The corrosion of nuclear fuel (UO 2) in oxygenated solutions

The corrosion mechanism of UO 2 (nuclear fuel) has been studied in 0.1 mol l −1 sodium perchlorate (pH = 9.5), with and without added sodium carbonate. The corrosion potential was followed for various exposure times. Subsequently, the electrode was either subjected to a cathodic stripping scan from the corrosion potential to −2.0 V to determine the presence and measure the thickness of surface films formed; or removed and examined by X-ray photo-electron spectroscopy to determine the composition of the electrode surface. In both perchlorate and perchlorate plus carbonate solutions two films were formed on the UO 2 prior to the establishment of steady-state dissolution conditions. A layer of UO 2.33 was formed over the first 10 h of exposure. The outer layers of this film slowly converted to hydrated UO 3 (or uranyl carbonate) over the next ∼90 h. This conversion appeared to be concentrated at the grain boundaries. Corrosion rates were measured by extrapolating from the Tafel region for steady-state anodic dissolution to the corrosion potential. The corrosion process appears to be controlled by the kinetics of the anodic dissolution step.