Advertisement

Deterministic and Probabilistic Safety Analysis

  • Mohammad Modarres
  • Inn Seock Kim

Abstract

The main theme of this chapter is the process and evolution of deterministic and probabilistic safety analyses that have played a backbone role in assuring public health and safety in the peaceful uses of nuclear power. The chapter begins with a discussion of the origin of nuclear power safety analysis together with the overall perspectives of both deterministic and probabilistic approaches that are still prevalent, although there is an increasing trend in application of probabilistic safety analysis in safety-related decision making. Deterministic approaches, such as the defense-in-depth or safety margin, are regarded as a means to cope with uncertainties associated with adequacy of safety features. As probabilistic methods and applications gain maturity and acceptance, the uncertainties associated with safety features are measured and described probabilistically. The chapter concludes with a detailed discussion of the probabilistic safety assessment and its uses in nuclear power safety analysis.

Keywords

Nuclear Power Plant International Atomic Energy Agency Atomic Energy Commission Fault Tree Nuclear Regulatory Commission 
These keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

References

  1. Ahearne J et al (2001) The regulatory process for the nuclear power reactors: a review. Report of the CSIS Nuclear Regulatory Process Review Steering CommitteeGoogle Scholar
  2. Ahn SK, Kim IS, Oh KM (2010) Deterministic and risk-informed approaches for safety analysis of advanced reactors: part I, deterministic approaches. To appear, Reliab Eng Syst SafetyGoogle Scholar
  3. AICHE (1989) Guidelines for process equip- ment data. Center for Chemical Process Safety, American Institute of Chemical Engineers, New YorkGoogle Scholar
  4. Aldemir T, Siu N (1996) Reliability and safety analysis of dynamic process systems. Reliab Eng Syst Safety (Special Issue) 52:181–337CrossRefGoogle Scholar
  5. ANS (1973) Nuclear safety criteria for the design of stationary pressurized water reactor plants, ANSI N18.2-1973. American National Standards Institute, American Nuclear Society, HinsdaleGoogle Scholar
  6. ANS (1983) Nuclear safety criteria for the design of stationary pressurized water reactor plants, ANSI/ANS-51.1-1983. American National Standards Institute, American Nuclear Society, La Grange ParkGoogle Scholar
  7. Ascher H, Feingold H (1984) Repairable systems reliability: modeling and inference, misconception and their causes. Marcel Dekker, New YorkGoogle Scholar
  8. Azarkhail M (2007) Agent autonomy approach to physics based reliability modeling of structures and mechanical systems. Ph.d. Dissertation, University of Maryland, College ParkGoogle Scholar
  9. Azarkhail M, Modarres M (2004) A Study of implications of using importance measures in risk-informed decisions. In: PSAM-7, ESREL 04 Joint Conference, Berlin, Germany, 2004Google Scholar
  10. Azarkhail M, Modarres M (2006) An intelligent-agent-oriented approach to risk analysis of complex dynamic systems with applications in planetary missions. In: Proceedings of the 8th international conference on probabilistic safety assessment and management, PSAM8, New Orleans, USA, 2006Google Scholar
  11. Birnbaum ZW (1969) On the importance of different components in a multicomponent system. In: Krishnaiah PR (ed) Multivariate analysis II. Academic Press, New YorkGoogle Scholar
  12. Carlisle R (1997) Probabilistic risk assessment in nuclear reactors: engineering success, public relations failure. Technol Culture 38:920–941CrossRefGoogle Scholar
  13. Chang YH, Mosleh A, Dang V (2003) Dynamic probabilistic risk assessment: framework, tool, and application. In: Proceedings of the society for risk analysis annual meeting, Baltimore, 2003Google Scholar
  14. Colglazier E, Weatherwas R (1986) Failure estimates for the space shuttle. In: Abstracts of the Society for Risk Analysis Annual Meeting, Boston, 1986Google Scholar
