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
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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。Abstract: 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.