References
- 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.
- 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.
- 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
- 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
- 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.
- 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.
- P. Vladislav, M. Pota, D. Lauerova, Probabilistic assessment of pressurised thermal shock, Nucl. Eng. Des 272 (2013) 84-91.
- 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
- 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.
- 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.
- 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.
- 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
- 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
- US Welding Research Council, Recommendations on Toughness Requirements for Ferritic Materials, US Welding Research Council, Washington, D.C., 1972.
- International Atomic Energy Agency, Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessments, International Atomic Energy Agency, Vienna, 2010. IAEA-TECDOC-1627.
- US Nuclear Regulatory Commission, Fracture Toughness Requirements, 10 CFR 50 Appendix G, US Nuclear Regulatory Commission, Washington, D.C., 1984.
- 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.
- 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
- 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
- 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.
- 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.
- 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
- 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.
- 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.
- 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.
Cited by
- Engineering critical assessment of RPV with nozzle corner cracks under pressurized thermal shocks vol.52, pp.11, 2016, https://doi.org/10.1016/j.net.2020.04.019
- Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels vol.13, pp.19, 2021, https://doi.org/10.3390/su131910510