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国产压水堆核电站机组主管道疲劳裂纹扩展特性实验研究 被引量:5

Experiment Study on Fatigue Crack Propagation Behavior of Primary Pipe in China Pressurized-water Reactor
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摘要 在室温条件下,对国产压水堆核电站主管道母材及其TIG接头试样进行疲劳裂纹扩展试验,并采用光学显微镜观测裂纹扩展路径,结合扫描电镜观察试样断口微观形貌。试验结果显示TIG接头的裂纹扩展速率高于母材,基于简化的四参数全范围Forman模型可以表征主管道母材与焊材全范围的疲劳裂纹扩展规律。疲劳裂纹在奥氏体与铁素体相内主要呈穿晶扩展,但在部分区域裂纹沿?/?或?/?相界产生分支。 A series of fatigue crack growth rate tests were carried out on the primary pipe of the pressurized water reactor nuclear power plant at room temperature, the research objects include base material and its TIG welded joint specimens. The fatigue crack growth rate was measured, and the crack propagation path was observed using an optical microscope.Furthermore, the sample fracture morphology was analyzed with a scanning electron microscope. The crack-growth curve identify that the crack growth rate of TIG welded joint specimen is higher than the base material. It is found that the Forman model can be used to describe the crack growth behavior for two materials. The crack predominantly shows a transgranular mode, but branch crack along the α/α or α/γ phase boundary was observed in the crack growth area.
出处 《中国电机工程学报》 EI CSCD 北大核心 2014年第8期1310-1317,共8页 Proceedings of the CSEE
基金 国家重点基础研究发展计划项目(973计划)(2011CB610506) 国家科技重大专项项目(2011ZX06004002)~~
关键词 国产压水堆核电站 主管道 自动焊 裂纹扩展 China pressurized-water reactor nuclearpower plant primary pipe automatic welding crack growth
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  • 1孙造占.LBB对材料性能的要求[J].核安全,2006,5(3):6-10. 被引量:5
  • 2李强,岑鹏,甄洪栋.核电厂高能管道LBB分析技术概述[J].核动力工程,2011,32(S1):189-191. 被引量:9
  • 3贺寅彪,曲家棣,窦一康.反应堆压力容器承压热冲击分析[J].压力容器,2004,21(10):5-9. 被引量:17
  • 4苏小萍.CFRP层板的疲劳损伤研究方法[J].纤维复合材料,2004,21(3):60-61. 被引量:5
  • 5李云龙,庄传晶,冯耀荣,霍春勇.油气输送管道疲劳寿命分析及预测[J].油气储运,2004,23(12):41-43. 被引量:25
  • 6U.S.Nuclear Regulatory Commission.10 CFR 50.61,Fracture toughness requirements for protection against pressurized thermal shock events[S].Washington DC:Nuclear Regulatory Commission,1984.
  • 7Qian G,Niffenegger M.Integrity analysis of a reactor pressure vessel subjected to pressurized thermal shocks by considering constraint effect[J].Engineering Fracture Mechanics,2013,112-113:14-25.
  • 8Wallin K.Quantifying T stress controlled constraint by the master curve transition temperature T 0 [J].Engineering Fracture Mechanics,2001,68(3):303-328.
  • 9Scheuerer M,Weis J.Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions[J].Nuclear Engineering and Design,2012,253:343-350.
  • 10IAEA 1627.Pressurized thermal shock in nuclear power plants:good practices for assessment[R].Austria:IAEA,2010:3-63.

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