Petrangeli Nuclear Safety

1. Auflage 2006
ISBN: 978-0-08-046078-9
Verlag: Elsevier Science & Techn.
Format: EPUB
Kopierschutz: 6 - ePub Watermark

E-Book, Englisch, 447 Seiten

Nuclear Safety provides the methods and data needed to evaluate and manage the safety of nuclear facilities and related processes using risk-based safety analysis, and provides readers with the techniques to assess the consequences of radioactive releases. The book covers relevant international and regional safety criteria (US, IAEA, EUR, PUN, URD, INI). The contents deal with each of the critical components of a nuclear plant, and provide an analysis of the risks arising from a variety of sources, including earthquakes, tornadoes, external impact and human factors. It also deals with the safety of underground nuclear testing and the handling of radioactive waste.
Covers all plant components and potential sources of risk including human, technical and natural factors.Brings together information on nuclear safety for which the reader would previously have to consult many different and expensive sources.Provides international design and safety criteria and an overview of regulatory regimes.Includes case studies and analysis of major accidents with data and calculations on accompanying website.

Autoren/Hrsg.

Petrangeli, Gianni

Inhaltsverzeichnis

1;Front cover;1
2;Title page;4
3;Copyright;5
4;Table of contents;6
5;Preface;14
6;Acknowledgements;16
7;Chapter 1 Introduction;18
7.1;1-1. Objectives;18
7.2;1-2. A short history of nuclear safety technology;19
7.2.1;1-2-1. The early years;19
7.2.2;1-2-2. From the late 1950s to the Three Mile Island accident;19
7.2.3;1-2-3. From the Three Mile Island accident to the Chernobyl accident;24
7.2.4;1-2-4. The Chernobyl accident and after;25
7.3;References;27
7.4;Chapter notes;27
8;Chapter 2 Inventory and localization of radioactive products in the plant;30
8.1;References;32
9;Chapter 3 Safety systems and their functions;34
9.1;3-1. Plant systems;34
9.2;3-2. Safety systems and accidents;35
9.3;3-3. Future safety systems and plant concepts;40
9.3.1;3-3-1. General remarks;40
9.3.2;3-3-2. Some passive safety systems for nuclear plants;44
9.3.3;3-3-3. Inherently safe systems in the process industries;47
9.4;References;49
9.5;Chapter notes;49
10;Chapter 4 The classification of accidents and a discussion of some examples;52
10.1;4-1. Classification;52
10.2;4-2. Design basis accidents;52
10.2.1;4-2-1. Some important data for accident analysis;52
10.2.2;4-2-2. Example of a category 2 accident: spurious opening of a pressurizer safety valve;57
10.2.3;4-2-3. Example of a category 3 accident: instantaneous power loss to all the primary pumps;58
10.2.4;4-2-4. Example of a category 4 accident: main steam line break;60
10.2.5;4-2-5. Example of a category 4 accident: sudden expulsion of a control rod from the core;61
10.2.6;4-2-6. Example of a category 4 accident: break of the largest pipe of the primary system (large LOCA);63
10.2.7;4-2-7. Example of a category 4 accident: fuel handling accident;64
10.2.8;4-2-8. Area accidents;67
10.3;4-3. Beyond design basis accidents;68
10.3.1;4-3-1. Plant originated accidents;68
10.3.2;4-3-2. Accidents due to human voluntary actions;68
10.4;4-4. External accidents of natural origin;68
10.5;References;68
10.6;Chapter notes;68
11;Chapter 5 Severe accidents;70
11.1;5-1. Existing plants;70
11.2;5-2. Future plants: extreme and practicable solutions;72
11.3;5-3. Severe accident management: the present state of studies and implementations;74
11.4;5-4. Data on severe accidents;75
11.5;5-5. Descriptions of some typical accident sequences;75
11.5.1;5-5-1. Loss of station electric power supply (TE = transient + loss of electrical supply);75
11.5.2;5-5-2. Loss of electric power with LOCA from the pump seals (SE = small LOCA + loss of electric power);78
11.5.3;5-5-3. Interfacing systems LOCA (V);78
11.5.4;5-5-4. Large LOCA with failure of the recirculation (ALFC);79
11.5.5;5-5-5. Small LOCA with failure of the recirculation (SLFC);79
