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

8th International Scientific Conference "Issues of Physics and Technology in Science, Industry and Medicine"

物理学;工业技术;医药卫生

Kotlyarevskiy, S.G.^1 ; Pavliuk, A.O.^1,2 ; Zakharova, E.V.^3 ; Volkova, A.G.^3

Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors, Seversk, Tomsk region, Russia^1

Tomsk Polytechnic University, Tomsk, Russia^2

Frumkin IPCE RAS, Moscow, Russia^3

关键词: Activation products;    Capability assessment;    Internal surfaces;    Radionuclide migration;    Retention properties;    Safety assessments;    Sorption properties;    Uranium-graphite reactors;   

Others  :  https://iopscience.iop.org/article/10.1088/1757-899X/135/1/012020/pdf
DOI  :  10.1088/1757-899X/135/1/012020

来源: IOP

【 摘 要 】

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:93mNb and the long-lived94Nb,93Zr radionuclides. Radionuclides of60Co,137Cs,90Sr, 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 for60Co and137Cs 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.