Effect of cement water on UO 2 solubility
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Effect of cement water on UO2 solubility C.Cachoir1, Th. Mennecart1, and K. Lemmens1 SCK•CEN, Boeretang 200, B-2400 Mol, Belgium
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ABSTRACT To assess the stability of spent fuel in the highly alkaline chemical environment of the Belgian Supercontainer design, static leach experiments were performed with depleted UO2 and 238Pu-doped UO2 at different SA/V ratios for 1.5 years in cement waters (11.7< pH < 13.5) at ambient temperature and under argon atmosphere. The influence of the calcium concentration on the uranium release was also investigated. While the ultrafiltered U(IV) concentration was 10-9-10-8 mol.L-1 and independent of the pH, the U(VI) release from the UO2 surface was enhanced by the OH- concentration, leading to soluble U concentrations up to 10-5 mol.L-1 at high SA/V and under the influence of radiolysis. Together with the high Ca concentration, this can lead to the formation of Ca-U(VI) colloids as precursor of Ca-U(VI) secondary phases, decreasing the soluble U concentration. The precipitation of Ca-U secondary phases was however not clearly evidenced by surface analyses. INTRODUCTION The Supercontainer design is the current reference design for the geological disposal of spent nuclear fuel in Belgium [1]. In the case of spent UOX fuel, four assemblies are placed inside a Supercontainer with a 30 mm thick carbon steel overpack and a 540 mm thick concrete buffer. The Supercontainer will thus provide a highly alkaline chemical environment. After perforation of the overpack, the high pH of the infiltrating water may have an impact on the radionuclide release from the spent fuel. As the majority of published data related to radionuclide release is reported at neutral pH [2-3], a research program was started at the Belgian Nuclear Research Centre (SCK•CEN), supported by ONDRAF/NIRAS, to evaluate the stability of UO2, as analogue of real spent fuel, in such alkaline environment. Our experimental program was defined to determine the UO2 dissolution rate, the UO2 solubility and the influence of α-radiation on UO2 behavior. Complementary experiments with fresh spent fuel (SF) were done by KIT-INE [4, 5]. The current paper summarizes the results of static experiments with depleted UO2 and Pu-doped UO2 performed at SCK•CEN to study the effect of high pH on the UO2 solubility. EXPERIMENTAL All experiments were performed at 25 – 30°C under Ar atmosphere with pO2 below 1 ppm and pCO2 below 0.1 ppm. Static experiments were performed with depleted and 238Pu-doped UO2 powders (table I), previously annealed (Argon/5% H2 gas at 1000 °C), at four different ratios of fuel surface area to leachant volume (SA/V of 6, 17, 130 and 257 m-1) and for more than one year with and without metallic iron as redox buffer. Three different cement waters were used for the experiments (table II), i.e. Young Cement Water with Calcium (YCWCa-pH 13.5-[Ca] 6.5u10-4 mol.L-1), Evolved Cement Water (ECW-pH 12.5-[Ca] 1.3u10-2 mol.L-1) and Old Cement Water (OCW-pH 11.7-[Ca] 7.3u10-4 mol.L-1). As Ca may strongly interact with uranium [4-7], a var
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