11.6;5-6. ‘Source terms’ for severe accidents;79
11.7;References;81
12;Chapter 6 The dispersion of radioactivity releases;82
12.1;6-1. The most interesting releases for safety evaluations;82
12.2;6-2. Dispersion of releases: phenomena;83
12.3;6-3. Release dispersion: simple evaluation techniques;87
12.4;6-4. Formulae and diagrams for the evaluation of atmospheric dispersion;88
12.5;Reference;93
12.6;Chapter notes;93
13;Chapter 7 Health consequences of releases;96
13.1;7-1. The principles of health protection and safety;96
13.2;7-2. Some quantities, terms and units of measure of health physics;96
13.3;7-3. Types of effects of radiation doses and limits;97
13.4;7-4. Evaluation of the health consequences of releases;98
13.4.1;7-4-1. Evaluation of inhalation doses from radioactive iodine;98
13.4.2;7-4-2. Evaluation of doses due to submersion in a radioactive cloud;98
13.4.3;7-4-3. Evaluation of the doses of radiation from caesium-137 deposited on the ground (‘ground-shine’ dose);98
13.4.4;7-4-4. Evaluation of the dose due to deposition of plutonium on the ground;98
13.4.5;7-4-5. Indicative evaluation of long distance doses for very serious accidents to nuclear reactors;99
13.4.6;7-4-6. Direct radiation doses;99
13.5;Reference;100
13.6;Chapter notes;100
14;Chapter 8 The general approach to the safety of the plant-site complex;102
14.1;8-1. Introduction;102
14.2;8-2. The definition of the safety objectives of a plant on a site;102
14.2.1;8-2-1. The objectives and limits of release/dose;102
14.3;8-3. Some plant characteristics for the prevention and mitigation of accidents;103
14.4;8-4. Radiation protection characteristics;103
14.5;8-5. Site characteristics;104
15;Chapter 9 Defence in depth;106
15.1;9-1. Definition, objectives, levels and barriers;106
15.2;9-2. Additional considerations on the levels of Defence in Depth;106
16;Chapter 10 Quality assurance;110
16.1;10-1. General remarks and requirements;110
16.2;10-2. Aspects to be underlined;110
16.3;Reference;111
17;Chapter 11 Safety analysis;112
17.1;11-1. Introduction;112
17.2;11-2. Deterministic safety analysis;112
17.3;11-3. Probabilistic safety analysis;114
17.3.1;11-3-1. Event trees;115
17.3.2;11-3-2. Fault trees;116
17.3.3;11-3-3. Failure rates;122
17.4;References;122
17.5;Chapter notes;122
18;Chapter 12 Safety analysis review;124
18.1;12-1. Introduction;124
18.2;12-2. The reference points;124
18.3;12-3. Foreseeing possible issues for discussion;124
18.4;12-4. Control is not disrespectful;125
18.5;12-5. Clarification is not disrespectful;126
18.6;12-6. Designer report;127
18.6.1;12-6-1. Introduction;127
18.6.2;12-6-2. Conclusions;127
18.6.3;12-6-3. Hydrodynamic aspects;127
18.6.4;12-6-4. Effective mass of oscillating system;128
18.6.5;12-6-5. Evaluation of fluid damping;128
18.6.6;12-6-6. Vibration analysis;128
18.7;12-7. Discussion;131
18.8;References;132
18.9;Chapter notes;132
19;Chapter 13 Classification of plant components;134
19.1;Reference;135
20;Chapter 14 Notes on some plant components;136
20.1;14-1. Reactor pressure vessel;136
20.1.1;14-1-1. Problems highlighted by operating experience;136
20.1.2;14-1-2. Rupture probability of non-nuclear vessels;137
20.1.3;14-1-3. Failure probability of nuclear vessels;139
20.1.4;14-1-4. Vessel material embrittlement due to neutron irradiation;141
20.1.5;14-1-5. Pressurized thermal shock;143
20.1.6;14-1-6. The reactor pressure vessel of Three Mile Island 2;143
20.1.7;14-1-7. General perspective on the effect of severe accidents on the pressure vessel;144
20.1.8;14-1-8. Recommendations for the prevention of hypothetical accidents generated by the pressure vessel;145
20.2;14-2. Piping;147
20.2.1;14-2-1. Evolution of the regulatory positions;147
20.2.2;14-2-2. Problems indicated by experience;147
20.2.3;14-2-3. Leak detection in water reactors;149
20.2.4;14-2-4. Research programmes on piping;150
20.3;14-3. Valves;151
20.3.1;14-3-1. General remarks;151
20.3.2;14-3-2. Some data from operating experience;151
