Skip to main content

Safety and Risk of Light Water Reactors and their Fuel Cycle Facilities

  • Chapter
  • First Online:
Sustainable and Safe Nuclear Fission Energy

Part of the book series: Power Systems ((POWSYS))

  • 3085 Accesses

Abstract

The safety of light water reactors is based on long term international research programs. The objective is to protect the operational personnel, the environment and the population against radioactivity releases during normal operation and in case of accidents. The safety concept is based on multiple containment structures (multi-barriers) as well as engineered safeguards components and other measures combined in a staggered-in-depth concept of four safety levels. The light water reactor plant and its protection system must be designed and built according to the design basis concept. Those design basis accidents which are part of the licensing process must be accommodated by the protection system, the inherent safety features and by the emergency cooling systems of the nuclear plant. Probabilistic safety analysis are supplements to this deterministic approach. They show that European light water reactors have a frequency of occurrence of about \(10^{-5}\) to \(10^{-6}\) per reactor year for core meltdown. Reactor risk studies which had been performed during the 1970s (USA) and 1980s (Europe) showed that the risk arising from light water reactors as a result of core melt down is well below the risk of other power generating or traffic systems. However, the Chernobyl accident in 1986 resulted—in addition to a not well known number of fatalities—in large scale land contamination by cesium-137 with a half-life of about 29 years. Similarly, the Fukushima accident (2011) resulted in land contamination by radioactive cesium isotopes. New research programs on severe accident consequences were initiated around 1990s. Their results lead to a revision of the results of the early risk studies of the 1980s and a new safety concept for modern light water reactors, e.g. the European Pressurized Water Reactor (EPR) and the European Boiling Water Reactor (SWR-1000). This new reactor safety concept allows to limit the severe accident consequences to the plant site itself. Also the introduction of additional severe accident management measures for existing light water reactors resulted in a considerable improvement of the prevention and mitigation of severe accident consequences. The safety concept of fuel cycle plants, e.g. spent fuel storage facilities, reprocessing facilities and waste treatment facilities is based on similar containment and engineered safeguards measures. However, the risk of these fuel cycle facilities is much smaller as the fuel is at much lower temperatures in reprocessing and refabricataion plants.

This is a preview of subscription content, log in via an institution to check access.

Access this chapter

Chapter
USD 29.95
Price excludes VAT (USA)
  • Available as PDF
  • Read on any device
  • Instant download
  • Own it forever
eBook
USD 129.00
Price excludes VAT (USA)
  • Available as EPUB and PDF
  • Read on any device
  • Instant download
  • Own it forever
Softcover Book
USD 169.99
Price excludes VAT (USA)
  • Compact, lightweight edition
  • Dispatched in 3 to 5 business days
  • Free shipping worldwide - see info
Hardcover Book
USD 169.99
Price excludes VAT (USA)
  • Durable hardcover edition
  • Dispatched in 3 to 5 business days
  • Free shipping worldwide - see info

Tax calculation will be finalised at checkout

Purchases are for personal use only

Institutional subscriptions

References

  1. Smidt D (1979) Reaktorsicherheitstechnik, Sicherheitssysteme und Störfallanalyse für Leichtwasserreaktoren und Schnelle Brüter. Springer, Berlin

    Google Scholar 

  2. Emendörfer D et al (1993) Theorie der Kernreaktoren, band 2: der instationäre reaktor. BI Wissenschaftsverlag, Mannheim

    Google Scholar 

  3. US Nuclear Regulatory Commission (1999) Regulatory guide 3.54, spent fuel heat generation in an independent spent fuel storage installation (revision 1, Jan 1999). http://www.nrc.gov/NRC/RG/03/03-054rl.html

  4. Xu Z et al (2005) Impact of high burnup on PWR spent fuel characteristics. Nucl Eng Des 151:261–273

    Google Scholar 

  5. Deutsche Risikostudie Kernkraftwerke Phase B (1990) Verlag TÜV Rheinland, Köln

    Google Scholar 

  6. Märkl M (1976) Core engineering and performance of pressurized water reactors. Kraftwerk Union AG, Erlangen

    Google Scholar 

  7. Tong LS et al (1979) Thermal analysis of pressurized water reactors. American Nuclear Society, LaGrange Park

    Google Scholar 

  8. Lahey RT et al (1977) The thermal hydraulics of a boiling water nuclear reactor. American Nuclear Society, LaGrange Park

    Google Scholar 

  9. RSK-Leitlinien für Druckwasserreaktoren (1996) Fassung 11.96, BAnz Nr. 214 vom 05.11.1996

    Google Scholar 

  10. Czech J et al (1999) European pressurized water reactor: safety objectives and principles. Nucl Eng Des 187:25–32

    Article  Google Scholar 

  11. Kersting E et al (1993) Safety analysis for boiling water reactors, a summary, GRS-98. Gesellschaft für Anlagen- und Reaktorsicherheit, Garching

