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

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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.