The review is devoted to overview of the current status on some radiochemical studies and their technological applications in Russian Federation. Multiple view on the problem is given including achievements of recent years in this field. The collected data reflect importance of studies connected with spent nuclear fuel and explain how complicated is this problem.

Results of experiments to test the modes of plutonium and neptunium displacement re-extraction with a nitrate solution of uranium (VI) are presented. The developed method of plutonium displacement re-extraction was tested at the refining extraction and crystallization facility of JSC SChC as part of the comprehensive program “Development of equipment, technologies, and scientific research in the field of nuclear energy use in the Russian Federation”. Due to a high accumulation of plutonium in the fuel of fast reactors, the main task of plutonium re-extraction consists in ensuring the re-extract ratio of Pu / (Pu + U) specified by the manufacturers of oxide fuel. According to the results of the performed tests, the developed method of plutonium displacement re-extraction with a solution of uranyl (VI) nitrate allows uranium, plutonium, and neptunium to be re-extracted in the ratio specified by fuel manufacturers. In this case, the completeness of plutonium extraction into the re-extract and its purification according to the proposed method are comparable to the results obtained in the process of reductive plutonium re-extraction.

Проведена оценка эффективности волоксидации смешанного нитридного уран-плутониевого отработавшего ядерного топлива (СНУП ОЯТ) для отделения топливной композиции от оболочек твэлов и удаления 3 H и 14 C. Показано, что полнота отделения ОЯТ от оболочек твэлов в оптимальных условиях находится на уровне 98-99%. Остаточное содержание трития в волоксидированном топливе не превышает 0.2% от его содержания в исходном образце ОЯТ, радиоуглерод удаляется на 98%.

A model is proposed for the mixing chamber of a centrifugal extractor, which is used in existing technological plants for the processing of waste nuclear fuel and designed for the recovery of target components from a solution with the use of a solvent (extragent). The model is developed by the freely distributed OpenFoam software package, which provides an option of modifying the program code via the addition of new dependences to the equations to be solved.

Data were obtained on the dissolution of voloxidized mixed nitride uranium-plutonium spent nuclear fuel (MNUP SNF), the amount and elemental composition of the insoluble residue formed during the operation in the mode corresponding to flowsheet of the MNUP SNF hydrometallurgical reprocessing. The effectiveness of the procedures for transferring the insoluble SNF residue into solution using nitric acid, by electrochemical dissolution in the presence of electrogenerated Ag 2+ ions, and also by fusion with a mixture of potassium hydroxide and nitrate, was evaluated.

The kinetics of Np(V) reduction with carbohydrazide in nitric acid medium in the presence of Tc(VII) and U(VI) ions was studied by spectrophotometry. The reduced form of Np(IV) is unstable in time: reverse oxidation occurs, which accelerates with increase in the Tc(VII) ion concentration and temperature. The rate of the reduction stage is significantly higher than the rate of reverse oxidation. The orders of both stages of the reaction were found for all components.

The results of experiments are presented on testing the separation of uranium from U-Be nuclear fuel solutions by extraction. It was shown that the degree of uranium extraction using a solution of 10 vol% TBP in RED-3 from a nitric acid solution of this fuel is 98.7%. Transfer of uranium into the re-extract is 97.9%. More than 99.99 % of beryllium goes into the extraction raffinate intended for further vitrification. The results of the study were used for the preparation of technological regulations on reprocessing of U-Be SNF at the RT-1 plant of the Mayak Production Association. The industrial-scale reprocessing of this fuel confirmed the values determined in the study.

The results of experiments on development of dissolution modes of the uranium–beryllium nuclear fuel produced on an enlarged batch of unirradiated fuel elements, are reported. The feasibility of a two-stage fuel dissolution mode was confirmed, which makes it possible to reduce the total time of the process and reduce the rate of gas evolution. The composition of the formed gas phase was determined. The weight of undissolved solid residue was estimated. The results of the study were used in the preparation of technological regulations for the reprocessing of uranium–beryllium SNF at the RT-1 plant of the Mayak Production Association.

The valence behavior of neptunium in the reaction of Np(V) with U(IV) in HNO 3 solutions containing hydrazine and Tc(VII) ions has been studied. It was found that Np(V) is reduced to Np(IV) in two successive stages differing in the rates, after which the formed Np(IV) is oxidized to the pentavalent state. Simultaneously with the Np(V) reduction, complete oxidation of U(IV) occurs. Kinetic equations describing the rates of Np(V) reduction and U(IV) oxidation are established, and possible schemes of these reactions are studied.

A full-scale semi-industrial crystallization and refining stand was created on the basis of the Siberian Chemical Combine to develop and optimize the hydrometallurgical technology. The main purpose of this stand was to study the crystallization refining of uranium, plutonium and neptunium nitrates at a full-scale stand using significant amounts of nuclear materials. A bench experiment on crystallization refining of an undivided U-Pu-Np mixture was carried out. As a result of crystallization purification, mixed crystals of hexahydrate nitrates of hexavalent actinides were obtained. The yield of U, Pu, Np into the crystalline phase was 60, 85, and 60%, respectively. The degree of purification of the target product from elements of stable isotopes imitating fission products was about 102.

