核电厂稳压器贯穿接管的修复和结构完整性评估
Structural Integrity Analysis of Nuclear Power Plant Pressure Vessel Penetration Nozzle Repaired

作者: 马琳伟 , 何家胜 , 舒安庆 , 郑小涛 , 徐建民 :武汉工程大学,湖北 武汉;

关键词: 贯穿接管应力强度疲劳裂纹扩展稳压器焊接修复Penetration Nozzle Stress Intensity Fatigue Crack Growth Pressurizer Weld Repair

摘要: 较早期核电厂的压力容器封头上的贯穿接管材料多为600合金,由于600合金对一回路水环境应力腐蚀开裂的抗性较差,易产生应力腐蚀裂纹贯穿压力边界而导致泄漏,可采用修复方法使用690合金接管更换损坏的接管,重构压力边界。为确保新压力边界能够达到原始设计要求,对重构压力边界的新接管及其焊缝进行结构完整性分析。对稳压器顶盖上的贯穿接管,进行重构压力边界的焊接修复和结构完整性评估,基于有限元分析获得工况瞬态热应力和焊接残余应力的基础上,依据ASME规范进行应力强度校核,并通过断裂力学计算分析假想裂纹在核电厂40年设计寿命内的扩展及稳定性,分析结果表明修复后的结构满足设计和使用要求。

Abstract: Many NPP pressure vessel head nozzles were manufactured of Alloy 600. Since Alloy 600 is sus-ceptible to PWSCC, the degradation mechanism has been observed in these nozzles. The industry has used several methods to mitigate PWSCC, including replacement of nozzles fabricated of Alloy 690. After the replacement of the nozzle, the structural integrity analysis of new nozzle and welds should be performed to ensure the pressure boundary compliance with the original design re-quirement. The PWR pressurizer top head instrument nozzle is evaluated. Thermal stress of the transients was obtained from 3D FEM analysis and residual stress of J-groove weld was obtained from 2D FEM analysis. Stress intensities and residual stress were conservatively determined and used for the ASME Code, stress intensity analysis, fatigue crack growth analysis and fracture me-chanics analysis. All of the analysis showed that the repaired nozzle satisfies the ASME Code design requirement.

文章引用: 马琳伟 , 何家胜 , 舒安庆 , 郑小涛 , 徐建民 (2017) 核电厂稳压器贯穿接管的修复和结构完整性评估。 核科学与技术, 5, 36-48. doi: 10.12677/NST.2017.51005

参考文献

[1] US.NRC. (2005) U.S. Plant Experience with Alloy 600 Cracking and Boric Acid Corrosion of Light-Water Reactor Pressure Vessel Materials. USNRC Report, NUREG 1823. US:USNRC.

[2] Sun, H.T., Shen, C.Y., Gao, C., Wang, C., Ling, L.G. and Jia, P.P. (2015) The Application of Weld Overlay in the Maintenance of Nuclear Equipment. Welding & Joining, 94, 53-56.

[3] US.NRC. (1989) Residual Life Assessment of Major Light Water Reactor Components. US.NRC Report, NUREG/CR 4731. US:USNRC.

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