反应堆压力容器用钢断裂韧度的表征方法及其统计模型
Characterization Methods and Statistical Models of Fracture Toughness of Reactor Pressure Vessel Steels
摘 要
简要介绍了目前反应堆压力容器(RPV)用钢断裂韧度的两种表征方法——基于参考无延性转变温度(TNDT)的下限曲线法和基于参考温度(T0)的主曲线法, 并阐述了断裂韧度的典型统计模型, 以表征断裂韧度试验数据的分散性。基于所开发的概率断裂力学(PFM)分析程序, 分析了不同统计模型对承压热冲击(PTS)条件下含缺陷RPV失效概率的影响, 为国产RPV用钢断裂韧度数据的统计分析提供参考。
概率断裂力学
反应堆压力容器
承压热冲击
fracture toughness
probability fracture mechanics
reactor pressure vessel
pressurized thermal shock
Abstract
Two characterization methods of fracture toughness of reactor pressure vessel (RPV) steels are briefly introduced. One is the Lower-Bound Curve based on the reference temperature of nil-ductility transition temperature (TNDT), another is the Master Curve based on reference temperature (T0). Typical statistical models of fracture toughness data are stated, to represent the scatter of these data. The failure probability of RPV at pressurized thermal shock (PTS) is calculated with different statistic models, based on a probability fracture mechanics (PFM) analysis program developed independently. The results may provide reference to establish the fracture toughness statistic model of homemade steels for RPV.
中图分类号 TL341 TL351
所属栏目 物理模拟与数值模拟
基金项目 “十二五”国家科技支撑计划项目(2011BAK06B02-03)
收稿日期 2013/1/15
修改稿日期 2013/12/25
网络出版日期
作者单位点击查看
备注李曰兵(1987-), 男, 山东五莲人, 博士研究生。
引用该论文: LI Yue-bing,GAO Zeng-liang,LEI Yue-bao. Characterization Methods and Statistical Models of Fracture Toughness of Reactor Pressure Vessel Steels[J]. Materials for mechancial engineering, 2014, 38(4): 91～95
李曰兵,高增梁,雷月葆. 反应堆压力容器用钢断裂韧度的表征方法及其统计模型[J]. 机械工程材料, 2014, 38(4): 91～95
被引情况:
【1】董泽忠,杜鸣杰,史 科,顾宙,王文东, "核电反应堆压力容器支座减摩板往复摩擦磨损性能",理化检验－物理分册 52, 552-556(2016)
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参考文献
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【3】ASME Boiler and Pressure Vessel Code, Section XI. Rules for inservice inspection of nuclear power plant components ［S］.
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【5】SATTARI-FAR I, WALLIN K. Application of master curve methodology for structural integrity assessments of nuclear components, SKI report 2005:55［R］.Stockholm: Swedish Nuclear Power Inspectorate,2005.
【6】杨文斗.核电厂压力容器安全评估的新方法——主曲线简介［J］.核安全,2011(2): 7-12.
【7】张亚平,王东辉,钟志民.RPV用钢美国常用断裂韧性KIc表达式的对比分析［J］.压力容器,2011,28(3):16-22.
【8】ASME Boiler and Pressure Vessel Code, Section III. Rules for construction of nuclear facility components［S］.
【9】RG 1.99 (Revision 2). Radiation embrittlement of reactor vessel materials［S］.
【10】ASTM E1921. Test method for the determination of reference temperature T0 for ferritic steels in the transition range［S］.
【11】方颖, 李辉, 惠虎, 等. 基于Master Curve方法的A508-III钢断裂韧性研究［J］.核动力工程,2011,32(增1):31-34.
【12】ASME Boiler and Pressure Vessel Code, Code Case N-629. Use of fracture toughness test data to establish reference temperature for pressure retaining materials, section XI, division 1［S］.
【13】WILLIAMS P T, BOWMAN K O, BASS B. R, et al. Weibull statistical models of KIc/KIa fracture toughness databases for pressure vessel steels with an application to pressurized thermal shock assessments of nuclear reactor pressure vessels［J］.International Journal of Pressure Vessels and Piping,2001,78(2/3):165-178.
【14】BASS B R, WILLIAMS P T, PUGH C E. An updated correlation for crack-arrest fracture toughness for nuclear reactor pressure vessel steels ［J］. International Journal of Pressure Vessels and Piping,2005,82(6):489-495.
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【16】李曰兵,金伟娅,高增梁,等.基于概率断裂力学的承压热冲击条件下含周向裂纹圆筒体的结构完整性研究［J］.核技术, 2013,36(4):1-6.
【17】BALKEY K, WITT F J, BISHOP B A. Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: parts 1 and 2［R］.California: Electric Power Research Institute,1995.
【2】PUGH C E, BASS B R, DICKSON T L. Role of probabilistic analysis in integrity assessments of reactor pressure vessels exposed to pressurized thermal-shock conditions［J］.Engineering Failure Analysis,2007,14(3):501-517.
【3】ASME Boiler and Pressure Vessel Code, Section XI. Rules for inservice inspection of nuclear power plant components ［S］.
【4】WALLIN K. Master curve analysis of the "Euro" fracture toughness dataset［J］.Engineering Fracture Mechanics,2002,69(4):451-481.
【5】SATTARI-FAR I, WALLIN K. Application of master curve methodology for structural integrity assessments of nuclear components, SKI report 2005:55［R］.Stockholm: Swedish Nuclear Power Inspectorate,2005.
【6】杨文斗.核电厂压力容器安全评估的新方法——主曲线简介［J］.核安全,2011(2): 7-12.
【7】张亚平,王东辉,钟志民.RPV用钢美国常用断裂韧性KIc表达式的对比分析［J］.压力容器,2011,28(3):16-22.
【8】ASME Boiler and Pressure Vessel Code, Section III. Rules for construction of nuclear facility components［S］.
【9】RG 1.99 (Revision 2). Radiation embrittlement of reactor vessel materials［S］.
【10】ASTM E1921. Test method for the determination of reference temperature T0 for ferritic steels in the transition range［S］.
【11】方颖, 李辉, 惠虎, 等. 基于Master Curve方法的A508-III钢断裂韧性研究［J］.核动力工程,2011,32(增1):31-34.
【12】ASME Boiler and Pressure Vessel Code, Code Case N-629. Use of fracture toughness test data to establish reference temperature for pressure retaining materials, section XI, division 1［S］.
【13】WILLIAMS P T, BOWMAN K O, BASS B. R, et al. Weibull statistical models of KIc/KIa fracture toughness databases for pressure vessel steels with an application to pressurized thermal shock assessments of nuclear reactor pressure vessels［J］.International Journal of Pressure Vessels and Piping,2001,78(2/3):165-178.
【14】BASS B R, WILLIAMS P T, PUGH C E. An updated correlation for crack-arrest fracture toughness for nuclear reactor pressure vessel steels ［J］. International Journal of Pressure Vessels and Piping,2005,82(6):489-495.
【15】MASAKI K, NISHIKAWA H, OSAKABE K, et al. User′s manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL3 for reactor pressure vessel［R］. Japan Ibaraki-ken: Japan Atomic Energy Agency,2010.
【16】李曰兵,金伟娅,高增梁,等.基于概率断裂力学的承压热冲击条件下含周向裂纹圆筒体的结构完整性研究［J］.核技术, 2013,36(4):1-6.
【17】BALKEY K, WITT F J, BISHOP B A. Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: parts 1 and 2［R］.California: Electric Power Research Institute,1995.
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