三维复杂力学条件下核电关键构件环境致裂预测方法研究
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摘要
为了提高核电设备的耐腐蚀性，在轻水核反应堆中除核燃料包壳等少量材料外，结构材料大量选用奥氏体不锈钢和镍基合金，其耐腐蚀性主要是由钢表面的富络氧化膜(钝化膜)的形成。然而，奥氏体不锈钢和镍基合金长期处于核压力容器髙温髙压及辐照等环境中，在一定的残余应力和工作应力作用下将会产生应力腐蚀裂纹(SCC)，这种应力腐蚀裂纹随时间不断扩展，最终会将核电站中关键结构和设备带到一个极其危险边缘。因此，奥氏体不锈钢和镍基合金材料在高温水环境中以SCC为代表的环境致裂(EAC)是影响核电设备长期安全运行的关键问题之一。
鉴于这个问题的重要性，长期以来，主要工业发达国家相关研究机构在基于标准断裂力学试样和模拟高温水实验环境下，完成了大量核电关键材料的EAC裂纹扩展实验，储备了丰富的实验数据。目前，许多在役核电结构被检出EAC问题，急需评估和预测。由于实际核电结构裂纹区域几何和力学状态的复杂性，仅利用相关材料的EAC实验数据来评估核电结构中的EAC扩展状况显然是不够的。
为了解决材料EAC扩展速率实验数据和实际核电构件EAC扩展评价的衔接问题，本论文完成的主要工作有：在对核反应堆高温水环境中奥氏体不锈钢和镍基合金环境致裂的机理研究的基础上，对于各国科学家相继提出的不同应力腐蚀破裂裂纹扩展速率预测模型进行了分析比较，重点研究了氧化滑移模型的机理、适用条件和存在的问题；在对奥氏体不锈钢和镍基合金材料在高温水环境中EAC机理探讨的基础上，确定了影响核电结构EAC扩展主要因素，为建立实际轻水堆构件EAC扩展状况的预测模型奠定了基础；在对含有内表面裂纹管试样的EAC实验结果分析、数值模拟和理论分析的基础上，建立了一种基于裂尖塑性应变率和弹塑性有限元相结合的，能用于预测具有复杂裂纹形状和力学状态的实际轻水堆构件EAC扩展状况的预测模型，并将其与实验结果进行了比较验证；通过理论分析和电场有限元模拟，建立了直流电位降法(DCPD)实时测量裂纹长度(深度)的理论基础，为进一步研制直流电位降裂纹测深仪和重要参数标定提供了保证；利用有限元子模型技术对含有内表面裂纹管试样裂纹前沿区域主要应力应变参量和断裂力学分布规律进行了详细研究，为建立具有复杂几何形状和力学状态的实际轻水堆构件裂纹区域的主要力学参量的获取奠定了基础；利用二维CDCB试样模型和二维含有内表面裂纹管试样模型，对一次超载对EAC裂纹扩展速率的影响进行了比较详细的分析，为掌握在EAC实验过程中的载荷波动及地震及其他超载情况对构件EAC扩展状况的影响奠定了基础。
In order to improve the corrosion resistance of nuclear power equipment，austeniticstainless steel and nickel alloy are mainly selected as structural materials, whose corrosionresistance is formed due to the steel surface of the chromium-rich oxide film (passivationfilm). However, austenitic stainless steel and nickel alloy long-term serve in high temperatureand high pressure and irradiation environments of nuclear pressure vessel, and the stresscorrosion cracking (SCC) would be induced by the residual stress and work loading, whichmight birng the nuclear power equipment to an extremely dangerous edge as SCC crackgrowing. Therefore, stress corrosion cracking (SCC), as a representative of environmentallyassisted cracking (EAC) in austenitic stainless steel and nickel-based alloy is one of the keyissues in long term safe service of equipment and structure in nuclear power plants.
Because of the importance of this issue, many research institutes of the majorindustiralized countires performed a large number of EAC crack growth experiments of keynuclear materials based on standard rfacture mechanics specimens and simulatedhigh-temperature water environments, and obtained a lot of experimental data. Currently, theEAC was detected in many in-service nuclear structures, which urgently need to assess andpredict. Due to the complexity of the crack geometry and mechanical state in the actualnuclear structure, it is obviously not enough to evaluate EAC expanding situation in thenuclear structures only by using the EAC experimental data of related material.
Li order to improve the interface between EAC growth rate data of the mateiralsobtained in laboratory and EAC growth of actual components in nuclear power plants, theresearches done in this dissertation are as follows. The EAC mechanism of austenitic stainlesssteel and nickel-based alloy in high temperature water environment of nuclear pressure vesseland piping was studied, and current prediction models of EAC growth rate proposed by scientists were analyzed and compared, and the oxidation mechanism and applicablecondition of sliding model was specially investigated. Based on discussions of EACmechanism of austenitic stainless steel and nickel based alloys in high temperature waterenvironments, the main factors affecting EAC growth of the nuclear structure wereinvestigated to establish a foundation for evaluating EAC growth situation of the actualcomponents in light water reactors. Based on experimental result analysis，numeircalsimulation and theoretical analysis of an EAC expeirment by using a tube-shape specimenwith inner surface cracks，an EAC prediction model was established incorporated by thecrack tip plastic strain rate and elastic-plastic finite element combining, which can be used topredict EAC growth statue of the actual component with complex crack shape and themechanical state in light-water reactors, and it was verified by expeirment. Through atheoretical analysis and ifnite element simulation on related electric ifeld, an in-situmeasurement approach of crack growing length (depth) by current potential drop (DCPD)was investigated, which provides a theoreitcal basis of developing new crack monitor andcalibrating parameters in DCPD. It was detail investigated that the distirbution of the stressand strain and the rfacture mechanics parameters along the crack rfont of a tube~shapespecimen with inner surface cracks, which establishes a foundation for obtaining the mainmechanical parameters nearby crack region of the actual components with complex geometryand mechanical state in light water reactors. By using a two dimension CDCB specimen anda three dimension tube-shape specimen with inner surface cracks, the effect of a single tensileoverload on EAC crack growth rate was investigated in detail，which establishes a foundationto understand the effect of load diversification on EAC experiments, and the effect of seismicand other overloading on EAC growth status.
鉴于这个问题的重要性，长期以来，主要工业发达国家相关研究机构在基于标准断裂力学试样和模拟高温水实验环境下，完成了大量核电关键材料的EAC裂纹扩展实验，储备了丰富的实验数据。目前，许多在役核电结构被检出EAC问题，急需评估和预测。由于实际核电结构裂纹区域几何和力学状态的复杂性，仅利用相关材料的EAC实验数据来评估核电结构中的EAC扩展状况显然是不够的。
为了解决材料EAC扩展速率实验数据和实际核电构件EAC扩展评价的衔接问题，本论文完成的主要工作有：在对核反应堆高温水环境中奥氏体不锈钢和镍基合金环境致裂的机理研究的基础上，对于各国科学家相继提出的不同应力腐蚀破裂裂纹扩展速率预测模型进行了分析比较，重点研究了氧化滑移模型的机理、适用条件和存在的问题；在对奥氏体不锈钢和镍基合金材料在高温水环境中EAC机理探讨的基础上，确定了影响核电结构EAC扩展主要因素，为建立实际轻水堆构件EAC扩展状况的预测模型奠定了基础；在对含有内表面裂纹管试样的EAC实验结果分析、数值模拟和理论分析的基础上，建立了一种基于裂尖塑性应变率和弹塑性有限元相结合的，能用于预测具有复杂裂纹形状和力学状态的实际轻水堆构件EAC扩展状况的预测模型，并将其与实验结果进行了比较验证；通过理论分析和电场有限元模拟，建立了直流电位降法(DCPD)实时测量裂纹长度(深度)的理论基础，为进一步研制直流电位降裂纹测深仪和重要参数标定提供了保证；利用有限元子模型技术对含有内表面裂纹管试样裂纹前沿区域主要应力应变参量和断裂力学分布规律进行了详细研究，为建立具有复杂几何形状和力学状态的实际轻水堆构件裂纹区域的主要力学参量的获取奠定了基础；利用二维CDCB试样模型和二维含有内表面裂纹管试样模型，对一次超载对EAC裂纹扩展速率的影响进行了比较详细的分析，为掌握在EAC实验过程中的载荷波动及地震及其他超载情况对构件EAC扩展状况的影响奠定了基础。
In order to improve the corrosion resistance of nuclear power equipment，austeniticstainless steel and nickel alloy are mainly selected as structural materials, whose corrosionresistance is formed due to the steel surface of the chromium-rich oxide film (passivationfilm). However, austenitic stainless steel and nickel alloy long-term serve in high temperatureand high pressure and irradiation environments of nuclear pressure vessel, and the stresscorrosion cracking (SCC) would be induced by the residual stress and work loading, whichmight birng the nuclear power equipment to an extremely dangerous edge as SCC crackgrowing. Therefore, stress corrosion cracking (SCC), as a representative of environmentallyassisted cracking (EAC) in austenitic stainless steel and nickel-based alloy is one of the keyissues in long term safe service of equipment and structure in nuclear power plants.
Because of the importance of this issue, many research institutes of the majorindustiralized countires performed a large number of EAC crack growth experiments of keynuclear materials based on standard rfacture mechanics specimens and simulatedhigh-temperature water environments, and obtained a lot of experimental data. Currently, theEAC was detected in many in-service nuclear structures, which urgently need to assess andpredict. Due to the complexity of the crack geometry and mechanical state in the actualnuclear structure, it is obviously not enough to evaluate EAC expanding situation in thenuclear structures only by using the EAC experimental data of related material.
