镍基合金中的氢对其在高温高压水中应力腐蚀萌生的影响及机理研究

Stress corrosion cracking (SCC) is one of the main failure modes of steam generator (SG) heat transfer tubes, which are the pressure boundary of primary and secondary circuits in pressurized water reactor (PWR) nuclear power plant. However, the influence of hydrogen diffused from the primary solution to the material on the SCC initiation behavior in the secondary remains unclear. This project intends to adopt a specially designed tubular specimen made of nickel-based alloy, which is used as heat transfer tubes material. Hydrogen gas will be injected inside the tubular specimen and specimen will be exposed to high temperature high pressure water to simulate the service environment of SG tubes. Immersion and slow strain rate tensile (SSRT) test at high temperature water will be conducted. Surface analysis methods, such as glow discharge optical emission spectrometry, X-ray photoelectron spectroscopy, scanning electron microscopy and transmission electron microscopy, combined with electrochemical methods, such as electrochemical impedance spectroscopy and Mott-Schottky methods, will be used to characterize the structure, composition, thickness and defect concentration of oxide film formed on the surface or before the crack tips of specimen after immersion and SSRT. The influence of hydrogen inside the material on the integrity and protection of the oxide film will be obtained and the interaction between hydrogen and alloying elements will be revealed. Combined with the measurement of crack density, length and depth, the influence of hydrogen in the nickel-based alloys on SCC initiation behavior in high-temperature water will be evaluated. The influence of hydrogen in material on the SCC initiation behavior of nickel-based alloy in high temperature aqueous environment. The understanding of the mechanism of hydrogen enhanced oxidation will be further deepened. These above results will offer a new perspective when deciding the optimization of the primary coolant of pressured water reactor nuclear power plant.

应力腐蚀开裂（SCC）是压水堆核电站一、二回路的压力边界——蒸汽发生器（SG）传热管的主要失效形式之一。但是，由一次侧本体溶液扩散至传热管材料内部的氢对其二次侧SCC萌生行为的影响尚不清楚。本项目拟采用传热管用镍基合金管状试样，通过管内充氢、外部浸泡在高温高压水中来模拟传热管的服役环境，通过浸泡及慢应变速率拉伸（SSRT）试验，结合辉光放电发射光谱、X-射线光电子能谱、扫描电镜、透射电镜等表面分析方法及电化学阻抗谱、Mott-Schottky等电化学方法，表征浸泡后试样表面及SSRT试验后裂纹尖端氧化膜的结构、成分、厚度及缺陷浓度，获得材料内部氢对氧化膜完整性和保护性的影响，揭示氢与合金元素之间的相互作用，结合测量裂纹密度、长度和深度，评价镍基合金内部氢对其在高温水中SCC萌生行为的影响，深化对“氢加速氧化”作用机制的认识，为压水堆一回路水化学优化决策的制定提供科学依据。

蒸汽发生器传热管作为压水堆核电站一、二回路的热交换枢纽，服役环境苛刻且复杂，是应力腐蚀开裂（SCC）的敏感部位。本项目以管状690合金以及其基本组成元素Fe、Cr、Ni为研究对象，通过管内充氢、外部浸泡在高温高压水中的双暴露实验来模拟传热管的服役环境中的氢浓度梯度，通过浸泡及慢应变速率拉伸试验，研究了内部充氢对材料高温水中氧化及SCC萌生行为的影响。结果表明，四种材料在内部充氢或真空情况下均形成双层结构氧化膜。内部充氢改变了Fe的外层氧化膜结构，而其他三种材料则未受充氢影响。不同元素对氢作用的敏感程度差异显著，其中Fe对基体中氢加速氧化作用最为敏感，在内部持续充氢条件下，内层氧化膜厚度几乎翻倍。在690合金中，外层氧化物和外层氧化物的厚度都有明显的降低，而 Cr 的厚度只有轻微的降低。对于690合金，外层氧化物表现接近金属Ni，而内层表现接近Cr。在持续充氢的过程中，溶液中的氢对外层氧化膜的影响起主导作用，不仅影响了氧化物的热稳定性，而且阻碍了外层氧化物的沉淀/沉积。将内部充氢过程中不同位置的氢的作用考虑进去，改进了混合导电模型（MCM），结合MCM对内部充氢材料氧化机理进行了深入讨论。研究结果对一回路水化学中溶解氢控制决策提供了实验依据和参照。