  15. Crow LH (1990) Evaluating the reliability of repairable systems. In: IEEE Proceedings of the annual reliability and maintainability sym- posium, pp. 275–279CrossRefGoogle Scholar
  16. Delaney MJ, Apostolakis GE, Driscoll MJ (2005) Risk-informed design guidance for future reactor systems. Nuc Eng Des 235:1537–1556CrossRefGoogle Scholar
  17. Dezfuli H, Modarres M (1984) A truncation methodology for evaluation of large fault trees. IEEE Trans Reliab R-33:325–328Google Scholar
  18. DOD (1995) Military handbook, reliability prediction of electronic equipment. MIL-HDBK-217F, Department of DefenseGoogle Scholar
  19. Dugan J, Bavuso S, Boyd M (1993) Dynamic fault tree models for fault tolerant computer systems. IEEE Trans Reliab 40(3):363Google Scholar
  20. EPRI (1995) PSA applications guide. Electric Power Research Institute, TR-105396, Palo AltoGoogle Scholar
  21. EPRI (2005) EPRI/NRC-RES fire PRA methodology for nuclear power facilities. EPRI 1011989, NUREG/CR-6850, Palo Alto, US Nuclear Regulatory Commission, Washington DCGoogle Scholar
  22. Ericson C (1999) Fault tree analysis – A history. In: Proceedings of the 17th international system safety conference, Orlando, 1999Google Scholar
  23. Farmer FR (1967) Reactor safety and siting: A proposed risk criterion. Nucl Safety 8:539–548Google Scholar
  24. Fischer, D (1997) History of the international atomic energy agency: The first forty years. International Atomic Energy AgencyGoogle Scholar
  25. Fleming KN (2003) Issues and recommendations for advancement of pra technology in risk-informed decision making, NUREG/CR-6813. US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  26. Fleming KN, Silady FA (2002) A risk informed defense-in-depth framework for existing and advanced reactors. Reliab Eng Syst Safety 78:205–225CrossRefGoogle Scholar
  27. Ford D (1977) A history of Federal nuclear safety assessments: From WASH 740 through the reactor safety study. Union of Concerned Scientists, WashingtonGoogle Scholar
  28. Frankel E (2002) Systems reliability and risk analysis, 2nd edn. Kluwer Academic Publishers, BostonGoogle Scholar
  29. Fussell J (1975) How to hand calculate system reliability and safety characteristics. IEEE Trans Reliab 24(3):169–174CrossRefGoogle Scholar
  30. Green A, Bourne A (1972) Reliability technology. Wiley, LondonGoogle Scholar
  31. Hu YS, Modarres M (1999) Evaluating system behavior through dynamic master logic diagram (DMLD) modeling. Reliab Eng Syst Safety 64:241–269CrossRefGoogle Scholar
  32. Hunt RN, Modarres M (1984) Integrated economic risk management in a nuclear power plant. In: Proceedings of the Annual Meeting of the Society for Risk Analysis, Knoxville, TN, October, 1984; published in Risk Abstracts, Vol. 2, No. 2.Google Scholar
  33. IAEA (1991) Safety culture. 75-INSAG-4, A report by the international nuclear safety advisory group, International Atomic Energy AgencyGoogle Scholar
  34. IAEA (1996) Defense in depth in nuclear safety. INSAG-10, A report by the international nuclear safety advisory group, International Atomic Energy AgencyGoogle Scholar
  35. IEEE (1984) IEEE guide to the collection and presentation of electrical, electronic, sensing component, and mechanical equipment reliability data for nuclear-power generating stations, IEEE Std 500–1984. Institute of Electrical and Electronics Engineers, New YorkGoogle Scholar
  36. Kaplan S, Garrick J (1981) On the quantitative definition of risk. Risk Anal 1:11–28CrossRefGoogle Scholar
  37. Kapur KC, Lamberson LR (1977) Reliability in engineering design. Wiley, New YorkGoogle Scholar
  38. Kemeny JG, Babbitt B, Haggerty PE, Lewis C et al (1979) Staff reports to the President’s commission on the accident at three mile island. Reports of the Technical Assessment Task Force, Washington, DCGoogle Scholar
