Our reputation for zirconium technology expertise continues to grow, and we maintain a position as having the largest university-based Zr research group in the world.
In order to improve efficiency of nuclear reactors and minimise nuclear waste produced during operation, it is the desire to take these fuel assemblies to the highest possible burnup without compromising safety. This requires a complete and in the future more physical understanding of how the manufacturing route affects the performance of the clad and the mechanisms that determine the degradation during in-reactor performance.
The MPC now has the largest research group working in the field of zirconium alloys in the UK. Our interest ranges from manufacturing of zirconium cladding material and fuel channels to a number of performance aspects. The general aim of our research is to understand the underlying mechanisms responsible for the microstructure evolution during processing and degradation modes during performance. Zirconium has a hexagonal close packed crystal structure, which makes the understanding of deformation mechanisms and the aim of predicting microstructure evolution during processing particularly challenging.
One of our current research efforts is to develop crystal plasticity models that will be able to predict texture evolution during processing. On the performance side, the MPC undertakes very significant research related to understanding corrosion and hydrogen pick-up mechanisms, hydride precipitation and irradiation growth and creep.
Irradiation-induced precipitation has been observed in a Zr-0.1Fe alloy and Zircaloy-2 in close proximity to Fe-containing secondary phase particles (SPP). Fe redistributes from said particles and precipitates along the trace of the basal plane.
However, beyond 500 nm from an SPP no Fe precipitation has been found. More irradiation-induced precipitation was observed in Zircaloy-2 than in the Zr-0.1Fe alloy. This increased irradiation-induced precipitation has been correlated to a decrease in dislocation line density. Therefore it is proposed that the irradiation-induced precipitates act as annihilation sites for point defects, resulting in fewer point defects forming dislocation loops. If this hypothesis is correct it could be vital in influencing future alloy development.
The pellet-cladding interaction (PCI) – an iodine stress corrosion cracking failure of fuel rods – is a key-limiting factor in load-following ability. The search for a mechanistic understanding of the phenomenon is a current industrial priority and over the last 12-18 months Manchester has built a significant strength in this area. We have a PDRA and 4 PhD students working on developing experimental apparatus to explore PCI and on modelling-based studies.
The demands of the next generation of nuclear power plants have made higher strength, dual phase a+ß Zr alloys attractive candidate materials. Microstructure control during processing will be paramount to the success of these materials. We have studied the microstructure evolution in these alloys during thermomechanical processing. We have analysed material from controlled rolling trials via neutron diffraction texture measurements and detailed EBSD microtexture analysis. This is complemented by in-situ loading synchrotron experiments at high strain rates. Recent highlights include the discovery of a new alpha a variant selection mechanism in these materials, driven by anisotropic ß grain breakup.
In-reactor conditions, including water chemistry and irradiation, are known to have a significant effect on the corrosion of fuel clad, but systematic investigation has thus far been focussed on understanding ex-situ corrosion and hydrogen pick-up. We have made strong progress in applying characterisation techniques honed in the MUZIC-2 programme to samples of both in-reactor and experimentally irradiated alloys.
Apparatus suitable for concurrent oxidation and proton irradiation are being developed at the Dalton Cumbrian Facility and will be used to undertake controlled investigation of the effect of irradiation damage on oxide growth and microstructure.
This project has evolved into a study of friction stir welds for advanced reactor applications, which is an essential aspect for future component fabrication using ODS materials. This work has explored the development of residual stresses and microstructures developed during the FSW process.
This study involves an international collaboration with CIEMAT (Madrid).