Nuclear Engineering and Technology

  • 제40권5호
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  • Pages.349-364
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  • 2008
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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PRA RESEARCH AND THE DEVELOPMENT OF RISK-INFORMED REGULATION AT THE U.S. NUCLEAR REGULATORY COMMISSION

  • Siu, Nathan (Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission) ;
  • Collins, Dorothy (Department of Communication Studies, Texas A&M University, College Station)
  • 발행 : 2008.08.31
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초록

Over the years, probabilistic risk assessment (PRA) research activities conducted at the U.S. Nuclear Regulatory Commission (NRC) have played an essential role in support of the agency's move towards risk-informed regulation. These research activities have provided the technical basis for NRC's regulatory activities in key areas; provided PRA methods, tools, and data enabling the agency to meet future challenges; supported the implementation of NRC's 1995 PRA Policy Statement by assessing key sources of risk; and supported the development of necessary technical and human resources supporting NRC's risk-informed activities. PRA research aimed at improving the NRC's understanding of risk can positively affect the agency's regulatory activities, as evidenced by three case studies involving research on fire PRA, human reliability analysis (HRA), and pressurized thermal shock (PTS) PRA. These case studies also show that such research can take a considerable amount of time, and that the incorporation of research results into regulatory practice can take even longer. The need for sustained effort and appropriate lead time is an important consideration in the development of a PRA research program aimed at helping the agency address key sources of risk for current and potential future facilities.

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참고문헌

  1. U.S. Nuclear Regulatory Commission, 'Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants,' WASH-1400 (NUREG-75/014), 1975
  2. Murphy, J.A., and M.A. Cunningham, 'Probabilistic risk assessment development in the United States: 1972-1995,' Proceedings of Twenty-Eighth Water Reactor Safety Information Meeting, NUREG/CP-0172, 2001, pp. 27-34
  3. Modarres, M., 'Technology-neutral nuclear power plant regulation: implications of a safety goals-driven performance-based regulation,' Nuclear Engineering and Technology, 37, 221-230(2005)
  4. Kadak, A., and T. Matsuo, 'The nuclear industry's transition to risk-informed regulation and operation in the United States,' Reliability Engineering and System Safety, 92, 609-618(2007) https://doi.org/10.1016/j.ress.2006.02.004
  5. Gaertner, J., D. True, and I. Wall, 'Safety benefits of risk assessment at U.S. nuclear power plants,' Nuclear News, 96, 28-36(Jan. 2003)
  6. U.S. Nuclear Regulatory Commission, 'Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,' NUREG-1150, 1990
  7. U.S. Nuclear Regulatory Commission, 'Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance,' NUREG-1560, 1997
  8. U.S. Nuclear Regulatory Commission, 'Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program,' NUREG-1742, 2002
  9. U.S. Nuclear Regulatory Commission, 'Use of Probabilistic Risk Assessment Methods in Nuclear Activities: Final Policy Statement,' Federal Register, Vol. 60, p. 42622 (60 FR 42622), August 16, 1995
  10. U.S. Nuclear Regulatory Commission, 'An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,' Regulatory Guide 1.174, Rev. 1, 2001
  11. U.S. Nuclear Regulatory Commission, 'Reactor Oversight Program,' NUREG-1649, Rev. 3, 2000
  12. U.S. Code of Federal Regulations, 'Fire Protection,' 10 CFR 50.48, November 10, 1980, last amended June 16, 2004
