Irradiation-induced creep in UO2: The role of grain boundaries

The impact of defect generation during irradiation on grain boundary (GB) properties and associated transport phenomena lacks fundamental understanding, despite the influence GBs are known to exert over properties such as swelling, sintering, and creep. In this study, a cluster dynamics model is developed that predicts steadystate concentrations of defects at UO2 GBs by tracking the rate of point defect production and depletion, under irradiation conditions. Fast U interstitial self-diffusivity in bulk UO2 under irradiation results in their flux to GBs exceeding that of slower U vacancies. Incorporation of these interstitials at fission gas bubbles located at the GB results in their over-pressurization and cessation as effective sinks. As a consequence, U interstitials are predicted to be significantly enhanced in concentration under irradiation at UO2 GBs, and their fast mobility relative to defects in the bulk results in GB U self-diffusion orders of magnitude greater than that in the bulk. Creep controlled by diffusion of defects at GBs-Coble creep-is calculated using the predicted enhanced U self-diffusivity. The modeled creep rates compare favorably with experimental irradiation creep measurements, exhibiting the correct athermal behavior when GB defects are within a sink-limited regime. UO2 was studied here-the standard nuclear fuel-however, we expect the physical processes governing our results to extend beyond UO2.