequences on human health and the environment................................... 91534.6.1. Direct effects of radiation...................................................................... 91534.6.2. Thyroid cancer in children..................................................................... 91734.6.3. Long-term contamination in the Dnieper Basin............................. 91934.7. Radioactive fallout in France and its consequences.......................................... 92134.7.1. Doses attributable to the plume......................................................... 92134.7.2. External doses due to soil deposition................................................ 92134.7.3. Doses due to ingestion of contaminated foodstuffs.................... 92234.7.4. Overall levels.............................................................................................. 92334.7.5. Thyroid cancer........................................................................................... 92434.7.5.1. Monitoring thyroid cancer in France............................ 92534.7.5.2. Assessment of the number of cancer casesinduced in France by the Chernobyl accident........... 92634.8. Lessons learned by the international communityfrom a general viewpoint and with regard to RBMK reactors....................... 92734.9. Lessons learned in France........................................................................................... 92834.10. Keeping the public informed..................................................................................... 93134.11. After the Chernobyl accident.................................................................................... 933Chapter 35Options and Control of Reactivity Insertionin Pressurized Water Reactors35.1. Research and study of event sequences................................................................ 93735.1.1. Cooling accidents..................................................................................... 93835.1.2. Incidents and accidents related to rod clustercontrol assemblies.................................................................................... 94035.1.3. Boron dilution accidents........................................................................ 94335.1.4. Inserting a cold water plug in the core............................................. 95035.2. Changes in criteria........................................................................................................ 95135.3. The case of outage states.......................................................................................... 95335.4. Regulations...................................................................................................................... 957Chapter 36The Reactor Accident at the Fukushima Daiichi Nuclear Power Plantand Lessons Learned in France36.1. Reactor units at the Fukushima Daiichi nuclear power plant........................ 96236.1.1. General operation of a boiling water reactor.................................. 96236.1.2. Containment.............................................................................................. 96236.1.3. Emergency cooling systems.................................................................. 96436.2. Sequence of events during the accident............................................................... 96636.3. Radioactive release....................................................................................................... 97236.3.1. Airborne radioactive release, residual caesium depositsand contamination of foodstuffs........................................................ 97236.3.1.1. Airborne radioactive release............................................ 97236.3.1.2. Persistent caesium deposits............................................ 97236.3.1.3. Contamination of foodstuffs.......................................... 97336.3.2. Release of radioactive substances in the Pacific Ocean.............. 97536.3.3. Long-distance atmospheric dispersionof the radioactive plume........................................................................ 97636.4. Action taken to control facilities and released contaminated water.......... 97836.5. Socioeconomic and health impact in numbers................................................... 98136.5.1. Socioeconomic impact............................................................................ 98136.5.2. Health impact............................................................................................ 98236.6. Lessons learned from the accident.......................................................................... 98536.6.1. Complementary safety assessments carried out in Europeand France following the Fukushima Daiichi nuclear powerplant accident............................................................................................ 98636.6.2. Complementary safety assessments carried out in France........ 98836.6.3. Procedure for complementary safety assessmentscarried out in France................................................................................ 98936.6.4. Conclusions of the complementary safety assessmentscarried out in France................................................................................ 99036.6.5. The ‘hardened safety core’.................................................................... 99136.6.5.1. Purpose................................................................................... 99136.6.5.2. Principles................................................................................ 99136.6.5.3. Illustrations........................................................................... 99336.6.6. Nuclear Rapid Response Force (FARN).............................................. 99536.6.7. Deployment of post-Fukushima measuresin French nuclear power plants............................................................ 99736.7. Other lessons learned in France from the Fukushima Daiichi nuclearpower plant accident................................................................................................... 