  13. U. S. Nuclear Regulatory Commission, 'Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10CFR50.61): Summary Report,' NUREG-1806, 2005
  14. Drouin, M., and J. Monninger, 'The development of PRA quality standards and use in risk-informed decision making,' Proceedings of PSAM 9, Ninth International Conference of Probabilistic Safety Assessment and Management, Hong Kong, China, May 18-23, 2008
  15. Drouin, M., 'Feasibility Study for a Risk-Informed and Performance-Based Regulatory Structure for Future Plant Licensing,' NUREG-1860, 2007
  16. Farmer, F.R., 'Reactor safety and siting: a proposed risk criterion,' Nuclear Safety, 8, 539-548(1967)
  17. Organization for Economic Cooperation and Development, 'Use and Development of Probabilistic Safety Assessment,' NEA/CSNI/R(2007)12, 2007
  18. Lewis, H., et al., 'Risk Assessment Review Group Report to the U.S. Nuclear Regulatory Commission,' NUREG/CR-0400, 1978
  19. U.S. Nuclear Regulatory Commission, 'Status of the accident sequence precursor program and the development of standardized plant analysis risk models,' SECY-07-0176, October 3, 2007
  20. Siu, N., 'Risk assessment for dynamic systems: an overview,' Reliability Engineering and System Safety, 43, No. 1, 43-73(1994) https://doi.org/10.1016/0951-8320(94)90095-7
  21. U.S. Nuclear Regulatory Commission, 'Strategic Plan, Fiscal Years 2008-2013,' NUREG-1614 Vol. 4, 2008
  22. Advisory Committee on Reactor Safeguards, 'Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program,' NUREG-1635, Vol. 6, 2004
  23. U.S. Nuclear Regulatory Commission , 'U.S. Nuclear Regulatory Commission Long-Term Research: Fiscal Year 2009 Activities,' Agencywide Documents Access and Management System (ADAMS) Accession No. ML0801501211, 2007
  24. Kuritzky, A., et al., 'Use of traditional PRA methods for digital system reliability modeling to support regulatory decision-making,' Proceedings of PSAM 9, Ninth International Conference of Probabilistic Safety Assessment and Management, Hong Kong, China, May 18-23, 2008
  25. U.S. Nuclear Regulatory Commission, 'Draft Advanced Reactor Research Plan,' ADAMS ML070600065, 2007
  26. Siu, N., 'Current applications of PRA in emergency management: a literature review,' Proceedings of PSAM 8, International Conference on Probabilistic Safety Assessment and Management, New Orleans, LA, May 14-19, 2006
  27. Smith, C.L., J.K. Knudsen, K. Kvarfordt, and S.T. Wood, 'Key Attributes of the SAPHIRE Risk and Reliability Analysis Software for Risk-Informed Probabilistic Applications,' Reliability Engineering and System Safety, 93, 1151-64(2008) https://doi.org/10.1016/j.ress.2007.08.005
  28. Appignani, P., R. Sherry, and R. Buell, 'The NRC's SPAR models: current status, future development, and modeling issues,' Proceedings of International Topical Meeting on Probabilistic Safety Analysis (PSA08), Knoxville, TN, Sep. 7-11, 2008
  29. U.S. Nuclear Regulatory Commission, 'Fire Protection for Operating Nuclear Power Plants,' Regulatory Guide 1.189, 2001
  30. Scott, R.L., 'Browns Ferry Nuclear Power Plant Fire on Mar. 22, 1975,' Nuclear Safety, 17, 592-611(Sep.-Oct. 1976)
  31. U.S. Code of Federal Regulations, 'Appendix A to Part 50- General Design Criteria for Nuclear Power Plants,' February 20, 1971, last amended December 23, 1999
  32. U.S. Code of Federal Regulations, 'Appendix R to Part 50- Fire Protection Program for Nuclear Power Plants Operating Prior to January 1, 1979,' November 19, 1980, last amended June 20, 2000
  33. Apostolakis, G., M. Kazarians, and D.C. Bley, 'Methodology for assessing the risk from cable fires,' Nuclear Safety, 23, 391-407(1982)
  34. 'Zion Probabilistic Safety Study,' Commonwealth Edison Co., Chicago (1981)
  35. 'Indian Point Probabilistic Safety Study,' Consolidated Edison Company of New York, Inc., and Power Authority of the State of New York, New York (1982).