Abstract Russian Federation is developing the key technologies for closed nuclear fuel cycle with FR used mixed uranium-plutonium nitride or oxide (MOX) fuel within the framework of the project “Proryv”. This is a new integrated comprehensive product for nuclear power, which provides in the future - no accumulation of spent nuclear fuel; - radioactive waste management based on the principles of radiation-equivalent disposal; - technological support for the non-proliferation treaty; - competitiveness with other large-scale power technologies. The experimental and demonstration energy complex is under construction now on the site of Siberian Chemical Combine for the industrial development of the closed nuclear fuel cycle. ODEC will include a fast reactor BREST-OD-300 with lead coolant, facility for fabricating/refabrication of nitride fuel and the spent nuclear fuel processing facility. Up today, the key technology elements were demonstrated using macro quantities U, Pu, Np, irradiated mixed nitride fuel and real high-level waste. Within the framework of the project “Proryv” a prototype of the fuel cycle of fast reactors is developing.

Irradiation with accelerated electrons of an extraction mixture of 30% TBP in Isopar-M isoparaffin diluent on the lower temperature limit of flame propagation (Тl) has been studied. It was found that, under irradiation up to a dose of 2 MGy, the value of Тl of the extraction mixture may decrease to 45°С, which should be taken into account when organizing the technological process. It was shown that, when an irradiated extraction mixture is bubbled with air, the value of Тl nearly reaches the Тl of the starting solution.

The results of experiments on the clarification of solution simulators obtainable by means of dynamic crossflow filtration during the reprocessing of mixed uranium-plutonium nitride spent fuel are presented. The experiments were performed using full-scale, dynamic, cross-fl ow filtration stands at the RT-1 radiochemical plant at the Mayak Production Association. It is shown that this method yields high-quality clarification of solutions without the use of flocculants – the obtained permeates were completely free of the solid phase and silicic acid in the γ-form. In addition, the concentration of the solid phase can reach 100 g/cm3 without loss-of-clarification. It is shown that the method of dynamic cross-fl ow fi ltration is effective for clearing transuranium elements from liquid radwaste. After their co-precipitation with the carrier (trivalent iron) and clarification, the obtained permeates are low-level wastes according to the Basic Sanitary Rules for Radiation Safety (OSPORB-99/2010).

Behavior of carbamide in HNO 3 solutions on Pt and SnO 2 electrodes was studied. It was found that the oxidation rate of Co(NH 2 ) 2 on the SnO 2 electrode is approximately two orders of magnitude lower than that on the Pt electrode. The electrochemical reduction of U(VI) in a cell having no diaphragm on a Ti (cathode)–SbO 2 (anode) pair of electrodes was examined at its concentration of 10 to 100 g/L in HNO 3 solutions (0.5–2.0 M) containing carbamide (5–30 g/L). It was found that the concentration of these components affects the completeness and rate of U(IV) formation and the current efficiency. It was shown that the reduction efficiency decreases in the presence of technetium ions, and, after the current is switched off, the already formed U(IV) is oxidized to U(VI). The rate of this process grows with increasing concentration of Te and HNO 3 . The electrochemical behavior of carbamide on an insulated Ti cathode was examined in aqueous nitric acid solutions. It was found that, in this case, the carbamide solutions acquire on being subjected to electrolysis reducing properties toward Pu(IV) and Np(VI). A laboratory installation of mixers-settlers, assembled in accordance with the technological scheme of the first cycle of SNF processing at RT-1 plant was used to perform experiments with the use of electrochemically processed carbamide as a re-extractive agent in the procedure of Pu and Np separation from uranium. It was found that a high rate of U purification to remove Np is reached in this case, but a satisfactory mutual separation of U and Pu is not provided.

Experimental justification has been provided for the possibility of extractive isolation of technetium from raffinate of technical process of affinage of uranium–plutonium product resulting from hydrometallurgical reprocessing of spent nuclear fuel (SNF). Technetium (VII) from aqueous solution of Zr(IV) (10 g/L) and 4 mol/L HNO3 with small content of ZrOb+ and $${\text{Z}}{{{\text{r}}}_{{\text{2}}}}{\text{O}}_{{\text{3}}}^{{b + }}$$ oxocations is recovered by 80% in 30% TBP in hydrocarbon diluent over seven stages of counter-current continuous cascade at ratio O : A = 1 : 1. Variation in phase flow ratio for cascade part leads to virtually complete Tc(VII) extraction in organic phase over ten stages. Technetium(VII) is selectively back extracted from loaded organic phase with water virtually completely over eleven stages of counter-current continuous cascade. Technological extraction scheme for technetium isolation from SNF solutions has been proposed.

A procedure was developed for preparing powders of solid solutions of uranium dioxide with 3 or 10 wt % Ce (as Am surrogate) from nitric acid solutions using microwave radiation. The powders obtained consist of particle aggregates of size no larger than 400 µm; the fraction of particles of size smaller than 25 µm does not exceed 1 wt %. The tap density of the powders is 2.3–2.5 g cm−3, and their specific surface area is 2.2–2.5 m2 g−1. The powder characteristics meet the requirements to powders of ceramic quality for nuclear fuel fabrication. The method developed can be used for producing mixed U-Am oxides on a unit for spent nuclear fuel reprocessing at the Pilot Demonstration Power Engineering Complex with the aim of Am transmutation in the BREST-OD-300 reactor.

Oxidation of Pu(III) in a diaphragmless cell with a Ti cathode and a Pt anode in 1.0–3.9 M HNO 3 solutions containing 4.5 × 10 −2 –1.8 × 10 −1 M hydrazine was studied. The final solutions contain, along with Pu(IV), also a small amount of Pu(VI), increasing with an increase in the current density. Methods allowing minimization of the Pu(VI) amount in the solutions after the electrolysis were suggested. The possible process mechanism involving reactions of Pu ions at the electrodes and in the bulk of the solution with HNO 2 was considered.