20.3.3;14-3-3. The most commonly used types of valve;152
20.3.4;14-3-4. Types of valve: critical areas, design and operation;153
20.3.5;14-3-5. Valve standards;157
20.4;14-4. Containment systems;158
20.5;References;159
21;Chapter 15 Earthquake resistance;162
21.1;15-1. General aspects, criteria and starting data;162
21.2;15-2. Reference ground motion;165
21.3;15-3. Structural verifications;175
21.3.1;15-3-1. Foundation soil resistance;175
21.3.2;15-3-2. Resistance of structures;179
21.4;References;199
22;Chapter 16 Tornado resistance;202
22.1;16-1. The physical phenomenon;202
22.2;16-2. Scale of severity of the phenomenon;203
22.3;16-3. Design input data;203
22.4;Reference;204
23;Chapter 17 Resistance to external impact;206
23.1;17-1. Introduction;206
23.2;17-2. Aircraft crash impact;206
23.2.1;17-2-1. Effects of an aircraft impact;206
23.2.2;17-2-2. Overall load on a structure;206
23.2.3;17-2-3. Vibration of structures and components;208
23.2.4;17-2-4. Local perforation of structures;208
23.2.5;17-2-5. The effect of a fire;209
23.2.6;17-2-6. Temporary incapacity of the operating personnel;209
23.3;17-3. Pressure wave;209
23.4;17-5. Other impacts;210
23.5;References;211
24;Chapter 18 Nuclear safety criteria;212
24.1;18-1. General characteristics;212
24.2;18-2. The US general design criteria;212
24.3;18-3. IAEA criteria;213
24.4;18-4. EUR criteria;213
24.5;18-5. Other general criteria compilations;214
24.6;References;215
24.7;Chapter notes;215
25;Chapter 19 Nuclear safety research;216
25.1;Reference;216
26;Chapter 20 Operating experience;218
26.1;20-1. Introduction;218
26.2;20-2. Principal sources;218
26.3;20-3. Some significant events;218
26.3.1;20-3-1. Mechanical events;218
26.3.2;20-3-2. Electrical events;219
26.3.3;20-3-3. System events;219
26.3.4;20-3-4. Area events;220
26.3.5;20-3-5. Reactivity accidents;221
26.3.6;20-3-6. Possible future accidents;221
26.4;20-4. The International Nuclear Event Scale;222
26.5;References;224
27;Chapter 21 Underground location of nuclear power plants;226
27.1;References;229
28;Chapter 22 The effects of nuclear explosions;232
28.1;22-1. Introduction;232
28.2;22-2. Types of nuclear bomb;232
28.3;22-3. The consequences of a nuclear explosion;232
28.4;22-4. Initial nuclear radiation;234
28.5;22-5. Shock wave;234
28.6;22-6. Initial thermal radiation;235
28.7;22-7. Initial radioactive contamination (‘fallout’);235
28.8;22-8. Underground nuclear tests;235
28.8.1;22-8-1. Historical data on nuclear weapons tests;235
28.8.2;22-8-2. The possible effects of an underground nuclear explosion;236
28.8.3;22-8-3. The possible radiological effects of the underground tests;237
28.9;References;237
29;Chapter 23 Radioactive waste;238
29.1;23-1. Types and indicative amounts of radioactive waste;238
29.2;23-2. Principles;239
29.3;Reference;240
30;Chapter 24 Fusion safety;242
30.1;References;245
31;Chapter 25 Safety of specific plants and of other activities;246
31.1;25-1. Boiling water reactors;246
31.2;25-2. Pressure tube reactors;248
31.3;25-3. Gas reactors;248
31.4;25-4. Research reactors;249
31.5;25-5. Sodium-cooled fast reactors;249
31.6;25-6. Fuel plants;250
31.7;25-7. Nuclear seawater desalination plants;250
31.8;25-8. VVER plants;251
31.9;25-9. Ship propulsion reactors;251
31.10;25-10. Safe transport of radioactive substances;251
31.11;25-11. Safety of radioactive sources and of radiation generating machines;251
31.12;References;253
32;Chapter 26 Nuclear facilities on satellites;254
32.1;26-1. Types of plant;254
32.2;26-2. Possible accidents and their consequences;255
32.3;Reference;255
33;Chapter 27 Erroneous beliefs about nuclear safety;256
33.1;References;258
34;Chapter 28 When can we say that a particular plant is safe?;260
35;Chapter 29 The limits of nuclear safety: the residual risk;262
35.1;29-1. Risk in general;262
35.2;29-2. Risk concepts and evaluations in nuclear installation safety;262
35.2.1;29-2-1. Tolerable risk;262
35.2.2;29-2-2. Risk-informed decisions;263
