Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Application of the French Codes to the Pressurized Thermal Shocks Assessment

  • Chen, Mingya (Suzhou Nuclear Power Research Institute, Life Management Center) ;
  • Qian, Guian (Paul Scherrer Institute, Nuclear Energy and Safety Department, Laboratory for Nuclear Materials) ;
  • Shi, Jinhua (Amec Foster Wheeler, Clean Energy Department) ;
  • Wang, Rongshan (Suzhou Nuclear Power Research Institute, Life Management Center) ;
  • Yu, Weiwei (Suzhou Nuclear Power Research Institute, Life Management Center) ;
  • Lu, Feng (Suzhou Nuclear Power Research Institute, Life Management Center) ;
  • Zhang, Guodong (Suzhou Nuclear Power Research Institute, Life Management Center) ;
  • Xue, Fei (Suzhou Nuclear Power Research Institute, Life Management Center) ;
  • Chen, Zhilin (Suzhou Nuclear Power Research Institute, Life Management Center)
  • Received : 2016.03.17
  • Accepted : 2016.06.03
  • Published : 2016.12.25
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Abstract

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

Keywords

References

  1. G.A. Qian, M. Niffenegger, Investigation on constraint effect of a reactor pressure vessel subjected to pressurized thermal shocks, Jour. Pre. Ves. Tec 137 (2015) 1-7.
  2. M.Y. Chen, F. Lv, R.S. Wang, P. Huang, X.B. Liu, G.D. Zhang, The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading, Nucl. Eng. Des 288 (2015) 84-91.
  3. W.E. Pennell, Structural integrity assessment of aging nuclear reactor pressure vessels, Nucl. Eng. Des 172 (1997) 27-47. https://doi.org/10.1016/S0029-5493(96)00006-4
  4. US Nuclear Regulatory Commission, Format and Content of Plant-specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors, Washington, D.C., 1987. Regulatory Guide, No. 1.154
  5. US Nuclear Regulatory Commission, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, US Nuclear Regulatory Commission, Washington, D.C., 1984, 10 CFR 50.61.
  6. Oak Ridge National Laboratory, Fracture Analysis of Vessels-Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods and Correlations, Oak Ridge National Laboratory, Washington, D.C., 2006.
  7. P. Vladislav, M. Pota, D. Lauerova, Probabilistic assessment of pressurised thermal shock, Nucl. Eng. Des 272 (2013) 84-91.
  8. M.Y. Chen, F. Lv, R.S. Wang, W.W. Yu, The probabilistic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading, Nucl. Eng. Des 294 (2015) 93-102. https://doi.org/10.1016/j.nucengdes.2015.08.020
  9. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Section XI, Appendix E, Evaluation of unanticipated operating events, American Society of Mechanical Engineers, New York, NY, 2013.
  10. RCC-M. Design and construction for mechanical components of PWR nuclear island, Sec. I. Subsec. Z. Annex Z G, Fast Fracture Resistance, French association for design, construction and in-service inspection rules for nuclear island components, Paris, 2010.
  11. RSE-M, In-service Inspection Rules for the Mechanical Components of PWR Nuclear Islands, French association for design, construction and in-service inspection rules for nuclear island components, Paris, 2007.
  12. Y.B. He, T. Isozaki, Fracture mechanics analysis and evaluation for the RPV of the Chinese Qinshan 300 MW NPP under PTS, Nucl. Eng. Des 201 (2000) 121-137. https://doi.org/10.1016/S0029-5493(00)00271-5
  13. M.Y. Chen, F. Lv, R.S. Wang, Structural integrity assessment of the reactor pressure vessel under the pressurized thermal shock loading, Nucl. Eng. Des 272 (2014) 84-91. https://doi.org/10.1016/j.nucengdes.2014.01.021
  14. US Welding Research Council, Recommendations on Toughness Requirements for Ferritic Materials, US Welding Research Council, Washington, D.C., 1972.
  15. International Atomic Energy Agency, Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessments, International Atomic Energy Agency, Vienna, 2010. IAEA-TECDOC-1627.
  16. US Nuclear Regulatory Commission, Fracture Toughness Requirements, 10 CFR 50 Appendix G, US Nuclear Regulatory Commission, Washington, D.C., 1984.
  17. F. Lu, R.S. Wang, P. Huang, H.Y. Qian, Prediction of irradiation embrittlement for Chinese domestic A508-3 steel, in: Proceedings of the 18th International Conference on Nuclear Engineering. Beijing, 2010.
  18. G.A. Qian, M. Niffenegger, Procedures, methods, and computer codes for the probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks, Nucl. Eng. Des 258 (2013) 35-50. https://doi.org/10.1016/j.nucengdes.2013.01.030
  19. M. Scheuerer, J. Weis, Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions, Nucl. Eng. Des 253 (2012) 343-350. https://doi.org/10.1016/j.nucengdes.2011.08.063
  20. J.P. Fontes, C. Raynaud, A. Martin, Reactor pressure vessel: EDF R&D program to support lifetime management, in: Pressure Vessels & Piping Conference, 2011, pp. 1-5.
  21. T.L. Dickson, J.A. Keeney, J.W. Bryson, Validation of a linearelastic fracture methodology for postulated flaws embedded in the wall of a nuclear reactor pressure vessel, in: Proceedings of ASME Pressure Vessel and Pipings, Seattle, 2000.
  22. D. Moinereau, G. Bezdikian, C. Faidy, Methodology for the pressurized thermal shock evaluation: recent improvement in French RPV PTS assessment, Int. J. Pres. Ves. Pip 78 (2001) 69-83. https://doi.org/10.1016/S0308-0161(01)00023-0
  23. M. Brumovsky, M. Kytka, R. Kopriva, Cladding in RPV integrity and lifetime evaluation, in: 14th International Conference on Pressure Vessel Technology, 2015, pp. 1-10.
  24. J.S. Lee, I.S. Kimi, C.H. Jang, A. Kimura, Irradiation embrittlement of cladding and HZA of RPV steel, Nucl. Eng. Tech 38 (2005) 405-410.
  25. H. Churiep, G. Balard, E. Meister, F. Clemendot, P. Todeschini, French nuclear reactor pressure vessel integrity assessment and life management strategy, in: Pressure Vessels & Piping Conference, 2011, pp. 1-6.

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