The effect of irradiation on the mechanical properties of nuclear fuel cladding and insulating materials for the Gen. IV pressure tube super critical water reactor

与拟议的第四代压力管式超临界水反应堆(SCWR)相关的主要挑战之一是选择合适的核燃料包壳和隔热材料，使其能够在超过750℃的超临界水环境中承受高水平的中子辐照。领先的候选燃料包层材料是微结构改性的奥氏体不锈钢合金，而领先的候选绝缘材料是氧化钇稳定的氧化锆。到目前为止，还不知道长时间的高温中子照射对这些材料的机械性能的影响。拟议的研究项目将通过在温度高达700摄氏度的情况下对材料进行离子辐照来获得这些数据，作为模拟高温中子辐照的一种手段。使用电子显微镜对离子辐照材料进行微观机械测试和微观结构分析，将首次确定这些材料的强度、延展性和微观结构稳定性对辐照剂量和辐照温度的依赖关系。这些数据对于为拟议的第四代SCWR进行详细的材料选择是绝对必要的。

One of the primary challenges associated with the proposed Generation IV pressure tube type SuperCritical Water Reactor (SCWR) is the selection of suitable nucear fuel cladding and thermal insulating material capable of enduring exposure to high levels of neutron irradiation in a supercritical water environment at temperature in excess of 750C. The leading candidate fuel cladding materials are microstructurally-modified austenitic stainless steel alloys while the leading candidate insulating material is an yttrium-stabilized zirconia. To date nothing is known of the effect of prolonged high-temperature neutron irradiation on the mechanical properties of these materials. The proposed research project will obtain such data by performing ion-irradiation on the materials at temperature up to 700C as a means of simulating high-temperature neutron irradiation. Micro-mechanical tests and microstructural analyses, using electron microscopy, of the ion-irradiated material will allow determination, for the first time, of the dependence of the strength, ductility, and microstructural stability of these materials upon irradiation dosage and irradiation temperature. These data are absolutly essential for proceeding with detailed materials selection for the proposed Gen. IV SCWR.