ELEMENTS OF NUCLEAR SAFETY
PRESSURIZED WATER REACTORS
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Presentation
Resume
Contents
Preface............................................................................................................................................. III
Foreword......................................................................................................................................... VII
Editors, Contributors and Reviewers................................................................................... XXXVII
Introduction................................................................................................................................... XLIII
Part 1
General Background
Chapter 1
Biological and Health Effects of Ionizing Radiation – The Radiological Protection System
1.1. Biological and health effects of ionizing radiation............................................. 4
1.1.1. Biological processes................................................................................. 4
1.1.2. Review of units of measure................................................................... 6
1.1.3. Natural radioactivity............................................................................... 7
1.1.4. Health effects............................................................................................ 8
1.1.4.1. Deterministic effects, tissue reactions......................... 8
1.1.4.2. Stochastic or random effects.......................................... 9
1.1.4.3. Induction of diseases other than cancer..................... 11
1.1.5. Example of the limitations of epidemiology................................... 11
1.2. Radiological protection system................................................................................ 13
1.2.1. Types of exposure situations................................................................ 14
1.2.2. Exposure categories................................................................................. 14
1.2.3. Justification principle............................................................................... 15
1.2.4. Optimization (ALARA) principle ......................................................... 16
1.2.5. Principle of application of dose limits............................................... 21
Chapter 2
Organization of Nuclear Safety Control and Regulation for Nuclear Facilities and Activities in France
2.1. From the founding of CEA to the TSN Act........................................................... 23
2.2. A few definitions........................................................................................................... 28
2.3. The different contributors to nuclear safety and their missions.................. 30
2.4. A few basic principles and notions in the field of nuclear safety................. 46
2.5. Statutory and quasi-statutory frameworks applicable to basic nuclear installations.............................................. 48
Chapter 3
The International Dimension and the Social Dimension
3.1. International dimension.............................................................................................. 77
3.1.1. Introduction................................................................................................ 77
3.1.2. IAEA standards........................................................................................... 80
3.1.3. International Reporting System for Operating Experience (IRS).......................................................... 82
3.1.4. Services developed by the IAEA........................................................... 84
3.1.4.1. OSART reviews.................................................................... 84
3.1.4.2. IRRS reviews......................................................................... 87
3.1.4.3. Other services and study frameworks set up by the IAEA.............................................................. 88
3.1.5. WANO.......................................................................................................... 90
3.1.6. NEA............................................................................................................... 91
3.1.7. Organizations dedicated to radiation protection and health........................................................................ 93
3.1.8. From bilateral Franco-German cooperation to European structures for the exchange and capitalization of knowledge and practices, training and assessment services........................................................ 94
3.1.9. Nuclear regulator associations............................................................. 100
3.2. The social dimension................................................................................................... 102
3.2.1. Introduction – the context in France................................................. 102
3.2.2. Examples of initiatives and issues raised concerning reactor safety in the French nuclear power plant fleet........... 102
Chapter 4
Nuclear Reactors: Complex Sociotechnical Systems – the Importance of Human and Organizational Factors
4.1. The introduction of human and organizational factors in the field of nuclear power reactors and lessons learned from the Three Mile Island nuclear power plant accident.............................. 108
4.2. The accident at the Chernobyl nuclear power plant and the concept of ‘safety culture’................................. 109
4.3. The Fukushima Daiichi nuclear power plant accident: the social dimension and the concept of organization ‘resilience’... 113
4.4. Changes in the perception of the role of people in achieving a high level of reliability in complex sociotechnical systems...... 113
4.5. Main topics studied in the development of resources and skills pertaining to human and organizational factors............ 116
4.5.1. Resources and skills................................................................................. 116
4.5.2. Main topics studied.................................................................................. 118
4.6. Human and organizational factors in French regulations................................ 119
Part 2
Safety by Design
Chapter 5
The Development of Nuclear Power Using Uranium-235 Fission – A Few Notions of Physics Used in Pressurized Water Reactors
5.1. Important milestones in the development of nuclear power using fission of the uranium-235 isotope.......................... 123
5.2. Fission and important concepts in reactor kinetics........................................... 127
5.3. Removing power from the core during operation............................................. 134
5.4. Decay heat...................................................................................................................... 135
5.5. Main features of pressurized water reactor cores.............................................. 136
5.6. Control and monitoring of pressurized water reactor cores.......................... 137
5.7. Using uranium and plutonium mixed oxide (MOX) fuel................................. 147
Chapter 6
General Objectives, Principles and Basic Concepts of the Safety Approach
6.1. General approach to risks – General objectives................................................. 152
6.2. Fundamental safety functions.................................................................................. 156
6.3. Confinement barriers................................................................................................... 157
6.4. Defence in depth........................................................................................................... 161
6.4. Defence in depth........................................................................................................... 161
6.4.1. Levels of defence in depth..................................................................... 162
6.4.2. Elements common to the different levels of defence in depth............................................................................. 169
6.5. Events considered: terminology adopted for nuclear power reactors......... 169
6.6. WENRA reference levels............................................................................................. 171
6.7. Deterministic safety analysis and probabilistic safety assessments............ 171
6.8. Lessons learned from the accident at the Fukushima Daiichi nuclear power plant on the concept of defence in depth and deterministic analysis......................................................................................... 173
6.9. Safety culture – Quality control.............................................................................. 174
Chapter 7
Safety Options and Considerations at the Design Phase
7.1. Different types of design provisions associated with safety considerations................................................. 188
7.2. Single-failure criterion................................................................................................. 189
