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核压力容器接管安全端堆焊修复对失效评定曲线的影响 被引量:4

Effect of Dissimilar Metal Weld Overlay between Pipe-nozzle of Nuclear Pressure Vessel and Safe End on Failure Assessment Curves
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摘要 核压力容器接管安全端异种金属焊接接头在服役中通常会产生高温高压水环境下的应力腐蚀裂纹扩展。目前减轻和修复这种裂纹的技术为在安全端管接头外表面堆焊一层更抗腐蚀的镍基合金(Alloy52M)材料。通过ABAQUS有限元分析,构建了堆焊修复后结构的失效评定曲线,分析了堆焊层厚度、裂纹深度和裂纹位置对失效评定曲线的影响。结果表明:随着堆焊层厚度的增加、裂纹深度的减小及裂纹位置向接管嘴的移动,失效评定曲线上移,结构的安全性增加。当对堆焊修复接头区的裂纹进行评定时,需要建立与堆焊修复后安全端结构尺寸、裂纹尺寸、裂纹位置和材料性能相关的准确的失效评定曲线。 Primary water stress corrosion cracking(PWSCC) has often been produced in the dissimilar metal welded joints between pipe-nozzles of nuclear pressure vessels and safe ends.The weld overlay of Alloy52M with higher corrosion resistant has usually been made to repair or mitigate this PWSCC.Based on the detailed three-dimensional finite element analyses,the failure assessment curves(FACs) of the dissimilar metal welded overlay structure were constructed,and the effects of the weld overlay thickness,crack depth and crack location on the FACs have been analyzed.The results show that with increasing the weld overlay thickness,decreasing the crack depth and moving the crack location to the pipe-nozzle,the failure assessment curves shift upward,which increases the safety of the structure.To accurately assess the integrity of the DMWJ structure with weld overlay,the FACs related to the weld overlay size,the crack size,the crack location and the material properties should be constructed.
出处 《压力容器》 2013年第5期58-63,共6页 Pressure Vessel Technology
基金 国家自然科学基金项目(51075149) 中央高校基本科研业务费专项资金
关键词 堆焊修复 失效评定曲线 核电安全端 焊接接头 有限元 weld overlay failure assessment curve nuclear safe end dissimilar metal welded joint finite element method
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参考文献15

  • 1PWR Material Realiability Project,Interim Alloy 600Safety Assessment for U.S.PWR Plants,Part 1:Al-loy82/182 Pipe Butt Welds[R].EPRI Report TP-1001491,2001.
  • 2Celin R,Tehovnik F.Degradation of a Ni-Cr-Fe Al-loy in a Pressurised Water Nuclear Power Plant[J].Mater Technol.,2011,45(1):151-157.
  • 3Jenssen A,Norrgard K,Lagerstron J.Assessment ofCracking in Dissimilar Metal Welds[C]//Proc.ofTenth Int.Symp.on Envionmental Degradation of Mate-rials in Nuclear Power Systems-Water Reactors.USA:NACE International,2001:CD-ROM.
  • 4Farley S.An Overview of Non Destructive InspectionService in Nuclear Power Plants[C]//InternationalConference on Nuclear Energy for New Europe.Porto-raz,Slovenia,2004.
  • 5ASME.Boiler and Pressure Vessel Code,Code Case N-504-2.Alternative Rules for Repair of Classes 1,2and 3 Austenitic Stainless Steel Piping SectionⅪ,Divi-sion 1[S].1995.
  • 6Zhang T,Brust F W,Wilkowski G,et al.Weld ResidualStress Analysis and the Effects of Structural Overlay onVarious Nuclear Power Plant Nozzles[J].Journal ofPressure Vessel Technology,2012,134(6).
  • 7Tsai Y L,Wang Li H,Fan T W,et al.Welding OverlayAnalysis of Dissimilar Metal Weld Cracking of Feedwa-ter Nozzle[J].International Journal of Pressure Vesselsand Piping,2009,87(5):26-32.
  • 8Marlette S,Freyer P,Smith M,et al.Simulation andMeasurement of Through-wall Residual Stresses in aStructural Weld Overlaid Pressurizer Nozzle[C]//Pro-ceedings of the ASME 2010 Pressure Vessels&PipingDivision.July 18-22,2010,Bellevue,Washington,USA.
  • 9Killian D E.Design of a Weld Overlay for a Large BorePipe Nozzle to Optimize Residual Stress[C]//Proceed-ings of the ASME 2010 Pressure Vessels&Piping Divi-sion.July 18-22,2010,Bellevue,Washington,USA.
  • 10Liu Ru-Feng,Huang Chin-Cheng.Welding Residu-al Stress Analysis for Weld Overlay on a BWR Feed-water Nozzle[J].Nuclear Engineering and Design,2012.10.In Press.

共引文献47

同被引文献29

  • 1段远刚,许斌,唐传宝.围板连接螺栓的辐照促进应力腐蚀裂纹研究[J].核动力工程,2007,28(2):62-65. 被引量:13
  • 2International Atomic Energy Agency. Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Steam Generators, IAEA-TECDOC-981 [R]. Vienna: IAEA, 1997.
  • 3International Atomic Energy Agency. Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety Primary Piping in PWRs, IAEA-TECDOC-1361 [R]. Vienna: IAEA, 2003.
  • 4Nuclear Regulatory Commission. Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants, NUREG-0531 [R]. Washington, DC: U.S. NRC, 1979.
  • 5International Atomic Energy Agency. Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: BWR Pressure Vessels, IAEA-TECDOC-1470 JR]. Vienna: IAEA, 2005.
  • 6International Atomic Energy Agency. Stress Corrosion Cracking in Light Water Reactors: Good Practices and Lessons Learned, NP-T- 3.13 [R]. Vienna: IAEA, 2011.
  • 7Nuclear Regulatory Commission. Technical report on material selection and processing guidelines for BWR coolant pressure boun-dary piping, NUREG-0313 [R]. Washington, DC: U.S. NRC, 1988.
  • 8American Society of Mechanical Engineers. Rules for in-service in- spection of nuclear power plant components [S]. New York, 2004.
  • 9AFCEN. Design and construction rules for mechanical compo nents of PWR nuclear islands [S]. Paris, 1997.
  • 10Damico T. The applicability of MSIP TM for mitigating PWSCC in pressurizer nozzle to safe-endwelds [R]. New Mexico: EPRI, 2005.

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