Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case

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

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

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

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

The problems of utilizing graphite of stopped graphite-uranium reactors

A list of radioactive nuclides, the activity of which forms the main part of total activity of graphite stack and graphite elements of the construction of stopped industrial graphite-uranium reactors has been defined. The analysis of activity part contributed by these nuclides at different moments of time after stopping reactor was carried out. A set of construction graphite elements, in which there is a possibility of self-sustaining release of the energy stored (Wigner's energy) was determined. It was stated that the most value of the Wigner's energy is achieved in graphite constructions operated in low-temperature region or at high values of flux densities of damaging neutrons and concurrent gamma radiation

Experimental Simulation of the Radionuclide Behaviour in the Process of Creating Additional Safety Barriers in Solid Radioactive Waste Repositories Containing Irradiated Graphite

Results of the experimental modeling of radionuclide behavior when creating additional safety barriers in solid radioactive waste repositories are presented. The experiments were run on the repository mockup containing solid radioactive waste fragments including irradiated graphite. The repository mockup layout is given; the processes with radionuclides that occur during the barrier creation with a clayey solution and during the following barrier operation are investigated. The results obtained confirm high anti-migration and anti-filtration properties of clay used for the barrier creation even under the long-term excessive water saturation of rocks confining the repository