    Google Scholar 

  12. Aleite W et al (1987) Leistungsregeleinrichtungen und Begrenzungen von Druck- und Siedewasserreaktoren. Atomwirtschaft 32:129–134

    Google Scholar 

  13. Aleite W et al (1987) Leittechnik in Kernkraftwerken. Atomwirtschaft 32:122–128

    Google Scholar 

  14. Boland JF (1970) Nuclear reactor instrumentation (in-core). Gordon and Breach Science, New York

    Google Scholar 

  15. Bachmann G et al (1971) Leittechnik des Kernkraftwerks Stade. Atomwirtschaft 16:600–602

    Google Scholar 

  16. Aleite W et al (1971) Regeleinrichtungen des Kernkraftwerks Stade. Atomwirtschaft 16:597–599

    Google Scholar 

  17. ASME Boiler and Pressure Vessel Code (2007) Rules for construction of nuclear power plant components, Sect III, Div 1. The American Society of Mechanical Engineers, New York

    Google Scholar 

  18. Laufs P (2006) Die Entwicklung der Sicherheitstechnik für Kernkraftwerke im politischen und technischen Umfeld der Bundesrepublik Deutschland seit 1955. Dissertation, Historisches Institut, Abteilung Geschichte der Naturwissenschaften, Universität Stuttgart, Germany

    Google Scholar 

  19. Merkle JG et al (1975) An evaluation of the HSST program intermediate pressure vessel tests in terms of light water reactor pressure vessel safety, ORNL-TM-5090

    Google Scholar 

  20. Derby RW et al (1974) Test of six-inch-thick pressure vessel, Series I: intermediate test vessels V-1 and V-2, ORNL-4895

    Google Scholar 

  21. Bryan RH et al (1975) Test of six-inch-thick pressure vessels, Series II: intermediate test vessels V-3, V-4, and V-6, ORNL-5059

    Google Scholar 

  22. Merkle JG et al (1976) Test of six-inch-thick-pressure vessels, Series III: intermediate test vessels V-7, ORNL-5059

    Google Scholar 

  23. Whitman GD (1976) Heavy section steel technology program quarterly progress report for January–March 1976, ORNL-TM-28

    Google Scholar 

  24. Griffith AA (1921) The phenomena of rupture and flow in solids. Philos Trans R Soc Lond A 221:163–198

    Article  Google Scholar 

  25. Irwin GR (1958) Fracture. In: Flügge S (ed) Handbuch der Physik, Bd. VI, Elastizität und Plastizität. Springer, Berlin

    Google Scholar 

  26. Hahn HG (1976) Bruchmechanik. Teubner, Stuttgart

    MATH  Google Scholar 

  27. Kußmaul K (1978) Maßnahmen und Prüfkonzept zur weiteren Verbesserung der Qualität von Reaktordruckbehältern für Leichtwasser-Kernkraft-werke, 2. VGB-Kraftwerkstechnik 58(6):439–448

    Google Scholar 

  28. Hoffmann H et al (2007) Das Integritätskonzept für Rohrleitungen sowie Leck- und Bruchpostulate in deutschen Kernkraftwerken. VGB Power Technol 7:78–91

    Google Scholar 

  29. Kußmaul K (1984) German basis safety concept rules out possibility of catastrophic failure. Nucl Eng Int 12:41–46

    Google Scholar 

  30. Bruch CG (1976) RELAP4/MOD5, a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems, user’s manual, vol 1: RELAP4/MOD5 description, SRD-113-76, Aerojet Nuclear Company, New York

    Google Scholar 

  31. Vigil JC et al (1981) Accident simulation with TRAC. Los Alamos Sci (Summer/Fall) 3:36–52

    Google Scholar 

  32. Liles DR et al (1978) TRAC-P1, an advanced best-estimate computer program for PWR LOCA analysis: methods, models, user information and programming details, LA-7279-MS, vol 1, NUREG/CR-0063

    Google Scholar 

  33. Liles DR et al (1979) TRAC-P1A, an advanced best-estimate computer program for PWR LOCA analysis. Report LA-7777-MS, NUREG/CR-0665, Los Alamos Scientific Laboratory

    Google Scholar 

  34. Liles DR et al (1981) TRAC-PD2, an advanced best-estimate computer program for pressurized water reactors loss-of-coolant accident analysis. Report LA-8709-MS, NUREG/CR-2054, Los Alamos Scientific Laboratory

    Google Scholar 

  35. Kuczera B (1990) Aktueller stand der Reaktorsicherheitsforschung dargestellt anhand von Ergebnissen aus der deutschen Risikostudie Kernkraftwerke—Phase B, Radioaktivität–Risiko–Sicherheit, Herausgeber Kernforschungszentrum Karlsruhe (2. veränderte und aktualisierte Auflage 1991). VS Verlag für Sozialwissenschaften, Wiesbaden

