The pellets of uranium oxide fissile fuel in pressurised water reactors are contained within tubes of zirconium alloy - the fuel clad. Cladding materials present a particular opportunity for research with rapid impact because they are among the few parts of a nuclear reactor that are replaced during its lifecycle.
Among other degradation processes, the zirconium alloy cladding undergoes oxidation in the high-temperature pressurised water in which it sits. A better understanding of the corrosion mechanism would allow for the design of safer, more efficient fuels. The pattern of corrosion involves repeated cycles of initially rapid then slower corrosion, before eventually a rapid breakaway phase, with linear kinetics, takes hold. Recent work at The University of Manchester suggests that oxidation proceeds very differently in the presence of irradiation to without. In particular, the oxide forms much more quickly, with smaller grains with more random orientations.
This modelling project will focus on disentangling the roles of oxide grain nucleation and growth in giving rise to differences in oxide growth rate and texture with and without irradiation. The student will develop a mesoscale phase-field model of oxide nucleation and growth on realistic length and time scales, and allow for the exploration of the various different processes involved in oxide growth. The chosen scale of the model will allow for direct comparison with experiment. The student will work alongside an experimental PhD student and build on previous work in the Zirconium team at Manchester.
CDT name: GREENCDT
The programme is funded by EPSRC, industrial partners and participating institutions