690镍基合金/405不锈钢高温高压水缝隙腐蚀机理研究

The crevice between steam generator (SG) tube and support plate in secondary circuit in pressurized water reator (PWR) nuclear power plants (NPPs) is a key site of corrosion degradation of SG tube, where the environmentally assisted degradation issues frequently takes place during service. The present project first designs an artifical exposure testing device whose geometric size is accurately adjustable for crevice corrsoion in high temperature and high pressure water environments, and a compartmentalized cell consiting of large anode and small cathode for electrochemical tests in high temperature and high pressure water environments. The crevcie corrosion behaviors of SG tube material and support plate materials, namely, 690 nickel-base alloy and 405 stainless steel, are investeged jointly using exposure and electrochemical tests in high temeprature pressurized water, theoretical calculation of the distribution of water chemistry parameter and potential within the crevice, thermodynamic analyses and predictions of potential-pH relationship, in-situ Raman spectra measurements of crevcie corrosion prodcuts in high temperature pressurized water and ex-situ surface and microscopic analyses of crevice specimens after exposure tests. The effects of crevice-constructed materials, crevice geometry, water chemistry parameters, potential, temperature and time on characteristics of crevice corrosion are taken fully into account. The aim is to clarify crevice corrosion failure mechanism in high temperatue pressurized water and environmentally assisted degradation of SG tubes during service. The present work will be much helpful for finding the effective paths to resist crevice corrosion in high temperature pressurized water and ensuring the safe service of SG tubes in PWR NPPs.

压水堆（PWR）核电站二次侧蒸汽发生器（SG）传热管与支撑板之间的缝隙区域是SG传热管腐蚀损伤的关键位置，由此引发的环境失效问题在服役过程中频有发生。本项目以PWR核电站SG传热管用690镍基合金/支撑板用405不锈钢为研究对象，通过设计缝隙几何精确可调的高温高压水浸泡试验用人工缝隙模拟装置及大阴极、小阳极的电极分离的电化学测量系统，结合缝隙内水化学参数和电位分布的理论计算及高温高压水缝隙腐蚀产物相的热力学预测、缝隙腐蚀产物膜的原位光谱分析和非原位表面与微观分析验证，研究同材和异材构造的人工缝隙在模拟PWR核电站二次侧高温高压水中的缝隙腐蚀行为，考察缝隙构造材料与缝隙几何、水化学参数、电位、温度、时间等因素对缝隙腐蚀产物膜特征的影响规律，阐明高温高压水缝隙腐蚀损伤机制，为揭示PWR核电站SG传热管的环境失效机理，寻找有效的防护措施，保障其安全运行提供参考依据。

压水堆（PWR）核电站二次侧SG传热管与支撑板之间的缝隙区域是易发生缝隙腐蚀损伤的关键位置，由此引发的SG传热管环境失效问题在服役过程中频有发生。研究SG传热管/支撑板材料在高温高压水中的缝隙腐蚀行为与机理，澄清腐蚀产物膜的演化特征、稳定性及水化学的影响规律，至关重要。本项目以PWR核电站SG传热管用690镍基合金/支撑板用405不锈钢为研究对象，研制了适用于高温高压水环境中长期浸泡、稳定可靠且能精确调整缝隙几何尺寸的人工缝隙模拟装置；试验研究并澄清了690合金/690合金同种材料构造的人工缝隙及690合金/405不锈钢异种材料构造的人工缝隙在模拟PWR二次侧高温高压水中的腐蚀行为及缝隙几何、水化学、温度、时间等因素的影响规律；研制了高温高压水中电化学测量的大阴极、小阳极的电极分离系统及其原位测量技术，揭示了电极材料（690合金/405不锈钢）、水化学、温度等因素变化对缝隙腐蚀电化学行为的影响规律；发展了高温高压水中模拟缝隙试样的原位光谱测量技术，结合原位电化学测量及非原位表面和微观分析技术，揭示了690合金/405不锈钢模拟缝隙试样不同区域腐蚀产物膜的形貌、结构和成分特征；理论计算了模拟缝隙内局部水化学参数（如溶解氧、pH值等）分布及缝隙几何的影响规律；结合计算的多元合金-复杂水溶液体系的高温电位-pH图，揭示了690合金/690合金同材和690合金/405不锈钢异材缝隙的高温高压水缝隙腐蚀损伤机理。研究结果对深入理解PWR核电站SG传热管的环境腐蚀失效机理，寻找有效的在线监测与防护措施，保障关键设备的服役安全，有重要意义。相关研究成果已形成了核学会团体标准，为核电企业提供了设备材料缝隙腐蚀模拟测试与损伤评价的基准，2018年获国家技术发明二等奖和中国科学院科技促进发展奖。

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