期刊文献+

基于结构安全裕量准则的反应堆压力容器承压热冲击分析 被引量:3

Analysis of the Pressurized Thermal Shock of Reactor Pressure Vessel Based on the Structure Safety Margin Criterion
原文传递
导出
摘要 针对反应堆压力容器承压热冲击分析中较普遍采用的结构临界失效状态下材料韧脆转变温度(RTNDT)参数存在局限性的问题,和同时分析热应力载荷与内压载荷时难以区分不同载荷类型对结构安全性能影响差别的问题,基于结构安全裕量准则,通过三维数值模拟,分别研究了热应力载荷与内压载荷对反应堆压力容器堆芯筒体段裂纹前沿最深点和界面点(母材与堆焊层界面处)处的起裂和止裂性能的影响。分析中考虑了裂纹尺寸、裂纹前沿温度和中子注量沿容器壁厚方向上分布等因素的影响。 Focused on the two questions of the pressured thermal shock (PTS) of the reactor pressure vessel (RPV): 1) there are some limitations for the commonly used parameter of the critical ductile-brittle transition temperature (RTNDT); 2) and it is difficult to distinguish the influence between the thermal load and the pressure load in the combined analysis, the safety margin (SM) parameter is used to assess the safety of the crack deepest points and interface points along the crack front through 3-D finite element method in this paper. Both crack initiation assessment and crack arrest assessment are considered. A series of crack models are established, and the influence of the distribution of neutron fluence and temperature along the thickness direction are also included in the analysis.
出处 《中国电机工程学报》 EI CSCD 北大核心 2015年第20期5272-5277,共6页 Proceedings of the CSEE
基金 国家自然科学基金项目(51435012 51275338)~~
关键词 反应堆压力容器 承压热冲击 确定性断裂力学 安全裕量 有限元 reactor pressure vessel pressured thermalshock deterministic fracture mechanics safety margin finiteelement
  • 相关文献

参考文献20

  • 1U.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.
  • 2Qian 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.
  • 3Wallin K.Quantifying T stress controlled constraint by the master curve transition temperature T 0 [J].Engineering Fracture Mechanics,2001,68(3):303-328.
  • 4Scheuerer 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.
  • 5IAEA 1627.Pressurized thermal shock in nuclear power plants:good practices for assessment[R].Austria:IAEA,2010:3-63.
  • 6Chen M Y,Lu F,Wang R S,et al.Structural integrity assessment of the reactor pressure vessel under the pressurized thermal shock loading[J].Nuclear Engineering and Design,2014,272:84-91.
  • 7Wang R S,Lu F,Chen M Y,et al.Use of the failure assessment diagram(FAD) to evaluate margins in the ASME code for P-T curves[C]//Proceedings of the ASME 2012 Pressure Vessels & Piping Division Conference,Toronto:ASME,2012:33-36.
  • 8The American Society of Mechanical Engineers.Boiler and pressure vessel code,section XI,Rules for Inservice inspection of nuclear power plant components,IWB 3000[S].New York:ASME,2013.
  • 9Lee T J,Choi J B,Kim Y J,et al.A parametric study on pressure-temperature limit curve using 3-D finite element analyses[J].Nuclear Engineering and Design,2002,214(1-2):73-81.
  • 10Pennell W E,Malik S N M.Structural integrity assessment of aging nuclear reactor pressure vessels [J].Nuclear Engineering and Design,1997,172(1-2):27-47.

二级参考文献11

共引文献4

同被引文献19

引证文献3

二级引证文献9

相关作者

内容加载中请稍等...

相关机构

内容加载中请稍等...

相关主题

内容加载中请稍等...

浏览历史

内容加载中请稍等...
;
使用帮助 返回顶部