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[해외논문] An event classification schema for evaluating site risk in a multi-unit nuclear power plant probabilistic risk assessment

Reliability engineering & system safety, v.117, 2013년, pp.40 - 51  

Schroer, S. ,  Modarres, M.

Abstract AI-Helper 아이콘AI-Helper

Today, Probabilistic Risk Assessments (PRAs) at multi-unit nuclear power plants consider risk from each unit separately and consider dependencies and interactions between the units informally and on an ad hoc basis. The accident at the Fukushima nuclear power station underlined the importance and po...

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참고문헌 (77)

  1. Reliability Engineering and Safety System Hakata 92 883 2007 10.1016/j.ress.2006.04.022 Seismic PSA method for multiple nuclear power plants in a site 

  2. U.S. Nuclear Regulatory Commission. Issues and recommendations for advancement of PRA technology in risk-informed decision making (NUREG/CR-6813). Washington, DC; 2003. 

  3. Fleming KN. On the issue of integrated risk-A PRA practitioner's perspective. In: Proceedings of the ANS international topical meeting on probabilistic safety analysis. San Francisco, CA; 2005. 

  4. Arndt S. Methods and strategies for future reactor safety goals. PhD thesis; 2010. 

  5. Omoto A. Design consideration on severe accident for future LWR (IAEA-TECDOC-1020). In: Proceedings of a technical committee meeting of the international atomic energy agency. Vienna; 1996. 

  6. Reliability Engineering and System Safety Jung 82 165 2003 10.1016/S0951-8320(03)00140-6 A new method to evaluate alternate AC power source effects in multi-unit nuclear power plants 

  7. Radiation and Nuclear Safety Authority (STUK) Sandberg 2009 Probabilistic safety analysis of non-seismic external hazards 

  8. U.S. Nuclear Regulatory Commission. Resolution of generic safety issues: task CH2: design (NUREG-0933). Washington, D.C.; 1981. 

  9. U.S. Nuclear Regulatory Commission. Use of probabilistic risk assessment methods in nuclear regulatory activities (60 FR 42622). Washington, D.C.; 1995. 

  10. U.S. Nuclear Regulatory Commission. Plan for resolving policy issues related to licensing non-light water reactor designs (SECY-02-0139). Washington, D.C.; 2002. 

  11. U.S. Nuclear Regulatory Commission. Policy issues related to licensing non-light-water reactor designs (SECY-03-0047). Washington, D.C.; 2003. 

  12. U.S. Nuclear Regulatory Commission. Policy issues related to new plant licensing and status of the technology-neutral framework for new plant licensing (SECY-05-0130); 2005. 

  13. U.S. Nuclear Regulatory Commission. Staff requirements memorandum-SECY-05-0130-policy issues related to new plant licensing and status of the technology-neutral framework for new plant licensing; 2005. 

  14. U.S. Nuclear Regulatory Commission, Staff plan to make a risk-informed and performance-based revision to 10 CFR Part 50 (SECY-06-0007). Washington, D.C.; 2006. 

  15. U.S. Nuclear Regulatory Commission. Feasibility study for a risk-informed and performance-based regulatory structure for future plant licensing (NUREG-1860). Washington, D.C.; 2007. 

  16. Idaho National Laboratory. The NRC's SPAR models: current status, future development, and modeling issues. In: Proceedings of the ANS international topical meeting on probabilistic safety analysis. Knoxville, TN; 2008. 

  17. U.S. Nuclear Regulatory Commission. Staff requirements memorandum-SECY-11-0089-options for proceeding with future Level 3 probabilistic risk assessment (PRA) activities. Washington, D.C.; 2011. 

  18. Journal of Power and Energy Systems Muramatsu 2 1 122 2008 10.1299/jpes.2.122 Effects of correlations of component failures and cross-connections of EDGs on seismically induced core damages of a multi-unit site 

  19. Muhlheim MD, Wood RT. Design strategies and evaluation for sharing systems at multi-unit plants phase I (ORNL/LTR/INERI-BRAZIL/06-01). Oak Ridge National Laboratory; 2007. 

  20. International Atomic Energy Agency. Development and application of Level 1 probabilistic safety assessment for nuclear power plants. Specific safety guide no. SSG-3. Vienna; 2010. 

  21. U.S. Nuclear Regulatory Commission. Risk assessment of operational events handbook: vol. 3-SPAR model reviews, Rev. 2. Washington, D.C.; 2010. 

  22. Bogazici University Nuclear Engineering Department. PSA Glossary, [Online]. Available from: 〈http://www.nuce.boun.edu.tr/psa/psaglossary.html〉; [accessed 15.05.12]. 

  23. Knochenhauer M, Holmberg J-E Guidance for the definition and application of probabilistic safety criteria. In: Proceedings of PSAM 10 international probabilistic safety assessment and management. Seattle, Washington; 2012. 

