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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  14 C, 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  14 С 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  14 С 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  14 C accumulated in graphite, the contribution of  14 C accumulation channel from  13 C(n, γ) 14 C reaction, as well as the total contribution of  14 N(n, p) 14 C 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  14 C 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  14 C 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  14 C accumulation channels in the irradiated reactor graphite. The methodology suggested can be applied for determining fraction of selectively separated  14 C 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.

Potential of application of IRT-T research reactor as the solution of the problem of graphite radwaste disposal