7.3. The specific nature of computer-based systems (based on instrumentation and control software)............... 193
7.4. Equipment safety classification................................................................................ 195
7.4.1. Importance of equipment for safety and safety classification...................................................................... 195
7.4.2. Generic requirements associated with the different safety classes.................................................................. 199
7.4.3. Qualification of equipment for accident conditions..................... 202
7.5. Information on designing nuclear pressure equipment.................................... 207
7.6. General considerations on provisions for hazards in facility design............ 210
7.7. Anticipating decommissioning in the design stage........................................... 212
Chapter 8
Study of Operating Conditions in the Deterministic Safety Analysis
8.1. Categories of operating conditions......................................................................... 218
8.2. Choice of operating conditions................................................................................ 222
8.2.1. Concept of ‘bounding’ incident or accident..................................... 223
8.2.2. Accident exclusion.................................................................................... 224
8.3. List and breakdown of operating conditions....................................................... 225
8.4. Methods for studying operating conditions......................................................... 227
8.4.1. Choice of initial conditions, conservatism....................................... 228
8.4.2. Consideration of an aggravating event in the study on operating conditions – ‘Passive’ failures............ 229
8.4.3. Conventional combinations.................................................................. 231
8.4.4. Preventing accident aggravation......................................................... 232
8.4.5. Operator response time......................................................................... 232
8.4.6. Using qualified simulation software................................................... 233
8.4.7. Main criteria to be met for fuel in the reactor core..................... 234
8.5. Concept of ‘design-basis situations’ for equipment.......................................... 236
8.6. Situations to be taken into account in application of pressure equipment regulations................................................. 237
8.7. Assessing the radiological consequences of incidents, accidents and hazards.............................................. 238
8.7.1. Assessing radioactive substances released from the facility...... 240
8.7.2. Assessing radiological consequences of radioactive release from the facility................................................ 242
8.7.3. Assessing radiological consequences................................................. 243
Chapter 9
Loss-of-Coolant Accident
9.1. Short- and medium-term aspects of a LOCA...................................................... 252
9.1.1. Mechanical effects on vessel internals and fuel assembly structures................................................................ 253
9.1.2. Thermal-hydraulic aspects and behaviour of fuel rods................ 255
9.1.2.1. Large-break LOCA............................................................... 255
9.1.2.2. Intermediate-break LOCA................................................ 257
9.1.3. Effects on reactor containment and internals................................ 258
9.1.4. Long-term aspect...................................................................................... 259
9.2. Safety demonstration.................................................................................................. 261
9.2.1. General information and background................................................ 261
9.2.2. Fuel assemblies and fuel rods, vessel internals, reactor coolant system components................................................. 263
9.2.2.1. Mechanical strength of vessel internals, fuel assembly structures and reactor coolant system components.............. 263
9.2.2.2. Fuel behaviour...................................................................... 265
9.2.3. Reactor containment and equipment located inside.................... 266
Chapter 10
A Special Issue: Steam Generator Tubes
10.1. Steam generator tube rupture as a Category 3 event...................................... 270
10.2. Preventing an SGTR accident, risk of multiple ruptures................................... 272
10.3. Steam generator tube rupture(s) studied as a Category 4 event................. 274
10.3.1. 900 MWe and 1300 MWe reactors.................................................... 274
10.3.2. 1450 MWe reactors and EPR (Flamanville 3).................................. 274
10.4. Provisions to mitigate the radiological consequences of SGTR accidents. 276
Chapter 11
Providing for Hazards: General Considerations and Internal Hazards
11.1. General considerations on providing for hazards............................................... 279
11.2. Potential projectiles inside the containment....................................................... 282
11.3. Effects of pipe breaks.................................................................................................. 284
11.4. Projectiles generated by a turbine rotor failure.................................................. 285
11.5. Protection against load drops................................................................................... 288
11.5.1. Risks related to spent fuel transport packaging............................. 288
11.5.2. Other handling risks................................................................................ 291
11.6. Fire protection............................................................................................................... 292
11.7. Explosion protection.................................................................................................... 297
11.8. Internal flooding............................................................................................................ 300
Chapter 12
Providing for External Hazards
12.1. General considerations on providing for external hazards.............................. 305
12.2. ‘Climate watch’ implemented by EDF.................................................................... 307
12.3. Earthquakes..................................................................................................................... 307
12.4. External floods............................................................................................................... 322
12.5. Extreme temperatures................................................................................................. 330
12.5.1. Extreme cold.............................................................................................. 330
12.5.2. Extreme heat.............................................................................................. 331
12.6. Possible heat sink hazards.......................................................................................... 332
12.7. Other naturally-occurring external hazards......................................................... 336
12.8. Accidental aeroplane crashes (excluding malicious acts)................................ 336
12.9. Risks related to the industrial environment (excluding malicious acts)..... 340
Chapter 13
Complementary Domain of Events
13.1. The origin of studies belonging to the complementary domain................... 344
13.2. Background of the complementary domain........................................................ 344
13.3. Analysis of complementary domain events......................................................... 351
13.4. ‘New complementary domain’................................................................................. 351
13.5. Case of the Flamanville 3 EPR.................................................................................. 354
Chapter 14
Development and Use of Probabilistic Safety Assessments
14.1. History and regulatory context................................................................................ 357
14.1.1. International situation............................................................................ 357
14.1.2. Situation in France................................................................................... 359
14.2. Level 1 PSA...................................................................................................................... 361
14.2.1. Scope............................................................................................................ 361
14.2.2. Method for carrying out a Level 1 PSA............................................. 362
