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Radioactive wastes are formed during the production of uranium-plutonium nuclear fuel at enterprises. In the present article the concentration ranges for uranium, plutonium, and neptunium in liquid and cemented wastes corresponding to different radioactivity classes are determined. The concentration ranges are determined for wastes formed in the production of nitride fuel for the initial load, as well as for waste formed on use of recycled uranium, plutonium, and neptunium in production. The effect of 241Am accumulation on the classification of radwaste is determined. The estimated cost of transferring the wastes formed on fabrication of mixed nitride fuel to the national operator for disposal is presented. The technological methods of reducing the cost of waste disposal, including the possibility of forming class-3 wastes for which surface disposal is admissible, are presented. The maximum concentration of plutonium and americium in liquid wastes at which the solidified forms of wastes correspond to class-3 is determined.

The formation of tritium in reactors, its behavior in nitride fuel and possibility of bonding with fission products, and the characteristics of migration out of the fuel-element cladding are examined. It is shown by a computational method that owing to diffusion through the steel cladding the stationary content of tritium in a fuel element is established over several hours of operation and most of the formed tritium escapes through the cladding, some being retained in it. The presence of hydrogen (tritium) in structural reactor materials in the course of operation of the reactor and thermal vibrations of the structure can significantly degrade the strength of steel structural materials. In this connection, in the BREST facility, the emphasis on the problem of catching tritium is shifted from reprocessing of spent nuclear fuel to handling the lead coolant and the gas medium in the interior of the reactor.

Salt cake radioactive waste is a remnant solid salt concentrate after deep evaporation of radioactive evaporator concentrate at WWER NPP’s. The traditional cementing of borate-containing liquid radioactive waste, to which the salt cake belongs, leads to a significant increase in the volume of the final product. This work describes borosilicate vitreous wasteforms developed to immobilize radioactive salt cake waste and comprises data on both glass synthesis and characterization. The composition of glass selected for the purpose of immobilisation of the salt cake radioactive waste allows to include up to 40 wt. % of the oxides contained in the salt cake and to reduce the volume of the final product by more than 2 times compared with the cement compound. The batches were melted in a cold crucible melter at 1200 °C. The normalized cesium leaching rate of the vitrified wasteform product was within range 3.0·10-5 – 3.7·10-6 g/(cm2·day).

The operation of a linear crystallizer, which performs crystallization extraction of uranium from a nitric acid solution of uranyl nitrate during nuclear reprocessing, is analyzed. Equations for the evaluation of the characteristics of the continuous operation of the apparatus are given, namely, the extraction efficiency of the crystal phase and the filling time of the storage container of the apparatus with the crystals of uranyl nitrate hexahydrate. The estimates are compared with the results of computer calculations and experimental data.

A mathematical model of the crystallization of uranium in a continuous linear crystallizer, designed for the crystallization separation of desired products in the processing of an irradiated nuclear fuel, is proposed. This model defines the dynamics of growth/dissolution of uranyl nitrate hexahydrate crystals in a nitric acid solution of uranyl nitrate. Results of a numerical simulation of the indicated process, pointing to the existence of stationary conditions in the working space of the crystallizer, are presented. On the basis of these results, the characteristic time of establishment of the stationary regime at different parameters of the process was estimated. The mathematical model proposed was validated on the basis of a comparison of the results of calculations carried out within its framework with experimental data.

A technology for reprocessing mixed uranium–plutonium nitride fiel (MUPN) from BREST reactor is considered. The technology should ensure reprocessing of spent nuclear fuel with the storage time after irradiation of no more than 1 year, 10–15% content of fissile materials (FM), and burn-up of 10% of heavy atoms. The target product of the technology is a mixture of actinide oxides separated from fission products with the separation factor of ~106. A PH (Pyro–Hydro) process was suggested for MUPN SNF reprocessing. It involves pyroelectrochemical fuel reprocessing with separation of U, Np, and Pu from the major fraction of fission products responsible for the heat release from the fuel and for the radiation load on process media, a series of hydrometallurgical operations for final purification of the target products (U–Pu–Np–Am), and radioactive waste (RW) management. The PH process is being developed since 2011 by the teams from the Bochvar High-Tech Research Institute of Inorganic Materials, Khlopin Radium Institute, and Research Institute of Atomic Reactors with active participation of the Leading Research Institute of Chemical Technology, Siberian Chemical Combine, and institutes of the Russian Academy of Sciences, primarily Frumkin Institute of Physical Chemistry and Electrochemistry and Vernadsky Institute of Geochemistry and Analytical Chemistry.

A mathematical description of the refined shape of the absorption lines in the Mossbauer spectrum that takes into account the distortion of the shape of the lines as a result of the effect of apparatus factors is suggested. Within the framework of the mathematical model suggested, processing of the absorption spectrum of iron in beryllium was done. The results may be used in the Mossbauer analysis of the phase composition of solid bodies with the aim of early identification of new phases or pre-phase precipitates.

The work shows that the friction coefficient μ during slip of polycrystalline electrolytic corundum, single-crystal silicon carbide and sintered corundum over hardened ShKh15 steel, and the wear m in an electrolytic corundum test-piece according to slip speed have a low point at speeds close to 100 m/sec. A hypothetical explanation is given for this behavior of μ and m.