  39. Kim IS (1996) Improving technical specifications from a risk perspective. Reliab Eng Syst Safety 54:83–87CrossRefGoogle Scholar
  40. Kim IS (2008) Feasibility study for development of human error pattern analysis methodology for operational experience feedback. Korea Institute of Nuclear Safety, Republic of KoreaGoogle Scholar
  41. Kim IS, Modarres M (1987) Application of goal tree-success tree model as the knowledge base of operator advisory systems. Nucl Eng Des 104: 67–81CrossRefGoogle Scholar
  42. Kim IS, Martorell S, Vesely WE, Samanta PK (1994) Risk analysis of surveillance requirements including their adverse effects. Reliab Eng Syst Safety 45:225–234CrossRefGoogle Scholar
  43. Kim IS, Ahn SK, Hong SJ, Lee HJ (2008) New insights on risk-informed performance-based approaches to technology-neutral regulatory framework for generation IV reactors. In: Proceedings of the 9th international probabilistic safety assessment and management conference, PSAM9, Hong Kong, 2008Google Scholar
  44. Kim IS, Ahn SK, Oh KM (2010) Deterministic and risk-informed approaches for safety analysis of advanced reactors: part II, risk-informed approaches. To appear, Reliab Eng Syst SafetyGoogle Scholar
  45. Kouts H (1998) History of safety research programs and some lessons to be drawn from it. In: 26th water reactor safety information meeting, Bethesda, 1998Google Scholar
  46. Kumamoto H, Henley EJ (1996) Probabilistic risk assessment for engineers and scientists. IEEE Press, New YorkGoogle Scholar
  47. Lewis H et al (1975) American physical society reactor study review group. Report on WASH-1400Google Scholar
  48. Meserve R (2001) The evolution of safety goals and their connection to safety culture. Speech at the American Nuclear Society topical meeting on safety goals and safety culture, Milwaukee, 2001Google Scholar
  49. Modarres M (1993) What every engineer should know about reliability and risk analysis. Marcel Dekker, New YorkGoogle Scholar
  50. Modarres M (2006) Risk analysis in engineering, techniques, tools and trends. CRC Press, Boca RatonGoogle Scholar
  51. Modarres M (2009) Advanced nuclear power plant regulation using risk-informed and performance-based methods. Reliab Eng Syst Safety 94:211–217CrossRefGoogle Scholar
  52. Modarres M, Kaminskiy M, Krivtsov V (1999) Reliability engineering and risk analysis: A practical guide. Marcel Dekker, New YorkGoogle Scholar
  53. Mosleh A, Fleming KN, Parry GW, Paula HM et al (1988) Procedure for treating common cause failures in safety and reliability studies, NUREG/CR-4780, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  54. Stamatelatos M, Apostolakis G, Dezfuli H, Everline C et al (2002) Probabilistic risk assessment procedures guide for NASA managers and practitioners, Vers. 1.1. National Aeronautics and Space Administration, Washington, DCGoogle Scholar
  55. NEI (2006) Risk-informed technical specifications initiative 4b, Risk-managed technical specifications (RMTS) guidelines, NEI-06–09, Rev. 0. Nuclear Energy Institute, Washington, DCGoogle Scholar
  56. NEI (2007) Risk-informed technical specifications initiative 5b, risk-informed method for control of surveillance frequencies, NEI-04–10, Rev. 1. Nuclear Energy Institute, Washington, DCGoogle Scholar
  57. Nelson W (1990) Accelerated testing: statistical models, test plans and data analyses. Wiley, New YorkGoogle Scholar
  58. NSAC (1979) Analysis of three mile island – unit 2 accident. Nuclear Safety Analysis Center, NSAC-1Google Scholar
  59. NUMARC (1993) Industry guideline for monitoring the effectiveness of maintenance at nuclear power plants, NUMARC 93–01. Nuclear Management and Resources CouncilGoogle Scholar
  60. Poucet A (1988) Survey of methods used to assess human reliability in the human factors reliability benchmark exercise. Reliab Eng Syst Safety 22:257–268CrossRefGoogle Scholar
  61. Reason J, Hobbs A (2003) Managing maintenance error. Ashgate, EnglandGoogle Scholar
