核级不锈钢堆焊材料腐蚀性能研究  

Investigation on the Corrosion Behavior of Nuclear Grade Stainless Cladding Materials

作　　者：邱绍宇[1] 孙丹琦[1] 李燕伶[1] 何艳春[1] 郑剑[1] 

机构地区：[1]中国核动力研究设计院核燃料及材料国家级重点实验室,四川成都610041

出　　处：《原子能科学技术》2002年第3期193-197,共5页Atomic Energy Science and Technology

摘　　要：采用GB43 3 4 7 84和法国RCC MMC1 3 1 0对国产两种堆焊材料进行了点腐蚀、晶间腐蚀试验 ,在模拟压水堆核电站介质 (温度 3 45℃ ,80 0mg/LB ,2mg/LLi)条件下 ,研究了堆焊材料的应力腐蚀和均匀腐蚀性能。试验结果表明 :在高温含B水中 ,U型试样试验 5 0 0 0h后无应力腐蚀破裂 ,静态月平均腐蚀速率小于 2mg/dm2 。两种堆焊材料均具有优良的耐腐蚀性。Tests of pitting corrosion and intergranular corrosion of two Chinese stainless cladding materials are carried out according to standards of GB4334 7 84 and French RCC M MC1310. Under the water chemistry condition simulating pressurized water reactors(PWRs) coolant(temperature,345 ℃;B,800 mg/L;Li,2 mg/L), their stress corrosion and uniform corrosion behaviors are studied. The results show that no stress corrosion cracking in U shape samples is found and average static uniform corrosion rate is less than 2 mg/dm 2 after testing in high temperature borated water for 5 000 h. Both of stainless cladding materials have better corrosion resistance.

关 键 词：腐蚀性能 堆焊材料 点腐蚀 晶间腐蚀 应力腐蚀 均匀腐蚀 核材料 

分 类 号：TL341[核科学技术—核技术及应用] TG171[金属学及工艺—金属表面处理]