辐照对奥氏体不锈钢在超临界水中的腐蚀行为影响研究

Supercritical water cooled reactors (SCWRs) have been designed to operate at high temperature (500~650℃) and high pressure （25~30MPa）, in which core internal materials will suffer extremely severe corrosion by supercritical water and severe neutron irradiation. The combination of these condidtions make the SCWR one of the most challenging reactors from the standpoint of materials selection. Zirconium alloys used in state-of-the-art light water reactors could not be used as fuel claddings in SCWRs because they could not meet the demand of high temperature strength and high corrosion resistance required for SCWRs. It is a key technique and scientific challenge to develope new type of core internal materials for the development of SCWRs.In this project, we conduct a systematic study on the ion irradiation damage behaviour of austenitic steels under high temperature and corrosion behaviour of irradiated austenitic steels in supercritical water. We shall study the effect of irradiation on corrosion behaviour of austenitic steels in supercritical water, and explore the mechanism of interaction between irradiation and corrosion, and, consequently, provide a scientific basis for the evaluation of the candidate materials used in SCWRs and the development of new materials.Up to now, the effect of irradiation on corrosion behavior of materials in supercritical water is rarely investigated in our country because of the equipment and technology limit. Irradiation was first introduced into the investigation of synergistic effects of irradiation and corrosion in supercritical water for nuclear materials in this study, which will promote the research progress of advanced nuclear materials in our country.

超临界水堆核电站的运行设计工况将为高温(500~650℃)、高压（25~30MPa）、强烈的超临界水腐蚀和强中子辐照等苛刻服役条件。原来用于轻水堆核电站的锆合金包壳管已无法满足超临界水堆高运行参数下的高温强度和耐蚀性要求，研究与开发新的燃料包壳材料成为发展超临界水堆的一个技术关键，也制约超临界水堆发展的一大科学挑战。本项目通过研究奥氏体不锈钢的高温离子辐照损伤行为和辐照对不锈钢材料在超临界水环境下的腐蚀行为的影响，获得辐照与超临界水腐蚀相互作用的机理，为超临界水堆候选材料的评估和新材料研发提供科学依据。目前国内由于设备和技术等原因关于辐照对材料在超临界水中腐蚀行为影响的研究极少，本项目首次将辐照引入到核电材料超临界水腐蚀领域，研究辐照与腐蚀的相互作用，将有力推进我国有关核电材料腐蚀方面的研究。

超临界堆严酷的服役环境对堆内材料的辐照和腐蚀性能提出了新的挑战。本项目开展了超临界水冷堆候选材料HR3C奥氏体不锈钢、718镍基合金和6XN奥氏体不锈钢的离子辐照损伤研究，并研究了辐照前后的AL-6XN奥氏体钢在高温高压水中的腐蚀行为以及辐照对腐蚀的影响。在对HR3C奥氏体不锈钢的辐照损伤研究中，第一次发现在位错环附近出现Cr23C6的析出，这与辐照下奥氏体不锈钢中常见的在晶界和位错环附近的Cr的贫化现象相反，我们从位错环与析出物形成顺序及基底中Cr含量的影响的角度阐述了这一现象的机理。在718镍基合金的离子辐照中，发现较低的剂量下基底中的超点阵结构就消失了，原有的有序r'和r''析出相被破坏，可导致合金屈服强度的降低。在AL-6XN奥氏体不锈钢的离子辐照研究中，发现高浓度的氢的滞留显著促进了位错环的长大，并使得晶界附近Cr的贫化区加宽，高浓度氢对奥氏体钢的辐照缺陷和微区化学成分进而对其力学能和腐蚀性能有重要影响。AL-6XN奥氏体钢在高温高压水中腐蚀时，表面产生对基底起保护作用的致密氧化膜。氢离子辐照产生的缺陷和辐照析出使氧化膜对基体的粘附性下降而极易发生脱落，导致了脱落区域基体被进一步腐蚀，使得辐照样品的腐蚀程度比未辐照样品更严重，且随着辐照剂量和温度的升高而加剧。应用辐照损伤速率理论,研究了质子辐照AL-6XN奥氏体钢产生的位错环的平均密度和尺寸随辐照温度和剂量的演化行为,从理论上对辐照缺陷的形核和长大规律进行了阐释。