Li order to improve the interface between EAC growth rate data of the mateiralsobtained in laboratory and EAC growth of actual components in nuclear power plants, theresearches done in this dissertation are as follows. The EAC mechanism of austenitic stainlesssteel and nickel-based alloy in high temperature water environment of nuclear pressure vesseland piping was studied, and current prediction models of EAC growth rate proposed by scientists were analyzed and compared, and the oxidation mechanism and applicablecondition of sliding model was specially investigated. Based on discussions of EACmechanism of austenitic stainless steel and nickel based alloys in high temperature waterenvironments, the main factors affecting EAC growth of the nuclear structure wereinvestigated to establish a foundation for evaluating EAC growth situation of the actualcomponents in light water reactors. Based on experimental result analysis，numeircalsimulation and theoretical analysis of an EAC expeirment by using a tube-shape specimenwith inner surface cracks，an EAC prediction model was established incorporated by thecrack tip plastic strain rate and elastic-plastic finite element combining, which can be used topredict EAC growth statue of the actual component with complex crack shape and themechanical state in light-water reactors, and it was verified by expeirment. Through atheoretical analysis and ifnite element simulation on related electric ifeld, an in-situmeasurement approach of crack growing length (depth) by current potential drop (DCPD)was investigated, which provides a theoreitcal basis of developing new crack monitor andcalibrating parameters in DCPD. It was detail investigated that the distirbution of the stressand strain and the rfacture mechanics parameters along the crack rfont of a tube~shapespecimen with inner surface cracks, which establishes a foundation for obtaining the mainmechanical parameters nearby crack region of the actual components with complex geometryand mechanical state in light water reactors. By using a two dimension CDCB specimen anda three dimension tube-shape specimen with inner surface cracks, the effect of a single tensileoverload on EAC crack growth rate was investigated in detail，which establishes a foundationto understand the effect of load diversification on EAC experiments, and the effect of seismicand other overloading on EAC growth status.
引文
[1]能源科学学科发展战略研究组.2011-2020年我国能源科学学科发展战略报告(第4稿).国家自然科学基金委员会，中国科学院，2010
[2]史永I核能发电的优点及世界核电发展动向[J].能源工程,2007, (1):1-6
[3] R.Pirce, J.Blaise. Nuclear fuel resources enough to last [J]. NEA NEWS,2002,20(2):10-13
[4]连培生.原子能工业[M].北京：原子能出版社，2002
[5]孙勤.第三代核电技术落在中国是历史的必然[N].科技日报，2009-03-20
[6]赵永新，张意轩.我国第三代核电产业体系基本成型[N].人民日报,2010七8-10
[7]李宗明.核电发展概述[J].求是,2011,(8): 54-55
[8]辛灵，刘艳.日核危机升级中国暂停审批核项目核电发展决心不改变，速度或将放缓[N].南方都市报，2011-03-18
[9]魏仁杰.核电与核电安全-三里岛和切尔诺贝利核电站事故的研究[J].核动力工程，1987,8(4):289-294
[10]武俊德.中国核电面临艰难抉择[J].节能与环保,2011,202(4): 34-36
[11]束国刚，陆念文.压水堆核电厂关键金属部件的老化和寿命评估[J].中国电力,2006,39(5):53-58
[12] F.A. Simonen，S.R. Gosselin. Life Prediction and monitoring of nuclear power plantcomponents for service-related degradation [J], Journal of Pressure Vessel Technology,2001，123(1):58-64
[13]毕岩，谢里阳.核电关键构件材料的寿命预测方法[J].失效分析与预防，2011，6(1):12-14
[14] R.G. Ballinger. Environmental degradation in lwr systems: current and emerging issues
[C]， Proceedings of5th IMR Symposium on Materials Science andEngineeirng-Mateirals and Reliability in Nuclear Power Plant,2009
[15]苏州热工研究院.与AECL的腐蚀技术交流一我国核电站面临的腐蚀问题[DB/OL],2005，http://wenku.baidu.com/view/f9d80ae69b89680203d82547.html
[16]柯伟，韩恩厚，王俭秋.核电镍基合金材料的腐蚀失效和预防[C].经济发展方式转变与自主创新-第十二届中国科学技术协会年会(第二卷)，2010
[17]刘建章.核结构材料[M],北京：化学工业出版社，2007
[18]李光福，杨武.核电站异材焊接件的破裂问题与应力腐蚀评价方法[J].核安全,2003,(2):37-40
[19]李继红，张建勋.断裂应变对含损伤焊接接头裂纹扩展驱动力影响的数值[J].得接学报,2004,25(1):52-56
[20]W.H. Hearns. Welding Handbook(Vol.4): Metals and Their Weldability [M], AmericanWelding Society,1982
[21]胡舜媛译.核电站部件受应力裂纹腐蚀[J].国外核新闻，1994，（10):27
[22]蒋浦宁.核电汽轮机防应力腐烛裂纹技术[J].热力透平，2010,39(2):89-92
[23]杨武.核电工程材料的应力腐蚀破裂研究[J].腐蚀科学与防护技术，1995，7(2):87-92
[24]P. Scott, M.L. Calvar. Some possible mechanism of intergranular stress, corrosioncracking of alloy600in PWR pirmary water [C]. Proceedings of Sixth InternationalConference of Environmental Degradation of Mateirals Nuclear Power Systems-WaterReactors, TMS,1993:657-665
[25]吕战鹏，杨武.遭受应力腐蚀开裂的设备寿命预测技术[J].腐烛科学与防护技术，1999,11(1):57-64
[26]D. Hamaguchi, Y. Dai. Micro-structural study of EC316LN and its welds irradiated inSINQ target-3[J], Journal of Nuclear Materials,2005，(343):262-266
[27]J.Yu, D.S. Gelles, F.A. Ganrer. The performance of Chinese316L and316Ti stainlesssteel irradiated at300,400,500and600°C in HFIR JP-23test capsule [J]. Journal ofNuclear Mateirals,2002,307-311:357-361
[28]A. Pardo, M.C. Merino, A.E. Coy. Influence of Ti，C and N concentration on theintergranular corrosion behaviour of AISI3i6Ti and321stainless steels [J]. Acta Mater,2007，55:2239-2251
[29]卢建树，王保峰，张九渊.高温水中不锈钢和镍基合金应力腐蚀破裂研究进展[J].核动力工程，2001,22(3):259-263
[30]环境断裂教育部重点实验室[EB/OL].http://www.ustb.edu.cn/esf/index.htm
[31]褚武扬，谷飚.应力腐蚀机理研究的新进展[J].腐蚀科学与防护技术,1995，7(2):97-101
[32]陆永浩，褚武扬，髙克玮等.T.Shoji.304L不锈钢在高温水中的应力腐蚀裂纹扩展[J],金属学报,2004,40(7):763-767
[33]李红梅.奥氏体不锈钢模拟辐照促进应力腐蚀破裂研究[D].博士论文，上海：上海交通大学,2005
[34]高欣，吴欣强，关辉等.高温高压水环境中腐蚀产物膜的研究现状[J].腐蚀科学与防护技术，2007,27(2):110-113
[35]徐松，吴欣强，韩恩厚等.316Ti不锈钢在模拟核电高温高压水中的腐蚀疲劳裂纹断口研究[J].中国腐蚀与防护学报，2010，30(2):119-123
[36]国家发展和改革委员会.重大工程材料服役安全研究评价设施国家重大科技基础设施建设项目批准立项[EB/OL].2007.02.16, http://www.sdpc.gov.cn/gjscy/cxtx/t20070216l17764_.htm
[37]科技部.国家重点基础研究发展计划经费预算拟安排情况汇总表[EB/OL].http://www.most.gov.cn/tztg/200611/P020061122376424687228.xls
[38]J. Hou, T. Shoji, Z.P. Lu et al. Residual strain measurement and grain boundarycharacteirzation in the heat-affected zone of a weld joint between Alloy690TT andAlloy52[J], Journal of Nuclear Materials,2010，397(1-3):109-115
[39]韩恩厚，王俭秋，吴欣强等.核电髙温高压水中不锈钢和镍基合金的腐蚀机制[J].金属学报,2010,46(11):1379-1390
[40]P.L. Andresen, K. Gott, J.L. Nelson. Stress corrosion cracking of sensitized type304stainless steel in288°C water: a ifve laboratory round bobbin [C]. Proceedings of the9thInternational Symposium on Environmental Degradation of Material in Nuclear PowerSystems-Water Reactors, S. Baremmer, et al eds., Newport Beach, California,1999.423-433
[41]沈长斌，陶晓杰，杨怀玉等.髙温高压水环境下传热管失效形式及防腐措施研究进展[J].腐蚀科学与防护技术，2003，15(4):223-227
[42]D.D. Macdonald, P.C. Lu. Theoretical estimation of crack growth rates in type304stainless steel in BWR coolant environments [J]. Corrosion，1996，52:768-785
[43]T. Kondo，H. Nakajina, R. Nagasak. Metallographic investigation on the claddingfailure in the pressure vessel of a BWR [J], Nuclear Engineeirng and Design,1971，16(8):205
[44]Y. Sato, H. Xue, Y. Takeda et al. Development of a stress corrosion crack testmethodology using tube-shaped specimen [J]. ASTM International-Journal of Testingand Evaluation,2007，35(3):254-258
