Corrosion Behavior of Pre-oxidized High Burnup Spent Fuel in Salt Brine
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&RUURVLRQ%HKDYLRURI3UHR[LGL]HG+LJK%XUQXS6SHQW)XHOLQ6DOW%ULQH Andreas Loida, Bernhard Kienzler, Horst Geckeis Forschungszentrum Karlsruhe, Institut für Nukleare Entsorgung, P.O.Box 3640, D-76021 Karlsruhe, Germany $%675$&7 During long-term interim storage of spent fuel, pre-oxidation of the UO2-matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered preoxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself. ,1752'8&7,21 In Germany it is intended to dispose of spent nuclear fuel in iron-based canisters. Rock salt is still one of the candidate host rock formations; clay or crystalline formations are presently under consideration, too. The overall safety of the repository relies on the protective effect of individual barriers such as the spent fuel matrix, the container, the backfill, the host rock and the overlying rock. They should prevent or retard brine contact with the waste and suppress mobilization and transport of radionuclides. To quantify the extent of radionuclide mobilization from a spent fuel waste package after an assumed contact with salt brine, corrosion tests with high burn-up spent fuel samples are being performed. These tests include different near-field materials, which are expected to be present in a repository. The overall alteration behavior of spent fuel will be controlled mainly by the availability and the consumption of oxidative radiolytic products, by the redox potential and by the pH of the brine. The oxidation state of the UO2 at the fuel surface also influences its alteration behavior. In the case of air contact prior to brine access, the UO2 matrix will be oxidized, resulting in the formation of easily-soluble surface layers. Assuming that long-term interim storage of spent fuel will be most likely, pre-oxidation of the UO2-matrix has to be considered, especially for access of small quantities of oxygen from air. The presumption of spent fuel element damage (cladding failure) resulted in studies of the leaching behavior of high burnup spent fuel rodlets with pre-set cladding defects in deionized water (DIW) [1]. It was shown that the instant release fraction (IRF) and the matrix dissolution rates were similar in comparison to the case if fuel pellets are used. However, in these tests the fuel was not pre-oxidized prior to water contact. Pre-oxidation of spent fuel due to cladding failure may generate a
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