14.2.2.1. General information........................................................... 362
14.2.2.2. Specific point: probabilistic human reliability analysis................................................................................... 364
14.2.3. Level 1 PSA results and lessons learned............................................ 369
14.3. Level 2 PSA...................................................................................................................... 373
14.3.1. Scope............................................................................................................ 373
14.3.2. Method for carrying out a Level 2 PSA............................................. 374
14.3.2.1. General information........................................................... 374
14.3.2.2. Probabilistic human reliability analysis for Level 2 PSAs................................................................... 379
14.3.3. Examples of lessons learned from Level 2 PSAs............................. 382
14.3.3.1. Steam explosion risk assessment.................................. 382
14.3.3.2. Mechanical integrity of the 900 MWe reactor containments........................................................ 383
14.3.3.3. Isolating penetrations in the containment................. 383
14.3.3.4. Modifying the pressure relief system of the reactor coolant system........................................ 384
14.3.3.5. Improvement of operating procedures to reduce risk of core melt under pressure................ 384
14.3.3.6. Contribution of Level 2 PSAs to emergency response measures............................................................. 385
14.4. Expanding the scope of PSA coverage................................................................... 385
14.5. Using probabilistic safety assessments.................................................................. 386
14.5.1. Using PSAs in the design phase........................................................... 386
14.5.1.1. Usefulness and particularities of PSAs in the design phase............................................................. 386
14.5.1.2. PSAs conducted to support the Flamanville 3 EPR design............................................................................. 387
14.5.2. Using PSAs in periodic reviews............................................................ 389
14.5.2.1. Level 1 PSA............................................................................ 390
14.5.2.2. Level 2 PSA............................................................................ 391
14.5.3. Using PSAs for reactor operation........................................................ 392
14.5.3.1. Using PSAs to analyse event severity.......................... 392
14.5.3.2. Using PSAs to analyse operational limits and conditions and temporary changes...................... 394
14.5.3.3. Using PSAs to analyse operating procedures............. 396
Chapter 15
Aspects Specific to PWR Spent Fuel Storage Pools
15.1. Spent fuel pool design................................................................................................. 399
15.1.1. Confinement barriers.............................................................................. 399
15.1.2. Initiating events defined at the design stage.................................. 400
15.2. Experience feedback..................................................................................................... 401
15.2.1. Loss of cooling........................................................................................... 401
15.2.1.1. Loss of heat sink.................................................................. 401
15.2.1.2. Risks related to maintenance during unit outages................................................................................... 402
15.2.1.3. Suction of foreign matter into the cooling system.................................................................................... 403
15.2.1.4. Exceeding the decay heat defined in facility design...................................................................................... 403
15.2.2. Water losses............................................................................................... 404
15.2.2.1. Gate or sluice gate failures.............................................. 404
15.2.2.2. Line-up errors....................................................................... 405
15.2.2.3. Failure of a reactor coolant system pipe nozzle dam.......................................................................................... 409
15.2.2.4. Rupture of a pipe connected to the spent fuel pool................................................................................. 411
15.3. Safety reassessments................................................................................................... 411
15.4. Experience feedback from the accident that affected the Unit 4 pool at the Fukushima Daiichi nuclear power plant..... 414
15.4.1. Events........................................................................................................... 414
15.4.2. Complementary safety assessments conducted in France......... 417
15.5. Measures adopted for the EPR.................................................................................. 419
15.6. Recommendations for new reactor designs......................................................... 420
15.7. New systems for storing spent fuel........................................................................ 423
Chapter 16
Taking into Account Human and Organizational Factors in Facility Design
16.1. Taking into account human and organizational factors in nuclear power reactor design....................................... 425
16.1.1. Importance of considering human and organizational factors at the design stage....................................... 425
16.1.2. Approach at the design stage............................................................... 431
16.1.2.1. Prior to the design phase: analysis of ‘existing elements’........................................ 432
16.1.2.2. Design objectives................................................................ 435
16.1.2.3. Definition of detailed design provisions...................... 436
16.1.2.4. Validation of design provisions...................................... 439
16.1.2.5. Assessments conducted during reactor startup and after commissioning.................................................. 442
16.1.3. Project management and human and organizational factors engineering programme..................................... 443
16.2. Considering human and organizational aspects when designing changes to nuclear power plants..................... 443
16.2.1. Importance of human and organizational factors in designing modifications..................................................................... 444
16.2.2. ‘Human, social and organizational approach’ implemented by EDF........................................................................... 445
16.2.3. Changes, a subject that always deserves special attention from a human and organizational factors perspective.... 447
16.3. Human and organizational factors for future nuclear power reactor projects.................................................. 448
Chapter 17
Studying Core-Melt Accidents to Enhance Safety
17.1. Core degradation and vessel failure....................................................................... 452
17.1.1. Core uncovery............................................................................................ 453
17.1.2. Fuel degradation....................................................................................... 454
17.1.3. Failure of the reactor coolant system............................................... 455
17.1.4. Phenomena that can cause early containment failure................. 455
17.1.5. Phenomena that can ultimately lead to containment failure....................................................... 457
17.2. Containment failure modes....................................................................................... 457
17.3. Classification of releases associated with core-melt accidents − ‘source terms’................................................. 460
17.4. Improving knowledge.................................................................................................. 462
17.5. Studies in France on containment failure modes............................................... 462
17.5.1. Introduction................................................................................................ 462
17.5.2. Initial containment leakage................................................................... 463
17.5.3. Direct heating of gases in the containment.................................... 463
17.5.4. Hydrogen explosion in the containment.......................................... 464
17.5.5. Steam explosion in the vessel or reactor pit................................... 465
17.5.6. Gradual pressure increase in the containment............................... 466