    Google Scholar 

  36. Rasmussen NC (ed) (1975) Reactor safety study: an assessment of accidents risks in US commercial nuclear power plants. US Nuclear Regulatory Commission, Washington, WASH-1400 (NUREG-75/014)

    Google Scholar 

  37. Deutsche Risikostudie Kernkraftwerke Phase A (1980) Gesellschaft für Reaktorsicherheit (GRS), TÜV Rheinland, Köln

    Google Scholar 

  38. Lewis EE (1977) Nuclear power reactor safety. Wiley, New York

    Google Scholar 

  39. Farmer FR (1967) Reactor safety and siting: a proposed risk criterion. Nucl Saf 8:539–548

    Google Scholar 

  40. Dunster HJ (1980) The approach of a regulatory authority to the concept of risk. IAEA Bull 22(5/6):123–128

    Google Scholar 

  41. Starr C (1969) Social benefits versus technological risks. Science 165(169):1232–1238

    Google Scholar 

  42. Bayer A, Heuser FW (1981) Basic aspects and results of the German risk study. Nucl Saf 22:695–709

    Google Scholar 

  43. Hassmann K et al (1987) Spaltproduktfreisetzung bei Kernschmelzen. TÜV Rheinland GmbH, Köln

    Google Scholar 

  44. Bunz H et al (1981) The role of aerosol behavior in light water reactor core melt accidents. Nucl Technol 53:141–146

    Google Scholar 

  45. COSYMA (1991) A new program package for accident consequence assessment, EUR-13028. European Commission, Brussels

    Google Scholar 

  46. Ehrhardt J et al (2000) RODOS: decision support system for off-site nuclear emergency management in Europe. Report EUR 19144, European Commission, Brussels

    Google Scholar 

  47. Ehrhardt J (1991) Probabilistische Unfallfolgenabschätzungen. In: Radioaktivität/Risiko—Sicherheit, pp 51–61. Kernforschungszentrum Karlsruhe

    Google Scholar 

  48. Hasemann I (2011) Personal communication. Karlsruhe Institute of Technology (KIT), Karlsruhe

    Google Scholar 

  49. Rahmenempfehlungen für den Katastrophenschutz in der Umgebung kerntechnischer Anlagen (2008) GMBI Nr. 62/63 vom 19, Dec 2008. http://www.bmu.de/files/pdfs/allgemein/application/pdf/rahmenempfehlung_katastrophenschutz.pdf

  50. Leitfaden für den Fachberater Strahlenschutz der Katastrophenschutzleitung bei kerntechnischen Notfällen (1989) Veröffentlichungen der Strahlenschutzkommission, Band 13. Gustav Fischer, Stuttgart; also: Radioaktivität und Strahlung Grenzwerte und Richtwerte. http://www.osiris22.pi-consult.de/userdata/I_20/P-105/Library/data/grenzwerte-und-richtwerte-04-03-internetversion.pdf

  51. Regulatory Guide 1.60 (1973) Design response spectra for seismic design of nuclear power plants (Revision 1). http://www.orau.org/ptp/PTP%20Library/library/NRC/Reguide/01-060.pdf

  52. Regulatory Guide 1.208 (2007) A performance-based approach to define the site-specific earthquake groundmotion, US Nuclear Regulatory Commission. http://www.nrc.gov/reading-rm/doc-collections/reg-guides/power-reactors/rg/01-208/01-208.pdf

  53. McGuire RK et al (2001) New seismic design spectra for nuclear power plants. Elsevier Science B.V., Amsterdam

    Google Scholar 

  54. KTA 2201.4 (1990) Auslegung von Kernkraftwerken gegen seismische Einwirkungen—Teil 4: Anforderungen an Verfahren zum Nachweis der Erdbebensicherheit für maschinen- und elektrotechnische Anlagenteile, Fassung 6/90, Kerntechnischer Ausschuss (KTA). http://www.kta-gs.de/d/regeln/2200/2201_4.pdf

  55. KTA 2201.1 (1990) Auslegung von Kernkraftwerken gegen seismische Einwirkungen, Teil 1. Grundsätze, Kerntechnischer Ausschuss (KTA). http://www.kta-gs.de/d/regeln/2200/2201_1.pdf

  56. Safety Guide IAEA NS-G-33 (2002) Evaluation of seismic hazards for nuclear power plants. International Energy Agency, Vienna

    Google Scholar 

  57. Safety Guide IAEA NS-G-1.6 (2003) Seismic design and qualification for nuclear power plants. International Energy Agency, Vienna

    Google Scholar 

  58. Forni M et al (2009) Seismic isolation of the IRIS nuclear plant. In: Proceedings of the 2009 ASME pressure vessel and piping conference, RVP 2009, 26–30 July 2009, Prague Czech Republic

    Google Scholar 

  59. Nakamura N et al (2008) Analyses of reactor building by 3D nonlinear FEM models considering basemat uplift for simultaneous horizontal and vertical ground motions. Nucl Eng Design 238:3551–3560