  24. U.S. Nuclear Regulatory Commission. Regulatory Guide, 1.174. An approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the licensing basis. Washington, D.C.; 2002. 

  25. U.S. Nuclear Regulatory Commission. Staff requirement memorandum-SECY-90-016-evolutionary light water reactor certification issues and their relationships to current regulatory requirements. Washington, D.C.; 1990. 

  26. Pickard Lowe and Garrick, Inc.. Seabrook station probabilistic safety assessment-section 13.3 risk of two unit station. Prepared for public service company of new hampshire, PLG-0300; 1983. 

  27. U.S. Nuclear Regulatory Commission. Industry-average performance for components and initiating events at U.S. Commercial Nuclear Power Plants (NUREG/CR-6928). Washington, D.C.; 2007. 

  28. International Atomic Energy Agency. Extreme external events in the design and assessment of nuclear power plants (IAEA-TECDOC-1341). Vienna; 2003. 

  29. Institute of Electrical and Electronics Engineers. IEEE standard application of the single-failure criterion to nuclear power generating station safety system (ANSI/IEEE Std. 379-1988). New York; 1988. 

  30. Tennessee Valley Authority 2011 Browns Ferry Nuclear Plant, Units 1, 2, and 3: licensee event report 50-259/2011-001-00 Letter to U.S. Nuclear Regulatory Commission 

  31. Point Beach Nuclear Plant 2000 Licensee event report 2000-010-00 Letter to U.S. Nuclear Regulatory Commission 

  32. Duke energy 2011 Duke energy carolinas, LLC: licensee event report 369/2011-01, Revision 1 Letter to U.S. Nuclear Regulatory Commission 

  33. Florida Power and Light 2005 Turkey point Units 3 and 4: reportable event 2005-006-01 Letter to U.S. Nuclear Regulatory Commission 

  34. PPL Susquehanna 2007 Licensee event report 50-387/2007-001-00 LCC, letter to U.S. Nuclear Regulatory Commission 

  35. Prairie Island Nuclear Generating Plant 2006 Licensee event report 2-06-02 Letter to U.S. Nuclear Regulatory Commission 

  36. Point Beach Nuclear Plant 2007 Licensee event report 266/301-2007-006-00 Letter to U.S. Nuclear Regulatory Commission 

  37. Duke Energy 2006 Licensee event report 413/06-003 Letter to U.S. Nuclear Regulatory Commission 

  38. Exelon Nuclear 2001 Licensee event report 2001-001-00, Peach Bottom Atomic Power Station Unit 2 and 3 Letter to U.S. Nuclear Regulatory Commission 

  39. Exelon Generation Company 2011 Licensee event report 2011-004-00 LLC. Letter to U.S. Nuclear Regulatory Commission 

  40. U.S. Nuclear Regulatory Commission. NRC inspection manual: manual chapter 0609-significance determination process. Washington, D.C.; 2011. 

  41. U.S. Nuclear Regulatory Commission. NRC enforcement policy. Washington, D.C.; 2011. 

  42. Tennessee Valley Authority 2009 Licensee event report (LER) 50-259/2009-003 Letter to U.S. Nuclear Regulatory Commission 

  43. Point Beach Nuclear Plant 2002 Licensee event report 266/2002-003-01 Letter to U.S. Nuclear Regulatory Commission 

  44. PPL Susquehanna 2002 Licensee event report 50-387/2002-005-00 LCC, Letter to U.S. Nuclear Regulatory Commission 

  45. Tennessee Valley Authority 2000 Licensee event report (LER) 50-260/2000-002-00 Letter to U.S. Nuclear Regulatory Commission 

  46. Point Beach Nuclear Plant 2001 Licensee event report (LER) 266/2001-005-00 Letter to U.S. Nuclear Regulatory Commission 

  47. Duke Energy 2002 Licensee event report 269/2002-01 Letter to U.S. Nuclear Regulatory Commission 

  48. Palo Verde Nuclear Generating Station 2004 Licensee event report 2004-009-01 Letter to U.S. Nuclear Regulatory Commission 

  49. Florida Power and Light. Letter to U.S. Nuclear Regulatory Commission. Turkey point Units 3 and 4: reportable event 2005-004-00, 2005 

  50. PSEG Nuclear LLC. Letter to U.S. Nuclear Regulatory Commission. LER 272/02-001-00 Salem Generating Station-Unit 1; 2002. 

