Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Integrated Level 1-Level 2 decommissioning probabilistic risk assessment for boiling water reactors

  • Mercurio, Davide (Jensen Hughes) ;
  • Andersen, Vincent M. (Jensen Hughes) ;
  • Wagner, Kenneth C. (Dycoda LLC)
  • Received : 2017.11.08
  • Accepted : 2018.03.08
  • Published : 2018.06.25
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Abstract

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

Keywords

References

  1. M. Stamatelatos, G. Apostolakis, H. Dezfuli, C. Everline, S. Guarro, P. Moieni, A. Mosleh, T. Paulos, R. Youngblood, Probabilistic Risk Assessment Procedure Guide for NASA Managers and Practitioners, Prepared for Office of Safety and Mission Assurance NASA Headquarters Washington, August 2002, ww.hq.nasa.gov/office/codeq/doctree/praguide.pdf.
  2. R.O. Gauntt, et al., MELCOR Computer Code Manuals, vols. 1 and 2, Sandia National Laboratories, September 2005. NUREG/CR-6119, Revision 3, SAND2005-5713, Version 1.8.6.
  3. D. Mercurio, Discrete Dynamic Event Tree Modeling and Analysis of Nuclear Power Plant Crews for Safety Assessment, DISS. ETH NO. 19321, 2011.
  4. CAFTA Fault Tree Analysis System, Electric Power Research Institute, January 2009, Version 5.4.
  5. K.J. Kvarfordt, S.T. Wood, C.L. Smith, Systems Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) Data Loading Manual, 2008. INL/EXT-05-00643, Rev. 1.
  6. M. Drouin, G. Parry, J. Lehner, G. Martinez-Guridi, J. LaChance, T. Wheeler, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making vol. 1, 2009. NUREG-1855.
  7. A.L. Camp, et al., Light Water Reactor Hydrogen Manual, Sandia National Laboratories, August 1983. NUREG/CR-2726, SAND82-1137.
  8. D. Mercurio, K.C. Wagner, M.T. Leonard, Y.Y. Bayraktarli, Low power shutdown level 2 probabilistic risk assessment methodology for boiling water reactors, in: ANS PSA 2013 International Topical Meeting on Probabilistic Safety Assessment and Analysis Columbia, SC, September 22-26, 2013, on CD-ROM, American Nuclear Society, LaGrange Park, IL, 2013.
  9. D. Mercurio, V. Andersen, K.C. Wagner, Decommissioning level 2 probabilistic risk assessment methodology for boiling water reactors, in: 13th International Conference on Probabilistic Safety Assessment and Management (PSAM 13), 2-7 October, Sheraton Grande Walkerhill, Seoul, Korea, 2016. www.psam13.org.
  10. N.E. Todreas, M.S. Kazimi, Nuclear Systems: Vol. I, Thermal Hydraulic Fundamentals, Taylor and Francis, Hemisphere, NY, 2001, 1990, 3rd printing.
  11. N.E. Todreas, M.S. Kazimi, Nuclear Systems: Vol. II, Elements of Thermal Hydraulic Design, Hemisphere, NY, 1990.
  12. NUREG/CR-2815 BNL-NUREG-51559, Probabilistic Safety Analysis Procedure Guide, January 1984.
  13. MAAP, "Transmittal Document for MAAP5 Code Revision MAAP 5.02," Prepared by Fauske & Associates, LLC, FAI/13e0801, November 2013.
  14. ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, 2009.
  15. NUREG/CR-4467, D.J. Alpert, et al., Relative Importance of Individual Elements to Reactor Accident Consequences Assuming Equal Release Fractions, SAND85-2575, Sandia National Laboratory, March 1986.

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