    Article  Google Scholar 

  60. KTA 2207 (2004) Schutz von Kernkraftwerken gegen Hochwasser, Sicherheitstechnische Regel des KTA (Kerntechnischer Ausschuss), Fassung 11/04. http://www.kta-gs.de/d/regeln/2200/2207n.pdf

  61. Birkhofer A (1989) Anlageninterner Notfallschutz. In: Achtes deutsches Atomrechts symposium, 1–3 März 1989, München. Carl Heymann Verlag KG, Köln

    Google Scholar 

  62. Schenk H (1990) Maßnahmen zum anlageninternen Notfallschutz. Atomwirtschaft 11:514–520

    Google Scholar 

  63. Fritsche AF (1988) Gesundheitsrisiken von energieversorgungssystemen. TÜV Rheinland, Köln

    Google Scholar 

  64. Hubert P et al (1981) Les impacts sanitaires et écologiques de la production de l’électricité: le cas français, CEPN-R-34.3 (2ème version), Fontenay-aux-Roses, Paris

    Google Scholar 

  65. Rogner HH (1998) Comparing energy options, IAEA bulletin 40/1/1998. Sustainable Development, Nuclear Power, IAEA, Vienna

    Google Scholar 

  66. Frischknecht S et al (1998) Project GaBE, comprehensive assessment of energy systems, severe accidents in the energy sector, PSI-Bericht Nr. 98-16, ISS 1019-0643

    Google Scholar 

  67. Kessler G (2002) Requirements for nuclear energy in the 21\(^{st}\) century, nuclear energy as a sustainable energy source. Prog Nucl Energy 60(3–4):309–325

    Google Scholar 

  68. Natural disasters. http://en.wikipedia.org/wiki/Natural_disaster

  69. Le programme international sur la sécurité chimique, Aide-Mémoire number 87, révisé, Mars 1998. http://apps.who.int/inf-fs/fr/am87.html

  70. Farmer R (1980) What is risk. Atom 282:108

    Google Scholar 

  71. Ehrhardt J et al (1982) Der Einsatz des unfallfolgenmodells der Deutschen Risikostudie Kernkraftwerke bei Risikoabschätzungen zu verschiedenen Reaktortypen, KfK-Nachrichten. Jahrg 14-4(82):269–277

    Google Scholar 

  72. Knief RA (1992) Nuclear engineering, theory and technology of commercial nuclear power. Hemisphere Publishing Corporation, Washington, DC

    Google Scholar 

  73. Ireland JR et al (1981) Three Mile Island and multiple-failure accidents. Los Alamos Sci 3(Summer/Fall):75–85

    Google Scholar 

  74. Kemeny JG (1979) Report of The President’s commission on the accident at Three Mile Island: the need for change: the legacy of TMI, ISBN 0935758003. The Commission, Washington, DC. http://www.threemileisland.org/downloads/188.pdf

  75. Nuclear Energy Agency (1995) Chernobyl ten years on, radiological and health impact. Nuclear Energy Agency (OECD), Paris

    Google Scholar 

  76. IAEA (1991) The international Chernobyl project. Report by an International Advisory Committee, Vienna

    Google Scholar 

  77. Kovan D (2011) Chernobyl 25 years on: time for a giant leap forward. Nucl News (Am Nucl Soc) 54(5):57

    Google Scholar 

  78. Nuclear Energy Agency (2002) Update of Chernobyl: ten years on. Nuclear Energy Agency (OECD), Paris

    Google Scholar 

  79. Deaths due to the Chernobyl disasters (2011) http://www.en.wikipedia.org/wiki/Deaths_due_to_the_Chernobyl_disaster

  80. International Atomic Energy Agency (2011) Chernobyl: answers to long standing questions. International Atomic Energy Agency (IAEA), Vienna. http://www.161.5.1.75/newscenter/focus/chernobyl/faqs.shtml

    Google Scholar 

  81. Jaworowski Z (1999) Radiation risk and ethics. Phys Today 52(9):24–29 (American Institute of Physics)

    Google Scholar 

  82. Nuclear News Special Report (2011) Fukushima Daiichi after the earthquake and tsunami. Nucl News 54(4):17

    Google Scholar 

  83. IAEA International Fact Finding Expert Mission of the Nuclear Accident Following the Great East Japan Earthquake and Tsunami (2011) Preliminary summary. IAEA, Vienna, 24 May–1 June 2011

    Google Scholar 

  84. Nuclear News (2011) TEPCO said that Fukushima Daiichi-I endured a melt down. American Nuclear Society, La Grange Park

    Google Scholar 

  85. Nuclear News (2011) Fukushima Daiichi accident offers lessons for all. American Nuclear Society, La Grange Park

    Google Scholar 

  86. IRSN (2011) Assessment on the 66\(^{\rm th}\) day of projected external doses for the populations living in the North-West fallout zone of the Fukushima nuclear accident. Report DRPH/2011-10, IRSN, Paris

