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核电站一回路主管道材料热老化力学性能的预测概述 被引量:3

Review of Estimation for Thermal Aging Mechanical Properties of Material Used in Primary Coolant Pipes of Nuclear Power Plant
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摘要 从核电站一回路铸造奥氏体不锈钢热老化现象、热老化后力学性能变化趋势及力学性能预测流程等方面,对铸造奥氏体不锈钢热老化后冲击性能、拉伸性能、J-R曲线和断裂韧性力学性能预测的研究现状进行了综述。并对目前力学性能预测方法的不足提出自己的想法。 The research status of thermal aging cast austenitic stainless steels' charpy-impact property,tensile property,J-R curves,fracture toughness property estimation achievements from the changes trend and prediction process of thermal aging cast austenitic stainless steels mechanical properties are reviewed.And the opinion about the shortage of current mechanical estimation method is given.
出处 《材料导报》 EI CAS CSCD 北大核心 2014年第1期95-99,共5页 Materials Reports
基金 国家863计划项目(2012AA050901)
关键词 铸造奥氏体不锈钢 力学性能 热老化 cast austenitic stainless steel mechanical property thermal aging
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参考文献27

  • 1Chung H M. Aging and life prediction of cast duplex stain- less steel components[J]. Int J Pressure Vessels Piping, 1992,50(1) : 179.
  • 2Kawaguchi S, Sakamoto N, Takano G, et al. Microstruc- tural changes and fracture behavior of CFSM duplex stain- less steels after long-term aging[J]. Nucl Eng Des, 1997, 174(3) :273.
  • 3Chopra O K. Estimation of fracture toughness of cast stain- less steels during thermal aging in LWR systems [ R]. Washington, DC: Argonne National Lab(ANL), 1994.
  • 4王永强,李时磊,杨滨,王艳丽,王西涛.核电站一回路主管道铸造奥氏体不锈钢热老化研究现状与展望[J].材料导报,2012,26(3):101-105. 被引量:21
  • 5赵彦芬,遆文新,汪小龙,薛飞.核电站用钢管材料及其国产化[J].钢管,2007,36(2):11-14. 被引量:27
  • 6Chopra O K. Thermal aging of cast stainless steels in LWR Systems: Estimation of mechanical properties[R]. Washing- ton, DC: Argonne National Lab (ANL), 1991.
  • 7Chung H M, Chopra O K. Kinetics and mechanism of ther- mal aging embrittlement of duplex stainless steels[C]//Pro- ceedings 3: conference on Environmental Degradation of Materials in Nuclear Power Systems:Water Reactors. War- rendale, 1988.
  • 8Yamada T, Okano S, et al. Mechanical property and micro- structural change by thermal aging of SCS14A cast duplex stainless steel[J]. J Nucl Mater, 2006,350(1) : 47.
  • 9Pareige C, Novy S, Saillet S, et al. Study of phase transfor- mation and mechanical properties evolution of duplex stain- less steels after long term thermal ageing (: 20 years)[J]. J Nucl Mater,2011,411(1) : 90.
  • 10Back S, Seok C S, Koo J M. Evaluation of the degradation characteristics of CF-SA cast stainless steel using indentation techniques and EDS[J]. Key Eng Mater, 2006,306 : 869.

二级参考文献74

  • 1王勤湖,李社坤,卢文跃,于海峰,曹智鹏.压水堆核电站一回路工况变化对主泵主要机械性能的影响[J].核动力工程,2005,26(S1):103-108. 被引量:24
  • 2刘鹏,薛飞,戴忠华,陈世均,朱文彬,汪小龙,遆文新.轻水堆核电站奥氏体不锈钢铸件的热老化及其老化管理[J].核动力工程,2005,26(S1):93-96. 被引量:23
  • 3李元太,张春来,雷中黎.压水堆一回路管道的铸造工艺及其国产化[J].核动力工程,2009,30(S2):6-10. 被引量:13
  • 4赵彦芬,遆文新,汪小龙,薛飞.核电站用钢管材料及其国产化[J].钢管,2007,36(2):11-14. 被引量:27
  • 5夏生兰 顾世雄.压水堆一回路水质标准的腐蚀依据[J].核动力工程,1988,9(2):60-65.
  • 6SAMUEL K G. Evaluation of Ageing-induced Embrittlement in an Austenitic Stainless Steel by Instrumented Impact Testing[J]. Journal of Nuclear Materials, 1987,150 : 78.
  • 7Chung H M. Aging and Life Prediction of Cast Duplex Stainless Steel Components[J]. Int J Press Vessels Piping. 1992,50 : 179.
  • 8Iacoviello F, Casari F, Gialanella S. Effect of "475 ℃ Embrittlement" on Duplex Stainless Steels Localized Corrosion Resistance[J]. Corrosion Science, 2005,47 : 909.
  • 9French Association for Design, Construction and Inservice Inspection Rules for Nuclear Island Components. RCC-M-2000, Design and Construction Rules for Mechanical Components of PWR Nuclear Islands [S]. Paris : AFNOR, 2002.
  • 10Chopra O K, Sather A. Initial Assessment of the Mechanisms and Significance of Low Temperature Embrittlement of Cast Stainless Steels in LWR Systems. US Nuclear Regulatory Commission Report: NUREG/CR-5385, Argonne National Laboratory, 1990.

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