35.3;29-3. Residual risk: the concept of loss-of-life expectancy;264
35.4;29-4. Risk from various energy sources;264
35.5;29-5. Risk to various human activities;265
35.6;29-6. Are the risk analyses of nuclear power plants credible?;265
35.7;29-7. Proliferation and terrorism;267
35.8;References;267
36;Additional references;268
37;Appendix 1 The Chernobyl accident;296
37.1;A1-1. Introduction;296
37.2;A1-2. The reactor;296
37.3;A1-3. The event;298
37.4;References;301
38;Appendix 2 Calculation of the accident pressure in a containment;302
38.1;A2-1. Introduction;302
38.2;A2-2. Initial overpressure;302
38.3;A2-3. Containment pressure versus time;303
38.3.1;A2-3-1. Introductory remarks;304
38.3.2;A2-3-2. Calculation method;304
38.3.3;A2-3-3. Heat exchanged with the outside through the metal container;305
38.3.4;A2-3-4. Heat released by hot metals;305
38.3.5;A2-3-5. Heat exchanged with cold metals;306
38.3.6;A2-3-6. Heat exchanged with concrete layers;306
38.3.7;A2-3-7. Decay heat;307
38.3.8;A2-3.8. Heat removed by the spray system internal to the containment;308
38.3.9;A2-3-9. Solar heat;308
38.3.10;A2-3-10. Thermal balance in the interval ??;309
38.3.11;A2-3-11. Considerations on the performance of the calculation and on the choice of the input data;309
38.3.12;A2-3-12. Example calculation;310
38.4;References;313
39;Appendix 3 Table of safety criteria;314
40;Appendix 4 Dose calculations;332
40.1;A4-1. Introduction;332
40.2;A4-2. Virtual population dose in a severe accident;332
40.2.1;A4-2-1. The reactor and the released isotopes;332
40.2.2;A4-2-2. Source term at three days (I, Cs, Xe);332
40.2.3;A4-2-3. Dose at the fence after three days of exposure;333
40.2.4;A4-2-4. Ground shine long-term dose;333
40.3;A4-3. Explorative evaluation of the radiological consequences of a mechanical impact on a surface storage facility for category 2 waste;333
40.3.1;A4-3-1. Type of repository;333
40.3.2;A4-3-2. Reference impact;333
40.3.3;A4-3-3. Fragmentation and dispersion of material;334
40.3.4;A4-3-4. Doses;335
40.3.5;A4-3-5. Conclusions;336
40.4;A4-4. Explorative evaluation of the radiological consequences of a mechanical impact on a transport/storage cask containing spent fuel;336
40.4.1;A4-4-1. Characteristics of the cask;336
40.4.2;A4-4-2. Reference impact;336
40.4.3;A4-4-3. Amount of significant fission products in the internal atmosphere of the cask and external release in one day;336
40.4.4;A4-4-4. Effective committed doses;337
40.4.5;A4-4-5. Conclusions;338
40.5;References;338
41;Appendix 5 Simplified thermal analysis of an insufficiently refrigerated core;340
41.1;A5-1. Analysis of the core without refrigeration;340
41.2;A5-2. Other formulae and useful data for the indicative study of the cooling of a core after an accident;342
41.3;References;343
42;Appendix 6 Extracts from EUR criteria (December 2004);344
42.1;2-1-8-3. List of design basis conditions;344
42.2;2-1-8. TABLES;345
42.2.1;2-1-8-1. Table 1: Radiological criteria for radioactive releases in normal operation and incident conditions;345
42.2.2;2-1-B-1. Criteria for limited impact for DEC;345
42.2.3;2-1-B 1-1. Table B1: Criteria for limited impact for no emergency action beyond 800 m from the reactor;346
42.2.4;2-1-B 1-2. Table B2: Criteria for limited impact for no delayed action beyond 3 km from the reactor;347
42.2.5;2-1-B 1-3. Table B3: Criteria for limited impact for no long-term actions beyond 800 m from the reactor;347
42.2.6;2-1-B 1-4. Table B4: Criteria for limited impact for economic impact;347
42.2.7;2-1-B-2 Release targets for design basis category 3 and 4 conditions;347
42.2.8;2-1-B-2-1. Table B5: DBA release targets for no action beyond 800 m from the reactor;348
42.2.9;2-1-B-2-2. Table B6: DBA release targets for economic impact;348
42.2.10;2-1-2-3. Operational staff doses during normal operation and incidents;348
42.2.11;2-1-2-6. Probabilistic safety targets;349
42.2.12;2-1-3-4. Single failure criterion;349
42.2.13;2-1-6-8. Classification of the safety functions and categorisation of the equipment;350