17.5.7. Penetration of the concrete basemat of the containment by corium............................................ 467
17.5.8. ‘U4’ provisions........................................................................................... 468
17.5.9. Bypass of containment by outgoing pipes (the V mode).............................................................................................. 468
17.5.10. Fast reactivity insertion accidents...................................................... 469
17.6. Severe accident operating guidelines..................................................................... 469
17.7. Radiological consequences associated with the S3 source term and emergency response plans implemented by public authorities............ 470
17.8. Ultimate emergency operating procedures.......................................................... 473
17.9. On-site emergency plan............................................................................................. 473
17.10. Approach adopted for the EPR................................................................................. 476
17.10.1. General safety objectives....................................................................... 476
17.10.2. ‘Practical elimination’ of core-melt conditions that could lead to significant early releases............................. 476
17.10.3. Provisions for low-pressure core melt............................................... 480
Chapter 18
New-Generation Reactors
18.1. Organization and framework of Franco-German discussions........................ 485
18.2. Progression of safety objectives and design options for the EPR project.......................................................... 486
18.2.1. General safety objectives....................................................................... 486
18.2.2. Events to be taken into account at the design stage and in deterministic and probabilistic analyses.............................. 488
18.2.3. Main provisions for preventing incidents and accidents............. 490
18.2.4. Functional redundancy, independence between systems, system reliability................................................ 495
18.2.5. Confinement preservation..................................................................... 496
18.2.6. Radiological protection........................................................................... 497
18.2.7. Incorporating lessons learned from the Fukushima Daiichi nuclear power plant accident......................... 497
18.3. International context: general safety objectives for new-generation reactors............................................ 499
18.4. Concepts highlighted in new reactor designs...................................................... 501
18.4.1. AP1000: gravity systems........................................................................ 501
18.4.2. VVER: SPOT system................................................................................. 503
18.4.3. NM EPR: ‘multi-group’ technology, diversified heat sink............ 504
18.4.4. ATMEA 1: safety injection accumulators in the reactor coolant system.............................................................. 505
18.4.5. NuScale: common pool for modular reactors................................. 505
Part 3
Safety in Operation
Chapter 19
Startup Tests for Pressurized Water Reactors
19.1. Introduction.................................................................................................................... 511
19.2. Commissioning.............................................................................................................. 514
19.2.1. Defining startup tests.............................................................................. 514
19.2.2. Phasing of startup tests.......................................................................... 515
19.2.2.1. Preliminary and pre-operational tests......................... 515
19.2.2.2. Operational tests................................................................ 517
19.2.2.3. General principles for test sequencing and execution....................................................................... 517
19.2.3. Documentation for startup tests......................................................... 518
19.2.3.1. Integrated system test procedures and startup test procedures............................................ 518
19.2.3.2. Test programmes, test procedures, standard test guidelines.................................................... 518
19.2.3.3. Completeness analysis, adequacy analysis................. 518
19.2.3.4. Acceptance criteria............................................................. 519
19.3. Objectives and general rules to take into account for startup tests........... 520
19.4. Key lessons learned from startup tests on nuclear power reactors in France............................................................ 521
19.4.1. Qualification tests and on-site tests.................................................. 522
19.4.2. Long-term on-site testing...................................................................... 524
19.4.3. Test configurations and completeness, transpositions................ 525
19.4.4. Safety measures that cannot be verified by testing..................... 527
19.4.5. Criteria......................................................................................................... 527
19.4.6. Cleanness, keeping system lines clean, foreign matter................ 528
19.4.7. Piping support structures and displacement................................... 532
19.4.8. Pump and piping vibrations................................................................... 532
19.4.9. Validation of operating procedures and periodic tests................ 534
19.4.10. Uncertainty and ‘set points’.................................................................. 535
19.4.11. Condition of facilities during startup tests...................................... 536
19.4.12. Other aspects............................................................................................. 537
19.5. Examples of findings resulting from startup tests............................................. 538
Chapter 20
General Operating Rules
20.1. General Operating Rules............................................................................................ 552
20.1.1. Content of general operating rules..................................................... 553
20.1.2. Limits of general operating rules......................................................... 554
20.2. Operational limits and conditions........................................................................... 554
20.2.1. Content of operational limits and conditions................................. 555
20.2.1.1. Operating modes and standard states......................... 556
20.2.1.2. Requirements and unavailability................................... 558
20.2.1.3. Fallback states and time required to reach them.... 558
20.2.1.4. Events and event groups.................................................. 559
20.2.1.5. Combined types of unavailability.................................. 560
20.2.1.6. Concepts of ‘boundary condition’ and ‘specific requirement’................................................ 560
20.2.2. Average pressure and temperature range of the reactor coolant system.............................................................. 561
20.2.3. Changes in operational limits and conditions................................. 563
20.3. Initial and periodic tests............................................................................................. 564
20.4. Incident and accident operating procedures........................................................ 566
Chapter 21
Operating Experience Feedback from Events: Rules and Practices
21.1. Background...................................................................................................................... 569
21.2. Objectives of an operating experience feedback system................................ 571
21.3. Components of an operating experience feedback system – Regulations............................................................................... 572
21.4. Operating experience feedback practices adopted for the French nuclear power plant fleet................................. 576
Chapter 22
Operating Experience from Events Attributable to Shortcomings in Initial Reactor Design or the Quality of Maintenance
22.1. Events attributable to design shortcomings: core cooling deficiencies when reactor is shut down with water level in mid-loop operating range of the residual heat removal system (RHRS).......................................... 590
22.2. Recurrent loss of safety function events related to maintenance operations – Lessons learned........................... 595
22.2.1. Events........................................................................................................... 595
