Corrosion and hydrogen pick-up mechanisms in zirconium nuclear fuel cladding alloys in active environments

在过去几年中，许多发达国家和发展中国家开始了核材料研究的一个新阶段，因为人们认为核能在减少二氧化碳排放和确保能源供应方面起着至关重要的作用。人们认识到，虽然可再生能源的供应将会增加，但只有保持或扩大目前核能发电能力的比例，才能实现英国和其他地方对能源安全和碳减排的需求。世界上最常见的裂变反应堆类型是压水堆（PWR）。在过去的20年里，改进的压水堆设计已经开发出来（通常称为3+反应堆），以降低建设成本和提高运行效率。当欣克利角核电站开始建设时，这种类型的反应堆将成为英国25年来安装的第一座新发电厂。先进的沸水反应堆也在考虑用于威尔法等其他核电站。这两种反应堆设计的燃料组件都基于锆合金，提高未来反应堆效率的主要驱动因素之一是设计在更严格的燃料工作循环下运行的燃料，包括更长的堆芯停留时间，以实现更高的燃耗分数，从而增加从铀燃料中提取的能量。为了满足高性能燃料的需求，燃料供应商纷纷推出新型锆基合金，以满足对耐腐蚀覆层材料的需求。然而，这种合金的发展几乎完全基于经验研究，而不是任何控制氧化速率的一般模型-特别是在辐照下。从商业运营商的角度来看，这种理解尤为重要，因为只有在实际服务条件下对腐蚀过程进行物理预测，才能使他们有信心将燃料组件运行到非常高的燃点，从而降低反应堆的使用寿命成本，并最大限度地减少每GWh电力产生的核废料量。我们在之前的氧化锆研究工作中为理解在高压灭菌器条件下形成的氧化层的微观结构做出了贡献，首次使用最新一代的分析技术来回答长期存在的问题，即氧化物/金属界面上存在的相的性质，以及控制不同合金在特定时间腐蚀速率突然加速的转变行为的关键步骤。这涉及开发新的样品制备技术，新的分析方案，并报告了在氧化过程的不同阶段，在决定速率的金属/氧化物界面上发生的一些最详细的分析。我们现在打算将这些技术应用于来自真实反应堆环境的锆样品，在这些环境中，中子损伤可能会改变许多关键机制，目的是建议如何设计这些合金，以在使用中提供更好的降解性能。

In the past few years a new phase of nuclear materials research has started in many developed and developing countries since nuclear power is seen to play a vital role in reducing CO2 emission and securing energy supply. It is recognized that while the supply of energy from renewable sources will increase, the need for both energy security and carbon emission reduction in the UK and elsewhere will only be achieved if the current percentage of nuclear generation capacity is maintained or expanded. The most common fission reactor type worldwide is the pressurised water reactor (PWR). Over the last 2 decades improved PWR designs have been developed (often termed Gen 3+ reactors) to reduce the construction costs and improve operating efficiency. It is reactors of this type that will be the first new generating plant installed in the UK for 25 years when construction begins at Hinkley. Advanced Boiling Water Reactors are also under consideration for other sites like Wylfa. The fuel assemblies for both these reactor designs are based on zirconium alloys, and one of the main drivers to improve the efficiency of future reactors is to design fuel to operate under more severe fuel duty cycles, including longer in-core residence times to achieve higher burn-up fraction and so increase the energy extracted from the uranium fuel. Fuel vendors have responded to the need of more corrosion resistant cladding material by introducing new zirconium based alloys to meet the needs of high performance fuel. However, this alloy development has been based almost wholly on empirical research rather than any general model of what controls the oxidation rate - especially under irradiation. This understanding is particularly important from the point of view of the commercial operators, since only physically-based predictions of the corrosion process under real service conditions will give them the confidence to operate fuel assemblies to the very high degrees of burn-up needed to reduce the life-time costs of a reactor and minimise the volume of nuclear waste generated per GWh of power.We have in previous work on zirconium oxidation made contributions to understanding the microstructure of the oxide scales formed under autoclave conditions, using for the first time the latest generation of analytical techniques to answer longstanding questions on the nature of the phases present at the oxide/metal interface and what the critical steps might be in controlling the transition behaviour where the corrosion rate accelerates abruptly at specific times for different alloys. This involved the development of new sample preparation techniques, new analytical protocols and reporting some of the most detailed analyses of what is happening at the rate-determining metal/oxide interface at different stages of the oxidation process. It is these techniques that we now intend to apply to zirconium samples from real reactor environments where neutron damage may alter many of the key mechanisms, with the aim of suggesting how these alloys can be designed to give better degradation performance in service.