The prediction of nuclear fuel (UO 2) dissolution rates under waste disposal conditions

Abstract

Abstract An electrochemistry-based model to predict the dissolution of UO 2 fuel under both oxidizing and nonoxidizing conditions is presented and compared with other available models for fuel dissolution. Dissolution rates are predicted by extrapolating steady-state electrochemical currents for the anodic dissolution of UO 2 to the corrosion potentials measured in solutions containing various oxidants, including dissolved oxygen, hydrogen peroxide, and the products of the gamma or alpha radiolysis of water. Where possible, these predictions are compared with dissolution rates measured chemically and available in the literature. With a few exceptions, the agreement between our predictions and published rates is good. A threshold dissolution rate, below which the oxidative dissolution of UO 2 becomes negligible in comparison with the rate of chemically-controlled dissolution, is estimated to be∼2× 10 −4 μg cm −2 d −1. It is predicted that the rate for oxidative dissolution of used CANDU **fuel due to gamma radiolysis will fall below this threshold after ∼200 yr, a time period shorter than the anticipated lifetimes of titanium waste containers, which are expected to last for a period greater than ∼1200 yr. For dissolution due to alpha radiolysis, oxidative rates are uncertain, but could be above this threshold for a period of 500 to 10000 yr for CANDU fuel, and 500 to 30000 yr for light water reactor (PWR) fuel. The uncertainty in these ranges reflects the poor quality and limited number of corrosion potential measurements in the presence of alpha radiolysis. Areas of further research, necessary for the refinement of the UO 2 dissolution model, are outlined.