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ELEMENTS OF NUCLEAR SAFETY

PRESSURIZED WATER REACTORS

by Jean Couturier (editorial coordination)
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Presentation

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.

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: 85 Mo (PDF), 29 Mo (ePub)