Keywords:
Nuclear engineering -- Safety measures.
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Nuclear reactors -- Safety measures -- France.
;
Electronic books.
Description / Table of Contents:
This basically educational book is intended for all involved in nuclear facility safety. It dissects the principles and experiences conducive to the adoption of attitudes compliant with what is now known as safety culture. This book is accessible to a wide range of readers.
Type of Medium:
Online Resource
Pages:
1 online resource (558 pages)
Edition:
1st ed.
ISBN:
9782759801190
URL:
https://ebookcentral.proquest.com/lib/geomar/detail.action?docID=309866
DDC:
539.77
Language:
English
Note:
Intro -- Contents -- Introduction -- 1. Radioactivity and the biological effects of ionizing radiation -- 1.1. Units used -- 1.2. Natural radioactivity -- 1.3. Biological effects of ionizing radiation -- 1.4. Radiation protection principles -- 2. Nuclear safety organization -- 2.1. Nuclear security and safety -- 2.2. Nuclear safety organization and responsibility sharing -- 2.3. Safety analysis reports and regulations -- 2.4. Developments in safety goals -- 2.5. Safety Culture -- 3. Deterministic safety approach -- 3.1. Determination of specific risks -- 3.2. Potential risks, residual risks, acceptable risks -- 3.3. The barriers -- 3.4. The defense in depth concept -- 3.5. Quality Control -- 4. Analysis of operating conditions -- 4.1. Classification of operating conditions -- 4.2. Definition of design basis operating condition categories -- 4.3. Choice of operating conditions -- 4.4. Operating conditions: list and subdivisions -- 4.5. Operating condition analysis process -- 4.6. Consideration of internal or external hazards -- 5. Assessment of the radiological consequences of accidents -- 5.1. Quantities of radioactive products involved -- 5.2. Release rates -- 5.3. Transfer and deposit in reactor systems -- 5.4. Transfer and deposit in buildings -- 5.5. Leak rate to the outside atmosphere and filtering provisions -- 5.6. Environmental transport and deposit conditions -- 5.7 Pathways to man -- 5.8. Dose conversion factors -- 5.9. Changes in radiological consequence calculation methods -- 6. An example of accident analysis: LOCA -- 6.1. Physical effects of a large break -- 6.2. Assumptions adopted in safety analysis -- 6.3. Acceptability criteria and results -- 6.4. Evaluation of radiological consequences -- 6.5. Safety demonstration evolution -- 7. Assessment of safety justifications -- 7.1. Data drawn from operating condition studies.
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7.2. Checking the number of lines of defense -- 7.3. New safety demonstration requirements for the N4 series -- 8. A particular barrier point: the steam generator tubes -- 8.1. Steam generator tube rupture without human intervention -- 8.2. Complementary French studies -- 8.3. Dealing with the problem for the N4 series -- 9. Internal hazards -- 9.1. Missiles from inside the containment -- 9.2. The results of piping breaks -- 9.3. Turbogenerator bursting -- 9.4. Protection against load dropping -- 9.5. Fire protection -- 9.6. Internal flooding -- 10. External hazards -- 10.1. Determination of earthquake hazards -- 10.2. Protection against aircraft crashes -- 10.3. Industrial hazards -- 10.4. Floods -- 10.5. Protection against other external hazards -- 11. Complementary operating conditions -- 11.1. Origins -- 11.2. The position of the safety authorities -- 11.3. Complementary operating conditions -- 12. Probabilistic assessment of an accident sequence -- 12.1. Effects of failures and initial assumptions -- 12.2. Chronological list of the elements forming the scenario -- 12.3. Required data -- 12.4. Assessment results -- 12.5. Revision of scenarios and their probabilities -- 13. The accident at Three Mile Island -- 13.1. The accident -- 13.2. Causes of the accident -- 13.3. Lessons learned from the accident -- 14. The state-oriented approach -- 14.1. Limits of the event-related approach -- 14.2. Development of the state-oriented approach -- 14.3. First application of the state-oriented approach -- 14.4 Generalization of the state-oriented approach -- 14.5. Safety panels -- 15. Preparation for the management of severe accidents -- 15.1. Core and vessel degradation -- 15.2. The Rasmussen report -- 15.3. "Source terms -- 15.4. Severe accident management studies in France -- 15.5. Radiological consequences of source term S3 and intervention provisions.