22.2.2. General discussion initiated by EDF in the late 1980s on the quality of maintenance operations....................................... 605
22.2.3. Applying the defence-in-depth concept when working on a reactor in service.................................. 608
22.2.4. Problems that may recur....................................................................... 609
Chapter 23
Operating Experience from Events Related to Maintenance Operations, Electrical Power Sources and Distribution, Internal and External Hazards
23.1. Risks of failure related to equipment or maintenance..................................... 612
23.1.1. Risks of common-mode failure............................................................ 612
23.1.1.1. Risks of common-mode failure related to settings............................................................................. 612
23.1.1.2. Risks of common-mode failure on electrical switchboards.................................................613
23.1.1.3. Unavailability of two out of three high-head safety injection lines in the cold legs of the reactor coolant system.............. 614
23.1.1.4. Loss of electrical power supplies................................... 617
23.1.2. Introduction of non-borated water into the reactor coolant system.......................................................... 620
23.1.3. Cooling the reactor coolant system after inhibition of automatic actions............................................................................... 622
23.1.4. A temporary device prevents switching the safety injection system to the water recirculation mode................... 624
23.2. Events related to internal hazards........................................................................... 625
23.2.1. Risk of common-mode failure due to internal flooding.............. 625
23.2.2. Risk of failure due to fire....................................................................... 628
23.2.3. Risks associated with the use of hydrogen in 900 MWe reactors............................................................................... 631
23.3. External hazards: events related to periods of extreme cold......................... 635
Chapter 24
Enhanced Protection of Estuary and River Sites: Flooding at the Blayais Nuclear Power Plant and Obstruction of a Water Intake
at the Cruas-Meysse Nuclear Power Plant
24.1. Partial loss of engineered safety systems following flooding of the Blayais nuclear power plant............................... 642
24.2. Total loss of heat sink due to clogging of filter drums by a massive influx of plant matter at the Cruas-Meysse nuclear power plant...................................................... 648
Chapter 25
Taking into Account Human and Organizational Factors in Facility Operation
25.1. Skills management....................................................................................................... 653
25.1.1. Historical background............................................................................. 654
25.1.2. Managing training.................................................................................... 655
25.1.3. Strategic workforce planning................................................................ 656
25.1.4. Personnel certification............................................................................ 657
25.2. Safety and risk management.................................................................................... 658
25.2.1. Historical background............................................................................. 658
25.2.2. Decision-making and safety.................................................................. 660
25.2.3. Risk analyses applied to work activities............................................ 662
25.2.4. Operating experience feedback........................................................... 663
25.2.5. Managing organizational change......................................................... 664
25.3. Managing operational activities............................................................................... 666
25.3.1. Characteristics of operational activities............................................ 666
25.3.2. Monitoring by the operating crew in the control room.............. 669
25.3.3. Compliance with general operating rules......................................... 670
25.3.4. Line-up......................................................................................................... 671
25.3.5. Operation in extreme situations......................................................... 672
25.4. Management of maintenance activities................................................................ 672
25.4.1. Management of a scheduled reactor outage for refuelling
and maintenance...................................................................................... 672
25.4.2. Risks during reactor outages................................................................. 673
25.4.3. Preparation for scheduled reactor outages...................................... 674
25.4.4. Managing scheduled reactor outages................................................ 675
25.5. Supervising outsourced activities............................................................................ 675
25.5.1. Contractor qualification and contracting......................................... 676
25.5.2. Matching workload and resources...................................................... 676
25.5.3. Carrying out work.................................................................................... 677
25.5.4. Surveillance of outsourced activities................................................. 678
25.5.5. Operating experience feedback and assessing
outsourced activities............................................................................... 678
Chapter 26
Facility Maintenance
26.1. Maintenance objectives.............................................................................................. 681
26.2. Maintenance................................................................................................................... 682
26.2.1. Definition.................................................................................................... 682
26.2.2. Maintenance strategies........................................................................... 683
26.3. Optimizing maintenance............................................................................................ 684
26.3.1. Reliability-centred maintenance.......................................................... 684
26.3.2. Conditional maintenance....................................................................... 686
XXVI Elements of nuclear safety – Pressurized water reactors
26.3.3. Conditional maintenance by sampling –
Maintenance based on reference equipment items...................... 688
26.3.4. ‘AP-913’ method....................................................................................... 689
26.4. Maintenance baselines................................................................................................ 691
26.5. On-site maintenance................................................................................................... 694
26.5.1. The various stages of maintenance operations.............................. 694
26.5.2. Main conditions for successful maintenance.................................. 696
26.5.3. Examples of anomalies or deviations discovered during routine
maintenance, explained by an inadequate maintenance
baseline with regard to deterioration mechanisms....................... 705
26.5.4. Examples of events associated with non-quality
maintenance............................................................................................... 708
26.5.4.1. Example of an event explained by an incorrect
setting on redundant equipment................................... 708
26.5.4.2. Example of an event explained by an incorrect
setting of electrical protection thresholds................. 708
26.5.4.3. Examples of events explained by a failure
to return equipment to a compliant state
after maintenance or an error in performing
a work procedure................................................................ 709
Chapter 27
In-service Monitoring and Inspection of Equipment
27.1. Main internal equipment items on a pressurized water reactor vessel...... 719
27.1.1. ‘Core baffle’ around the core................................................................ 720
27.1.2. RCCA guide tubes..................................................................................... 722
27.2. Reactor vessel, nozzles and head............................................................................. 722
27.2.1. Vessel underclad defects........................................................................ 724
27.2.2. Cracking on vessel head adaptors....................................................... 725
27.2.2.1. Condition of other reactors............................................. 727