[45]K. Chopra, H.M. Chung. Current research on environmentally assisted cracking in lightwater reactor environments [J]. Nuclear Engineering and Design,1999，194:205-223
[46]Q.J. Peng，J. Kwon, T. Shoji. Development of a fundamental crack tip strain rateequation and its application to quantitative prediction of stress corrosion cracking ofstainless steels in high temperature oxygenated water [J]. Journal of Nuclear Mateirals,2004，324(1):52-61
[47]李金许，高克玮，宿彦京等.钝化膜应力导致不锈钢应力腐蚀的研究[lf.中国科学(E辑)，2003，33(9):796-803
[48]O.K. Chopra, H.M. Chung, T.F. Kassner et al. Current research on environmentallyassisted cracking in light water reactor environments [J], Nuclear Engineering andDesign,1999，194:205-223
[49]P.L. Andresen, F.P. Ford. Fundamental modeling of environmental cracking forimproved design and lifetime evaluation in BWRs [J]. International Journal of PressureVessels and Piping，1994，59:61-70
[50]M. Hall. Film rupture model for aqueous stress corrosion cracking under constant andvairable stress intensity factor [J]. Corrosion Science,2009,51:225-233
[51]F.P. Ford. Mechanisms of environmentally-assisted cracking [J]. International Journalof Pressure Vessels and Piping,1989，40(55):343-362
[52]P.L. Andresen, F.P. Ford. Life prediction by mechanistic modeling and systemmonitoirng of environmental cracking of iron and nickel alloys in aqueous systems [J].Mateirals Science and Engineering,1988, A103:167-184
[53]Y.C. Gao, K.C. Hwang. Elastic-plastic fields in steady crack growth in a strainhardening material [C]. Proceedings of Fitfh Intenrational Conference on Fracture,1981，669-682
[54]T. Shoji，S. Suzuki, R.G. Ballinger. Theoretical prediction of SCC growthbehavior-threshold and plateau growth rate [C]. Proceedings of the Seventh InternationalSymposium on Environmental Degradation of Materials in Nuclear Power Systems,Breckinridge,1995,881-889
[55]T. Hashimoto，M. Koshishi. Modiifcation of the FRI crack growth model formulationrfom a mathematical viewpoint [J]. Journal of Nuclear Science and Technology,2009,46:295-302
[56]M. Vankeerberghen, G. Weyns, S. Gavrilov et al. Crack propagation rate modelling for316SS exposed to PWR-relevant conditions [J]. Journal of Nuclear Mateirals，2009，384(3):274-285
[57]M.M. Hall. An alternative to the Shoji crack tip strain rate equation [J]. CorrosionScience.2008，50:2902-2905
[58]T. Shoji, Z.P. Lu，H. Xue et al. Quantification of the effects of crack tip plasticity onenvironmentally-assisted crack growth rates in LWR environments [M]. In:Environment-Induced Cracking of Mateirals，2008:107-122
[59]Q.J. Peng，J. Kwon, T. Shoji. Development of a fundamental crack tip strain rateequation and its application to quantitative prediction of stress corrosion cracking ofstainless steels in high temperature oxygenated water [J], Journal of Nuclear Materials,2004，324(1):52-61
[60]H_Xue, Y. Sato，T.Shoji. Quantitative estimation of the growth of environmentallyassisted cracks at flaws in light water reactor components [J]. Transactions of theASME-Journal of Pressure Vessel and Technology,2009，131(1):61-70
[61]Y.C. Gao. Exponential stress and strain singularity at tip of mode I growing crack inpower hardening plastic material [J]. Theoretical and Applied Fracture Mechanics，1996,25(2):103-111
[62]T. Shoji, Z.P. Lu，H. Murakami. Formulating stress corrosion cracking growth rates bycombination of crack tip mechanics and crack tip oxidation kinetics [J]. CorrosionScience,2010，52(3):769-779
[63]H. Xue, T. Shoji. Quantitative prediction of EAC crack growth rate of sensitized type304stainless steel in boiling water reactor environments based on EPFEM [J],Transactions of the ASME-Journal of Pressure Vessel and Technology,2007，129(3):254-258
[64]左景伊.应力腐烛破裂[M].西安：西安交通大学出版社，1985
[65]李濒.断裂力学[M].济南：山东科技出版社，1979
[66]冒小萍，郎福元，柯显信.断裂力学的数值计算方法的研究现状与展望[J].商丘师范学院学报，2004，20(2):20-24
[67]周爱细，黄建龙，郎福元.计算断裂力学进展[J].甘肃工业大学学报，1998,24(1):111-116
[68]张国瑞.有限元法[M].北京：机械工业出版社，1992
[69]T.L. Andersen. Fracture mechanics: tfindamentals and applications, second edition[M]. Boca Raton，USA: CRC Press,1995
[70]P. Aaltonen，M. Bojiov, M. Helin et al. Facts and views on the role of anionic impuirties,crack tip chemistry and oxide films in environmentally assisted cracking [M]. Espoo，Finland: VTT industiral Systems，2002
[71]P. Siitonen. Passivation and the forms of corrosion [M]. In: Corrosion handbook,Hanko，Finland: Suomen Korroosioyhdistys SKY，1988
[72]Metals handbook [M]. Desk Edition (Second Edition). Mateirals Park, USA: ASMInternational,1998
[73]A. Toivonen. Stress corrosion crack growth rate measurement in high temperaturewater using small precracked bend specimens [D]. Ph.D. Thesis，Helsinki University ofTechnology,2004
[74]A.J. Sedirks. Corrosion testing made easy: stress corrosion cracking test methods [M].Houston，TX，USA: NACE，1990
[75]T.L. Andersen. Fracture mechanics: fundamentals and applications, second edition [M].Boca Raton, USA: CRC Press,1995
[76]R.A. Mayville，T.J. Warren, P.D. Hilton. The crack Velocity-Ki，relationship for AISI4340in seawater under fixed and irsing displacement [M]. In: Fracture Mechanics:Perspectives and Directions (wentieth Symposium). Philadelphia, USA: AmericanSociety for Testing and Materials,1989，605-614
[77]R.B. Rebak, Z.S. Smialowska. The mechanism of Stress corrosion cracking of alloy600in high temperature water [J], Corrosion Science，1996,38:971-988
[78]P.L. Andresen. Effects of temperature on crack growth rate in sensitized type304stainless steel and alloy600[J], Corrosion,1993，49:714-725
[79]T.M. Angeliu, P.L.Andresen and F.P.Ford. Applying slip-oxidation to the SCC ofaustenitic mateirals in BWR/PWR environments [C]. Proceedings of Corrosion’98conference, NACE International,1998,262
[80]P.L. Andresen. The effects of aqueous impuirties on intergranular stress corrosioncracking of sensitized type-304stainless steel [M]. In:Final report No. EPRI NP-3384,Palo Alto, CA: Electric Power Research Institute,1983
[81]P.L. Andresen. SCC growth rate behaviour in BWR water of increased purity. In:Proceedings of the eight international symposium on environmental degradation ofmateirals in nuclear power systems-water reactors [M]. Floirda, Illinois，USA: AmeircanNuclear Society,1997，675-684
[82]P.L. Andresen. A mechanism for the effects of ionic impuirties on SCC of austeniticiron and nickel base alloys in high temperature water [C]. Proceedings of Corrosion’85conference, NACE International,1985，101-120
[83]P.L. Andresen. Effect of dissolved oxygen, solution conductivity and stress intensity onthe interdendirtic stress corrosion cracking of Inconel182weld metal [C]. Proceedingsof Corrosion!87conference, NACE International,1987,84-98
[84]P.L. Andresen. The effects of nitrate on the stress corrosion cracking of sensitizedstainless steel in high temperature water [M], Proceedings of the seventh internationalsymposium on environmental degradation of mateirals in nuclear power systems-waterreactors. Breckenirdge, August7-10,1995.
