核电站一回路管道铸造奥氏体不锈钢热老化评估
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摘要
压水堆核电站中的一回路主冷却剂管道用铸造奥氏体不锈钢(CASS)长期在反应堆环境下服役,面临热老化脆化问题。本文利用时效参数外延法,对其主要用钢Z3CN20-09M钢在长期运行后的热老化脆化程度进行评估,并对材料的使用寿命进行预测。评估结果表明:在时效参数P=3.48的时效状态下,Z3CN20-09M钢中的铁素体发生了调幅分解,产生了热老化脆化倾向;同时,拟合得出该材料在288~327℃服役后的总冲击能量随时效参数P的变化关系,预测冷热腿管道用Z3CN20-09M钢在接近延寿期60满功率运行年限服役是安全的。
Cast austenitic stainless steel(CASS) used in main coolant pipes of the primary loop in pressurized-water reactor(PWR) nuclear power plants faces the problem of thermal aging embrittlement after long-time operation in the reactor environment. In this study, the aging parameter extrapolation method is used to evaluate the degree of thermal aging embrittlement and predict the service life of the mainly used Z3CN20-09 M steel after long-time operation. The evaluation results indicate that, when the aging parameter P is 3.48, the spinodal decomposition takes place on the ferrite of Z3CN20-09 M steels, which results in a thermal aging embrittlement tendency. In the meantime, the relationship between the total impact energy and aging parameter P of the Z3CN20-09 M steel after service at 288 ℃ to 327 ℃ is obtained by fitting. It is predicted that the Z3CN20-09 M steels used for the cold and hot leg pipe is safe when its extending service life is close to 60 effective full power years.
Cast austenitic stainless steel(CASS) used in main coolant pipes of the primary loop in pressurized-water reactor(PWR) nuclear power plants faces the problem of thermal aging embrittlement after long-time operation in the reactor environment. In this study, the aging parameter extrapolation method is used to evaluate the degree of thermal aging embrittlement and predict the service life of the mainly used Z3CN20-09 M steel after long-time operation. The evaluation results indicate that, when the aging parameter P is 3.48, the spinodal decomposition takes place on the ferrite of Z3CN20-09 M steels, which results in a thermal aging embrittlement tendency. In the meantime, the relationship between the total impact energy and aging parameter P of the Z3CN20-09 M steel after service at 288 ℃ to 327 ℃ is obtained by fitting. It is predicted that the Z3CN20-09 M steels used for the cold and hot leg pipe is safe when its extending service life is close to 60 effective full power years.
引文
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[8]张家倍,马琳伟,鲁红权.核电运行技术支持-基础及应用[M].上海:上海科学技术出版社,2010:53-54.ZHANG Jiabei,MA Liwei,LU Hongquan.Technical support for nuclear power operation-basics and applications[M].Shanghai:Shanghai Science and Technology Press,2010:53-54(in Chinese).
[9]王伟,孙振田,李超.核电站老化分析及延寿管理方法研究[J].机电产品开发与创新,2009,22(3):37-39.WANG Wei,SUN Zhentian,LI Chao.Study on aging analysis and service life extension management of nuclear power plant[J].Development and Innovation of Machinery and Electrical Products,2009,22(3):37-39(in Chinese).
[10]李承亮,张明乾.压水堆核电站反应堆压力容器材料概述[J].材料导报,2008,22(9):65-68.LI Chengliang,ZHANG Mingqian.Overview of reactor pressure vessel steel in PWR nuclear power plant[J].Materials Review,2008,22(9):65-68.
[11]黄均麟,刘向红,黄炳炎.核电厂主管道90°弯头热老化程度的评估[J].核动力评估,2015,36(2):105-108.HUANG Junlin,LIU Xianghong,HUANG Bingyan.Evaluation of thermal aging of RCP 90°elbow for nuclear power plants[J].Nuclear Power Engineering,2015,36(2):105-108.
[12]中华人民共和国国家质量监督检验检疫总局,中华人民共和国国家标准化管理委员会.钢材夏比V型缺口摆锤冲击试验仪器化试验方法:GB/T 19748-2005[S].北京:中国标准出版社,2005:11.General Administration of Quality Supervision,Inspection and Quarantine of the People’s Republic of China,Standardization Administration of the People’s Republic of China.Steel:Charpy V-notch pendulum impact test:Instrumented test method:GB/T 19748-2005[S].Beijing:Standards Press of China,2005:11(in Chinese).
[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.BI Yan,XIE Liyang.Fatigue life prediction of material of nuclear power key components[J].Failure Analysis and Prevention,2011,6(1):12-14.
