Corrosion Behavior of Pre-oxidized High Burnup Spent Fuel in Salt Brine

Abstract

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 pre-oxidation 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.

Radionuclide release from high burnup spent fuel during corrosion in salt brine in the presence of hydrogen overpressure

“…The amount of total matrix dissolution under externally applied H 2 overpressure was determined to be only slightly higher as found, when corroding pellets K3 and K4 in the presence of Fe powder. The significant higher Sr release obtained during the pre leaching procedure of pellet K8 under initial Ar-atmosphere during 293 days compared to results of tests with pellets K9 and K10 can be explained by the fact that pellet K8 suffered pre-oxidation in part due to air penetration into the storage capsules [11]. In contrast, pellets K9 and K10 were used in the respective experiments immediately after cutting in N 2 atmosphere.…”