辐照对奥氏体不锈钢在超临界水中的腐蚀行为影响研究

项目名称： 辐照对奥氏体不锈钢在超临界水中的腐蚀行为影响研究

项目编号： No.11275140

项目类型： 面上项目

立项/批准年度： 2013

项目学科： 数理科学和化学

项目作者： 郭立平

作者单位： 武汉大学

项目金额： 86万元

中文摘要： 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。

中文关键词： 辐照损伤；奥氏体不锈钢；位错环；腐蚀；超临界水冷堆

英文摘要： 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

英文关键词： Irradiation damage；Austenitic steels；Dislocation loops；Corrosion；Supercritical-water-cooled reactors