Release of Radiotoxic Elements from High Burn-Up UO2 and MOX Fuel in a Repository

Abstract

In a spent fuel repository the processes that govern the release of radionuclides are dissolution and transport in a possible groundwater flow. The cladding will be the last barrier before the water comes into contact with the fuel, namely with the outer rim of the pellet. Here the heterogeneity of the material due to the irradiation process is responsible for a complex release process. Fission products and minor actinides inventories are considerably higher at the pellet periphery as a result of increased epithermal neutron capture and of migration in the case of the volatile fission products.The present paper gives a review of experimental activities at the Institute for Transuranium Elements (ITU). Both single effects studies and integral tests are carried out to study the behavior of spent fuel under storage conditions.Leaching of irradiated UO2 (up to 50 GWd/tU) and MOX (up to 25 GWd/tU) fuel rods with preset cladding defects at 100°C under anoxic or reducing conditions should simulate the realistic case of groundwater coming into contact with a spent nuclear fuel repository. For all main radionuclides the release process can be described considering a two-step dissolution mechanism that includes the initial dissolution of an oxidized layer present on the fuel surface followed by a long-term oxidative matrix dissolution. By means of α-doped (238Pu) UO2 it could be demonstrated, that radiolysis has a significant influence on this dissolution. Especially high initial release rates were found for the volatile cesium and iodine for the reasons mentioned above.Besides the conventional leaching experiments electrochemical techniques are used to investigate for instance the complex corrosion behavior of the heterogeneous MOX fuel materials or the influence of α-radiolysis on spent fuel dissolution.In the integral tests mentioned above with large S/V values, reprecipitation of U is likely to happen. Therefore special dynamic test are conducted where this reprecipitation is prohibited and true U solubility can be determined.Thin layer of UO2 and (U,Pu)O2 doped with various fission products and minor actinides are prepared to study the influence of these elements on the matrix dissolution. When Cs is for instance co-deposited, the U oxidation state changes from U4+ to U6+ for the same O2 pressure possibly indicating a stable Cs uranate. This could be an indirect proof of the existence of such a species in irradiated fuel (e.g. at the grain boundaries).

Corrosion of irradiated MOX fuel in presence of dissolved H2

“…A few matrix corrosion studies under oxidizing conditions have been published for MOX fuels [23][24][25][26]. Since oxidative dissolution takes place in these studies, the release of matrix elements U and Pu is relatively high; however, during long term static leaching experiments, it becomes limited by the solubility of secondary phases formed during the test, thus making it complicated to determine true matrix dissolution rates.…”