    Google Scholar 

  87. DOE-NNSA (2011) DOE-NNSA Fukushima survey, PNG, 27–28 March 2011. http://www.en.wikipedia.org/wiki/File:DOE_NNSA_Fukushima_Survey_27-28.PNG

  88. Mohrbach L et al (2011) Earthquake and tsunami in Japan on March 11, 2011 and consequences for Fukushima and other nuclear power plants. http://www.vgb.org/vgbmultimedia/News/Fukushima_VGB_rev16.pdf

  89. Sekimura N (2011) Overview of the accident in Fukushima Daiichi nuclear power plants, IAEA, ICTP, Trieste, 8 Aug 2011. http://www.iaea.org/inisnkm/nkm/pages/2011/NEMschool2011/topics/topic0/fukushima%20Overview_Sekimura.pdf

  90. World Nuclear News (2011) Theory for Fukushima Daiichi 4 explosion: regulation and safety, 17 May 2011. http://www.world-nuclear-news.org/newsarticle.aspx?id=30068

  91. Iter Consult (2011) Independent technical evaluation and review, Fukushima Daiichi nuclear accident first considerations, preliminary report. http://www.iter-consult.it/ITER_Report_Fukushima_Accident.pdf

  92. Michel R (2011) Fukushima: eine vorläufige Bilanz im Juni 2011. http://www.zsr.uni-hannover.de/dokument/fubiju11.pdf

  93. Hennies HH, Kessler G, Eibl J (1989) Improved containment concept for future pressurized water reactors. In: 5\(^{\rm th}\) international conference on emerging nuclear energy systems (ICENES) 3–6 July 1989, Karlsruhe

    Google Scholar 

  94. Hennies HH, Kessler G, Eibl J (1992) Containments and core catchers in future reactors. Atomwirtschaft 37:238–247

    Google Scholar 

  95. Eibl J et al (1992) How to eliminate containment failure in tomorrows PWRs (pressurized water reactors). Nucl Eng Int 37(453):51–55

    Google Scholar 

  96. Berman M, Beck DF (1989), Steam explosion triggering and propagation: hypothesis and evidence. In: Proceedings of 3\(^{\rm rd}\) international seminar on containment of nuclear reactors, UCLA (SNL, report SAND89-1878C), Los Angeles, 10–11 Aug 1989

    Google Scholar 

  97. The SL-1 Reactor Accident. http://www.radiationworks.com/photos/sl1reactor1.htm; http://www.en.wikipedia.org/wiki/SL-1

  98. Corradini ML et al (1988) Vapor explosion in light water reactors: a review of theory and modelling. Prog Nucl Energy 22:1–117

    Article  Google Scholar 

  99. Magallon D (2005) FCI phenomena uncertainties impacting predictability of aquamic loading of reactor structures (from OECD SERENA programme). In: PSA-2 workshop, 7–9 Nov 2005, Aix-en-Provence

    Google Scholar 

  100. Jacobs H et al (1994) Untersuchungen zur Dampfexplosion, PSF Statusbericht, KfK 5326. Kernforschungszentrum Karlsruhe, pp 214–232, 23 März 1994

    Google Scholar 

  101. Cronenberg AW, Benz R (1980) Vapour explosion phenomena with respect to nuclear reactor safety assessment. Adv Nucl Sci Technol 12:247–335

    Google Scholar 

  102. Fletcher DF, Anderson RP (1990) A review of pressure-induced propagation models of the vapour explosion process. Prog Nucl Energy 23:137–179

    Article  Google Scholar 

  103. Board, S J et al. (1975), Detonation of coolant explosions, Nature 254(3):319–321.

    Article  Google Scholar 

  104. Diab A et al (2000) Long-term validation of the molten fuel-moderator interactions model. Nucl Technol 169:114–125

    Google Scholar 

  105. Berthoud G (2000) Vapor explosions. Annu Rev Fluid Mech 32:573–611

    Article  Google Scholar 

  106. Struwe D et al (1999) Consequence evaluation of in-vessel fuel coolant interaction in the European pressurized water reactor, FZKA 6316, Forschungszentrum Karlsruhe

    Google Scholar 

  107. The RELAP5 Development Team (1995) RELAP5/mod3 code manual, vol 1–7, NUREG/CR-5535, INEL-95/1074. Idaho National Engineering Laboratory, Idaho Falls

    Google Scholar 

  108. Allison M et al (1993) SCDAP/RELAP5 mod3.1 code manual, vol I–IV, NUREG/CR-6150, EGG-2720. Idaho National Engineering Laboratory, Idaho Falls

    Google Scholar 

  109. Coryell E et al (1997) SCDAP/RELAP5 mod3.2 code manual, vol I–V, NUREG/CR-6150, INEL-96/0422. Idaho National Engineering Laboratory, Idaho Falls