[85]P.L Andresen, F.P. Ford. Response to "On the modeling of stress corrosion cracking ofiron and nickel base alloys in high temperature environments”[J]. Corrosion Science,1996，38:1011-1016
[86]P.L. Andresen. Chemistry and Observation and prediction of the effects of watermechanics on environmentally assisted Inconels182weld metal and600[J]. Corrosion，1988，44:376-385
[87]F.P. Ford，P.L. Andresen. Development and use of a predictive model of crackpropagation in304/316L，A533B/A508and Incone)600/182allo°ys in288C water [C].Proceedings of the third international symposium on environmental degradation ofmateirals in nuclear power systems-water reactors，1988，789-800
[88]F.P. Ford, D.F. Taylor，P.L. Andresen, R.G.. Ballinger. Corrosion-assisted cracking ofstainless and low-alloy steels in LWR environments: final report. Palo Alto, CA: ElectricPower Research Institute,1987.Report No. EPRINP-5064M
[89]P.L. Andresen, T.M. Angeliu. Evaluation of the role of hydrogen in SCC in hot water
[C]. In: Proceedings of the Corrosion'97conference, NACE International,1997,195-210
[90]F.P. Ford. Overview of collaborative research into the mechanisms of environmentallycontrolled cracking in the low alloy pressure vessel steel/water system [C]. In: Cullen，W.H. Proceedings of the second international atomic energy agency specialists'meeting an subcritical crack growth. Sendai,1986.3-72
[91]F.P. Ford. Quantitative prediction of environmentally assisted cracking [J]. Corrosion,1996，52:375-395
[92]P.L. Andresen. SCC mitigation strategies in hot water [C]. In: Proceedings ofCorrosion!97conference, NACE International,1997.102
[93]R.M. Horn, G..M. Gordon, F.P Ford et al. Experience and assessment of stresscorrosion cracking in L-grade stainless steel BWR internals [J]. Nuclear Engineeirng andDesign,1997,174:313-325
[94]P.L. Andresen. Probabilistic interpretation of SCC initiation and growth [C]. In:Proceedings of Corrosion’99conference, NACE International,1999,446457
[95]P.L. Andresen, F.P. Ford, T.M. Angeliu et al. Prediction of environmentally assistedcracking and its relevance to life management in SWRs [C]. In: Baremmer，S” Ford，P.and Was, G. Proceedings of the ninth international symposium on environmentaldegradation of materials in nuclear systems-water reactors. Newport Beach, California,Warrendale. USA: The Minerals，Metals and Mateirals Society,1999,373-381
[96]P.L. Andresen, T.M. Angeliu, P.W. Emigh et al. Crack growth rate behaviour of Nialloys [C]. In: Proceedings of Corrosion12000conference, NACE International,2000,202-214
[97]P.L. Andresen, F.P. Ford. Fundamental quantification of crack advance for lifeprediction in energy systems [C]. In: Proceedings of Corrosion!96conference, MACEInternational,1996,51-99
[98]P.L. Andresen, T.M. Angeliu，L.M. Young et al. Mechanisms and kinetics of SCC instainless steels [C]. In: Proceedings of the tenth international symposium onenvironmental degradation of mateirals in nuclear systems-water reactors. Lake Tahoe,NY, NACE International, The Minerals, Metals and Materials Society, AmericanNuclear Society (CDROM)，2001
[99]P.L. Andresen，P.W. Emigh, M.M. Morra et al. Yield strength, corrosion potential, stressintensity factor，Effects of silicon and grain boundary character on the SCC of stainlesssteels [C], In: Proceedings of the eleventh intenrational conference on environmentaldegradation of mateirals in nuclear power systems-water reactors, Stevenson,Washington，USA (CDROM)，2003
[100]D.A. Yermilyea. A theory for the propagation of stress corrosion cracks in metals [J].J. Electrochem. Soc，1972,119:405-407
[101]J. Congleton，W. Yang. The effect of applied potential on the stress corrosioncracking of sensitized type316stainless steel in high temperature water[J]. Corrosion Science，1995，37:429-144
[102]T. Shoji, K.S. Raja, G.F. Li et al. Critical parameters of environmentally assitedcracking in nuclear systems [C]. In: Proceedings of Corrosion’2000conference, NACEInternational,2000.190-202
[103]N.S. Mclntyre, C.M. Huctwith, R.D. Davidson et al. M. ID cracking in UNS N0b600nuclear steam generator tubes: evidence for SCC from secondary ion mass spectrometricimages [J]. Corrosion Science,1998，40:1799-1802
[104]A. Tumbull. Modelling of crack chemistry in sensitized stainless steel in boiling waterreactor environments [J]. Corrosion Science，1997，39:789-805
[105]D.F. Taylor. Crevice corrosion of Alloy600in high-temperature aqueousenvironments [J], Corrosion,1979，35:550-559
[106]J. Congleton，R.N. Parkins. Stress corrosion cracking of steel in high temperaturewater [J]. Corrosion,1988,44:290-298
[107]W.E. Ruther，W.K. Soppet, T.F. Kassner. Effect of temperature and ionic impurities atvery low concentrations on stress corrosion cracking of type304stainless steel [J].Corrosion,1988，44:791-799
[108]D.D. Macdonald. Clairfication of issues raised by P.L andresen and F.P. Ford in theirresponse to "On the modeling of stress corrosion cracking of iron and nickel base alloysin high temperature environments"[J]. Corrosion Science,1997,39:1487-1490
[109]A. Tumbull, M. Psaila-Dombrowski. A review of electrochemistry of relevance toenvironment-assisted cracking in light water reactors [J]. Corrosion Science，1932，32:1925-1966
[110]A. Turnbull. Modelling of environment assisted cracking [J]. Corrosion Science,1993,34:921-960
[111]D.D. Macdonald，M. Urquidi-Macdonald. A coupled environment model for stresscorrosion cracking in sensitized type304stainless steel in LWR environments [J].Corrosion Science，1991，32:51-81
[112]D.D. Macdonald，P.C. Lu, M. Urquidi-Macdonaid et al. Theoretical estimation ofcrack growth rates in type304stainless steel in BWR coolant environments [J].Corrosion,1996，52:768-785
[113]D.D. Macdonald，I_Balachov. Modeling the accumulation and mitigation of SCCdamage in BWRs. In: Bruemmer, S_，Ford，P and Was，G. Proceedings of the ninthinternational symposium on environmental degradation of mateirals in nuclearsystems-water reactors [C]. Newport Beach, California, Warrendale, USA: The Minerals,Metals and Materials Society.1999，399-407
[114]G.R. Engelhardt, D.D. Macdonald. Modelling of corrosion fatigue chemistry insensitized steel in boiling water reactor environments [C]. In: Proceedings ofCorrosion!2000conference, MACE International,2000.227-241
[115]M.P. Manahan, D.D. Macdonaid，A.J. Peterson. Determination of the fate of thecurrent in the stress corrosion cracking of sensitized type304SS in high temperatureaqueous systems [J], Corrosion Science，1995，37:189-208
[116]M. Vankeerberghen, D. Macdonald. Calculating the temperature-maximum and thelower potential limit for the crack growth rate in type304SS using the CEFM [C]. In:Proceedings of Corrosion’2003conference, NACE International,2003,3520-3530
[117]G.R. Engelhardt，D.D. Macdonald，M. Urduidi-Macdonald. Development of fastalgorithms for estimating stress corrosion crack growth rate [J], Corrosion Science,1999,41:2267-2302
[118]P.M. Scott, M. LeCalvar. On the role of oxygen in stress corrosion cracking as afunction of temperature [C], In: Magnin, T. Proceedings of the second intenrationalconference on corrosion-deformation interactions in conjuction with EUROCORR!96.Nice, France, London: The Institute of Mateirals,1996,384-393
[119]P.M. Scott, M. LeCalvar. Some possible mechanisms of intergranular stress corrosioncracking of alloy600in PWR pirmary water [C]. In: Gold，R.E and Simonen，E.P.Proceedings of the sixth intenrational symposium on environmental degradation ofmateirals in nuclear power systems-water reactors. San Diego, CA，Warrendale, PA,USA: The Minerals, Metals and Mateirals Society.1993,657-667
[120]P.M. Scott. An overview of internal oxidation as a possible explanation ofintergranular stress corrosion cracking of alloy600in PWRs. In: Bruemmer, S?，Ford, Pand Was，G. Proceedings of the ninth international symposium on environmentaldegradation of materials in nuclear systems-water reactors [C]. Newport Beach，California, Warrendale, PA, USA: The Minerals, Metals and Materials Society.1999,3-11
[121]P. Combrade, P.M. Scott. The role of surface films on alloy600corrosion inpressuirsed water reactors [C]. In: Magnin, T. Proceedings of the secondinternational conference on corrosion-deformation interactions in conyuction withEUROCORR'96. Nice，France, London，UK: The Institute of Mateirals,1996,394-409
[122]R. Staehle，Z. Fang. Comments on a proposed mechanism of internal oxidation foralloy600as applied to low potential SCC [C]. In: Baremmer, S.，Ford, P and Was, G.Proceedings of the ninth international symposium on environmental degradation ofmateirals in nuclear systems-water reactors，Newport Beach, Califonria，Warrendale, PA,USA: The Minerals, Metals and Materials Society,1999,69-77
[123]J.R. Galvele. A stress corrosion crack mobility [J]. Corrosion Science,1987，27mechanism based on surface1-33
[124]J.R. Galvele. Recent developments in the surface-mobility stress corrosion crackingmechanism [J], Electrochim. Acta,2000，45:3537-3541
[125]J.R. Galvele. Comments on”Notes on the surface mobility mechanism of stress-corrosion cracking"[J]. By K. Sieradzki and F.J. Firedersdorf. Corrosion Science,1994,36(5):901-910
