UO 2 dissolution in high pH conditions of the Belgian Supercontainer
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UO2 dissolution in high pH conditions of the Belgian Supercontainer Th. Mennecart1, C. Cachoir1 and K. Lemmens1 1 SCK•CEN, Boeretang 200, B-2400 Mol, Belgium ABSTRACT To assess the long-term behavior of spent fuel in alkaline conditions representative for the Belgian Supercontainer design, static and dynamic dissolution tests were performed with depleted and Pu-doped UO2 , simulating medium burn-up UOX fuels of different fuel ages. The experiments were performed under argon atmosphere at 25 – 30°C in cement waters in the pH range 11.7 – 13.5 and at different SA/V ratios. This paper presents the observed UO2 matrix dissolution rates based on the (238U or 233U) release, and proposes a selection of reference dissolution rates for performance assessment. We demonstrate that the dissolution rates at high pH are equivalent to the dissolution rates reported in the literature for neutral pH conditions. The α-activity threshold below which radiolytical fuel oxidation becomes negligible, seems to be close to the threshold reported for anoxic media at neutral pH. 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 carbon steel overpack and a concrete buffer of Ordinary Portland Cement and limestone aggregates. The overpack has to prevent contact of the spent fuel with the cementitious pore water at least during the thermal phase (~2000 years). After perforation of the overpack, the high pH of the infiltrating water may have an impact on the radionuclide release from the spent fuel. Because the majority of published data related to RN release is reported at neutral pH [2, 3], a research program was started to assess the UO2 matrix dissolution rate in a highly alkaline environment. This paper summarizes the results of experiments with depleted and Pu-doped UO2 performed at the Belgian Nuclear Research Centre (SCK•CEN). EXPERIMENTAL DETAILS All experiments by SCKxCEN were performed with depleted UO2 and Pu-doped UO2 at 25 – 30°C under Ar atmosphere with pO2 below 1 ppm and pCO2 below 0.1 ppm. Complementary experiments with fresh spent fuel (SF) were done by KIT-INE, and reported in [4, 5]. Pu-doped fuels (F1, F2 and F4) were manufactured with 238Pu to generate the α-dose and with 233U (0.3%) as corrosion indicator [6]. The α-activity was respectively 244, 36 and 17 MBq.g-1 UO2 for fuel F1, F2 and F4, simulating a medium burn-up UOX fuel with ages of 150, 2000, and 11000 years. Undoped depleted UO2 is considered as a surrogate of very old spent fuel, although the α-activity of depleted UO2 (~0.01 MBq.g-1) is about 200 times lower than the expected α -activity of fuel 106 years after discharge. The experiments were performed with UO2 powder with a grain size of 100 – 200 μm and 50 – 100 μm for depleted UO2 and F1-doped fuel, and with a grain size of less than 100 μm for the doped fuels F2 and F4. The chemical evolution of the concrete pore water after contact w
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