镍基690合金在溶解氢/溶解氧交替变化高温高压水中应力腐蚀萌生行为和机制研究

由于补给水中的溶解氧（DO）、停堆时充入的氧等原因，压水堆（PWR）核电站一回路设备材料会服役于溶解氢（DH）/DO交替变化高温高压水环境。目前研究表明交替变化的DH/DO环境会损伤材料氧化膜的保护性，但是对应力腐蚀（SCC）萌生的影响及作用机制并不清楚。本项目以PWR核电站蒸汽发生器（SG）传热管用镍基690合金为研究对象，通过高温高压水腐蚀热力学计算、原位电化学实验、浸泡实验以及管状样品慢应变速率拉伸实验，结合表面与微观分析方法，研究DH/DO交替变化环境中合金表面、截面和裂纹尖端氧化膜成分、结构特征及其耐蚀性与晶界迁移程度、裂纹密度、长度和深度的关系，揭示DH/DO交替变化对690合金氧化膜保护性的影响及SCC萌生的作用机制，阐明掺入DO浓度对氧化膜保护性的损伤程度，为深入理解DH和DO对材料服役安全的影响、优化PWR一回路水化学控制、确保SG结构完整性和长期服役可靠性提供依据。

As a result of the dissolved oxygen (DO) in the supply water, the oxygen ingress during shutdown etc., the primary circuit equipment materials of pressurized water reactor (PWR) nuclear power plant will be served in the high temperature and high pressure water environment with alternating dissolved hydrogen (DH)/DO. However, the current studies show alternating DH/DO environment can degrade the protectiveness of oxide film on material. However, the effects and mechanism on stress corrosion initiation behavior are not clear. This project uses the materials for PWR steam generator(SG) heat transfer tube, namely nickel-based alloy 690 as the research objective. High temperature thermodynamics calculation, in situ electrochemical experiments, immersion test and slow strain rate tensile test with tube specimen are used to study the composition, structure, corrosion resistance etc. on the surface, in cross section and in the crack tip and grain boundary migration, crack density, length and depth under the alternating DH/DO environments combined with surface analysis methods. The project is to clarify the protective evolution and the mechanism of SCC initiation of alloy 690 caused by alternating DH/DO environment. The damage degree of DO with different concentrations in the primary water to the oxidation film protective is also to be clarified. The above results are helpful for optimizing PWRs primary circuit water chemical control and ensuring the SG structural integrity and reliability for a long time.