Issues associated with handling irradiated graphite of uranium-graphite nuclear reactors are examined. It is demonstrated that selection of approaches, methods and means for handling irradiated graphite are determined by the form of occurrence and binding energy of long-lived 14C radionuclide with graphite crystalline lattice. The purpose of the present study is the determination of possible chemical compounds in which 14C can be found and assessment of fastness of its binding in the structure of irradiated graphite. Indigent and foreign experience of handling graphite radioactive wastes was analyzed, calculations and measurements were performed. Information was provided on the channels of accumulation of 14C in the structure of reactor graphite and it was demonstrated that the largest quantities of the radionuclide in question are generated according to the reaction 14N(n, p)14C. Here, most part of radioactive carbon is generated on 14N nuclei found in the form of impurities in non-irradiated graphite and in the composition of gas used for purging nuclear reactor in the process of operation. 14C radionuclide generated according to 14N(n, p)14C nuclear reaction is localized in the near subsurface graphite layer (in the near subsurface layer of pores) at the depth of not more than 50 nm. Analysis was performed of possible chemical compounds which may incorporate radioactive carbon. It was established that the form of occurrence is determined by the operational properties of specific graphite element in the reactor core. 14C binding energy in the structure of irradiated graphite was evaluated and depth of its penetration in the structure was calculated. It was established that selective extraction of this radionuclide is possible only under elevated temperatures in weakly oxidizing environment which is explained by the binding energy reaching up to 800 kJ/mole in the process of chemical sorption of 14C on the surface of graphite and depth of its occurrence equal to ~ 70 nm in the course of ion implantation. It was demonstrated that radioactive carbon generated according to 13C(n, γ)14C nuclear reaction is uniformly distributed among graphite elements and possesses binding energy ~477 kJ/mole. Its selective extraction is possible only under the condition of destruction of graphite crystalline lattice and organization of the process of isotopic separation. The obtained results allow recommending the most efficient methods of handling irradiated graphite during decommissioning uranium-graphite reactors.

The paper presents the results of the best material selection for additional safety barrier construction during uranium-graphite reactors decommissioning. The research findings show that the best material for safety barriers is clay-containing natural material of various types with counter-migrational and counter-filtrational qualities. We have demonstrated a technology for void-free filling of cavities in the reactor space of under-decommissioning uranium-graphite reactors on the site of JSC Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors (PDC UGR). This will make it possible to construct reliable man-made geo-barriers and prevent the release of radionuclides from the repository into the environment for thousands of years.

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Aspects of handling irradiated graphite during decommissioning uranium-graphite reactors (UGR) of different types were investigated. It was demonstrated that handling reactor graphite is complicated by the presence in the composition of graphite of long-lived radionuclides, especially 14C, which may get entrained in biological cycles since carbon constitutes one of the main components of biological chains. Practical implementation of the process of selective separation of 14С can significantly reduce potential danger represented by graphite radioactive wastes due to the reduction of graphite activity as related to the isotope in question, as well as due to the reduction of the leaching rate by separating 14С isotope which is the most weakly bound within the graphite structure. Conclusion was formulated that analytical measurement methodologies and calculation methods allow reliably estimating only the total quantity of 14C accumulated in graphite, the contribution of 14C accumulation channel from 13C(n, γ)14C reaction, as well as the total contribution of 14N(n, p)14C reaction on nitrogen impurities and on nitrogen contained in purge gas. Method was suggested for estimating the values of contributions of different channels of accumulation on nitrogen impurities and nitrogen contained in purge gas using IRT-T research reactor (Tomsk, Tomsk Region). Parallel irradiation of batches of samples of non-irradiated (fresh) reactor-grade graphite contained in different gaseous media constitutes the basis of the study. Algorithm was suggested for calculating contributions of all channels of 14C accumulation according to the results of measurements to be obtained in the proposed studies. Recommendations were formulated on the use of all brands of graphite applied for manufacturing elements of graphite stacks of uranium-graphite reactors designed in Russia for determining selectively separated fraction of 14C for all types of graphite radioactive wastes by the companies in the RF which operated (are operating) the uranium-graphite reactors. Time of exposure of samples of irradiated graphite in the GEK-4 horizontal experimental channel of the IRT-T reactor was calculated and was found to be equal to ~ 10 days. Methodology was suggested for conducting a series of experiments for determining the values of contributions of 14C accumulation channels in the irradiated reactor graphite. The methodology suggested can be applied for determining fraction of selectively separated 14C in irradiated graphite elements of practically all uranium-graphite nuclear reactors, including reactors operated abroad Russia, under the condition of maintaining carbon dioxide gas atmosphere in one of the irradiated containers.