27.2.2.2. Impact on safety................................................................. 728
27.2.2.3. Prevention, monitoring and mitigation....................... 729
27.2.2.4. Developing inspection tools............................................ 729
27.2.2.5. Repairs.................................................................................... 730
Contents XXVII
27.2.2.6. Leak detection...................................................................... 730
27.2.2.7. Anti-ejection devices......................................................... 731
27.2.2.8. Current situation................................................................. 731
27.2.2.9. Cracks observed on reactor vessel heads
in other countries............................................................... 731
27.2.2.10. Implementation of special monitoring
for ‘Inconel areas’ beginning in 1992........................... 732
27.2.3. Cracking on vessel lower head penetrations detected
in 2011......................................................................................................... 733
27.2.4. Monitoring the ‘beltline’ region of the vessel................................. 735
27.2.5. Defects observed on reactor vessels in Belgium............................ 736
27.3. Steam generators.......................................................................................................... 738
27.3.1. The different types of defects.............................................................. 738
27.3.2. Associated risks......................................................................................... 740
27.3.3. Monitoring during operation and inspection during outages.... 741
27.3.3.1. Monitoring during operation........................................... 741
27.3.3.2. Inspection during reactor outages................................. 742
27.3.4. Steps to be taken when a defect is detected.................................. 742
27.3.4.1. Tube wear due to foreign matter.................................. 743
27.3.4.2. Wear due to contact with anti-vibration bars.......... 743
27.3.4.3. Cracking in U-bend tubes................................................. 744
27.3.4.4. Tube deformation and cracking..................................... 744
27.3.5. Steam generator replacement.............................................................. 745
27.3.6. Clogging observed in the 2000s.......................................................... 746
27.3.7. Conclusion.................................................................................................. 748
27.4. Steam lines...................................................................................................................... 748
27.5. Auxiliary systems: cracks induced by local thermal-hydraulic
phenomena..................................................................................................................... 750
27.5.1. Cracking in non-isolatable sections connected
to the reactor coolant loops................................................................. 750
27.5.2. RHRS thermal fatigue at Unit 1 of the Civaux nuclear
power plant................................................................................................ 751
27.6. Civil works: containment structures....................................................................... 755
27.6.1. Anticipated degradation phenomena................................................. 756
27.6.2. Devices for direct monitoring of containment building
concrete walls............................................................................................ 757
27.6.3. Leak tests and measurements.............................................................. 759
27.6.4. Main anomalies......................................................................................... 759
Chapter 28
Fuel Management, Monitoring and Developments
28.1. Procedures for monitoring fuel rod integrity....................................................... 768
28.1.1. Radiochemical specifications for reactor coolant.......................... 768
28.1.2. Inspections and measurements carried out directly
on fuel assemblies.................................................................................... 777
28.1.2.1. Liquid penetrant testing in the refuelling
machine mast....................................................................... 778
28.1.2.2. Liquid penetrant testing in the FB cell......................... 779
28.1.2.3. Inspections performed on fuel rods.............................. 780
28.2. Operating experience feedback and changes in cladding material.............. 782
28.3. Anomalies and significant events involving fuel assemblies.......................... 786
28.3.1. Baffle jetting............................................................................................... 786
28.3.2. Fretting........................................................................................................ 787
28.3.3. Events encountered during handling operations............................ 789
28.3.4. Lateral deformation of fuel assemblies interfering
with RCCA drop......................................................................................... 791
Chapter 29
Facility Compliance
29.1. Introduction.................................................................................................................... 795
29.2. Detection and treatment of compliance deviations for pressurized
water reactors in the nuclear power plant fleet................................................. 796
29.2.1. Process for handling compliance gaps............................................... 796
29.2.2. Examples of compliance gaps............................................................... 798
29.2.2.1. Compliance gap in electrical connection boxes
qualified for accident conditions................................... 798
29.2.2.2. Failure in the seismic resistance of metal floors
in electrical and auxiliary buildings of 900 MWe
reactors (CPY series).......................................................... 799
29.2.2.3. Risk of containment sump screen blockage............... 799
29.2.2.4. Anomaly found in engines of emergency and SBO
diesel generators for 900 MWe reactors.................... 800
29.2.2.5. Temperature resistance fault in the high-head
safety injection pumps...................................................... 802
29.2.2.6. Mixed lubricants in equipment required
for accident situations....................................................... 802
29.2.2.7. Flow imbalance between safety injection lines
of 900 MWe reactors......................................................... 804
29.2.2.8. Anomaly in CATHARE software modelling
of natural circulation in the upper part
of the vessel.......................................................................... 805
29.2.2.9. Vibrations and rotor lift on engineered
safety motor-driven pump units.................................... 805
Chapter 30
Periodic Reviews
30.1. Introduction.................................................................................................................... 807
30.2. History of periodic reviews in France for nuclear power reactors................ 809
30.2.1. Reactors other than PWRs in the French nuclear power
plant fleet.................................................................................................... 809
30.2.2. PWRs in the French nuclear power plant fleet
(900 MWe, 1300 MWe and 1450 MWe).......................................... 811
30.3. Periodic review process for PWRs in the French nuclear power
plant fleet........................................................................................................................ 815
30.3.1. Regulations................................................................................................. 815
30.3.2. Outline of a PWR periodic review....................................................... 817
30.4. Case of the review associated with the third ten-yearly outage
of 900 MWe reactors................................................................................................... 822
30.4.1. Plant unit compliance reviews, the complementary
investigation and ageing management............................................. 824
30.4.1.1. Plant unit compliance reviews........................................ 824
30.4.1.2. Complementary investigation programme................ 825
30.4.1.3. Ageing management.......................................................... 826
30.4.2. Compliance studies on the design of civil works systems
and structures............................................................................................ 826
30.4.3. Studies to reassess system design...................................................... 827
30.4.4. Reassessment of reactor resistance to internal
and external hazards................................................................................ 830