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15.6. List of ultimate emergency procedures -- 15.7. Summary of procedures -- 15.8. Internal Emergency Plan -- 15.9. The fourth level of defense in depth -- 16. Special risks associated with criticality accidents -- 16.1. Theoretical scenario -- 16.2. A plausible scenario and corrective measures -- 16.3. Identification of other dilution scenarios -- 16.4. Other criticality accident hazards -- 16.5. International information -- 17. Emergency preparedness and IPSN resources -- 17.1. Emergency preparedness -- 17.2. Role of the IPSN crisis team -- 17.3. Method and tools of the assessment cell -- 17.4. Methods and tools of the radiological consequences cell -- 17.5. Conclusion on the method and tools -- 17.6. External Emergency Plan -- 17.7. Environmental transfer and deposit conditions -- 18. Severe accident research and development work -- 18.1. Thermal hydraulic codes -- 18.2. Fission product codes -- 18.3. Fission product experiments -- 18.4. Corium and containment building behavior studies -- 18.5. Other on-going surveys -- 19. Probabilistic safety assessment -- 19.1. Initiation of the studies -- 19.2. Aims and organization of the studies -- 19.3. Core meltdown probability assessment method -- 19.4. Specificities of French studies -- 19.5. Results of the 900 PSA survey -- 19.6. Results of the 1300 PSA -- 19.7. Comparison with studies undertaken abroad -- 20. Applications and development of probabilistic studies -- 20.1. Use of probabilistic safety studies -- 20.2. Development of these studies and tools -- 20.3. Probabilistic assessment of radioactive release -- 20.4. Conclusions on the probabilistic safety studies -- 21. The Chernobyl accident -- 21.1. The Chernobyl plant and the RBMK reactors -- 21.2. The accident -- 21.3. The release and its consequences -- 21.4. Causes of the accident and lessons learned.
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21.5. Future of the other Chernobyl units -- 21.6. Lessons drawn in France -- 21.7. Information of the general public and communication -- 21.8. After Chernobyl -- 22. General operating rules -- 22.1. General operating rules -- 22.2. Technical Operating Specifications -- 22.3. Initial and periodic tests -- 22.4. Emergency operating procedures -- 23. Incident analysis -- 23.1. Incident selection -- 23.2. Significant incident analysis methods -- 23.3. Case of a repetitive incident -- 24. Detailed analysis of incidents involving human factors -- 24.1. Pressurizer heater damage at Flamanville 2 -- 24.2. Isolation of pressurizer level sensors at Cruas 2 -- 24.3. Isolation of pressurizer level sensors at Gravelines 4 -- 24.4. Analysis and lessons -- 24.5. Check on sensor operability -- 24.6. General considerations on maintenance activity quality -- 24.7. Defense in depth applied to operation -- 25. Preventive maintenance and in-service surveillance -- 25.1. In-service surveillance for large components -- 25.2. Preventive maintenance of equipment -- 25.3. Steam generators -- 25.4. Steam line defects -- 25.5. Closure head adapter cracking -- 26. Some French precursors -- 26.1. Incidents -- 26.2. Latent nonconformances revealed by inspections -- 27. Periodic safety review -- 27.1. Safety review methodology -- 27.2. Fessenheim and Bugey plant safety reviews -- 27.3. Safety review of the CP1 and CP2 standardized 900 MWe plant series -- 28. The international dimension -- 28.1. The IAEA standards and guides program -- 28.2. The Incident Reporting System -- 28.3. French-German comparisons -- 28.4. Services proposed by the IAEA -- 28.5. Plants of soviet design -- 29. The next generation of reactors -- 29.1. Setting up of French-German safety options -- 29.2. Changes in safety objectives -- 29.3. Application of the defense in depth concept.
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29.4. Preliminary characteristics of the EPR project -- 29.5. Illustration of defense in depth provisions -- 30. Safety considerations on other nuclear installations -- 30.1. Safety organization changes at the CEA -- 30.2. General safety approach -- 30.3. Safety objectives, notion of acceptability -- 30.4. Risk potentials -- 30.5. Design bases -- 30.6. Safety analysis of an installation -- 30.7. Operating safety -- 30.8. Plant end of life -- 30.9. Conclusion of this chapter -- Conclusion -- Appendix A - Basic safety rules -- A.1 Rules concerning pressurized water reactors (June 1995) -- A.2 Rules concerning basic nuclear installations other than reactors (June 1995) -- Appendix B - Regulatory texts related to quality -- B.1. Order of August, 10, 1984 -- B.2. Circular of August, 10, 1984 -- Appendix C - French nuclear power plants -- C.1. Graphite-moderated, gas-cooled reactors (GCR) -- C.2. Heavy water reactor (HWR) -- C.3. fast breeder reactors (FBR) -- C.4. Pressurized water reactor (PWR) -- Appendix D - Basic Nuclear Installations -- D.1. Experimental reactors in service -- D.2. Fuel cycle basic nuclear installations -- D.3. Other CEA basic nuclear installations -- D.4. Other nuclear installations -- D.5. Particle accelerators considered as basic nuclear installations.
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