Radiolytic Corrosion of 238 Pu-doped UO 2 Pellets in 5 M NaCl Solution
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Radiolytic Corrosion of 238Pu-doped UO2 Pellets in 5 M NaCl Solution M. Kelm and E. Bohnert Institut für Nukleare Entsorgung, Forschungszentrum Karlsruhe Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen, Germany e-mail: [email protected] ABSTRACT Deaerated 5 M NaCl solution is alpha-irradiated up to 700 days in the presence of UO2 pellets. Experiments are conducted with 238Pu-doped pellets in pure brine and with undoped UO2 pellets in a 238Pu containing brine. The long-lived radiolysis products H2, O2 and ClO3- are formed in all cases proportional to the dose applied on the solution and with yields corresponding to 200, 80 and 7 nMol/(L*Gy). The U concentrations increase with time but do not exceed about 10-4 Mol/L thus indicating Uranium to be solubility limited. Moreover, the rinse solutions of the vessels at the end of the experiments contain up to one order of magnitude more U than found in the solutions. The total amounts of mobilized U deviate by less than a factor 10 from each other regardless of the large ratio of surface dose rates (ratio up to 2300) or mean dose rates (ratio up to 100), applied while the reference experiment (UO2 pellet in pure brine) yields only some 10-7 Mol/L U which is over 3 orders of magnitude lower than observed in experiments with radiation present. The simulation of the radiation chemical processes using a kinetic reaction model can reasonably reproduce the findings of the experiments with respect to the formation of H2 and O2 and give the order of magnitude for the concentrations of chlorine species and of oxidized UO2. INTRODUCTION In Germany spent nuclear fuel is supposed to be buried in the deep underground in a final disposal in rock salt. As an accident scenario, intrusion of brine to the disposal is considered. The failure of container and fuel rods by corrosion is assumed to occur 500 years after disposal at the earliest. At this time alpha radiation from the spent fuel will be dominant over beta/gamma and the geological surrounding and metals from the container material etc. will have established reducing conditions in the brine coming into contact with the fuel. As the UO2 matrix of spent fuel has a very low solubility under reducing conditions, and is therefore considered resistant to corrosion, the fuel should strongly limit the release of fission products under these conditions. However, brine in contact with spent fuel could produce sufficient radiolytically formed oxidants to convert tetravalent U in the UO2 fuel into hexavalent U, thereby degrading the UO2 matrix and releasing radionuclides. Many works about the radiolytical decomposition of ground waters from final disposal sites and about the radiolytically conveyed UO2 corrosion were published in the past (review with references in [1], [2- 7]). We investigated the gamma- and alpha-radiolysis of concentrated NaCl brine under anoxic conditions experimentally and simulated the processes by means of a kinetic reaction model [8 - 10]. This work, from which first results were published earlier [11]
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