Capability assessment for application of clay mixture as barrier material for irradiated zirconium alloy structure elements long-term processing for storage during decommissioning of uranium-graphite nuclear reactors

Abstract

The radionuclide composition and the activity level of the irradiated zirconium alloy E110, the radionuclide immobilization strength and the retention properties of the mixed clay barrier material with respect to the radionuclides identified in the alloy were investigated to perform the safety assessment of handling structural units of zirconium alloy used for the technological channels in uranium-graphite reactors. The irradiated zirconium alloy waste contained the following activation products:{93m}Nb and the long-lived {94}Nb, {93}Zr radionuclides. Radionuclides of {60}Co, {137}Cs, {90}Sr, and actinides were also present in the alloy. In the course of the runs no leaching of niobium and zirconium isotopes from the E110 alloy was detected. Leach rates were observed merely for {60}Co and {137}Cs present in the deposits formed on the internal surface of technological channels. The radionuclides present were effectively adsorbed by the barrier material. To ensure the localization of radionuclides in case of the radionuclide migration from the irradiated zirconium alloy into the barrier material, the sorption properties were determined of the barrier material used for creating the long-term storage point for the graphite stack from uranium-graphite reactors.