  36. American Nuclear Society and Institute of Electrical and Electronics Engineers, 'PRA Procedures Guide,' NUREG/CR-2300, 1983
  37. Siu, N., J.S. Hyslop, and S.P. Nowlen, 'Fire risk analysis for nuclear power plants,' The Society for Fire Protection Engineers Handbook of Fire Protection Engineering, 4th Edition, National Fire Protection Association, in publication
  38. Advisory Committee on Reactor Safeguards, 'Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program,' NUREG-1635, Vol. 1, 1998
  39. Siu, N., and H. Woods, 'The U.S. Nuclear Regulatory Commission's fire risk research program - an overview,' Proceedings from International Workshop on Fire Risk Assessment, NEA/CSNI/R(99)26, June 2000, pp. 32-44
  40. Fleming, K.N., W.T. Houghton, and F.P. Scaletta, 'A Methodology for Risk Assessment of Major Fires and Its Application to an HTGR Plant,' GA-A15402, General Atomic Company, San Diego, CA, 1979
  41. Organization for Economic Cooperation and Development, Proceedings from International Workshop on Fire Risk Assessment, NEA/CSNI/R(99)26, June 2000
  42. Berry, D.L., and E.E. Minor, 'Nuclear Power Plant Fire Protection - Fire-Hazard Analysis (Subsystems Study Task 4),' NUREG/CR-0654, Sandia National Laboratories, 1979
  43. Apostolakis, G., 'The concept of probability in safety assessments of technological systems,' Science, 250, 1359-1364(1990) https://doi.org/10.1126/science.2255906
  44. Siu, N. and G. Apostolakis, 'Probabilistic models for cable tray fires,' Reliability Engineering, 3, 213-227(1982) https://doi.org/10.1016/0143-8174(82)90031-2
  45. Ho, V., N. Siu, and G. Apostolakis, 'COMPBRN III - a fire hazard model for risk analysis,' Fire Safety Journal, 13, 137-154(1988) https://doi.org/10.1016/0379-7112(88)90009-4
  46. Kazarians, M. and G. Apostolakis, 'Modeling rare events: the frequencies of fires in nuclear power plants,' Proceedings of Workshop on Low Probability/High Consequence Risk Analysis, Society for Risk Analysis, Arlington, VA, 1982
  47. Siu, N., 'Probabilistic Models for the Behavior of Compartment Fires,' NUREG/CR-2269, University of California at Los Angeles, 1981
  48. Siu, N. and G. Apostolakis, 'A methodology for analyzing the detection and suppression of fires in nuclear power plants,' Nuclear Science and Engineering, 94, 213-226(1986) https://doi.org/10.13182/NSE86-A17264
  49. Sideris, A.G., R.W. Hockenbury, M.L. Yeater and W.E. Vesely, 'Nuc!ear plant fire incident data file,' Nuclear Safety, 20 308(May-June 1979)
  50. Lambright, J., et al., 'Evaluation of Generic Issue 57: Effects of Fire Protection System Actuation on Safety- Related Equipment,' NUREG/CR-5580, Sandia National Laboratories, 1992
  51. U.S. Nuclear Regulatory Commission, 'Regulatory Issue Summary 2007-19: Process for Communicating Clarifications of Staff Positions Provided in Regulatory Guide 1.205 Concerning Issues Identified During the Pilot Application of National Fire Protection Association Standard 805,' ADAMS ML071590227, 2007
  52. Nowlen, S.P. and F.J. Wyant, 'Cable Response to Live Fire (CAROLFIRE),' NUREG/CR-6931, Sandia National Laboratories, 2008
  53. Salley, M. and R.P. Kassawara, 'Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications,' NUREG-1824/EPRI 1011999, U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research and Electric Power Research Institute, 2007
  54. National Fire Protection Association, 'Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants,' NFPA 805, 2001
  55. Henneke, D., E. Kleinsorg, and K. Zee, 'Risk-informed fire protection and fire PRA for Duke Power's Oconee, Catawba and McGuire nuclear plants,' Proceedings of International Topical Meeting on Probabilistic Safety Assessment (PSA 05), San Francisco, 2005
  56. American Nuclear Society, 'Fire PRA Methodology,' ANSI/ANS-58.23-2007, 2007
  57. Najafi, B. and S.P. Nowlen, 'EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,' EPRI 1011989/NUREG/CR-6850, Electric Power Research Institute and U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research, 2005
