核电用304L不锈钢包壳的慢应变速率拉伸试验
Slow strain rate tensile test of 304L stainless steel cladding for nuclear power
摘 要
对核电用304L不锈钢包壳进行慢应变速率拉伸试验,用扫描电子显微镜对试样的断口进行观察。结果表明:核电用304L不锈钢包壳的应力腐蚀开裂敏感性系数接近1,在高温氮气和高温、高压水中测试后,试样断口的宏观形貌基本一致,呈韧性断裂特征;核电用304L不锈钢包壳在高温、高压水中的应力腐蚀敏感性较低。
慢应变速率拉伸试验
304L不锈钢
包壳
应力腐蚀
nuclear power
slow strain rate tensile test
304L stainless steel
cladding
stress corrosion
Abstract
The slow strain rate tensile test of 304L stainless steel cladding for nuclear power was carried out, and the fracture of the sample was observed by scanning electron microscope. The results show that the stress corrosion cracking sensitivity coefficient of 304L stainless steel cladding for nuclear power was close to 1, and the macroscopic morphology of the fracture after testing in high temperature nitrogen, high temperature and high pressure water was basically the same, showing ductile fracture characteristics. The stress corrosion sensitivity of 304L stainless steel cladding for nuclear power was low in high temperature and high pressure water.
中图分类号 TB31 TG115.5 DOI 10.11973/lhjy-wl202306007
所属栏目 试验与研究
基金项目
收稿日期 2022/9/14
修改稿日期
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备注钟多军(1983-),硕士,高级工程师,主要从事核燃料元件制造及其工艺技术研究工作,limu1983@126.com
引用该论文: ZHONG Duojun,HE Jun,CHEN Yao,LI Sha. Slow strain rate tensile test of 304L stainless steel cladding for nuclear power[J]. Physical Testing and Chemical Analysis part A:Physical Testing, 2023, 59(6): 25～28
钟多军,何君,陈耀,李莎. 核电用304L不锈钢包壳的慢应变速率拉伸试验[J]. 理化检验－物理分册, 2023, 59(6): 25～28
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参考文献
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