    Google Scholar 

  110. SCDAP/RELAP5 mod 3.2. http://www.relap5.inel.gov/scdap/home.html

  111. Power Research Institute (1979) InstituteStatus report on EPRI fuel cycle accident risk assessment, EPRI-NP-1128. Power Research Institute, Palo Alto

    Google Scholar 

  112. Valette M (1997) MC3D V3.0 directions for use. Commissariat a l’energie atomique Grenoble, STR/LTEM, STR-LTEM-96-52

    Google Scholar 

  113. Berthoud G, Valette M (1994) Development of a multidimensional model for the premixing phase of a fuel coolant interaction. Nucl Eng Des 149:409–418

    Article  Google Scholar 

  114. Jacobs H et al (1995) Multifield simulations of premixing experiments. In: Proceedings of a multidisciplinary international seminar on intense multiphase interactions, pp 56–69, Santa Barbara, 9–13 June 1995

    Google Scholar 

  115. Krieg R (1995) Missiles caused by severe pressurized-water reactor accidents. Nucl Saf 36:299–309

    Google Scholar 

  116. Krieg R et al (1995) Slug impact loading on the vessel head during a postulated in-vessel steam explosion in pressurized water reactors—assessments and discussion of the investigation strategy. Nucl Technol 111:369–385

    Google Scholar 

  117. Hirt A (1998) Rechenmodell zum Aufprall von Kernschmelze auf die oberen Einbauten und den Deckel eines Reaktordruckbehälters, FZKA 6054, Forschungszentrum Karlsruhe

    Google Scholar 

  118. Malmberg T (1995) Aspects of similitude theory in solid mechanics, Part I: Deformation behavior, FZKA 5657, Forschungszentrum Karlsruhe

    Google Scholar 

  119. Krieg R et al (2003) Load carrying capacity of a reactor vessel under molten core slug impact. Final report including recent experimental findings. Nucl Eng Des 293:237–253

    Article  Google Scholar 

  120. Stach T (1997) Zur Skalierung von Modellversuchen zum Aufprall flüssiger Massen auf deformierbare Strukturen, FZKA 5903, Forschungszentrum Karlsruhe

    Google Scholar 

  121. Krieg R et al (1980) Transient, three-dimensional potential flow problems and dynamic response of the surrounding structures, Part I. Comput Phys 8(2):139

    MathSciNet  Google Scholar 

  122. Krieg R et al (2000) Load carrying capacity of a reactor vessel head under a corium slug impact from a postulated in-vessel steam explosion. Nucl Eng Des 202:179–196

    Article  Google Scholar 

  123. Krieg R (1997) Mechanical efficiency of the energy release during a steam explosion. Nucl Technol 117:151–157

    Google Scholar 

  124. Theofanous B et al (1997) An assessment of steam explosion-induced containment failure, Part I: Probabilistic effects. Nucl Sci Eng 97:259–281

    Google Scholar 

  125. Amarasooriya WH et al (1987) An assessment of steam-explosion-induced containment failure. Part III: Expansion and energy partition. Nucl Sci Eng 97:296–315

    Google Scholar 

  126. Eibl J (1994) Zur bautechnischen Machbarkeit eines alternativen Containments für Druckwasserreaktoren—Stufe 3, KfK 5366, Kernforschungszentrum Karlsruhe

    Google Scholar 

  127. Travis JR et al (1998) GASFLOW-II: A three-dimensional-finite-volume fluid-dynamics code for calculating the transport, mixing, and combustion of flammable gases and aerosols in geometrically complex domains, theory and computational model, vol 1, FZKA-5994 and LA-13357-MS

    Google Scholar 

  128. Redlinger R (1999) DET3D: a code for calculating detonations in reactor containments. In: Proceedings of Jahrestagung Kerntechnik 99, Kerntechnische Gesellschaft e.V. Deutsches atomforum e.V. annual meeting on nuclear technology 99, Karlsruhe, 18–20 May 1999

    Google Scholar 

  129. Kotchourko AS et al (1999) Reactive flow simulations in complex 3D geometries using the COM3D code. In: Proceedings of Jahrestagung Kerntechnik 99, Kerntechnische Gesellschaft e.V. Deutsches atomforum e.V. annual meeting on nuclear technology 99, Karlsruhe, 18–20 May 1999

    Google Scholar 

  130. Veser A et al (1999) Experiments on turbulent combustion and COM3D verification. In: Proceedings of Jahrestagung Kerntechnik 99, Kerntechnische Gesellschaft e.V. Deutsches atomforum e.V. annual meeting on nuclear technology 99, Karlsruhe, 18–20 May 1999

    Google Scholar 

  131. Dorofeev SB et al (2001) Evaluation of limits for effective flame acceleration in hydrogen mixtures. J Loss Prev Process Ind 14:583–589