  62. Rhodes R (1986) The making of the atomic bomb. Simon and Schuster, New YorkGoogle Scholar
  63. Rogovin M, Frampton GT Jr. (1980) Three mile island – A report to the commissioners and to the public. Nuclear Regulatory Commission Special Inquiry Group, NUREG/CR-1250, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  64. Samanta PK, Kim IS, Mankamo T, Vesely WE (1994) Handbook of methods for risk-based analyses of technical specifications. NUREG/CR-6141, US Nuclear Regulatory Commission, Washington, DCCrossRefGoogle Scholar
  65. Sattison MB et al (1990) Analysis of core damage frequency: zion, unit 1 internal events. NUREG/CR-4550, Vol. 7, Rev. 1, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  66. Smidts C (1996) Software reliability. In: Whitaker JC (ed) The electronics handbook. CRC Press, Boca RatonGoogle Scholar
  67. Sorensen J, Apostolakis G, Kress T, Powers D (1999) On the role of defense-in-depth in risk-informed regulation. In: Proceedings of the probabilistic safety assessment PSA’99, Washington, DC., American Nuclear Society, La Grange Park, ILGoogle Scholar
  68. Stamatis DH (2003) Failure mode and effect analysis: FMEA from theory to execution, 2nd edn. ASQ Quality Press, Wisconsin, USAGoogle Scholar
  69. Starr C (1969) Social Benefit versus technological risk. Science 19:1232–1238CrossRefGoogle Scholar
  70. Swain AD, Guttmann HE (1983) Handbook of human reliability analysis with emphasis on nuclear power plant applications, NUREG/CR-1278, US Nuclear Regulatory Commission (USNRC), Washington DCGoogle Scholar
  71. USAEC (1957) WASH-740, Theoretical possibilities and consequences of major accidents in large nuclear power plants. US Atomic Energy Commission, AKA The Brookhaven ReportGoogle Scholar
  72. USAEC (1966) Minutes of the AEC general advisory committee. US Atomic Energy CommissionGoogle Scholar
  73. USNRC (1956) Domestic licensing of production and utilization facilities, Title 10. Code of Federal Regulations, Part 50, 21FR355Google Scholar
  74. USNRC (1975) Reactor safety study – an assessment of accident risks in US commercial nuclear power plants. WASH-1400 (NUREG-75/014, US Nuclear Regulatory Commission, Washington, DC)Google Scholar
  75. USNRC (1977) Single failure criterion. SECY-77–439Google Scholar
  76. USNRC (1978) Standard format and content of safety analysis reports for nuclear power plants – LWR edition. Regulatory Guide 1.70, Revision 3Google Scholar
  77. USNRC (1980a) NRC action plan developed as a result of the TMI-2 accident. NUREG-0660, Rev. 1, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  78. USNRC (1980b) Severe accident risks: An assessment for five US Nuclear Power Plants. NUREG-1150, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  79. USNRC (1983) PRA procedures guide: a guide to the performance of probabilistic risk assessments for nuclear power plants. NUREG/CR-2300, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  80. USNRC (1986) Safety goals for the operation of nuclear power plants; policy statement. 51 FR 30028Google Scholar
  81. USNRC (1988) Individual plant examination for severe accident vulnerabilities – 10 CFR 50.54(f). Generic Lett 1:88–20Google Scholar
  82. USNRC (1991) Individual plant examination of external events (ipeee) for severe accident vulnerabilities – 10 CFR 50.54(f). Generic Lett 4: 88–20Google Scholar
  83. USNRC (1993) Final policy statement on technical specifications improvements for nuclear power plants. 58FR39132Google Scholar
  84. USNRC (1995) Use of probabilistic risk assessment methods in nuclear regulatory activities; final policy statement. 60FR42622Google Scholar
  85. USNRC (1997a) Maintenance rule status, results, and lessons learned. SECY-97–055Google Scholar
  86. USNRC (1997b) Monitoring the effectiveness of maintenance at nuclear power plants. Regulatory Guide 1.160Google Scholar
  87. USNRC (1998a) White paper on risk-informed and performance-based regulation. SECY-98–144Google Scholar
