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中，使用具有更高浓缩度和燃耗的氧化铀燃料具有经济性。更高的燃耗将导致不锈钢燃料棒内的物理变化，可能影响其在燃料循环结束时的性能，特别是在最终处置在地质处置设施（GDF）之前的排放后处理和储存期间。较高的燃耗将导致由于燃料中次锕系元素的α衰变而产生的大量氦，以及储存期间较高的管线温度。在不锈钢系统内，这些温度可能足以使氦原子在中子诱导的缺陷簇、位错和晶界处扩散和聚集，可能导致诸如开裂、表面剥落、硬化和分解的特征，特别是在用于最终储存的放电后处理期间。类似地，这些聚集体可能充当氢吸气剂，这将在以后的生活中成为问题。这种影响可能也适用于铝合金包层在某种程度上，这个项目的首要目标是提高我们的理解的基本机制，影响包层的行为在存储中，集中在英国的系统。如果成功的话，该项目的总结将允许提出一个模型的气泡形成，气泡团聚和分布在辐照不锈钢。具体而言，该项目将：1。使用透射和扫描电子显微镜（TEM和SEM）以及退役耐热合金（如果可用）确定退役AGR材料内的放电后微观结构。使用冷加工、离子注入和热退火来模拟放电后的微结构，以产生非活性模拟样品，3。确定放电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.