    Google Scholar 

  132. Dorofeev SB et al (1999) Effect of scale and mixture properties on behavior of turbulent flames in obstructed areas, FZKA 6268, Forschungszentrum Karlsruhe

    Google Scholar 

  133. Kuznetsov M et al (1999) Effect of obstacle geometry on behaviour of turbulent flames, FZKA 6328, Forschungszentrum Karlsruhe

    Google Scholar 

  134. Breitung W et al (2005) Innovative methoden zur analyse und Kontrolle des Wasserstoffverhaltens bei Kernschmelzunfällen, FZKA 7085, Forschungszentrum Karlsruhe

    Google Scholar 

  135. Krieg R et al (2003) Assessment of the load-carrying capacities of a spherical pressurized water reactor steel containment under a postulated hydrogen detonation. Nucl Technol 141:109–121

    Google Scholar 

  136. Rohde J et al (1997) Selection of representative accidents and evaluation of \({\rm H}_{2}\)-control measures in PWR containments. In: \(141^{\rm st}\) session of RSK light water reactor safety committee, Jan 1997

    Google Scholar 

  137. ABAQUS (1989) A general purpuse linear and nonlinear finite element code, user manual standard 5.8. Hibbit, Karlson and Sorenson, Providence

    Google Scholar 

  138. Bung H et al (1993) A new method for the treatment of impact and mechanics in reactor technology (SMIRT12), Stuttgart

    Google Scholar 

  139. Krieg R (2005) Failure strains and proposed limit strains for a reactor pressure vessel under severe accident conditions. Nucl Eng Des 235:199–212

    Article  Google Scholar 

  140. Jeschke J et al (2011) Critical strains and melting phenomena for different steel sheet specimens under uniaxial loading. Nucl Eng Des 241:2045–2052

    Article  Google Scholar 

  141. Jacobs G (1995) Dynamic loads from reactor pressure vessel core melt through under high primary pressure. Nucl Technol 111:351–356

    Google Scholar 

  142. Plank H et al (2009) http://www.sacre.web.psi.ch/ISAMM2009/oecd-sami2001/Papers/p20-Plank/SAM-Paper-b.pdf

  143. Stosic ZV et al (2008) Boiling water reactor with innovative safety concept: the generation III+ SWR-1000. Nucl Eng Des 238:1863–1901

    Article  Google Scholar 

  144. Thinnes GL et al (1989) Comparison of thermal and mechanical responses of the Three Mile Island unit 2 vessel. Nucl Technol 87:1036–1049

    Google Scholar 

  145. Tong LS (1968) Core cooling in a hypothetical loss of cooling accident. Estimate of heat transfer in core meltdown. Nucl Eng Des 8:309–312

    Article  Google Scholar 

  146. Henry RE et al (1993) External cooling of a reactor vessel under severe accident conditions. Nucl Eng Des 139:31–43

    Article  Google Scholar 

  147. Rempe JL et al (1993) Light water reactor lower head failure, NUREG/CR-5642, EGG-2618. Idaho National Engineering Laboratory, Idaho

    Google Scholar 

  148. Hagen S et al (1987) LWR fuel rod behaviour during severe accidents. Nucl Eng Des 103:85–106

    Article  Google Scholar 

  149. Kolev NI (2004) External cooling—the SWR-1000 severe accident management strategy. In: 12th international conference on nuclear engineering—ICONE-12, Arlington, 25–29 April 2004

    Google Scholar 

  150. Nazaré S et al (1975) Über theoretische und experimentelle Möglichkeiten zur Bestimmung der Stoffwerte von Corium, Abschlußbericht Teil II, KFK 2217

    Google Scholar 

  151. Schneider H et al (1975) Zur Bestimmung der Zusammensetzung verschiedener Corium-Schmelzen, KFK 2227

    Google Scholar 

  152. Reimann M et al (1981) The WECHSL-code: a computer program for the interaction of a core melt with concrete, KfK 2980, Kernforschungszentrum Karlsruhe

    Google Scholar 

  153. Reimann M (1987) Verification of the WECHSL-code on melt/concrete interaction and application to the core melt accident. Nucl Eng Des 103:127–137

    Article  Google Scholar 

  154. Alsmeyer H et al (1987) BETA-experiments in verification of the WECHSL-code: experimental results on the melt-concrete interaction. Nucl Eng Des 103:115–125

    Article  Google Scholar 

  155. Krieg R et al (1987) Failure pressure and failure mode of the latest type of German PWR containments. Nucl Eng Des 104:381–390

    Article  Google Scholar 

  156. Göller B et al (1988) Failure pressure and failure mode of the bolted connection for the large component port in German PWR containments. Nucl Eng Des 106:35–45

    Article  Google Scholar 

  157. Tromm W et al (1991) Radionuclide dispersion after core-concrete melt leaching by groundwater. Kerntechnik 56(6):7–12