42.2.14;2-1-6-6-3. Requirements according to level of safety functions;351
42.2.15;2-1-6-8-4. Assignment of equipment and structures to a safety category;351
42.2.16;2-1-6-8-5. Requirements on equipment and structures according to safety category;352
42.2.17;2-1-6-8-6. Classification of structures and equipment according to the design and construction codes;352
42.2.18;2-1-6-8-7. The relation of seismic categorisation to safety level of functions;352
42.2.19;2-1-6-13. Accident management;352
42.2.20;2-1-6-14. Radiation protection;353
43;Appendix 7 Notes on fracture mechanics;354
43.1;A7-1. Introduction;354
43.2;A7-2. Current practice;355
43.3;References;358
44;Appendix 8 US general design criteria;360
44.1;A8-1. Introduction;361
44.2;A8-2. Definitions and explanations;362
44.3;A8-3. Criteria;362
44.3.1;A8-3-1. Overall requirements;362
44.3.2;A8-3-2. Protection by Multiple Fission Product Barriers;363
44.3.3;A8-3-3. Protection and Reactivity Control Systems;365
44.3.4;A8-3-4. Fluid Systems;366
44.3.5;A8-3-5. Reactor Containment;368
44.3.6;A8-3-6. Fuel and Radioactivity Control;369
44.4;Notes;370
45;Appendix 9 IAEA criteria;372
46;Appendix 10 Primary depressurization systems;374
46.1;A10-1. Initial studies;374
46.2;A10-2. Depressurization systems for modern design reactors;376
46.3;References;380
47;Appendix 11 Thermal-hydraulic transients of the primary system;382
47.1;A11-1. General remarks;382
47.2;A11-2. General program characteristics;383
47.3;A11-3. Program description;383
47.3.1;A11-3-1. Macro Stampa dati;383
47.3.2;11-3-2. Macro Copia_dati;385
47.3.3;11-3-3. Macro HF;385
47.3.4;11-3-4. Macro HFG;386
47.3.5;11-3-5. Macro VF;386
47.3.6;11-3-6. Macro VFG;387
47.3.7;11-3-7. Macro QS;387
47.3.8;11-3-8. Macro GU;387
47.3.9;11-3-9. Macro GE;389
47.3.10;11-3-10. Macro DT;390
47.3.11;11-3-11. Macro PS;390
47.4;A11-4. Using the program;394
47.5;A11-5. Other formulae for the expanded use of the program;394
47.5.1;A11-5-1. ATWS;394
47.5.2;A11-5-2. Pressure in a depressurization water discharge tank;395
47.6;References;395
48;Appendix 12 The atmospheric dispersion of releases;396
49;Appendix 13 Regulatory framework and safety documents;402
49.1;A13-1. Regulatory framework;402
49.2;A13-2. Safety documents;402
49.2.1;A13-2-1. The safety report;403
49.2.2;A13-2-2. The probabilistic safety assessment;405
49.2.3;A13-2-3. The environmental impact assessment;405
49.2.4;A13-2-4. The external emergency plan;405
49.2.5;A13-2-5. The operation manual, including the emergency procedures;405
49.2.6;A13-2-6. Operation organization document;407
49.2.7;A13-2-7. The pre-operational test programme;407
49.2.8;A13-2-8. The technical specifications for operation;407
49.2.9;A13-2-9. The periodic safety reviews;408
49.3;References;408
50;Appendix 14 USNRC Regulatory Guides and Standard Review Plan;410
50.1;A14-1. Extracts from a regulatory guide;410
50.2;A14-2. List of contents and extracts from a sample chapter of the Standard Review Plan;412
50.3;A14-3. Sample chapter;417
51;Appendix 15 Safety cage;422
51.1;A15-1. General remarks;422
51.2;A15-2. Available energy;422
51.3;A15-3. Mechanical energy which can be released;422
51.4;A15-4. Overall sizing of a structural cage around the pressure vessel;423
51.5;A15-5. Experimental tests on steel cages for the containment of vessel explosions;425
51.6;Reference;425
52;Appendix 16 Criteria for the site chart (Italy);426
52.1;A16-1. Population and land use;426
52.2;A16-2. Geology, seismology and soil mechanics;426
52.3;A16-3. Engineering requirements;427
52.4;A16-4. Extreme events from human activities;427
52.5;A16-5. Extreme natural events;427
53;Appendix 17 The Three Mile Island accident;428
53.1;A17-1. Summary description of the Three Mile Island no.2 Plant;428
53.2;A17-2. The accident;430
53.3;A17-3. The consequences of the accident on the outside environment;436
53.4;A17-4. The actions initiated after the accident;438
53.5;References;439
54;Glossary;440
55;Web sites;442
56;Index;444