核级商用690合金和800合金在模拟压水堆核电站一回路高温高压水中的腐蚀行为研究

CORROSION BEHAVIOR OF NUCLEAR GRADE ALLOYS 690 AND 800 IN SIMULATED HIGH TEMPERATURE AND HIGH PRESSURE PRIMARY WATER OF PRESSURIZED WATER REACTOR

在自行搭建的高温高压水循环回路系统中,通过原位电化学测量,结合SEM观察和XPS分析,研究了核级商用690合金和800合金在模拟压水堆核电站一回路高温高压水环境中的腐蚀行为.结果表明,690合金和800合金的自腐蚀电位随浸泡时间的延长而逐渐降低,浸泡时间对690合金和800合金的阻抗谱影响并不明显.经过408 h浸泡后,690合金表面生成了大量针状氧化物,800合金表面则同时生成了针状氧化物和颗粒状氧化物.690合金表面富Cr氧化层位于表面膜内侧,而800合金的富Cr氧化层位于表面膜外侧.在高温高压水中,690合金比800合金表现出更为优异的抗腐蚀能力.浸泡实验后,溶液中主要金属离子Ni^(2+),Cr^(3+)和Fe(3+)的含量分别为0.1×10^(-6),0.1×10^(-6)和0.3×10^(-6).

The corrosion behaviors of nuclear grade commercial alloys 690 and 800 were studied by in situ electrochemical measurements using a self-built high temperature and high pressure water loop system, combining with SEM observation and XPS analysis. The results show that the corrosion potentials of alloys 690 and 800 decrease gradually with immersion time increasing, while the immersion time has no obvious impact on the result from electrochemical impedance spectroscopy （EIS）. A large number of needle-like oxides have been found on the surface of alloy 690 after being exposed to high temperature and high pressure water for 408 h. For alloy 800, except needle-like oxides, many particle oxides are also observed. For alloy 690, Cr is rich at inner side of oxide film, while it is rich at outer side of oxide film for alloy 800. Alloy 690 shows better corrosion resistance than alloy 800 in high temperature and high pressure water. After immersion experiment, the contents of Ni2＋, Cr3＋ and Fe3＋ ions in the test water solutions are 0.1 x 10^-6, 0.1 x 10^-6 and 0.3 x 10^-6, respectively.