[14]CHELLAPANDI P,SRINIVASAN R,CHETAL S C,et al.Experimental creep life assessment of tubular structures with geometrical imperfections in welds with reference to fast reactor plant life[J].International Journal of Pressure Vessels and Piping,2006,83:556-564.
[15]Applied Science and Technology,Framatome Technologies,Inc.Evaluation of thermal aging embrittlement for cast austenitic stainless steel components in LWR reactor coolant systems[R].California:Electric Power Research Institute,1997:56-86.(责任编辑杜亚勤)
[2]KAWAGUCHI S,SAKAMOTO N,TAKANO G,et al.Microstructural changes and fracture behavior of CF8Mduplex stain-less steels after long-term aging[J].Nuclear Engineering and Design,1997,174(3):273-285.
[3]束国刚,路念文.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.SHU Guogang,LU Nianwen.Aging problems and life evaluation for the key metallic components in PWRnuclear power plant[J].Electric Power,2006,39(5):53-58.
[4]CHOPRA O K.Estimation of fracture toughness of cast stainless steels during thermal aging in LWR systemsRevison 1[R].Washington:U.S.Nuclear Regulatory Commission,1994:56-69.
[5]CHOPRA O K,CHUNG H M.Aging of cast duplex stainless steels in LWR systems[J].Nuclear Engineering and Design,1985,89(2/3):305-318.
[6]王理.压水堆核蒸汽供应系统主要设备材料的典型老化问题[C]//中国核学会2009年学术年会论文集.北京:原子能出版社,2009:441-446.WANG Li.The typical aging problem of the main component materials in PWR nuclear steam supply system[C]//Proceedings of the Progress Report on China Nuclear Science and Technology.Beijing:Atomic Energy Press,2009:441-446.
[7]郑凯,王镇,钟振前,等.长期热老化后核电站主管道材料的力学行为[J].钢铁研究学报,2014,26(7):1-7.ZHENG Kai,WANG Zhen,ZHONG Zhenqian,et al.Mechanical behavior of long-term thermal aged nuclear power plant’s primary circuit pipes[J].Journal of Irion and Steel Research,2014,26(7):1-7.
[8]张家倍,马琳伟,鲁红权.核电运行技术支持-基础及应用[M].上海:上海科学技术出版社,2010:53-54.ZHANG Jiabei,MA Liwei,LU Hongquan.Technical support for nuclear power operation-basics and applications[M].Shanghai:Shanghai Science and Technology Press,2010:53-54(in Chinese).
[9]王伟,孙振田,李超.核电站老化分析及延寿管理方法研究[J].机电产品开发与创新,2009,22(3):37-39.WANG Wei,SUN Zhentian,LI Chao.Study on aging analysis and service life extension management of nuclear power plant[J].Development and Innovation of Machinery and Electrical Products,2009,22(3):37-39(in Chinese).
[10]李承亮,张明乾.压水堆核电站反应堆压力容器材料概述[J].材料导报,2008,22(9):65-68.LI Chengliang,ZHANG Mingqian.Overview of reactor pressure vessel steel in PWR nuclear power plant[J].Materials Review,2008,22(9):65-68.
[11]黄均麟,刘向红,黄炳炎.核电厂主管道90°弯头热老化程度的评估[J].核动力评估,2015,36(2):105-108.HUANG Junlin,LIU Xianghong,HUANG Bingyan.Evaluation of thermal aging of RCP 90°elbow for nuclear power plants[J].Nuclear Power Engineering,2015,36(2):105-108.
[12]中华人民共和国国家质量监督检验检疫总局,中华人民共和国国家标准化管理委员会.钢材夏比V型缺口摆锤冲击试验仪器化试验方法:GB/T 19748-2005[S].北京:中国标准出版社,2005:11.General Administration of Quality Supervision,Inspection and Quarantine of the People’s Republic of China,Standardization Administration of the People’s Republic of China.Steel:Charpy V-notch pendulum impact test:Instrumented test method:GB/T 19748-2005[S].Beijing:Standards Press of China,2005:11(in Chinese).
[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.BI Yan,XIE Liyang.Fatigue life prediction of material of nuclear power key components[J].Failure Analysis and Prevention,2011,6(1):12-14.
[14]CHELLAPANDI P,SRINIVASAN R,CHETAL S C,et al.Experimental creep life assessment of tubular structures with geometrical imperfections in welds with reference to fast reactor plant life[J].International Journal of Pressure Vessels and Piping,2006,83:556-564.
[15]Applied Science and Technology,Framatome Technologies,Inc.Evaluation of thermal aging embrittlement for cast austenitic stainless steel components in LWR reactor coolant systems[R].California:Electric Power Research Institute,1997:56-86.(责任编辑杜亚勤)