Cladding behaviour of high enrichment and high burn-up nuclear fuel: determining the impact of radiation damage on storage and corrosion.

在目前的英国核反应堆机群中，有一种经济驱动，即使用具有更高浓缩和燃耗的氧化铀燃料。较高的燃耗将导致不锈钢燃料棒内部发生物理变化，可能影响其在燃料循环结束时的性能，特别是在排放后的处理和最终处置在地质处置设施（GDF）之前的储存期间。由于燃料中少量锕系元素的α衰变，更高的燃耗将导致产生更多的氦，并且在存储期间产生更高的在线温度。在不锈钢系统中，这些温度可能足以使氦原子在中子诱导的缺陷团簇、位错和晶界处扩散和团聚，可能导致裂纹、表面剥落、硬化和解体等特征，特别是在放电后处理最终储存时。同样地，这些团块可能充当氢吸收体，这将在以后的生活中成为问题。这种影响在一定程度上也适用于锆合金复层。该项目的总体目标是加强我们对影响存储中包层行为的基本机制的理解，重点是英国系统。如果成功，该项目的总结将允许提出辐照不锈钢内气泡形成、气泡聚集和分布的模型。具体而言，本项目将：1。确定退役AGR材料放电后的微观结构，使用透射电镜和扫描电子显微镜（TEM和SEM），如果有的话，使用退役锆合金。2 .利用冷加工、离子注入和热退火来模拟放电后的微观组织，以产生非活性模拟样品；确定排放后微观结构对机械和腐蚀性能的影响，以评估高燃耗燃料对长期储存和处理的影响。

There is an economical drive to use uranium oxide fuel with higher enrichment and burn-up in the current UK nuclear reactor fleet. Higher burn-up will result in physical changes within stainless steel fuel rods, potentially impacting their performance at the end of the fuel cycle, specifically during post-discharge handling and storage prior to final disposal in a Geological disposal facility (GDF). Higher burn-up will result in larger quantities of helium generated due to the alpha-decay of the minor actinides in the fuel, and greater in-line temperatures during storage. Within the stainless steel system, these temperatures may be sufficient for helium atom diffusion and agglomeration at neutron-induced defect clusters, dislocations and grain boundaries, possibly resulting in features such as cracking, surface exfoliation, hardening and disintegration, especially during post-discharge handling for final storage. Similarly, these agglomerations may act as hydrogen getters, which will become problematic in later life. This affect may also be applicable to zircaloy cladding to some extent.The overarching aim of this project is to enhance our understanding of the fundamental mechanisms which influence cladding behaviour in storage, focusing on the UK system. If successful, the summation of the project will allow proposal of a model for bubble formation, bubble agglomeration and distribution within irradiated stainless steels. Specifically, this project will: 1. Determine post-discharge microstructures within ex-service AGR material, using transmission and scanning electron microscopies (TEM and SEM), and ex-service zircaloy if available.2. Simulate post-discharge microstructures using cold work, ion-implantation and thermal annealing to produce non-active analogue samples, 3. Determine the impact of post-discharge microstructure on mechanical and corrosion properties, to assess the impact of high burn-up fuels on long-term storage and handling.