    Google Scholar 

  158. Al-Omari I (1990) Abschätzung der Strahlenexposition infolge störfallbedingter radionuklideinleitungen von kerntechnischen Anlagen in Fließgewässer unter Berücksichtigung der Zeitabhängigkeit Relevanter Parameter, KfK 4793, Kernforschungszentrum Karlsruhe

    Google Scholar 

  159. Alsmeyer H (1989) Containment loadings from melt-concrete interaction. Nucl Eng Des 117:45–50

    Article  Google Scholar 

  160. Turricchia A (1992) How to avoid molten core/concrete interaction (and steam explosions). In: Alsmeyer H (ed) Proceedings of 2\(^{\rm nd}\) OECD(NEA) CSNI specialist meeting on molten core debris-concrete interaction, KfK 5108, NEA/CSNI/R(92)10. KfK, Karlsruhe, p 503

    Google Scholar 

  161. Seiler JM et al (1992) Conceptual studies of core catchers for advanced LWRs. In: Proceedings of international conference on design and safety of advanced nuclear power plants, Tokyo, Oct 1992, vol III, p 23.3-1

    Google Scholar 

  162. Alsmeyer H et al (1998) Beherrschung und Kühlung von Kernschmelzen außerhalb des Druckbehälters. Nachrichten Forschungszentrum Karlsruhe, Jahrg 29(4):327–335

    Google Scholar 

  163. Tromm W et al (1993) Fragmentation of melts by water inlet from below. In: \(6^{\rm th}\) international topical meeting on nuclear reactor thermal hydraulics (NURETH-6), Grenoble, 5–8 Oct 1993

    Google Scholar 

  164. Fieg G et al (1996) Simulation experiments on the spreading behavior of molten core melts. In: Proceedings of the 1996 national heat transfer conference, vol 9, Houston, 3–6 Aug 1996. American Nuclear Society, La Grange Park, pp 121–130

    Google Scholar 

  165. Lewis BJ (2008) Overview of experimental programs on core melt progression and fission product release behaviour. J Nucl Mater 380:126–143

    Article  Google Scholar 

  166. Meyer L et al (2009) Direct containment heating integral effects tests in geometrics of European nuclear power plants. Nucl Eng Des 239:2070–2084

    Article  Google Scholar 

  167. Meyer L et al (2003) Low-pressure corium dispersion experiments with simulant fluids in a scaled annular cavity. Nucl Technol 141:257–274

    Google Scholar 

  168. IPSN-GRS Proposals for the development of technical guidelines for future PWRs (1998) Structuring GPR-RSK recommendations as guidelines. Common report IPSN/GRS No. 42, vol 5, Institut de Protection et de Sûreté Nucléaire, Saclay, Gesellschaft für Reaktorsicherheit, Garching

    Google Scholar 

  169. Faude D et al (1989) Plutonium-Handhabung in Brennstoff-Kreislauf in Plutonium, KfK 4516, Kernforschungszentrum Karlsruhe

    Google Scholar 

  170. Schleisiek K (1980) Nukleare Sicherheit von Wiederaufarbeitungsanlagen. In: Grupe H (ed) Wie sicher ist die Entsorgung? Vortrag einer Informationsveranstaltung über Fragen der Kernenergie. Kernforschungszentrum Karlsruhe

    Google Scholar 

  171. Hennies HH et al (1976) Nukleare Sicherheit bei Wiederaufarbeitungsanlagen. Jahreskolloquium 1976 Projekt Nukleare Sicherheit, KfK 2399, Kernforschungszentrum Karlsruhe

    Google Scholar 

  172. Baumgaertel G et al (1988) Nukleare Sicherheit von Wiederaufarbeitungsanlagen. Grupe H (ed) Vortrag einer Informationsveranstaltung über Fragen der Kernenergie. Kernforschungszentrum Karlsruhe, letzte Aktualisierung

    Google Scholar 

  173. Summers RM et al (1995) MELCOR computer code manuals, vols 1–2 (Vers: 1.8.3), NUREG/CR-6119, SAND93-2185. Sandia National Laboratories, Albuquerque

    Google Scholar 

Download references

Author information

Authors and Affiliations

Authors

Rights and permissions

Reprints and permissions

Copyright information

© 2012 Springer-Verlag Berlin Heidelberg

About this chapter

Cite this chapter

Kessler, G. (2012). Safety and Risk of Light Water Reactors and their Fuel Cycle Facilities. In: Sustainable and Safe Nuclear Fission Energy. Power Systems. Springer, Berlin, Heidelberg. https://doi.org/10.1007/978-3-642-11990-3_11

Download citation

  • DOI: https://doi.org/10.1007/978-3-642-11990-3_11

  • Published:

  • Publisher Name: Springer, Berlin, Heidelberg

  • Print ISBN: 978-3-642-11989-7

  • Online ISBN: 978-3-642-11990-3

  • eBook Packages: EngineeringEngineering (R0)

Publish with us

Policies and ethics