  51. Southern Company 2002 Vogtle Electric Generating Plant-Unit 1 and 2 licensee event report 1-2005-002 Letter to U.S. Nuclear Regulatory Commission 

  52. Tennessee Valley Authority 2002 Sequoyah Nuclear Power Plant Units 1 and 2-licensee event report 50-327/2002001 Letter to U.S. Nuclear Regulatory Commission 

  53. Westinghouse Nuclear Safety Advisory Letter 02-3. Steam generator mid-deck plate pressure loss issue; 2002. 

  54. Exelon Generation Company 2010 Braidwood Station, Units 1 and 2, licensee event report 2010-007-00 Letter to U.S. Nuclear Regulatory Commission 

  55. Exelon Generation Company 2011 Braidwood Station, Units 1 and 2, licensee event report 2010-007-00 Letter to U.S. Nuclear Regulatory Commission 

  56. Luminant Power 2011 Comanche Peak Nuclear Power Plant, licensee event report 445/11-001-00 Letter to U.S. Nuclear Regulatory Commission 

  57. Westinghouse Nuclear Safety Advisory Letter 09-8. Presence of vapor in emergency core cooling system/residual heat removal system in modes 3/4 loss-of-coolant accident conditions; 2009. 

  58. Exelon Generation Company. Letter to U.S. Nuclear Regulatory Commission. Braidwood Station, Unit 1, licensee event report 2005-001-00, 2005 

  59. Exelon Generation Company 2010 Braidwood Station, Units 1 and 2, licensee event report 2010-006-00 Letter to U.S. Nuclear Regulatory Commission 

  60. U.S. Nuclear Regulatory Commission. Guidelines on modeling common-cause failures in probabilistic risk assessment (NUREG/CR-5485). Washington, D.C.; 1998. 

  61. Reliability Engineering and System Safety Mohaghegh 94 1000 2009 10.1016/j.ress.2008.11.006 Incorporating organizational factors into probabilistic risk assessment of complex socio-technical systems: a hybrid technique formalization 

  62. Reliability Engineering and System Safety Hu 64 241 1999 10.1016/S0951-8320(98)00066-0 Evaluating system behavior through dynamic master logic diagram modeling 

  63. IEEE Transactions on Computers Akers C-27 509 1978 10.1109/TC.1978.1675141 Binary decision diagrams 

  64. Safety Science Mohaghegh 47 1139 2009 10.1016/j.ssci.2008.12.008 Incorporating organizational factors into probabilistic risk assessment of complex socio-technical systems: principles and theoretical foundations 

  65. 10.1115/POWER2011-55324 Mohaghegh Z, Modarres M, Christou A. Physics-based common cause failure modeling in probabilistic risk analysis: a mechanistic approach. In: Proceedings of the ASME 2011 power conference. Denver, Colorado; 2011. 

  66. Mohaghegh Z, Modarres M. A probabilistic physics-of-failure approach to common cause failures in reliability assessment of structures and components. In: American nuclear society, 2011 winter meeting. Washington, D.C.; October 30-November 3, 2001. 

  67. Reliability Engineering and System Safety Trucco 93 823 2008 10.1016/j.ress.2007.03.035 A bayesian belief network modelling of organisational factors in risk analysis: a case study in maritime transportation 

  68. 10.1007/3-540-45416-0_4 Gran BA, Helminen A. A Bayesian belief network for reliability assessment. In: SAFECOMP'01 Proceedings of the 20th international conference on computer safety, reliability, and security. London; 2001. 

  69. U.S. Nuclear Regulatory Commission. PRA procedures guide (NUREG/CR-2300). Washington, D.C.; 1983. 

  70. U.S. Nuclear Regulatory Commission. Procedures for the external event core damage frequency analyses for NUREG-1150 (NUREG/CR-4840). Washington, D.C.; 1990. 

  71. U.S. Nuclear Regulatory Commission. Technical review of risk-informed, performance-based methods for nuclear power plant fire protection analyses-draft report for comment (NUREG-1521). Washington, D.C.; 1998. 

  72. U.S. Nuclear Regulatory Commission. EPRI/NRC-RES fire PRA methodology for nuclear power facilities (NUREG/CR-6850). Washington, D.C.; 2005. 

  73. International Atomic Energy Agency. Probabilistic safety assessment for seismic events. IAEA-TECDOC-724. Austria; 1993. 

  74. Chokshi N, Johnson JJ. Methodology of seismic margin assessment and evaluation examples in the US. In: Presentations from the first Kashiwazaki international symposium. Kashiwazaki, Japan; November 24-26; 2010. 

  75. International Atomic Energy Agency. IAEA safety standards series: external events excluding earthquakes in the design of nuclear power plants (no. NS-G-1.5). Vienna; 2003. 

  76. Reliability Engineering and System Safety Chang 92 997 2007 10.1016/j.ress.2006.05.014 Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents, part 1: overview of the IDAC model 

  77. Princeton University. WordNet Search-3.1. [Online]. Available from: 〈http://wordnetweb.princeton.edu/perl/webwn?s=fuzzy%20logic〉; 2005 [accessed 13.04.12]. 

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