The dissolution of unirradiated UO 2 fuel pellets under simulated disposal conditions

Abstract

The rate of dissolution of unirradiated UO 2 fuel pellets under mildly oxidizing conditions has been investigated. A static leaching procedure was used, and the extent of dissolution was monitored by radiochemical analysis for uranium in the leachate. At 70 °C, the average dissolution rate is ∼5 × 10 −6 kg·m −2·day −1, being smaller in granite groundwater than in distilled deionized water. The rate exhibits a first order dependence on the oxygen content ( P O 2 = 25–75 kPa) and on the carbonate and bicarbonate concentrations (< 10 −3 mol/L) of the leachant. The rate of dissolution increases with increasing temperature (30–90 °C) in distilled deionized water, but shows the opposite behavior in granite groundwater. The rate increases with decreasing pH or pOH, and is minimal in the pH range 5–10. These data will aid in the characterization of used UO 2, fuel as a waste form, and in establishing reliable source terms at the vault-geosphere interface of a geologic nuclear fuel waste disposal vault.