国产A508-3钢辐照性能
Mechanical Property of China A508-3 Steel after Neutron Irradiation
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摘要: 反应堆压力容器（RPV）作为反应堆寿期内不可更换的核心设备，是防止堆芯放射性泄漏的最主要屏障。本文针对国产压力容器材料A508-3钢，开展了一定剂量水平（约10×1019 cm-2，E≥1 MeV）的研究堆加速辐照试验，并进行了辐照后力学性能测试分析，包括拉伸性能和冲击性能测试。结果显示，辐照后在-100、20、288 ℃下，A508-3钢的屈服强度分别增加了83、108、52 MPa，抗拉强度分别增加了58、61、49 MPa，韧脆转变温度T41J增加了68 ℃，上平台能量降低了61 J。A508-3钢辐照前后性能测试结果表明，在中子辐照至60 a寿期后，A508-3钢仍能满足反应堆使用要求。Abstract: Reactor pressure vessel (RPV) is the critical un-changeable component of the reactor during its service lifetime, which prevents the radioactive leak of the nuclear power plant core. The irradiation test (about 10×1019 cm-2, E≥1 MeV) of the pressure vessel material of China A508-3 steel in research reactor was carried out, and the mechanics performance tests were carried out after the neutron irradiation, including tensile property and impact property. The results show that the yield strength increases by 83, 108 and 52 MPa, and the tensile strength increases by 58, 61 and 49 MPa at -100, 20 and 288 ℃, respectively. The ductile-brittle transition temperature T41J increases by 68 ℃, and the upper shelf energy decreases by 61 J. Meanwhile, by comparing the property of un-irradiated and irradiated material, after irradiated to the level of 60 a service life, A508-3 steel still meets the reactor operation requirement.
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Keywords:
- A508-3 steel ,
- neutron irradiation ,
- irradiation embrittlement
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[1] 刘向阳. 我国核电技术发展现状及存在的问题[J]. 电器工业,2010(6):59-66.LIU Xiangyang. Current situation and existing problems of nuclear power technology in China[J]. China Electrical Equipment Industry, 2010(6): 59-66(in Chinese). [2] 万强茂. 美国压水堆RPV延寿分析研究及中国RPV延寿之关键问题[J]. 压力容器,2010,27(6):46-51.WAN Qiangmao. Analysis on life extension of PWR-RPV in USA and key issues on life extension of RPV in China[J]. Pressure Vessel Technology, 2010, 27(6): 46-51(in Chinese). [3] 杨文斗. 反应堆材料学[M]. 北京:原子能出版社,2000:160-175. [4] U. S. Nuclear Regulatory Commission Regulations: Title 10: Code of federal regulations: Appendix G: Fracture toughness requirements[S]. U.S.: [s. n.], 2014. [5] 杨文斗. 核电厂压力容器安全评估的新方法——主曲线简介[J]. 核安全,2011(2):7-13.YANG Wendou. New method of safety assessment for pressure vessel of nuclear power plant—Brief introduction of master curve approach[J]. Nuclear Safety, 2011(2): 7-13(in Chinese). [6] ASME.BPVC section Ⅺ: Appendix G: Fracture toughness criteria for protection against failure[S]. U.S.: ASME, 2010.
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