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

在过去的几年里，许多发达国家和发展中国家开始了核材料研究的新阶段，因为核能被视为在减少二氧化碳排放和确保能源供应方面发挥着至关重要的作用。人们认识到，虽然可再生能源的供应将会增加，但只有维持或扩大目前的核能发电能力百分比，才能在英国和其他地方实现能源安全和减少碳排放的需要。全世界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(通常被称为第三代+反应堆)被开发出来，以降低建设成本和提高运行效率。当欣克利角核电站开工建设时，这种类型的反应堆将成为25年来英国安装的第一座新发电厂。先进的沸水反应堆也在考虑用于其他地点，如Wylfa。这两个反应堆设计的燃料组件都是基于锆合金的，提高未来反应堆效率的主要驱动因素之一是设计燃料在更严格的燃料占空比下运行，包括更长的堆芯停留时间，以实现更高的燃尽率，从而增加从铀燃料中提取的能量。燃料供应商已经通过引入新的锆基合金来满足高性能燃料的需求，以满足对更耐腐蚀的覆层材料的需求。然而，这种合金的发展几乎完全建立在经验研究的基础上，而不是任何控制氧化速度的一般模型--特别是在辐射下。从商业运营商的角度来看，这一理解尤其重要，因为只有对实际使用条件下的腐蚀过程进行基于物理的预测，才能使他们有信心将燃料组件操作到降低反应堆寿命成本和将每千瓦时产生的核废料最小化所需的非常高的燃耗程度。我们在先前关于锆氧化的工作中为了解在高压灭菌条件下形成的氧化膜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，以及在控制不同合金的腐蚀速度在特定时间突然加速的转变行为中可能采取的关键步骤。这包括开发新的样品制备技术、新的分析方案，并报告在氧化过程的不同阶段决定速度的金属/氧化物界面上正在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，在这些环境中，中子损伤可能会改变许多关键的机制，目的是建议如何设计这些合金，使其在使用中具有更好的降解性能。

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.