[126]K. Sieradzki, F.J. Firedersdorf. Notes on the surface mobility mechanism ofstress-corrosion cracking [J]. Corrosion Science,1994，36:669-675
[127]P. Aaltonen, T. Saairo, P. Karjalainen-Roikonen et al. Vacancy-creep model for EACof metallic materials in high temperature water [C]. In: Proceedings of Corrosion!96conference, NACE International,1996.81-95
[128]L. Balachov，M. Bojinov, T. Laitinen et al. Correlations between the stability of oxidefil2ms in simulated BWR crack conditions，S04" enrichment and cracking susceptibility
[C]. In: Proceedings of water chemistry of nuclear reactor systems8. Bournemouth，London，UK: Birtish Nuclear Energy Society,2000，170-172
[129]M. Bojinov, G. Fabircius，T. Laitinen et al. Coupling between ionic defect structureand electronic conduction in passive films on iron, chromium and iron-chromium alloys[J]. Electrochim Acta,2000，45:2029-2048
[130]M. Bojinov, G. Fabricius，T. Laitinen et al. Influence of molybdenum on theconduction mechanism in passive films on iron-chromium alloys in sulphuric acidsolutions Electrochim [J]. Acta，2001，46:1339-1358
[131]B_Beverskog，M. Bojinov, P. Kinnunen et al. A mixed-conduction model for oxidefilms on Fe，Cr and Fe-Cr alloys in high-temperature aqueous electrolytes-II. Adaptationand justification of the model [J]. Corrosion Science，2002，44:1923-1940
[132]Y. Jagodzinski，P. Aaltonen，S. Smuk et al. Internal friction study of environmentaleffects on metals and alloys [J]. J. Alloys and Compounds,2000,310:256-260
[133]P. Aaltonen，Y_Jagodzinski, O. Tarasenko et al. Low-frequency internal friction ofpure copper atfer anodic polairzation in sodium nitrite solution [J]. Corrosion Science，1998，40:903-908
[134]P. Aaltonen, Y. Jagodzinski, O. Tarasenko et al. Intenral rfiction in brasses atferoxidation in tap water by anodic polairzation [J], Acta Mater,1998，46:2039-2046
[135]Q. Peng，J. Kwon，T_Shoji. Inverse analysis of crack growth kinetics of sensitized304L stainless steel in a simulated boiling water reactor environment for quantitativeprediction of stress corrosion cracking behavior [C]. In: Proceedings of the tenthinternational symposium on environmental degradation of materials in nuclearsystems-water reactors. Lake Tahoe，NV: NACE Intenrational, The Minerals, Metals andMaterials Society, American Nuclear Society (CDROM),2001
[136]S. Suzuki, T. Shvji, Y.S. Yie tc. Theoretical SCC crack growth prediction undervairous loading modes in high temperature water [C], In: Proceedings of the eighthinternational symposium on environmental degradation of materials in nuclearsystems-water reactors. Florida, Illinois, USA: American Nuclear Society,1997,695-703
[137]T. Shoji. Progress in the mechanistic understanding of BWR SCC and its implicationto the prediction of SCC growth behavior in plants [C]. In: Proceedings of the eleventhinternational conference on environmental degradation of mateirals in nuclear powersystems-water reactors，Stevenson, Washington, USA (CDROM)，2003
[138]G.A. Webster, R.A. Ainsworth. High temperature component life assessment. Landon
[M]. UK: Chapman and Hall，1994，327-340
[139]K.R.W. Wallin. Cirtical assesment of the standard ASTM E399[C]. In: Daniewciz,Sth.R., Newman, J.C.and Schwalbe，K.-H. Fatigue and rfacture mechanics:34volume.West Conshohocken, PA, USA. ASTM International,2004
[140]俞铭华，吴剑国.有限元法与C程序设计[M].科学出版社,1998
[141]张洪武，关振群等.有限元分析与CAE技术基础[M].北京：清华大学出版社，2004
[142]庄茁，张帆，岑松等.ABAQUS非线性有限元分析与实例[M].北京：科学技术出版社，2005
[143]石亦平，周玉蓉.ABAQUS有限元分析实例详解[M].北京：机械工业出版社，2006
[144]孙戬.多尺度下裂纹断裂过程区力学特性分析[D].硕士论文，西安：西安科技大学，2009
[145]T. Shoji, G..F. Li. Quantification of yield strength effects on IGSCC of austeniticstainless steels in high temperature water [C], Proceedings of the11th Conference ofEnvironmental Degradation of Mateirals in Materials in Nuclear Systems, Stevenson,Atevenson，Washington,2003，3834-3844
[146]H. Xue, K. Ogawa，T. Shoji. Effect of welded mechanical heterogeneity on local stressand strain in stationary and growing crack tips [J]. Nuclear Engineeirng and Design,2009，236(5):628-640
[147]H. Xue, X.Y. Gong, L.Y. Zhao et al. Effect of mechanical parameter selection inquantitative estimation of the growth of environmentally assisted cracks at flaws in lightwater reactor components with complex mechanical condition [C]. Proceesings ofASME2011Pressure Vessels and Piping Conference, Baltimore Maryland,2011
[148]李智军，薛河.直流电位降裂纹测深仪的数值标定[J]，西安科技大学学报，2012,32(1):116-120
[149]G. Sposito，P. Cawley，P. Nagy. An approximate model for three-dimensionalalternating current potential drop analyses using a commercial finite element code [J].NDT&E International,2010，43:134-140
[150]I. Cerny. The use of DCPD method for measurement of growth of cracks in largecomponents at normal and elevated temperatures [J]. Engineering Fracture Mechanics,2004,71(4-6):837-848
[151]E. Madhi，P. Nagy. Sensitivity analysis of a directional potential drop sensor for creepmonitoirng [J], NDT&E International,2011，44:708-717
[152]Y. Satoa，K. Watanabe, T. Shoji. Simulation of stress corrosion cracking behavior in atube-shaped specimen of nickel-based alloy600[J]. Nuclear Engineeirng and Design,2008，238:1-7
[153]D.M. Syahkal, D.H. Michael，R. Collins. Parasitic voltages induced by artificial flawswhen measured using the ac field technique [J]. Nondestruct. Eval,1981,2:195-202
[154]R. Collins, D.H. Michael, R. Clark. Measurement of crack depth in a transition weldusing ACPD [J]. Rev. Prog. Quant. Nondestruct. Eval，1992，545-552
[155]ABAQUS/Standard User's Manual. Version6.7, Hibbitt, Karlsson and Sorensen，Inc，Pawtucket, RI, U.S，2007
[156]H. Xue，Z. Li, Z. Lu,T. Shoji. The effect of a single tensile over load on stresscorrosion cracking growth of stainless steel in a light water reactor environment [C].Nuclear Engineering and Design,2011
[157]Z.P. Lu，T. Shoji, F.J. Meng，Y.B. Qiu, T.C. Dan, H. Xue. Effects of water chemistryand loading conditions on stress corrosion cracking of cold-rolled316NG stainless steelin high temperature water [J]. Corrosion Science,2011，53(1):247-262
[158]ASTM Standard E399-90. Standard test method for plane strain rfacture toughness ofmetallic materials. Annual book of ASTM Standards, ASTM Intenrational, Vol.03.01,2002
[159]K. Miyazaki, M. Mochizulki，S. kanno，M. Hayashi，M. Shiratoir，Q.Yu. Analysis ofstress intensity factor due to surface crack propagation in residual stress fields caused bywelding [J]. JSME Intenrational Journal Series A,2002,45(2):199-207
[160]Z.P. Lu, T. Shoji” F.J. Meng，Y.B. Qiu, T.C. Dan. and H. Xue. Effects of waterchemistry and loading conditions on stress corrosion cracking of cold-rolled316NGstainless steel in high temperature water [J]. Corrosion Science，2011，53(1):247-262
[161]Y. Ito, M. Saito，Q. Peng, J. Kuniya. and T. Shoji. Influencing factors for improvingaccuracy in prediction of stress corrosion crack growth rate in boiling water reactor
operational conditions-2-The effect of single overload on SCC growth behavior [C]. In:
Proceedings of the2010ASME Pressure Vessels and Piping Division Conference.
Bellevue. Washington.2010. PVP2010-26121
[2]史永I核能发电的优点及世界核电发展动向[J].能源工程,2007, (1):1-6
[3] R.Pirce, J.Blaise. Nuclear fuel resources enough to last [J]. NEA NEWS,2002,20(2):10-13
[4]连培生.原子能工业[M].北京：原子能出版社，2002
[5]孙勤.第三代核电技术落在中国是历史的必然[N].科技日报，2009-03-20
[6]赵永新，张意轩.我国第三代核电产业体系基本成型[N].人民日报,2010七8-10
[7]李宗明.核电发展概述[J].求是,2011,(8): 54-55
[8]辛灵，刘艳.日核危机升级中国暂停审批核项目核电发展决心不改变，速度或将放缓[N].南方都市报，2011-03-18
[9]魏仁杰.核电与核电安全-三里岛和切尔诺贝利核电站事故的研究[J].核动力工程，1987,8(4):289-294
[10]武俊德.中国核电面临艰难抉择[J].节能与环保,2011,202(4): 34-36
[11]束国刚，陆念文.压水堆核电厂关键金属部件的老化和寿命评估[J].中国电力,2006,39(5):53-58
[12] F.A. Simonen，S.R. Gosselin. Life Prediction and monitoring of nuclear power plantcomponents for service-related degradation [J], Journal of Pressure Vessel Technology,2001，123(1):58-64
[13]毕岩，谢里阳.核电关键构件材料的寿命预测方法[J].失效分析与预防，2011，6(1):12-14
[14] R.G. Ballinger. Environmental degradation in lwr systems: current and emerging issues
[C]， Proceedings of5th IMR Symposium on Materials Science andEngineeirng-Mateirals and Reliability in Nuclear Power Plant,2009
[15]苏州热工研究院.与AECL的腐蚀技术交流一我国核电站面临的腐蚀问题[DB/OL],2005，http://wenku.baidu.com/view/f9d80ae69b89680203d82547.html
[16]柯伟，韩恩厚，王俭秋.核电镍基合金材料的腐蚀失效和预防[C].经济发展方式转变与自主创新-第十二届中国科学技术协会年会(第二卷)，2010
[17]刘建章.核结构材料[M],北京：化学工业出版社，2007
[18]李光福，杨武.核电站异材焊接件的破裂问题与应力腐蚀评价方法[J].核安全,2003,(2):37-40
[19]李继红，张建勋.断裂应变对含损伤焊接接头裂纹扩展驱动力影响的数值[J].得接学报,2004,25(1):52-56
[20]W.H. Hearns. Welding Handbook(Vol.4): Metals and Their Weldability [M], AmericanWelding Society,1982
[21]胡舜媛译.核电站部件受应力裂纹腐蚀[J].国外核新闻，1994，（10):27
[22]蒋浦宁.核电汽轮机防应力腐烛裂纹技术[J].热力透平，2010,39(2):89-92
[23]杨武.核电工程材料的应力腐蚀破裂研究[J].腐蚀科学与防护技术，1995，7(2):87-92
[24]P. Scott, M.L. Calvar. Some possible mechanism of intergranular stress, corrosioncracking of alloy600in PWR pirmary water [C]. Proceedings of Sixth InternationalConference of Environmental Degradation of Mateirals Nuclear Power Systems-WaterReactors, TMS,1993:657-665
[25]吕战鹏，杨武.遭受应力腐蚀开裂的设备寿命预测技术[J].腐烛科学与防护技术，1999,11(1):57-64
[26]D. Hamaguchi, Y. Dai. Micro-structural study of EC316LN and its welds irradiated inSINQ target-3[J], Journal of Nuclear Materials,2005，(343):262-266
