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

作　　者：钟多军 何君 陈耀 李莎 ZHONG Duojun;HE Jun;CHEN Yao;LI Sha(CNNC Jianzhong Nuclear Fuel Co.,Ltd.,Yibin 644000,China)

机构地区：[1]中核建中核燃料元件有限公司,宜宾644000

出　　处：《理化检验（物理分册）》2023年第6期25-28,共4页Physical Testing and Chemical Analysis(Part A:Physical Testing)

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。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.

关 键 词：核电 慢应变速率拉伸试验 304L不锈钢 包壳 应力腐蚀 

分 类 号：TB31[一般工业技术—材料科学与工程] TG115.5[金属学及工艺—物理冶金]