Table of Contents
International Journal of Nuclear Energy
Volume 2015, Article ID 785041, 9 pages
http://dx.doi.org/10.1155/2015/785041
Research Article

Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

Institute of Nuclear Energy Research, Taoyuan 32546, Taiwan

Received 5 February 2015; Accepted 26 October 2015

Academic Editor: Arkady Serikov

Copyright © 2015 Hsoung-Wei Chou and Chin-Cheng Huang. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

Linked References

  1. BWRVIP-05, “BWR reactor pressure vessel shell weld inspection recommendations (BWRVIP-05),” BWR Vessel and Internals Project EPRI TR-105697, 1995. View at Google Scholar
  2. USNRC, “Final safety evaluation of the BWR vessel and internals project BWRVIP-05,” USNRC TAC M93925, U.S. Nuclear Regulatory Commission, 1998. View at Google Scholar
  3. M. EricksonKirk, M. Junge, W. Arcieri et al., “Technical basis for revision of the pressurized thermal shock (PTS) screening limit in the PTS rule (10 CFR 50.61),” Tech. Rep. NUREG-1806, US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington, DC, USA, 2007. View at Google Scholar
  4. M. T. EricksonKirk and T. L. Dickson, “Recommended screening limits for pressurized thermal shock (PTS),” Report NUREG-1874, US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington, DC, USA, 2010. View at Google Scholar
  5. US Nuclear Regulatory Commission, “Code of federal regulations, title 10, section 50.61a, alternate fracture toughness requirements for protection against pressurized thermal shock events,” 2010.
  6. H.-W. Chou, C.-C. Huang, B.-Y. Chen, R.-F. Liu, and H.-C. Lin, “Probabilistic fracture analysis for boiling water reactor pressure vessels subjected to low temperature over-pressure event,” in Proceedings of the ASME Pressure Vessels and Piping Division/K-PVP Conference (PVP '10), pp. 157–164, Bellevue, Wash, USA, July 2010. View at Publisher · View at Google Scholar · View at Scopus
  7. C.-C. Huang, H.-W. Chou, B.-Y. Chen, R.-F. Liu, and H.-C. Lin, “Probabilistic fracture analysis for boiling water reactor pressure vessels subjected to low temperature over-pressure event,” Annals of Nuclear Energy, vol. 43, pp. 61–67, 2012. View at Publisher · View at Google Scholar · View at Scopus
  8. F. A. Simonen, S. R. Doctor, G. J. Schuster, and P. G. Heasler, “A generalized procedure for generating flaw-related inputs for the FAVOR code,” Tech. Rep. NUREG/CR-6817 (PNNL14268), Pacific Northwest National Laboratory, Richland, Wash, USA, 2003. View at Google Scholar
  9. P. T. Williams, T. L. Dickson, and S. Yin, “Fracture analysis of vessels—Oak Ridge FAVOR, v09.1, computer code: theory and implementation of algorithms, methods, and correlations,” Tech. Rep. ORNL/TM-2010/5, Oak Ridge National Laboratory, Oak Ridge, Tenn, USA, 2009. View at Google Scholar
  10. P. T. Williams, T. L. Dickson, and S. Yin, “Fracture analysis of vessels-Oak Ridge FAVOR, v09.1, computer code: user's guide,” Tech. Rep. ORNL/TM-2010/4, Oak Ridge National Laboratory, Oak Ridge, Tenn, USA, 2009. View at Google Scholar
  11. G. Qian and M. Niffenegger, “Integrity analysis of a reactor pressure vessel subjected to pressurized thermal shocks by considering constraint effect,” Engineering Fracture Mechanics, vol. 112-113, pp. 14–25, 2013. View at Publisher · View at Google Scholar · View at Scopus
  12. G. Qian and M. Niffenegger, “Deterministic and probabilistic analysis of a reactor pressure vessel subjected to pressurized thermal shocks,” Nuclear Engineering and Design, vol. 273, pp. 381–395, 2014. View at Publisher · View at Google Scholar · View at Scopus
  13. H.-W. Chou and C.-C. Huang, “Effects of fracture toughness curves of ASME section XI-appendix G on a reactor pressure vessel under pressure-temperature limit operation,” Nuclear Engineering and Design, vol. 280, pp. 404–412, 2014. View at Publisher · View at Google Scholar · View at Scopus
  14. E. D. Eason, G. R. Odette, R. K. Nanstad, and T. Yamamoto, “A physically based correlation of irradiation-induced transition temperature shifts for RPV steels,” Report ORNL/TM-2006/530, Oak Ridge National Laboratory, Oak Ridge, Tenn, USA, 2007. View at Google Scholar
  15. US Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2. Radiation Embrittlement of Reactor Vessel Materials, US Nuclear Regulatory Commission, 1988.
  16. H.-C. Chu, J.-Y. Huang, H.-H. Hsu et al., “Chinshan nuclear power station unit 2 reactor vessel surveillance materials testing and fracture toughness analysis,” Report INER-A0089, Institute of Nuclear Energy Research, Taipei, Taiwan, 1999. View at Google Scholar