[27]J.Yu, D.S. Gelles, F.A. Ganrer. The performance of Chinese316L and316Ti stainlesssteel irradiated at300,400,500and600°C in HFIR JP-23test capsule [J]. Journal ofNuclear Mateirals,2002,307-311:357-361
[28]A. Pardo, M.C. Merino, A.E. Coy. Influence of Ti，C and N concentration on theintergranular corrosion behaviour of AISI3i6Ti and321stainless steels [J]. Acta Mater,2007，55:2239-2251
[29]卢建树，王保峰，张九渊.高温水中不锈钢和镍基合金应力腐蚀破裂研究进展[J].核动力工程，2001,22(3):259-263
[30]环境断裂教育部重点实验室[EB/OL].http://www.ustb.edu.cn/esf/index.htm
[31]褚武扬，谷飚.应力腐蚀机理研究的新进展[J].腐蚀科学与防护技术,1995，7(2):97-101
[32]陆永浩，褚武扬，髙克玮等.T.Shoji.304L不锈钢在高温水中的应力腐蚀裂纹扩展[J],金属学报,2004,40(7):763-767
[33]李红梅.奥氏体不锈钢模拟辐照促进应力腐蚀破裂研究[D].博士论文，上海：上海交通大学,2005
[34]高欣，吴欣强，关辉等.高温高压水环境中腐蚀产物膜的研究现状[J].腐蚀科学与防护技术，2007,27(2):110-113
[35]徐松，吴欣强，韩恩厚等.316Ti不锈钢在模拟核电高温高压水中的腐蚀疲劳裂纹断口研究[J].中国腐蚀与防护学报，2010，30(2):119-123
[36]国家发展和改革委员会.重大工程材料服役安全研究评价设施国家重大科技基础设施建设项目批准立项[EB/OL].2007.02.16, http://www.sdpc.gov.cn/gjscy/cxtx/t20070216l17764_.htm
[37]科技部.国家重点基础研究发展计划经费预算拟安排情况汇总表[EB/OL].http://www.most.gov.cn/tztg/200611/P020061122376424687228.xls
[38]J. Hou, T. Shoji, Z.P. Lu et al. Residual strain measurement and grain boundarycharacteirzation in the heat-affected zone of a weld joint between Alloy690TT andAlloy52[J], Journal of Nuclear Materials,2010，397(1-3):109-115
[39]韩恩厚，王俭秋，吴欣强等.核电髙温高压水中不锈钢和镍基合金的腐蚀机制[J].金属学报,2010,46(11):1379-1390
[40]P.L. Andresen, K. Gott, J.L. Nelson. Stress corrosion cracking of sensitized type304stainless steel in288°C water: a ifve laboratory round bobbin [C]. Proceedings of the9thInternational Symposium on Environmental Degradation of Material in Nuclear PowerSystems-Water Reactors, S. Baremmer, et al eds., Newport Beach, California,1999.423-433
[41]沈长斌，陶晓杰，杨怀玉等.髙温高压水环境下传热管失效形式及防腐措施研究进展[J].腐蚀科学与防护技术，2003，15(4):223-227
[42]D.D. Macdonald, P.C. Lu. Theoretical estimation of crack growth rates in type304stainless steel in BWR coolant environments [J]. Corrosion，1996，52:768-785
[43]T. Kondo，H. Nakajina, R. Nagasak. Metallographic investigation on the claddingfailure in the pressure vessel of a BWR [J], Nuclear Engineeirng and Design,1971，16(8):205
[44]Y. Sato, H. Xue, Y. Takeda et al. Development of a stress corrosion crack testmethodology using tube-shaped specimen [J]. ASTM International-Journal of Testingand Evaluation,2007，35(3):254-258
[45]K. Chopra, H.M. Chung. Current research on environmentally assisted cracking in lightwater reactor environments [J]. Nuclear Engineering and Design,1999，194:205-223
[46]Q.J. Peng，J. Kwon, T. Shoji. Development of a fundamental crack tip strain rateequation and its application to quantitative prediction of stress corrosion cracking ofstainless steels in high temperature oxygenated water [J]. Journal of Nuclear Mateirals,2004，324(1):52-61
[47]李金许，高克玮，宿彦京等.钝化膜应力导致不锈钢应力腐蚀的研究[lf.中国科学(E辑)，2003，33(9):796-803
[48]O.K. Chopra, H.M. Chung, T.F. Kassner et al. Current research on environmentallyassisted cracking in light water reactor environments [J], Nuclear Engineering andDesign,1999，194:205-223
[49]P.L. Andresen, F.P. Ford. Fundamental modeling of environmental cracking forimproved design and lifetime evaluation in BWRs [J]. International Journal of PressureVessels and Piping，1994，59:61-70
[50]M. Hall. Film rupture model for aqueous stress corrosion cracking under constant andvairable stress intensity factor [J]. Corrosion Science,2009,51:225-233
[51]F.P. Ford. Mechanisms of environmentally-assisted cracking [J]. International Journalof Pressure Vessels and Piping,1989，40(55):343-362
[52]P.L. Andresen, F.P. Ford. Life prediction by mechanistic modeling and systemmonitoirng of environmental cracking of iron and nickel alloys in aqueous systems [J].Mateirals Science and Engineering,1988, A103:167-184
[53]Y.C. Gao, K.C. Hwang. Elastic-plastic fields in steady crack growth in a strainhardening material [C]. Proceedings of Fitfh Intenrational Conference on Fracture,1981，669-682
[54]T. Shoji，S. Suzuki, R.G. Ballinger. Theoretical prediction of SCC growthbehavior-threshold and plateau growth rate [C]. Proceedings of the Seventh InternationalSymposium on Environmental Degradation of Materials in Nuclear Power Systems,Breckinridge,1995,881-889
[55]T. Hashimoto，M. Koshishi. Modiifcation of the FRI crack growth model formulationrfom a mathematical viewpoint [J]. Journal of Nuclear Science and Technology,2009,46:295-302
[56]M. Vankeerberghen, G. Weyns, S. Gavrilov et al. Crack propagation rate modelling for316SS exposed to PWR-relevant conditions [J]. Journal of Nuclear Mateirals，2009，384(3):274-285
[57]M.M. Hall. An alternative to the Shoji crack tip strain rate equation [J]. CorrosionScience.2008，50:2902-2905
[58]T. Shoji, Z.P. Lu，H. Xue et al. Quantification of the effects of crack tip plasticity onenvironmentally-assisted crack growth rates in LWR environments [M]. In:Environment-Induced Cracking of Mateirals，2008:107-122
[59]Q.J. Peng，J. Kwon, T. Shoji. Development of a fundamental crack tip strain rateequation and its application to quantitative prediction of stress corrosion cracking ofstainless steels in high temperature oxygenated water [J], Journal of Nuclear Materials,2004，324(1):52-61
[60]H_Xue, Y. Sato，T.Shoji. Quantitative estimation of the growth of environmentallyassisted cracks at flaws in light water reactor components [J]. Transactions of theASME-Journal of Pressure Vessel and Technology,2009，131(1):61-70
[61]Y.C. Gao. Exponential stress and strain singularity at tip of mode I growing crack inpower hardening plastic material [J]. Theoretical and Applied Fracture Mechanics，1996,25(2):103-111
[62]T. Shoji, Z.P. Lu，H. Murakami. Formulating stress corrosion cracking growth rates bycombination of crack tip mechanics and crack tip oxidation kinetics [J]. CorrosionScience,2010，52(3):769-779
[63]H. Xue, T. Shoji. Quantitative prediction of EAC crack growth rate of sensitized type304stainless steel in boiling water reactor environments based on EPFEM [J],Transactions of the ASME-Journal of Pressure Vessel and Technology,2007，129(3):254-258
[64]左景伊.应力腐烛破裂[M].西安：西安交通大学出版社，1985
[65]李濒.断裂力学[M].济南：山东科技出版社，1979
[66]冒小萍，郎福元，柯显信.断裂力学的数值计算方法的研究现状与展望[J].商丘师范学院学报，2004，20(2):20-24
[67]周爱细，黄建龙，郎福元.计算断裂力学进展[J].甘肃工业大学学报，1998,24(1):111-116
[68]张国瑞.有限元法[M].北京：机械工业出版社，1992
[69]T.L. Andersen. Fracture mechanics: tfindamentals and applications, second edition[M]. Boca Raton，USA: CRC Press,1995
[70]P. Aaltonen，M. Bojiov, M. Helin et al. Facts and views on the role of anionic impuirties,crack tip chemistry and oxide films in environmentally assisted cracking [M]. Espoo，Finland: VTT industiral Systems，2002
[71]P. Siitonen. Passivation and the forms of corrosion [M]. In: Corrosion handbook,Hanko，Finland: Suomen Korroosioyhdistys SKY，1988
[72]Metals handbook [M]. Desk Edition (Second Edition). Mateirals Park, USA: ASMInternational,1998
[73]A. Toivonen. Stress corrosion crack growth rate measurement in high temperaturewater using small precracked bend specimens [D]. Ph.D. Thesis，Helsinki University ofTechnology,2004
[74]A.J. Sedirks. Corrosion testing made easy: stress corrosion cracking test methods [M].Houston，TX，USA: NACE，1990
[75]T.L. Andersen. Fracture mechanics: fundamentals and applications, second edition [M].Boca Raton, USA: CRC Press,1995
[76]R.A. Mayville，T.J. Warren, P.D. Hilton. The crack Velocity-Ki，relationship for AISI4340in seawater under fixed and irsing displacement [M]. In: Fracture Mechanics:Perspectives and Directions (wentieth Symposium). Philadelphia, USA: AmericanSociety for Testing and Materials,1989，605-614
[77]R.B. Rebak, Z.S. Smialowska. The mechanism of Stress corrosion cracking of alloy600in high temperature water [J], Corrosion Science，1996,38:971-988
[78]P.L. Andresen. Effects of temperature on crack growth rate in sensitized type304stainless steel and alloy600[J], Corrosion,1993，49:714-725
[79]T.M. Angeliu, P.L.Andresen and F.P.Ford. Applying slip-oxidation to the SCC ofaustenitic mateirals in BWR/PWR environments [C]. Proceedings of Corrosion’98conference, NACE International,1998,262
[80]P.L. Andresen. The effects of aqueous impuirties on intergranular stress corrosioncracking of sensitized type-304stainless steel [M]. In:Final report No. EPRI NP-3384,Palo Alto, CA: Electric Power Research Institute,1983
[81]P.L. Andresen. SCC growth rate behaviour in BWR water of increased purity. In:Proceedings of the eight international symposium on environmental degradation ofmateirals in nuclear power systems-water reactors [M]. Floirda, Illinois，USA: AmeircanNuclear Society,1997，675-684
[82]P.L. Andresen. A mechanism for the effects of ionic impuirties on SCC of austeniticiron and nickel base alloys in high temperature water [C]. Proceedings of Corrosion’85conference, NACE International,1985，101-120
[83]P.L. Andresen. Effect of dissolved oxygen, solution conductivity and stress intensity onthe interdendirtic stress corrosion cracking of Inconel182weld metal [C]. Proceedingsof Corrosion!87conference, NACE International,1987,84-98
[84]P.L. Andresen. The effects of nitrate on the stress corrosion cracking of sensitizedstainless steel in high temperature water [M], Proceedings of the seventh internationalsymposium on environmental degradation of mateirals in nuclear power systems-waterreactors. Breckenirdge, August7-10,1995.