30.4.5. Accident studies........................................................................................ 832
30.4.6. Taking into account lessons learned during the review
associated with the VD3 900 outage
for subsequent reviews........................................................................... 837
30.5. Fourth ten-yearly outage of 900 MWe reactors: integrating
the extension of the operating lifetime of nuclear power reactors
in France.......................................................................................................................... 838
30.5.1. Background................................................................................................. 838
30.5.2. Periodic Review Strategic Plan – Setting objectives..................... 839
30.5.3. A few significant issues identified in reviews conducted
by safety organizations........................................................................... 842
30.6. Overview of international practices – IAEA Guides........................................... 845
30.6.1. International practices............................................................................ 846
30.6.2. IAEA Guides................................................................................................ 847
30.7. Multilateral practices................................................................................................... 848
Chapter 31
Optimizing Radiation Protection and Limiting Doses Received
by Workers During Operations in a Nuclear Power Plant
31.1. Sources of ionizing radiation in a nuclear power reactor................................ 852
31.2. Examples of optimization of worker radiation protection.............................. 853
31.3. Arrangements for ‘Major Refit’ operations........................................................... 855
31.4. Approach and objectives adopted for the EPR.................................................... 858
Part 4
The Accidents at Three Mile Island,
Chernobyl and Fukushima Daiichi Nuclear Power Plants,
Lessons Learned and Emergency Response Management
Chapter 32
The Three Mile Island Nuclear Power Plant Accident
32.1. Accident sequence – Reconstitution through simulation................................ 864
32.2. Accident consequences............................................................................................... 872
32.3. Analysis of the accident causes................................................................................ 874
32.3.1. Error in identifying the position of the relief valve....................... 874
32.3.2. Understanding the behaviour of the pressurizer............................ 875
32.3.3. Stopping safety injection....................................................................... 876
32.3.4. Human-machine interface..................................................................... 876
32.3.5. Isolating the reactor containment...................................................... 877
32.3.6. Confinement inside the auxiliary building....................................... 877
32.3.7. Emergency feedwater supply to the steam generators............... 877
32.4. Lessons learned from the Three Mile Island accident....................................... 877
32.4.1. The human factor in facility operation............................................. 878
32.4.2. Importance of precursor events........................................................... 881
32.4.3. Study of complex situations and core-melt accidents,
handling emergency situations............................................................ 882
32.5. Conclusions..................................................................................................................... 883
Chapter 33
Incident and Accident Operation:
from the Event-Oriented Approach to the State-Oriented Approach
33.1. Limits of the event-oriented approach.................................................................. 885
33.2. The State-Oriented Approach concept.................................................................. 886
33.3. First application of the state-oriented approach............................................... 888
33.4. Widespread application of the state-oriented approach................................. 891
33.5. ‘Stabilized’ state-oriented approach....................................................................... 892
33.6. State-oriented approach adopted for the EPR.................................................... 894
Chapter 34
The Chernobyl Nuclear Power Plant Accident
34.1. The Chernobyl nuclear power plant and RBMK reactors................................. 897
34.2. The accident sequence................................................................................................ 901
34.3. Analysis of the accident causes and changes made to RBMK units
soon after the accident............................................................................................... 906
34.4. The other units at the facility................................................................................... 908
34.5. Radioactive release and protection of the population..................................... 908
34.5.1. Radioactive release kinetics.................................................................. 908
34.5.2. Protection of the population................................................................ 911
34.6. Consequences on human health and the environment................................... 915
34.6.1. Direct effects of radiation...................................................................... 915
34.6.2. Thyroid cancer in children..................................................................... 917
34.6.3. Long-term contamination in the Dnieper Basin............................. 919
34.7. Radioactive fallout in France and its consequences.......................................... 921
34.7.1. Doses attributable to the plume......................................................... 921
34.7.2. External doses due to soil deposition................................................ 921
34.7.3. Doses due to ingestion of contaminated foodstuffs.................... 922
34.7.4. Overall levels.............................................................................................. 923
34.7.5. Thyroid cancer........................................................................................... 924
34.7.5.1. Monitoring thyroid cancer in France............................ 925
34.7.5.2. Assessment of the number of cancer cases
induced in France by the Chernobyl accident........... 926
34.8. Lessons learned by the international community
from a general viewpoint and with regard to RBMK reactors....................... 927
34.9. Lessons learned in France........................................................................................... 928
34.10. Keeping the public informed..................................................................................... 931
34.11. After the Chernobyl accident.................................................................................... 933
Chapter 35
Options and Control of Reactivity Insertion
in Pressurized Water Reactors
35.1. Research and study of event sequences................................................................ 937
35.1.1. Cooling accidents..................................................................................... 938
35.1.2. Incidents and accidents related to rod cluster
control assemblies.................................................................................... 940
35.1.3. Boron dilution accidents........................................................................ 943
35.1.4. Inserting a cold water plug in the core............................................. 950
35.2. Changes in criteria........................................................................................................ 951
35.3. The case of outage states.......................................................................................... 953
35.4. Regulations...................................................................................................................... 957
Chapter 36
The Reactor Accident at the Fukushima Daiichi Nuclear Power Plant
and Lessons Learned in France
36.1. Reactor units at the Fukushima Daiichi nuclear power plant........................ 962
36.1.1. General operation of a boiling water reactor.................................. 962
36.1.2. Containment.............................................................................................. 962
36.1.3. Emergency cooling systems.................................................................. 964
36.2. Sequence of events during the accident............................................................... 966
36.3. Radioactive release....................................................................................................... 972
36.3.1. Airborne radioactive release, residual caesium deposits
