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Z3CN20-09M铸造奥氏体不锈钢的热老化机理 被引量：17

Thermal aging mechanism of Z3CN20-09M cast austenite stainless steel
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摘要 研究了牌号为Z3CN20—09M的铸造奥氏体不锈钢在400℃下老化时间为100～3000h时的纳米压入硬度和铁素体含量的变化规律，并用透射电镜和场发射扫描电镜观察了老化后样品的组织结构．实验结果表明，该材料在3000h内的热老化脆化现象是由铁素体相区内发生调幅分解引起的． Z3CN20-09M cast austenite stainless steel was aged at 400 ℃ for 100 - 3 000 h and the change law of nano indenter hardness and ferrite content was investigated. Its microstructure after being aged was observed by both TEM and SEM. The results showed that the thermal aging embrittlement was induced by spinodal decomposition in ferrite.
作者 李时磊 王艳丽 程路 王西涛 林志 陈国良
机构地区 北京科技大学新金属材料国家重点实验室
出处 《北京科技大学学报》 EI CAS CSCD 北大核心 2008年第10期1117-1121,共5页 Journal of University of Science and Technology Beijing
基金 国家重点基础研究发展计划资助项目(No.2006CB605005)
关键词 铸造奥氏体不锈钢 压入硬度 铁素体含量 调幅分解 casting austenite stainless steel （CASS） indenter hardness ferrite content spinodal decomposition
分类号 TG142.1 [金属学及工艺—金属材料]
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参考文献10

  • 1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
  • 2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. 被引量：22
  • 3李颖,刘涛,栾培锋,史巨元.核电厂压水堆主管道材料性能的研究[J].物理测试,2006,24(5):12-13. 被引量：5
  • 4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
  • 5Baek S, Koo J M, Seok C S. Evaluation of the degradation characteristics of CF-8A cast stainless steel using indentation techniques and EDS. Key Eng Mater, 2006, 306/308:869
  • 6Chung H M. Aging and life prediction of cast duplex stainless steel components. Int J Pressure Vessels Piping, 1992, 50 (1/ 3) : 179
  • 7Chung H M, Chopra O K. Properties of stainless steels in elevated temperature service. Pressure Vessels Piping Div Publ PVP, 1987, 132(26): 17
  • 8Chung H M, Chopra O K. Environmental Degradation of Material in Nuclear Power Systems: Water Reactors. Traverse City: The Metallurgical Society, 1987
  • 9Seiichi K, Naruo S, Genta T, et al. Microstructural changes and fracture behavior of CF8M duplex stainless steels after long-term aging. Nucl Eng Des, 1997, 174(3):273
  • 10Takuyo Y, Satoshi O, Hisashi K. Mechanical property and microstructural change by thermal aging of SCS14A cast duplex stainless steel. J Nucl Mater, 2006, 350(1) : 47

二级参考文献18

  • 1IAEA-TECDOC-540.Safety aspects of nuclear power plant ageing[R].IAEA,1990.
  • 2NUREG/CR-4731.Residual life assessment of major LWR components-overview (Vol.1)[S].EGG-2469,1987.
  • 3IAEA.Assessment and management of ageing of major nuclear power plant components important to safety:primary piping in PWRs[R].IAEA-TECDOC-1361,2003.
  • 4NUREG-1801.Generic ageing lessons learned report (Vol.2)[S].2001.
  • 5NUREG/CR-4667.Environmentally assisted cracking in LWR(Vol.32)[S].2001.
  • 6NUREG/CR-6260.Application of NUREG/CR-5999 interim fatigue curves[S].1995.
  • 7NUREG/CR-6699.A review of large scale fracture experiments relevant to pressure vessel integrity under PTS conditions[S].2001.
  • 8MCCABE D E,MERKLE J G,WALLIN K.Technical basis for the master curve concept of fracture toughness evaluations in the transition range[A].ORNL,Prepared for the 30th National Symposium on Fatigue and Fracture Mechanics[C].1998.
  • 9NUREG/CR-5314.Life assessment procedures for major LWR components,cast stainless steel components (Vol.3)[S].1990.
  • 10美国国家标准，ASME锅炉及压力容器规范，核动力装置设备NCI分卷[S]．

共引文献67

同被引文献159

引证文献17

二级引证文献70

北京科技大学学报
2008年 第10期

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