[85]P.L Andresen, F.P. Ford. Response to "On the modeling of stress corrosion cracking ofiron and nickel base alloys in high temperature environments”[J]. Corrosion Science,1996，38:1011-1016
[86]P.L. Andresen. Chemistry and Observation and prediction of the effects of watermechanics on environmentally assisted Inconels182weld metal and600[J]. Corrosion，1988，44:376-385
[87]F.P. Ford，P.L. Andresen. Development and use of a predictive model of crackpropagation in304/316L，A533B/A508and Incone)600/182allo°ys in288C water [C].Proceedings of the third international symposium on environmental degradation ofmateirals in nuclear power systems-water reactors，1988，789-800
[88]F.P. Ford, D.F. Taylor，P.L. Andresen, R.G.. Ballinger. Corrosion-assisted cracking ofstainless and low-alloy steels in LWR environments: final report. Palo Alto, CA: ElectricPower Research Institute,1987.Report No. EPRINP-5064M
[89]P.L. Andresen, T.M. Angeliu. Evaluation of the role of hydrogen in SCC in hot water
[C]. In: Proceedings of the Corrosion'97conference, NACE International,1997,195-210
[90]F.P. Ford. Overview of collaborative research into the mechanisms of environmentallycontrolled cracking in the low alloy pressure vessel steel/water system [C]. In: Cullen，W.H. Proceedings of the second international atomic energy agency specialists'meeting an subcritical crack growth. Sendai,1986.3-72
[91]F.P. Ford. Quantitative prediction of environmentally assisted cracking [J]. Corrosion,1996，52:375-395
[92]P.L. Andresen. SCC mitigation strategies in hot water [C]. In: Proceedings ofCorrosion!97conference, NACE International,1997.102
[93]R.M. Horn, G..M. Gordon, F.P Ford et al. Experience and assessment of stresscorrosion cracking in L-grade stainless steel BWR internals [J]. Nuclear Engineeirng andDesign,1997,174:313-325
[94]P.L. Andresen. Probabilistic interpretation of SCC initiation and growth [C]. In:Proceedings of Corrosion’99conference, NACE International,1999,446457
[95]P.L. Andresen, F.P. Ford, T.M. Angeliu et al. Prediction of environmentally assistedcracking and its relevance to life management in SWRs [C]. In: Baremmer，S” Ford，P.and Was, G. Proceedings of the ninth international symposium on environmentaldegradation of materials in nuclear systems-water reactors. Newport Beach, California,Warrendale. USA: The Minerals，Metals and Mateirals Society,1999,373-381
[96]P.L. Andresen, T.M. Angeliu, P.W. Emigh et al. Crack growth rate behaviour of Nialloys [C]. In: Proceedings of Corrosion12000conference, NACE International,2000,202-214
[97]P.L. Andresen, F.P. Ford. Fundamental quantification of crack advance for lifeprediction in energy systems [C]. In: Proceedings of Corrosion!96conference, MACEInternational,1996,51-99
[98]P.L. Andresen, T.M. Angeliu，L.M. Young et al. Mechanisms and kinetics of SCC instainless steels [C]. In: Proceedings of the tenth international symposium onenvironmental degradation of mateirals in nuclear systems-water reactors. Lake Tahoe,NY, NACE International, The Minerals, Metals and Materials Society, AmericanNuclear Society (CDROM)，2001
[99]P.L. Andresen，P.W. Emigh, M.M. Morra et al. Yield strength, corrosion potential, stressintensity factor，Effects of silicon and grain boundary character on the SCC of stainlesssteels [C], In: Proceedings of the eleventh intenrational conference on environmentaldegradation of mateirals in nuclear power systems-water reactors, Stevenson,Washington，USA (CDROM)，2003
[100]D.A. Yermilyea. A theory for the propagation of stress corrosion cracks in metals [J].J. Electrochem. Soc，1972,119:405-407
[101]J. Congleton，W. Yang. The effect of applied potential on the stress corrosioncracking of sensitized type316stainless steel in high temperature water[J]. Corrosion Science，1995，37:429-144
[102]T. Shoji, K.S. Raja, G.F. Li et al. Critical parameters of environmentally assitedcracking in nuclear systems [C]. In: Proceedings of Corrosion’2000conference, NACEInternational,2000.190-202
[103]N.S. Mclntyre, C.M. Huctwith, R.D. Davidson et al. M. ID cracking in UNS N0b600nuclear steam generator tubes: evidence for SCC from secondary ion mass spectrometricimages [J]. Corrosion Science,1998，40:1799-1802
[104]A. Tumbull. Modelling of crack chemistry in sensitized stainless steel in boiling waterreactor environments [J]. Corrosion Science，1997，39:789-805
[105]D.F. Taylor. Crevice corrosion of Alloy600in high-temperature aqueousenvironments [J], Corrosion,1979，35:550-559
[106]J. Congleton，R.N. Parkins. Stress corrosion cracking of steel in high temperaturewater [J]. Corrosion,1988,44:290-298
[107]W.E. Ruther，W.K. Soppet, T.F. Kassner. Effect of temperature and ionic impurities atvery low concentrations on stress corrosion cracking of type304stainless steel [J].Corrosion,1988，44:791-799
[108]D.D. Macdonald. Clairfication of issues raised by P.L andresen and F.P. Ford in theirresponse to "On the modeling of stress corrosion cracking of iron and nickel base alloysin high temperature environments"[J]. Corrosion Science,1997,39:1487-1490
[109]A. Tumbull, M. Psaila-Dombrowski. A review of electrochemistry of relevance toenvironment-assisted cracking in light water reactors [J]. Corrosion Science，1932，32:1925-1966
[110]A. Turnbull. Modelling of environment assisted cracking [J]. Corrosion Science,1993,34:921-960
[111]D.D. Macdonald，M. Urquidi-Macdonald. A coupled environment model for stresscorrosion cracking in sensitized type304stainless steel in LWR environments [J].Corrosion Science，1991，32:51-81
[112]D.D. Macdonald，P.C. Lu, M. Urquidi-Macdonaid et al. Theoretical estimation ofcrack growth rates in type304stainless steel in BWR coolant environments [J].Corrosion,1996，52:768-785
[113]D.D. Macdonald，I_Balachov. Modeling the accumulation and mitigation of SCCdamage in BWRs. In: Bruemmer, S_，Ford，P and Was，G. Proceedings of the ninthinternational symposium on environmental degradation of mateirals in nuclearsystems-water reactors [C]. Newport Beach, California, Warrendale, USA: The Minerals,Metals and Materials Society.1999，399-407
[114]G.R. Engelhardt, D.D. Macdonald. Modelling of corrosion fatigue chemistry insensitized steel in boiling water reactor environments [C]. In: Proceedings ofCorrosion!2000conference, MACE International,2000.227-241
[115]M.P. Manahan, D.D. Macdonaid，A.J. Peterson. Determination of the fate of thecurrent in the stress corrosion cracking of sensitized type304SS in high temperatureaqueous systems [J], Corrosion Science，1995，37:189-208
[116]M. Vankeerberghen, D. Macdonald. Calculating the temperature-maximum and thelower potential limit for the crack growth rate in type304SS using the CEFM [C]. In:Proceedings of Corrosion’2003conference, NACE International,2003,3520-3530
[117]G.R. Engelhardt，D.D. Macdonald，M. Urduidi-Macdonald. Development of fastalgorithms for estimating stress corrosion crack growth rate [J], Corrosion Science,1999,41:2267-2302
[118]P.M. Scott, M. LeCalvar. On the role of oxygen in stress corrosion cracking as afunction of temperature [C], In: Magnin, T. Proceedings of the second intenrationalconference on corrosion-deformation interactions in conjuction with EUROCORR!96.Nice, France, London: The Institute of Mateirals,1996,384-393
[119]P.M. Scott, M. LeCalvar. Some possible mechanisms of intergranular stress corrosioncracking of alloy600in PWR pirmary water [C]. In: Gold，R.E and Simonen，E.P.Proceedings of the sixth intenrational symposium on environmental degradation ofmateirals in nuclear power systems-water reactors. San Diego, CA，Warrendale, PA,USA: The Minerals, Metals and Mateirals Society.1993,657-667
[120]P.M. Scott. An overview of internal oxidation as a possible explanation ofintergranular stress corrosion cracking of alloy600in PWRs. In: Bruemmer, S?，Ford, Pand Was，G. Proceedings of the ninth international symposium on environmentaldegradation of materials in nuclear systems-water reactors [C]. Newport Beach，California, Warrendale, PA, USA: The Minerals, Metals and Materials Society.1999,3-11
