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舰队中，有一种经济上的动力，即使用浓缩度和燃耗更高的氧化铀燃料。更高的燃耗将导致不锈钢燃料棒内部的物理变化，潜在地影响其在燃料循环结束时的性能，特别是在地质处置设施(GDF)最终处置之前的排放后处理和储存期间。更高的燃耗将导致更多的氦气产生，这是由于燃料中的次要辐系元素的阿尔法衰变，以及在储存过程中更高的在线温度。在不锈钢系统中，这些温度可能足以使氦原子在中子引起的缺陷团簇、位错和晶界上扩散和聚集，可能导致破裂、表面剥落、硬化和解体等特征，特别是在最后储存的放电后处理过程中。同样，这些凝聚体可能会起到吸氢剂的作用，这在以后的生活中会成为问题。这一影响也可能在一定程度上适用于锆合金覆层。本项目的总体目标是加强我们对影响覆层在储存中行为的基本机制的理解，重点是英国的体系。如果成功，该项目的总结将允许提出一个关于辐照不锈钢内气泡形成、气泡聚集和分布的模型。具体地说，本项目将：1.使用透射和扫描电子显微镜(透射电子显微镜和扫描电子显微镜)以及退役锆合金(如果可用)来确定退役AGR材料中的放电后微结构。用冷加工、离子注入和热退火等方法模拟放电后的微观组织，产生非活性的模拟样品；3.确定放电后的微观组织对力学性能和腐蚀性能的影响，评估高燃耗燃料对长期储存和搬运的影响。

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.