SIMFUEL and UO 2 Solubility and Leaching Behavior Under Anoxic Conditions
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ABSTRACT Most of performance assessment models for spent fuel repository safety consider radiolysis-self-oxidation to describe fuel matrix release. Nevertheless, due to radioactive decay, the matrix dissolution process under reducing conditions would be controlled by the solubility limit of the steady U solid phase. In this work, leaching behaviour under anoxic and reducing conditions of spent fuel unirradiated chemical analogues (natural U0 2 and SIMFUEL) in simulated groundwater is studied. The trial procedure was performed taking into account the possibility that the uranium oxide to be leachated had an initial outer layer with an oxidation state higher than the matrix. This oxidised layer would produce an overestimation on U concentration in solution for the solid studied. In order to avoid this effect, a complete replacement of the leaching solutions was carried out after several days of experimentation. After this initial experimental step, the steady state concentrations obtained in all tests were more than one order of magnitude lower than before. Uranium concentrations found in reducing and anoxic experiments for both U0 2 and SIMFUEL tests were very close. This fact is attributed to similarity in environmental conditions (pH, Eh, etc.). From that, it can be assured that steady state concentration obtained is independent of solid leached (U0 2 or SIMFUEL). In order to assess which is the solid phase that could control the solubility of U, the experimental concentration obtained was compared with results from geochemical code EQ3/6. According with the theoretical calculations U40 9 would be the controlling pure phase formed in whole experimental tests described in this work. Comparisons with bibliography data from leaching experiments of spent nuclear fuel were made as well. INTRODUCTION Spent fuel performance assessment requires thorough evaluation of its long-term ability to isolate and immobilise individual radionuclides in solid phase upon groundwater contact. In order to demonstrate the robust performance of the engineered barrier is necessary to know the waste behaviour under repository conditions. The first mechanism to define is the radionuclide release from spent fuel by groundwater's dissolution and also the retention factor in case of long lasting spent fuel and water contact. Other factors that are necessary to consider are: dissolution of gap and grain boundary inventories of segregated radionuclides; radiolytic production of oxidant; matrix dissolution; sorption of radionuclides on surface of solid phases, formation of secondary solid alteration products. Other point that it is necessary for performance assessment is that the repository is not a static system. It will continue evolve after waste storage. Considering them, the chemical environment (pH, Eh or pe, dissolution ionic strength, etc.) will become the main controlling factor of radionuclide release. 247
Mat. Res. Soc. Symp. Proc. Vol. 506 01998 Materials Research Society
The study of the behaviour of unirradiated U-based materials
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