Understanding the role of radiation damage on the environmental degradation of zirconium alloys is essential in providing cost effective and safe operation of nuclear fuel cladding. Since obtaining data from post-irradiation examinations is costly and infrequent, there is a drive to use ion irradiation as a surrogate to simulate aspects of in-service irradiation. In this study, ion irradiations have been used to generate specimens for corrosion testing in an autoclave. Multiple He+ ion irradiations at different energies have been used to generate a more uniform damage profile of ~10 dpa, to a depth of ~3 μm, in multiple specimens of recrystallisation annealed (RXA) Zircaloy-4. The radiation damage layer has been characterised using transmission electron microscopy (TEM) to investigate the nature of the damage and determine changes to the microstructure as a result of the ion irradiation. Autoclave
corrosion studies in simulated PWR chemistry have demonstrated that ion irradiation has significantly enhanced the corrosion rate compared to control specimens. Local oxide thickness measurements using Specular Reflectance Fourier Transform Infrared (SR-FTIR) spectroscopy have confirmed significantly thicker oxide on the irradiated face of the specimens compared to the non-irradiated face. Additional characterisation of accelerated oxide regions has been performed on selected samples
using a combination of focused ion beam (FIB) trenching and scanning electron microscopy (SEM) to investigate potential mechanisms of corrosion acceleration as a result of irradiation damage to the substrate, with the ultimate aim of correlating to in-reactor observations to better understand inservice corrosion behaviour.