Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Round robin analysis of vessel failure probabilities for PTS events in Korea

  • Jhung, Myung Jo (Department of Nuclear Safety Research, Korea Institute of Nuclear Safety) ;
  • Oh, Chang-Sik (Department of Nuclear Safety Research, Korea Institute of Nuclear Safety) ;
  • Choi, Youngin (Department of Nuclear Safety Research, Korea Institute of Nuclear Safety) ;
  • Kang, Sung-Sik (Department of Nuclear Safety Research, Korea Institute of Nuclear Safety) ;
  • Kim, Maan-Won (Mechanical Engineering Lab., Central Research Institute, Korea Hydro & Nuclear Power Co., Ltd) ;
  • Kim, Tae-Hyeon (Mechanical Engineering Lab., Central Research Institute, Korea Hydro & Nuclear Power Co., Ltd) ;
  • Kim, Jong-Min (Safety Materials Technology Development Division, Korea Atomic Energy Research Institute) ;
  • Kim, Min Chul (Safety Materials Technology Development Division, Korea Atomic Energy Research Institute) ;
  • Lee, Bong Sang (Safety Materials Technology Development Division, Korea Atomic Energy Research Institute) ;
  • Kim, Jong-Min (Mechanical Engineering Group, KEPCO Engineering & Construction Co., Inc.) ;
  • Kim, Kyuwan (Mechanical Engineering Group, KEPCO Engineering & Construction Co., Inc.)
  • Received : 2019.12.04
  • Accepted : 2020.01.26
  • Published : 2020.08.25
Download PDF
⟨ Previous Next ⟩

Abstract

Round robin analyses for vessel failure probabilities due to PTS events are proposed for plant-specific analyses of all types of reactors developed in Korea. Four organizations, that are responsible for regulation, operation, research and design of the nuclear power plant in Korea, participated in the round robin analysis. The vessel failure probabilities from the probabilistic fracture mechanics analyses are calculated to assure the structural integrity of the reactor pressure vessel during transients that are expected to initiate PTS events. The failure probabilities due to various parameters are compared with each other. All results are obtained based on several assumptions about material properties, flaw distribution data, and transient data such as pressure, temperature, and heat transfer coefficient. The realistic input data can be used to obtain more realistic failure probabilities. The various results presented in this study will be helpful not only for benchmark calculations, result comparisons, and verification of PFM codes developed but also as a contribution to knowledge management for the future generation.

Keywords

References

  1. M.J. Jhung, S.H. Kim, Y.H. Choi, Y.S. Chang, X. Xu, J.M. Kim, J.W. Kim, C. Jang, Probabilistic fracture mechanics round robin analysis on reactor pressure vessels during pressurized thermal shock, J. Nucl. Sci. Technol. 47 (12) (2010) 1131-1139. https://doi.org/10.3327/jnst.47.1131
  2. OECD/NEA, Comparison Report of RPV Pressurized Thermal Shock International Comparative Assessment Study (PTS ICAS), NEA/CSNI/R(99)3, NEA Committee on the Safety of Nuclear Installations, Paris, 1999.
  3. OECD/NEA, Proceedings of the CSNI Workshop on Structural Reliability Evaluation and Mechanical Probabilistic Approaches of NPP Components, NEA/CSNI/R(2007), vol. 18, NEA Committee on the Safety of Nuclear Installations, Paris, 2008.
  4. IAEA, Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessment, IAEA-TECDOC-1627, International Atomic Energy Agency, Vienna, 2010.
  5. KINS, PFM Round Robin Analysis on RPV Integrity During PTS by Korean Participants, KINS/RR-686, Korea Institute of Nuclear Safety, Daejeon, 2009.
  6. KINS, 2016, A-Pro2: Phase 2 ASINCO Project for Probabilistic Fracture Mechanics Analysis e Korea Results, KINS/RR-1483, Korea Institute of Nuclear Safety, Daejeon.
  7. KINS, Reactor Probabilistic Integrity Evaluation (R-PIE) Code: User's Guide, KINS/RR-545, Korea Institute of Nuclear Safety, Daejeon, 2008.
  8. F.A. Simonen, K.I. Johnson, A.M. Liebetrau, D.W. Engel, E.P. Simonen, VISA-II: A Computer Code for Predicting the Probability of Reactor Vessel Failure, NUREG/CR-4486(PNL-5775), Pacific Northwest Laboratory, Richland, Washington, 1986.
  9. J.M. Kim, B.S. Lee, T.H. Kim, Y.S. Chang, Development of probabilistic fracture mechanics analysis codes for reactor pressure vessels considering recent embrittlement model and calculation method of SIF - progress of the work, in: PVP2016-63128, 2016 ASME Pressure Vessels and Piping Conference, Vancouver, Canada, 2016.
  10. P.T. Williams, T.L. Dickson, B.R. Bass, H.B. Klasky, Fracture Analysis of Vessels - Oak Ridge FAVOR, v16.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations, ORNL/LTR-2016-309, Oak Ridge National Laboratory, Oak Ridge, 2016.
  11. USNRC, Radiation Embrittlement of Reactor Vessel Materials, Regulatory Guide 1.99, Rev. 2, US Nuclear Regulatory Commission, Washington, DC, 1988.
  12. M. Kirk, M. Erickson, W. Server, G. Stevens, R. Cipolla, "Assessment of Fracture Toughness Models for Ferritic Steels Used in Section XI of the ASME Code Relative to Current Data-Based Models," PVP2014-28540, 2014 ASME Pressure Vessels and Piping Conference, Anaheim, California, 2014.
  13. USNRC, Format and content of plant-specific pressurized thermal shock safety analysis reports for pressurized water reactor, in: Regulatory Guide 1.154, US Nuclear Regulatory Commission, Washington, DC, 1987.