  58. U.S. Nuclear Regulatory Commission 'Risk-Informed Approach for Post-Fire Safe-Shutdown Circuit Inspections,' Regulatory Issue Summary 2004-03, Rev. 1, 2005
  59. U.S. Nuclear Regulatory Commission, 'Fire Protection Significance Determination Process,' Inspection Manual, Chapter 0609, Appendix F, 2005
  60. U. S. Nuclear Regulatory Commission, 'Clarification of TMI Action Plan Requirements,' NUREG-0737, 1980
  61. U. S. Code of Federal Regulations, 'Contents of construction permit and operating license applications; technical information, Additional TMI requirements,' 10 CFR 50.34(f), December 17, 1968, amended February 16, 1982, last amended August 28, 2007
  62. U. S. Nuclear Regulatory Commission, 'Policy on factors causing fatigue of operating personnel at nuclear reactors,' Federal Register, 47 FR 23836, June 1, 1982
  63. U.S Code of Federal Regulations, 'Fitness for Duty Programs,' 10 CFR 26, June 7, 1989
  64. U.S. Code of Federal Regulations, 'Conditions of Licenses,' 10 CFR 50.54(i,j,k,l,m), January 19, 1956
  65. U.S. Code of Federal Regulations, 'Training and qualification of nuclear power plant personnel,' 10 CFR 50.120, last amended July 21, 1993
  66. U.S. Code of Federal Regulations, 'Operators' Licenses,' 10 CFR 55, March 25, 1987
  67. U. S. Nuclear Regulatory Commission, 'Policy statement about engineering expertise on shift,' Federal Register, 50 FR 43621, October 28, 1985
  68. U. S. Nuclear Regulatory Commission, 'Education for senior reactor operators and shift supervisors at nuclear power plants,' Federal Register, 54 FR 33639, August 15, 1989
  69. Gertman, D.I. and H.S. Blackman, 'Human Reliability and Safety Analysis Data Handbook,' John Wiley and Sons, 1994
  70. Bley, D., et al., 'Untangling the Causes of Human Error: Predicting the Likelihood of Human Error in High-Risk Industries,' letter report to the U.S. Nuclear Regulatory Commission, 2005
  71. Forester, J. et al., 'ATHEANA User's Guide,' NUREG-1880, 2007
  72. Swain, A.D., 'Human reliability analysis: need, status, trends and limitations,' Reliability Engineering and System Safety, 29, 301-313(1990) https://doi.org/10.1016/0951-8320(90)90013-D
  73. Swain, A.D., and H.E. Guttmann, 'Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications,' NUREG/CR-1278, Sandia National Laboratories, 1983
  74. Hall, R.E., J. Fragola, and J. Wreathall, 'Post Event Human Decision Errors: Operator Action Tree/Time Reliability Correlation,' NUREG/CR-3010, Brookhaven National Laboratory, 1982
  75. Embrey, D.E., 'The Use of Performance Shaping Factors and Quantified Expert Judgment in the Evaluation of Human Reliability: An Initial Appraisal,' NUREG/CR-2986, Brookhaven National Laboratory, 1983
  76. Siegel, A.I., et al., 'Maintenance Personnel Performance Simulation (MAPPS) Model,' NUREG/CR-3626, 1984
  77. U. S. Nuclear Regulatory Commission, 'Technical Basis and Implementation Guidelines for A Technique for Human Event Analysis (ATHEANA),' NUREG-1624, Rev. 1, 2000
  78. Barriere, M.T., et al., 'Multidisciplinary Framework for Human Reliability Analysis with an Application to Errors of Commission and Dependencies,' NUREG/CR-6265, Brookhaven National Laboratory, 1995
  79. Acosta, C., and N. Siu, 'Dynamic event trees In accident sequence analysis: application to steam generator tube rupture,' Reliability Engineering and System Safety, 41, 135-154(1993) https://doi.org/10.1016/0951-8320(93)90027-V
  80. Woods, D.D., E.M. Roth, and H. Pople, Jr., 'Cognitive Environment Simulation: An Artificial Intelligence System for Human Performance Assessment,' NUREG/CR-4862, 1987
  81. Huang, Y., N. Siu, D. Lanning, J. Carroll, and V. Dang, 'Modeling Control Room Crews for Accident Sequence Analysis,' MITNE-296, Massachusetts Institute of Technology, 1991
  82. Coyne, K., and A. Mosleh, 'Implementation of a dynamic PRA approach for the prediction of operator errors during abnormal nuclear power plant events,' Proceedings of PSAM 9, Ninth International Conference of Probabilistic Safety Assessment and Management, Hong Kong, China, May 18-23, 2008