  88. USNRC (1998b) An approach for plant-specific, risk-informed decisionmaking: inservice testing. Regulatory Guide 1.175Google Scholar
  89. USNRC (1998c) An approach for plant-specific, risk-informed decisionmaking: technical specifications. Regulatory Guide 1.177Google Scholar
  90. USNRC(1999a) General design criteria for nuclear power plants, Appendix A of 10 CFR 50Google Scholar
  91. USNRC (1999b) General requirements for monitoring the effectiveness of maintenance at nuclear power plants, Title 10, Code of Federal Regulations, Part 65. 64FR72001Google Scholar
  92. USNRC (1999c) Staff Briefing on reactor inspection, enforcement and assessmentGoogle Scholar
  93. USNRC (2000) Consolidated line item improvement process for adopting standard technical specifications changes for power reactors. Regulatory Issue Summary (RIS) 2000–06Google Scholar
  94. USNRC (2001) Modified reactor safety goal policy statement. SECY-01–0009Google Scholar
  95. USNRC (2002a) Perspectives gained from the individual plant examination of external events (IPEEE) program. NUREG-1742, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  96. USNRC (2002b) An approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the licensing basis. Regulatory Guide 1.174, Rev. 1Google Scholar
  97. USNRC (2003a) Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, Title 10, Code of Federal Regulations. Part 46, 68FR54142Google Scholar
  98. USNRC (2003b) An approach for plant-specific risk-informed decisionmaking for inservice inspection of piping. Regulatory Guide 1.178, Rev. 1Google Scholar
  99. USNRC (2003c) NUREG-CR-6813, Issues and recommendations for advancement of PRA technology in risk-informed decision making. Letter of ACRS Chairman M.V. Bonaca to EDO Director W.D. Travers, ACRSR-2034, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  100. USNRC (2004) An approach for estimating the frequencies of various containment failure modes and bypass events. NUREG/CR-6595, Rev.1, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  101. USNRC (2005a) Technical work to support evaluation of a broader change to the single-failure criterion. Technical reportGoogle Scholar
  102. USNRC (2005b) Independent verification of the mitigating systems performance index (MSPI) results for the pilot plants. NUREG-1816, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  103. USNRC (2006) Reactor oversight process. NUREG-1649, Rev. 4, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  104. USNRC (2007a) Combined license applications for nuclear power plants (LWR edition). Regulatory Guide 1.206Google Scholar
  105. USNRC (2007b) Standard review plan for the review of safety analysis reports for nuclear powerplants. NUREG-0800, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  106. USNRC (2007c) Feasibility study for a risk-informed and performance-based regulatory structure for future plant licensing. NUREG-1860, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  107. USNRC (2007d) 19.0 Probabilistic risk assessment and severe accident evaluation for new reactors. In: Standard review plan for the review of safety analysis reports for nuclear power plants. NUREG-0800, US Nuclear Regulatory Commission, Washington, DCGoogle Scholar
  108. USNRC (2007e) An approach for determining the technical adequacy of psrobabilistic risk assessment results for risk-informed activities. Regulatory guide 1.200, Rev. 1Google Scholar
  109. USNRC (2007f) Operating reactor assessment program, inspection manual chapter 0305Google Scholar
  110. Wood W (1983) Nuclear safety, risks and regulation. American Enterprise Institute – Public Policy ResearchGoogle Scholar

Copyright information

© Springer Science+Business Media, LLC 2010

Authors and Affiliations

  • Mohammad Modarres
    • 1
  • Inn Seock Kim
    • 2
  1. 1.Department of Mechanical EngineeringUniversity of MarylandCollege ParkUSA
  2. 2.ISSA TechnologyInc.GermantownUSA

Personalised recommendations