U.S. Nuclear Regulatory Commission．10 CFR 50.61,Fracture toughness requirements for protection against pressurized thermal shock events[S]．Washington DC：Nuclear Regulatory Commission,1984．
[2]
Qian G,Niffenegger M．Integrity analysis of a reactor pressure vessel subjected to pressurized thermal shocks by considering constraint effect[J]．Engineering Fracture Mechanics,2013,112-113：14-25．
[3]
Wallin K．Quantifying T stress controlled constraint by the master curve transition temperature T 0 [J]．Engineering Fracture Mechanics,2001,68(3)：303-328．
[4]
Scheuerer M,Weis J．Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions[J]．Nuclear Engineering and Design,2012,253：343-350．
[5]
IAEA 1627．Pressurized thermal shock in nuclear power plants：good practices for assessment[R]．Austria：IAEA,2010：3-63．
[6]
Chen M Y,Lu F,Wang R S,et al．Structural integrity assessment of the reactor pressure vessel under the pressurized thermal shock loading[J]．Nuclear Engineering and Design,2014,272：84-91．
[7]
Wang R S,Lu F,Chen M Y,et al．Use of the failure assessment diagram(FAD) to evaluate margins in the ASME code for P-T curves[C]//Proceedings of the ASME 2012 Pressure Vessels & Piping Division Conference,Toronto：ASME,2012：33-36．
[8]
The American Society of Mechanical Engineers．Boiler and pressure vessel code,section XI,Rules for Inservice inspection of nuclear power plant components,IWB 3000[S]．New York：ASME,2013．
[9]
Lee T J,Choi J B,Kim Y J,et al．A parametric study on pressure-temperature limit curve using 3-D finite element analyses[J]．Nuclear Engineering and Design,2002,214(1-2)：73-81．
[10]
Pennell W E,Malik S N M．Structural integrity assessment of aging nuclear reactor pressure vessels [J]．Nuclear Engineering and Design,1997,172(1-2)：27-47．
[11]
RCCM．Design and Construction Rules for Mechanical Components of PWR Nuclear Islands[S]．French：RCCM,2007．
[12]
The American Society of Mechanical Engineers．Boiler and Pressure Vessel Code,Section XI,Rules for Inservice Inspection of Nuclear Power Plant Components,Appendix A：Analysis of flaw[S]．New York：ASME,2013．
[13]
Chen M Y,Lu F,Wang R S,et al．Use of the failure assessment diagram to evaluate the safety of the reactor pressure vessel[J]．Journal of Pressure Vessel Technology,2015,137(5)：051203．
[14]
U.S. Nuclear Regulatory Commission．Regulatory Guide,No.1.99,Revision 2,Radiation embrittlement of reactor vessel materials[S]．Washington DC：Nuclear Regulatory Commission,1988．
[15]
吕峰,陈明亚,吴洪,等．基于RCCM规范的压力-温度限值曲线计算与讨论[J]．核动力工程,2012,33(6)：1-5． Lü Feng,Chen Mingya,Wu Hong,et al．Calculation and discussion of pressure-temperature limit curves based on RCCM code[J]．Nuclear Power Engineering,2012,33(6)：1-5(in Chinese)．
[16]
Qian G,Niffenegger M．Procedures, methods and computer codes for the probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks [J]．Nuclear Engineering and Design,2013,258：35-50．
[17]
U.S. Nuclear Regulatory Commission．Format and content of plant-specific pressurized thermal shock safety analysis reports for pressurized water reactors[S]．Washington D C：Nuclear Regulatory Commission,1987．
[18]
Chen M Y,Lu F,Wang R S,et al．The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading [J]．Nuclear Engineering and Design,2015(288)：130-140．
[19]
He Y B,Isozaki T．Fracture mechanics analysis and evaluation for the RPV of the Chinese Qinshan 300 MW NPP under PTS[J]．Nuclear Engineering and Design,2000,201(2-3)：121-137．
[20]
薛飞,余伟炜,蒙新明,等．国产压水堆核电站机组主管道疲劳裂纹扩展特性实验研究[J]．中国电机工程学报,2014,34(8)：1310-1317． Xue Fei,Yu Weiwei,Meng Xinming,et al．Experiment study on fatigue crack propagation behavior of primary pipe in China pressurized-water reactor[J]．Proceedings of the CSEE,2014,34(8)：1310-1317(in Chinese)．