and contamination of foodstuffs........................................................ 972
36.3.1.1. Airborne radioactive release............................................ 972
36.3.1.2. Persistent caesium deposits............................................ 972
36.3.1.3. Contamination of foodstuffs.......................................... 973
36.3.2. Release of radioactive substances in the Pacific Ocean.............. 975
36.3.3. Long-distance atmospheric dispersion
of the radioactive plume........................................................................ 976
36.4. Action taken to control facilities and released contaminated water.......... 978
36.5. Socioeconomic and health impact in numbers................................................... 981
36.5.1. Socioeconomic impact............................................................................ 981
36.5.2. Health impact............................................................................................ 982
36.6. Lessons learned from the accident.......................................................................... 985
36.6.1. Complementary safety assessments carried out in Europe
and France following the Fukushima Daiichi nuclear power
plant accident............................................................................................ 986
36.6.2. Complementary safety assessments carried out in France........ 988
36.6.3. Procedure for complementary safety assessments
carried out in France................................................................................ 989
36.6.4. Conclusions of the complementary safety assessments
carried out in France................................................................................ 990
36.6.5. The ‘hardened safety core’.................................................................... 991
36.6.5.1. Purpose................................................................................... 991
36.6.5.2. Principles................................................................................ 991
36.6.5.3. Illustrations........................................................................... 993
36.6.6. Nuclear Rapid Response Force (FARN).............................................. 995
36.6.7. Deployment of post-Fukushima measures
in French nuclear power plants............................................................ 997
36.7. Other lessons learned in France from the Fukushima Daiichi nuclear
power plant accident................................................................................................... 998
Chapter 37
Lessons Learned from the Fukushima Daiichi Nuclear Power
Plant Accident: Work Conducted by the IAEA and WENRA,
Action Taken in Countries Other than France
37.1. Work conducted by the IAEA.................................................................................... 1002
37.2. Work conducted by WENRA..................................................................................... 1003
37.3. Japan................................................................................................................................. 1004
37.4. Belgium............................................................................................................................ 1006
37.4.1. Nuclear power plants in Belgium........................................................ 1006
37.4.2. General details on the design of Belgian nuclear power
plants............................................................................................................ 1006
37.4.3. Stress tests and main lessons learned............................................... 1007
37.4.3.1. Improving protection of facilities against
external hazards.................................................................. 1007
37.4.3.2. Improving protection of facilities against loss
of electrical power supplies or loss of heat sink....... 1010
Contents XXXV
37.4.3.3. Improving on-site emergency plans............................. 1011
37.4.3.4. Improving management of core-melt accidents...... 1012
37.5. USA.................................................................................................................................... 1013
Chapter 38
Emergency Preparedness and Response
38.1. Defining a radiological emergency and ‘response’ objectives........................ 1021
38.2. General organization of radiological emergency management..................... 1023
38.2.1. Organization and entities concerned................................................. 1023
38.2.2. Major Nuclear or Radiological Accident National Response
Plan and emergency plans..................................................................... 1025
38.2.2.1. Major Nuclear or Radiological Accident National
Response Plan...................................................................... 1025
38.2.2.2. Emergency plans................................................................. 1026
38.2.2.3. Provisions for protecting the public in the event
of an accidental release of radioactivity..................... 1028
38.3. Management by the operator................................................................................... 1029
38.4. Prefectural authorities and mayors......................................................................... 1031
38.5. ASN, the Nuclear Safety Authority......................................................................... 1032
38.6. IRSN................................................................................................................................... 1033
38.7. Assessment approach in the event of an accident affecting a reactor
in the nuclear power plant fleet.............................................................................. 1038
38.7.1. ‘3D/3P’ method......................................................................................... 1039
38.7.2. The ‘aggravated prognosis’ approach................................................ 1041
38.7.3. Extending the 3D/3P method to severe accidents
(the ‘D/P AG’ method)............................................................................ 1041
38.8. Emergency preparedness............................................................................................ 1042
38.8.1. Emergency response exercises............................................................. 1043
38.8.2. Operating experience feedback........................................................... 1045
Part 5
PWR Safety Studies, R&D and Simulation Software
Chapter 39
PWR Safety Studies and R&D
39.1. Contribution of studies to the improvement of pressurized water
reactor safety................................................................................................................. 1050
39.2. Purpose and overview of R&D work, dedicated programmes
and organizations involved, and research facilities in France........................ 1052
39.2.1. Purpose and overview of R&D............................................................. 1052
39.2.2. Dedicated frameworks and organizations involved....................... 1063
39.2.3. Facilities in France used for research and development.............. 1065
Chapter 40
Examples of Simulation Software Developed for Safety Analysis
of Pressurized Water Reactors
40.1. Simulation software for neutronics........................................................................ 1076
40.2. Simulation software for thermal hydraulics (and mechanics)....................... 1079
40.3. Simulation software for thermal mechanics........................................................ 1084
40.4. Software for simulating core-melt situations..................................................... 1085
40.5. Simulation software for mechanics........................................................................ 1088
40.6. Fire simulation software............................................................................................. 1089
List of Acronyms.......................................................................................................................... 1093
Acronyms for institutions, bodies and groups..................................................................... 1093
Technical acronyms and abbreviations.................................................................................. 1103
Technical glossary......................................................................................................................... 1121
Compléments
Characteristics
Language(s): English
Audience(s): Professionals, Research, Students
Publisher: EDP Sciences
Collection: Institut de Radioprotection et de Sûreté Nucléaire
Published: 25 august 2022
EAN13 Paper book: 9782759827220
EAN13 eBook [PDF]: 9782759827237
EAN13 eBook [ePub]: 9782759827244
Interior: Colour
Format (in mm) Paper book: 160 x 240
Pages count Paper book: 1170
Pages count eBook [PDF]: 1168
Size: 84.9 MB (PDF), 29.5 MB (ePub)
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