  83. Organization for Economic Cooperation and Development, 'Errors Of Commission In Probabilistic Safety Assessment,' NEA/CSNI/R(2000)17, 2000
  84. Gertman, D.I., et al., 'The SPAR-H Human Reliability Analysis Method,' NUREG/CR-6883, Idaho National Laboratory, 2005
  85. Hallbert, B., et al., 'Human Event Repository and Analysis (HERA) System, Overview,' NUREG/CR-6903, Idaho National Laboratory, 2006
  86. U. S. Nuclear Regulatory Commission, 'Good Practices for Implementing Human Reliability Analysis,' NUREG-1792, 2005
  87. Dang, V.N., et al., 'Benchmarking HRA methods against simulator data - design and organization of the International HRA Empirical Study,' Proceedings of PSAM 9, Ninth International Conference of Probabilistic Safety Assessment and Management, Hong Kong, China, May 18-23, 2008
  88. Reason, J., 'Human Error,' Cambridge University Press, 1990
  89. Senders, J.W. and N.P. Moray, 'Human Error: Cause, Prediction, and Reduction,' Lawrence Erlbaum Associates, Hillsdale, NJ, 1991
  90. Blackman, H., N. Siu, and A. Mosleh, 'Human Reliability Models: Theoretical and Practical Challenges,' Center for Reliability Engineering, University of Maryland, College Park, MD, 1998
  91. Ryan, T.G., 'Task analysis linked approach for integrating the human factor in reliability assessments of nuclear power plants,' Reliability Engineering and System Safety, 22, 219-234(1988) https://doi.org/10.1016/0951-8320(88)90075-0
  92. Swain, A.D., 'Accident Sequence Evaluation Program Human Reliability Analysis Procedure,' NUREG/CR-4772, Sandia National Laboratories, 1987
  93. Embrey, D.E., et al., 'SLIM-MAUD: An Approach to Aassessing Human Error Probabilities Using Structured Expert Judgment,' NUREG/CR-3518, Brookhaven National Laboratory, 1984
  94. Gertman, D.I., et al., 'Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR),' NUREG/CR-4639, Idaho National Laboratory, 1990
  95. U.S. Code of Federal Regulations, 'Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,' 10 CFR 50.61, May 15, 1991; last amended July 29, 1996
  96. U. S. Nuclear Regulatory Commission, 'Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors,' Regulatory Guide 1.154, 1987
  97. U. S. Nuclear Regulatory Commission, 'Pressurized Thermal Shock,' SECY-82-465, November 23, 1982
  98. Burns, T.J., et al., 'Preliminary Development of an Integrated Approach to the Evaluation of Pressurized Thermal Shock as Applied to the Oconee Unit 1 Nuclear Power Plant,' NUREG/CR-3770, Oak Ridge National Laboratory, 1986
  99. Selby, D.L, and G.F. Flanagan, 'Pressurized Thermal Shock Evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant,' NUREG/CR-4022, Oak Ridge National Laboratory, 1985
  100. Selby, D.L, and G.F. Flanagan, 'Pressurized Thermal Shock Evaluation of the H.B. Robinson Unit 2 Nuclear Power Plant,' NUREG/CR-4183, Oak Ridge National Laboratory, 1985
  101. Westinghouse Electric Corporation, 'Palisades Reactor Vessel Integrity Study Final Report,' WP0677-1, prepared for Palisades Generating Company, 1991
  102. Kolaczkowski, A., et al., 'Field test of ATHEANA (A Technique for Human Event Analysis) in pressurized thermal shock probabilistic risk assessments,' Proceedings from International Workshop on Building the New HRA, NEA/CSNI/R(2002)3, 2002
  103. Apostolakis, G., 'The concept of probability in safety assessments of technological systems,' Science, 250, 1359-1364(1990) https://doi.org/10.1126/science.2255906
  104. Siu, N. and M. Stutzke, 'PSA research and development at the U.S. Nuclear Regulatory Commission,' Proceedings of PSAM 9, Ninth International Conference of Probabilistic Safety Assessment and Management, Hong Kong, China, May 18-23, 2008

피인용 문헌

  1. Methodological and Practical Comparison of Integrated Probabilistic Risk Assessment (I-PRA) with the Existing Fire PRA of Nuclear Power Plants pp.1943-7471, 2018, https://doi.org/10.1080/00295450.2018.1486159