Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

  • Received : 2015.09.19
  • Accepted : 2015.11.15
  • Published : 2016.04.25
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Abstract

The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition ($RT_{NDT}$). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

Keywords

References

  1. U.S. Nuclear Regulatory Commission, Fracture toughness requirements for protection against pressurized thermal shock events, Title 10, Section 50.61, USNRC, Washington (DC), 1984.
  2. U.S. Nuclear Regulatory Commission, Alternative fracture toughness requirements for protection against pressurized thermal shock events, Title 10, Section 50.61a, USNRC, Washington (DC), 2010.
  3. Y. Kanto, M.J. Jhung, K. Ting, Y.B. He, K. Onizawa, S. Yoshimura, Summary of international PFM round robin analyses among Asian countries on reactor pressure vessel integrity during pressurized thermal shock, Int. J. Pressure Vessels Piping 90 (2012) 46-55.
  4. M.J. Jhung, S.H. Kim, Y.H. Choi, Y.S. Chang, W.Y. Xu, J.M. Kim, J.W. Kim, C. Jang, Probabilistic fracture mechanics round robin analysis of reactor pressure vessels during pressurized thermal shock, J. Nucl. Sci. Technol. 47 (2010) 1131-1139. https://doi.org/10.1080/18811248.2010.9720980
  5. G. Qian, M. Niffenegger, Procedures, methods and computer codes for the probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks, Nucl. Eng. Design 258 (2013) 35-50. https://doi.org/10.1016/j.nucengdes.2013.01.030
  6. G. Qian, V.F. Gonzalez-Albuixech, M. Niffenegger, Probabilistic assessment of a reactor pressure vessel subjected to pressurized thermal shocks by using crack distributions, Nucl. Eng. Design 270 (2014) 312-324. https://doi.org/10.1016/j.nucengdes.2013.12.062
  7. G. Qian, M. Niffenegger, Deterministic and probabilistic analysis of a reactor pressure vessel subjected to pressurized thermal shocks, Nucl. Eng. Design 273 (2014) 381-395. https://doi.org/10.1016/j.nucengdes.2014.03.032
  8. C.C. Huang, H.W. Chou, B.Y. Chen, R.F. Liu, H.C. Lin, Probabilistic fracture analysis for boiling water reactor pressure vessels subjected to low temperature over-pressure event, Ann. Nucl. Energy 43 (2012) 61-67. https://doi.org/10.1016/j.anucene.2011.12.028
  9. H.W. Chou, C.C. Huang, Effects of fracture toughness curves of ASME Section XI-Appendix G on a reactor pressure vessel under pressure-temperature limit operation, Nucl. Eng. Design 280 (2014) 404-412. https://doi.org/10.1016/j.nucengdes.2014.09.002
  10. M.J. Jhung, Reactor probabilistic integrity evaluation (R-PIE) code, KINS/RR-545, Korea Institute of Nuclear Safety, Daejeon (Korea), 2008.
  11. U.S. Nuclear Regulatory Commission, Radiation embrittlement of reactor vessel materials, Regulatory Guide 1.99, Rev. 2, USNRC, Washington (DC), 1988.
  12. 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 (WA), 1986.
  13. H.W. Chou, C.C. Huang, B.Y. Chen, H.C. Lin, R.F. Liu, Failure probability assessment for a boiling water reactor pressure vessel under low temperature over-pressure event, PVP2012-78244, Proceedings of the ASME 2012 Pressure Vessels & Piping Conference, Toronto (Canada), 2012.
  14. American Society of Mechanical Engineers, Fracture toughness criteria for protection against failure, ASME Boiler and Pressure Vessel Code, Section XI, Appendix G, ASME, New York (NY), 2011.
  15. R.D. Cheverton, D.G. Ball, The role of crack arrest in the evaluation of PWR pressure vessel integrity during PTS transients, Oak Ridge National Laboratory, Oak Ridge (TN), 1984.
  16. I.S. Raju, J.C. Newman Jr., Stress intensity factors for internal and external surface cracks in cylindrical vessels, J. Pressure Vessel Technol. 104 (1982) 293-298. https://doi.org/10.1115/1.3264220
  17. D.A. Curry, A model for predicting the influence of warm prestressing and strain aging on the cleavage fracture toughness of ferritic steels, Int. J. Fracture 22 (1983) 149-159.

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