998Chapter 37Lessons Learned from the Fukushima Daiichi Nuclear PowerPlant Accident: Work Conducted by the IAEA and WENRA,Action Taken in Countries Other than France37.1. Work conducted by the IAEA.................................................................................... 100237.2. Work conducted by WENRA..................................................................................... 100337.3. Japan................................................................................................................................. 100437.4. Belgium............................................................................................................................ 100637.4.1. Nuclear power plants in Belgium........................................................ 100637.4.2. General details on the design of Belgian nuclear powerplants............................................................................................................ 100637.4.3. Stress tests and main lessons learned............................................... 100737.4.3.1. Improving protection of facilities againstexternal hazards.................................................................. 100737.4.3.2. Improving protection of facilities against lossof electrical power supplies or loss of heat sink....... 1010Contents XXXV37.4.3.3. Improving on-site emergency plans............................. 101137.4.3.4. Improving management of core-melt accidents...... 101237.5. USA.................................................................................................................................... 1013Chapter 38Emergency Preparedness and Response38.1. Defining a radiological emergency and ‘response’ objectives........................ 102138.2. General organization of radiological emergency management..................... 102338.2.1. Organization and entities concerned................................................. 102338.2.2. Major Nuclear or Radiological Accident National ResponsePlan and emergency plans..................................................................... 102538.2.2.1. Major Nuclear or Radiological Accident NationalResponse Plan...................................................................... 102538.2.2.2. Emergency plans................................................................. 102638.2.2.3. Provisions for protecting the public in the eventof an accidental release of radioactivity..................... 102838.3. Management by the operator................................................................................... 102938.4. Prefectural authorities and mayors......................................................................... 103138.5. ASN, the Nuclear Safety Authority......................................................................... 103238.6. IRSN................................................................................................................................... 103338.7. Assessment approach in the event of an accident affecting a reactorin the nuclear power plant fleet.............................................................................. 103838.7.1. ‘3D/3P’ method......................................................................................... 103938.7.2. The ‘aggravated prognosis’ approach................................................ 104138.7.3. Extending the 3D/3P method to severe accidents(the ‘D/P AG’ method)............................................................................ 104138.8. Emergency preparedness............................................................................................ 104238.8.1. Emergency response exercises............................................................. 104338.8.2. Operating experience feedback........................................................... 1045 Part 5PWR Safety Studies, R&D and Simulation SoftwareChapter 39PWR Safety Studies and R&D39.1. Contribution of studies to the improvement of pressurized waterreactor safety................................................................................................................. 105039.2. Purpose and overview of R&D work, dedicated programmesand organizations involved, and research facilities in France........................ 105239.2.1. Purpose and overview of R&D............................................................. 105239.2.2. Dedicated frameworks and organizations involved....................... 106339.2.3. Facilities in France used for research and development.............. 1065Chapter 40Examples of Simulation Software Developed for Safety Analysisof Pressurized Water Reactors40.1. Simulation software for neutronics........................................................................ 107640.2. Simulation software for thermal hydraulics (and mechanics)....................... 107940.3. Simulation software for thermal mechanics........................................................ 108440.4. Software for simulating core-melt situations..................................................... 108540.5. Simulation software for mechanics........................................................................ 108840.6. Fire simulation software............................................................................................. 1089List of Acronyms.......................................................................................................................... 1093Acronyms for institutions, bodies and groups..................................................................... 1093Technical acronyms and abbreviations.................................................................................. 1103Technical glossary......................................................................................................................... 1121 21 00 06 01 https://laboutique.edpsciences.fr/produit/1280/9782759827237/elements-of-nuclear-safety 01 00 03 02 https://laboutique.edpsciences.fr/system/product_pictures/data/009/982/690/original/9782759827220-PressurizedWaterReactors_couv-sofedis.jpg 17 14 20240913T180957+0200 01 P1 EDP Sciences 01 01 P1 EDP Sciences 17 avenue du Hoggar - PA de Courtaboeuf - 91944 Les Ulis cedex A http://publications.edpsciences.org/ 04 11 00 20220707 01 00 20220707 19 00 20220825 2026 06 3052868830012 01 WORLD 13 03 9782759827220 15 9782759827220 06 03 9782759827237 15 9782759827237 WORLD 08 EDP Sciences 04 01 00 20220707 03 01 D1 EDP Sciences 20 04 05 01 02 1 00 01 EUR