Understanding Radiation Damage in Zircaloy-4 for Current and Future Nuclear Applications

锆-4是一种锆基合金，对压水反应堆的核燃料包壳非常重要。在高温辐照下，它会产生包括位错环和气泡在内的微观结构缺陷。氢化物也可能由于水腐蚀而形成。这种缺陷可能对材料的性能产生重大影响，因此需要准确量化，以便继续使用和未来开发将这种合金纳入其设计的反应堆技术。这个博士项目将支持与罗尔斯·罗伊斯公司（将提供锆合金-4样品）的初步合作，该合作来自地平线2020“IL TROVATORE”项目。

Zircaloy-4 is a Zr-based alloy of major importance for nuclear fuel cladding in pressurised water reactors. Under irradiation at high temperatures, it develops microstructural defects including dislocation loops and gas bubbles. Hydrides can also develop due to aqueous corrosion. Such defects can have significant effects on the performance of the material and hence needs to be accurately quantified for the continued use and future development of reactor technologies incorporating such alloys into their design. This PhD project will support a fledging collaboration with Rolls Royce (who will provide Zircaloy-4 samples) which has emerged from the Horizon 2020 "IL TROVATORE" project.