In-situ loading during irradiation of materials for fusion applications

材料将在聚变反应堆内经历的条件是严峻的，包括高温、负荷和辐照。为了成功地设计和建造一个商业上可行的聚变反应堆，我们必须很好地了解我们使用的材料在服役期间的性能。不幸的是，我们对结构合金在热、负荷和辐照条件下的行为的理解目前是有限的，并且对改变这一点有相当大的兴趣。该项目旨在开发和利用新的先进实验能力，以提高我们对聚变反应堆中合金行为的理解。该项目将结合曼彻斯特道尔顿坎布里亚设施（DCF）质子和/或离子辐照期间的原位机械载荷和辐照后的高分辨率表征。该项目将开发辐照光束线上的机械钻机能力，并使用世界领先的显微镜、微观结构分析和原位高分辨率数字图像相关（HRDIC）设备作为一种新的变形映射技术，见图1。这将评估辐照对材料力学的影响，为STEP的工程设计提供数据帮助。通过改变辐照条件——剂量、温度和应变速率——将有可能演示核聚变反应堆内部的不同辐照条件如何影响材料性能，从而影响组件寿命。此前，HRDIC技术已被证明在量化未辐照和辐照锆合金在低损伤水平[1]下的变形差异方面非常有效。将这种分析技术与原位辐照-机械载荷相结合将对该领域做出重大贡献，并使人们能够更详细地了解辐照诱导效应的机制，有助于改进材料-优化它们用于聚变-并验证材料性能模型，以便这些模型可以用作聚变发电厂恶劣环境的预测工具。博士学位将成为UKAEA能源生产球形托卡马克（STEP）计划的一部分，该计划旨在设计和建造一个原型聚变能源工厂。该博士项目将有助于评估材料在聚变过程中通常经历的恶劣环境下的性能，并为UKAEA的实验和建模活动提供信息。可能应用于聚变反应堆的候选材料包括马氏体钢、Cu-Cr-Zr合金和钨合金。这些材料的性能取决于其机械性能如何受到聚变反应堆极端操作条件的影响，特别是辐照过程中蠕变和疲劳的影响。

The conditions materials will experience inside a fusion reactor are severe, combining high temperatures, loads and irradiation. In order to successfully design and construct a commercially-viable fusion reactor, we must have a good understanding of how the materials we use will perform during service. Unfortunately, our understanding of how structural alloys behave in conditions combining heat, load and irradiation is somewhat limited at present, and there in considerable interest in changing this. This project will aim to develop and utilise new advanced experimental capabilities in order to improve our understanding of the behaviours of alloys in fusion reactors.The project will combine in-situ mechanical loading during proton and/or ion irradiation at Manchester's Dalton Cumbrian Facility (DCF) with high resolution characterisation post-irradiation. The project will develop mechanical rig capability on the irradiation beamline and use world-leading facilities for microscopy, microstructural analysis analysis and in-situ High Resolution Digital Image Correlation (HRDIC) as a novel deformation mapping technique, see Fig. 1. This will assess the effect of irradiation on material mechanics, providing data to aid in engineering design for STEP. By varying irradiation conditions - dose, temperature and strain rate - it will be possible to demonstrate how the variable irradiation conditions inside a fusion reactor affect material performance and therefore component lifetime. Previously, the HRDIC technique has proven highly effective in quantifying the differences in deformation between non-irradiated and irradiated zirconium alloys subjected to low damage levels [1]. Combining this analysis technique with in-situ irradiation-mechanical loading will make a significant contribution to the field and enable a more detailed understanding of the mechanisms of irradiation-induced effects, helping to improve materials - optimising them for fusion - and to validate models of material performance so that the models can be used as predictive tools for the harsh environment of a fusion power plant. The PhD will be part of UKAEA's Spherical Tokamak for Energy Production (STEP) programme, which has been created to design and construct a prototype fusion energy plant. This PhD project will help inform the assessment of material performance when subjected to the harsh environments typically experienced during the fusion process, feeding into the experimental and modelling activities at UKAEA. Likely candidate materials for application in fusion reactors include martensitic steels, Cu-Cr-Zr alloys and Tungsten alloys. The performance of these materials depends how their mechanical properties are affected by the extreme operating conditions in a fusion reactor, especially the impact of creep and fatigue during the irradiation process.