Corrosion and Hydrogen pick-up of Zr alloys in nuclear reactors

Zirconium alloys are the preferred fuel cladding material in light water nuclear reactors, and understanding the performance of the cladding materials is critical for safe, efficient operation. My DPhil work on zirconium alloys and its oxides made contributions to understanding the microstructure of the oxide scales formed/degraded under different conditions (autoclave, in-reactor or heavy-ion irradiated). Some key findings include the study of oxide porosity and its roles in hydrogen pick-up, the nature of phases present at the oxide/metal interface and their roles in controlling the overall corrosion behaviour and irradiation damage mechanisms in zirconium alloys and its oxides.