[121]P. Combrade, P.M. Scott. The role of surface films on alloy600corrosion inpressuirsed water reactors [C]. In: Magnin, T. Proceedings of the secondinternational conference on corrosion-deformation interactions in conyuction withEUROCORR'96. Nice，France, London，UK: The Institute of Mateirals,1996,394-409
[122]R. Staehle，Z. Fang. Comments on a proposed mechanism of internal oxidation foralloy600as applied to low potential SCC [C]. In: Baremmer, S.，Ford, P and Was, G.Proceedings of the ninth international symposium on environmental degradation ofmateirals in nuclear systems-water reactors，Newport Beach, Califonria，Warrendale, PA,USA: The Minerals, Metals and Materials Society,1999,69-77
[123]J.R. Galvele. A stress corrosion crack mobility [J]. Corrosion Science,1987，27mechanism based on surface1-33
[124]J.R. Galvele. Recent developments in the surface-mobility stress corrosion crackingmechanism [J], Electrochim. Acta,2000，45:3537-3541
[125]J.R. Galvele. Comments on”Notes on the surface mobility mechanism of stress-corrosion cracking"[J]. By K. Sieradzki and F.J. Firedersdorf. Corrosion Science,1994,36(5):901-910
[126]K. Sieradzki, F.J. Firedersdorf. Notes on the surface mobility mechanism ofstress-corrosion cracking [J]. Corrosion Science,1994，36:669-675
[127]P. Aaltonen, T. Saairo, P. Karjalainen-Roikonen et al. Vacancy-creep model for EACof metallic materials in high temperature water [C]. In: Proceedings of Corrosion!96conference, NACE International,1996.81-95
[128]L. Balachov，M. Bojinov, T. Laitinen et al. Correlations between the stability of oxidefil2ms in simulated BWR crack conditions，S04" enrichment and cracking susceptibility
[C]. In: Proceedings of water chemistry of nuclear reactor systems8. Bournemouth，London，UK: Birtish Nuclear Energy Society,2000，170-172
[129]M. Bojinov, G. Fabircius，T. Laitinen et al. Coupling between ionic defect structureand electronic conduction in passive films on iron, chromium and iron-chromium alloys[J]. Electrochim Acta,2000，45:2029-2048
[130]M. Bojinov, G. Fabricius，T. Laitinen et al. Influence of molybdenum on theconduction mechanism in passive films on iron-chromium alloys in sulphuric acidsolutions Electrochim [J]. Acta，2001，46:1339-1358
[131]B_Beverskog，M. Bojinov, P. Kinnunen et al. A mixed-conduction model for oxidefilms on Fe，Cr and Fe-Cr alloys in high-temperature aqueous electrolytes-II. Adaptationand justification of the model [J]. Corrosion Science，2002，44:1923-1940
[132]Y. Jagodzinski，P. Aaltonen，S. Smuk et al. Internal friction study of environmentaleffects on metals and alloys [J]. J. Alloys and Compounds,2000,310:256-260
[133]P. Aaltonen，Y_Jagodzinski, O. Tarasenko et al. Low-frequency internal friction ofpure copper atfer anodic polairzation in sodium nitrite solution [J]. Corrosion Science，1998，40:903-908
[134]P. Aaltonen, Y. Jagodzinski, O. Tarasenko et al. Intenral rfiction in brasses atferoxidation in tap water by anodic polairzation [J], Acta Mater,1998，46:2039-2046
[135]Q. Peng，J. Kwon，T_Shoji. Inverse analysis of crack growth kinetics of sensitized304L stainless steel in a simulated boiling water reactor environment for quantitativeprediction of stress corrosion cracking behavior [C]. In: Proceedings of the tenthinternational symposium on environmental degradation of materials in nuclearsystems-water reactors. Lake Tahoe，NV: NACE Intenrational, The Minerals, Metals andMaterials Society, American Nuclear Society (CDROM),2001
[136]S. Suzuki, T. Shvji, Y.S. Yie tc. Theoretical SCC crack growth prediction undervairous loading modes in high temperature water [C], In: Proceedings of the eighthinternational symposium on environmental degradation of materials in nuclearsystems-water reactors. Florida, Illinois, USA: American Nuclear Society,1997,695-703
[137]T. Shoji. Progress in the mechanistic understanding of BWR SCC and its implicationto the prediction of SCC growth behavior in plants [C]. In: Proceedings of the eleventhinternational conference on environmental degradation of mateirals in nuclear powersystems-water reactors，Stevenson, Washington, USA (CDROM)，2003
[138]G.A. Webster, R.A. Ainsworth. High temperature component life assessment. Landon
[M]. UK: Chapman and Hall，1994，327-340
[139]K.R.W. Wallin. Cirtical assesment of the standard ASTM E399[C]. In: Daniewciz,Sth.R., Newman, J.C.and Schwalbe，K.-H. Fatigue and rfacture mechanics:34volume.West Conshohocken, PA, USA. ASTM International,2004
[140]俞铭华，吴剑国.有限元法与C程序设计[M].科学出版社,1998
[141]张洪武，关振群等.有限元分析与CAE技术基础[M].北京：清华大学出版社，2004
[142]庄茁，张帆，岑松等.ABAQUS非线性有限元分析与实例[M].北京：科学技术出版社，2005
[143]石亦平，周玉蓉.ABAQUS有限元分析实例详解[M].北京：机械工业出版社，2006
[144]孙戬.多尺度下裂纹断裂过程区力学特性分析[D].硕士论文，西安：西安科技大学，2009
[145]T. Shoji, G..F. Li. Quantification of yield strength effects on IGSCC of austeniticstainless steels in high temperature water [C], Proceedings of the11th Conference ofEnvironmental Degradation of Mateirals in Materials in Nuclear Systems, Stevenson,Atevenson，Washington,2003，3834-3844
[146]H. Xue, K. Ogawa，T. Shoji. Effect of welded mechanical heterogeneity on local stressand strain in stationary and growing crack tips [J]. Nuclear Engineeirng and Design,2009，236(5):628-640
[147]H. Xue, X.Y. Gong, L.Y. Zhao et al. Effect of mechanical parameter selection inquantitative estimation of the growth of environmentally assisted cracks at flaws in lightwater reactor components with complex mechanical condition [C]. Proceesings ofASME2011Pressure Vessels and Piping Conference, Baltimore Maryland,2011
[148]李智军，薛河.直流电位降裂纹测深仪的数值标定[J]，西安科技大学学报，2012,32(1):116-120
[149]G. Sposito，P. Cawley，P. Nagy. An approximate model for three-dimensionalalternating current potential drop analyses using a commercial finite element code [J].NDT&E International,2010，43:134-140
[150]I. Cerny. The use of DCPD method for measurement of growth of cracks in largecomponents at normal and elevated temperatures [J]. Engineering Fracture Mechanics,2004,71(4-6):837-848
[151]E. Madhi，P. Nagy. Sensitivity analysis of a directional potential drop sensor for creepmonitoirng [J], NDT&E International,2011，44:708-717
[152]Y. Satoa，K. Watanabe, T. Shoji. Simulation of stress corrosion cracking behavior in atube-shaped specimen of nickel-based alloy600[J]. Nuclear Engineeirng and Design,2008，238:1-7
[153]D.M. Syahkal, D.H. Michael，R. Collins. Parasitic voltages induced by artificial flawswhen measured using the ac field technique [J]. Nondestruct. Eval,1981,2:195-202
[154]R. Collins, D.H. Michael, R. Clark. Measurement of crack depth in a transition weldusing ACPD [J]. Rev. Prog. Quant. Nondestruct. Eval，1992，545-552
[155]ABAQUS/Standard User's Manual. Version6.7, Hibbitt, Karlsson and Sorensen，Inc，Pawtucket, RI, U.S，2007
[156]H. Xue，Z. Li, Z. Lu,T. Shoji. The effect of a single tensile over load on stresscorrosion cracking growth of stainless steel in a light water reactor environment [C].Nuclear Engineering and Design,2011
[157]Z.P. Lu，T. Shoji, F.J. Meng，Y.B. Qiu, T.C. Dan, H. Xue. Effects of water chemistryand loading conditions on stress corrosion cracking of cold-rolled316NG stainless steelin high temperature water [J]. Corrosion Science,2011，53(1):247-262
[158]ASTM Standard E399-90. Standard test method for plane strain rfacture toughness ofmetallic materials. Annual book of ASTM Standards, ASTM Intenrational, Vol.03.01,2002
[159]K. Miyazaki, M. Mochizulki，S. kanno，M. Hayashi，M. Shiratoir，Q.Yu. Analysis ofstress intensity factor due to surface crack propagation in residual stress fields caused bywelding [J]. JSME Intenrational Journal Series A,2002,45(2):199-207
[160]Z.P. Lu, T. Shoji” F.J. Meng，Y.B. Qiu, T.C. Dan. and H. Xue. Effects of waterchemistry and loading conditions on stress corrosion cracking of cold-rolled316NGstainless steel in high temperature water [J]. Corrosion Science，2011，53(1):247-262
[161]Y. Ito, M. Saito，Q. Peng, J. Kuniya. and T. Shoji. Influencing factors for improvingaccuracy in prediction of stress corrosion crack growth rate in boiling water reactor
operational conditions-2-The effect of single overload on SCC growth behavior [C]. In:
Proceedings of the2010ASME Pressure Vessels and Piping Division Conference.
Bellevue. Washington.2010. PVP2010-26121