Mixed Uranium-plutonium Research Articles

Monitoring and simulation of the fuel irradiation behavior in nuclear reactors based on phononic crystal structure

We have presented in the current work a novel idea for simulating the irradiation behaviors of the nuclear fuel pellets in nuclear reactors by using a one-dimensional defective phononic crystal (1D-DPnC) design was presented. The transmission spectra of the incident mechanical waves were considered basic data for expressing the characteristics of different nuclear fuel-pellets. Herein, the density, sound speed, and Young’s modulus of the fuel-pellets represent the key parameters that are influenced by the irradiation behaviors of these pallets. Mixed plutonium–uranium oxide (MOX) nuclear fuel is considered the main fuel in the present study. In addition, a comparison is performed for this fuel with other types of nuclear fuels. Moreover, the mechanical properties of these MOX-pellets are dependent on the porosity, the ratio of oxygen-to-metal (O/M), and the plutonium (Pu-content). The theoretical treatments depend on the transfers matrix method to compute the transmission spectra through the 1D-DPnC. The numerical findings provided that the MOX-pellet has the highest performance compared to other fuel pellets and with sensitivity equal to 59.388 × 103 Hz s/m. It was also reported that the effects of the percentage of the O/M and Pu- content in MOX had a minor effect in a comparison with the impact of porosity. The theoretical simulation agreed extremely with the experimental data reported for these nuclear fuels. Because of the close relationship between sound speed and density, this sensor can be utilized to monitor the porosity, O/M, Pu-content, and density of fuel-pellets as a quick and non-destructive evaluation technique in a nuclear fuel fabrication laboratory. This article has proven theoretically that MOX fuel produced from nuclear waste of uranium dioxide and plutonium dioxide gives excellent results compared to other types of nuclear fuels, and this agrees with experimental researches. Thus, it may contribute in preserving the environment from nuclear waste, and this can be considered a novel kind of purification of environmental pollution treatment.

Robotics applied in the module for fabrication and refabrication of dense mixed uranium-plutonium fuel of the pilot demonstration energy complex of «SCP» JSC. Decommissioning of nuclear and radiation hazardous facilities as a driver of Russia’s and world robotics markets

The article is devoted to the topic of great current interest that is the issue of using robots in the nuclear industry, in particular, in the module for fabrication and refabrication (MFR) of dense mixed uranium-plutonium fuel of the Pilot Demonstration Energy Complex (PDEC) of “SCP” JSC. The author attributes this to the danger or impossibility of direct participation of a person in some operations associated with adverse effects of high radiation burden and toxicity on the human body. Therefore, the author emphasizes the need to use technologies for indirect control of a number of processes or their automation, partial or complete. Particular attention is paid to the fact that robots are not introduced into existing production lines, but the robotic fabrication of fuel is implemented from the very outset, i.e. at the design stage. The author predicts that one of the benefits gained from the employment of robots will be the ability to ensure product quality at a more stable level. The article clarifies that, at the MFR, robotic manipula-tors, automated guided carts (AGC), and transfer carts will be used in the principal process in operations con-cerned with the fabrication of fuel assemblies, when handling radioactive waste, and in automated transport sys-tem lines. The growing need to solve the problems of decommissioning of nuclear energy facilities (NEF) requires new approaches to their implementation, including the minimization of human work directly in conditions of in-creased radiation background. As such an approach, it is suggested to use modern robotic technologies that make it possible to efficiently resolve problems at all stages of decommissioning of nuclear energy facilities, thereby adopting a comprehensive problem-solving approach and laying the groundwork for the future. The introduction of robotic complexes within the framework of the federal project not only has practical value for the NEF decommis-sioning in terms of resource efficiency, cost optimization and speed of project work, but also enables the imple-mentation of Russia’s strategic tasks in the social and economic spheres.

Concept of Carbide Compaction of Mixed Uranium-Plutonium Fuel

Concept of Carbide Compaction of Mixed Uranium-Plutonium Fuel

Techno-Economic Prognosis for the Manufacture of Fuel Assemblies with Mixed Uranium-Plutonium Fuel for the VVER-SKD Power Reactor

Techno-Economic Prognosis for the Manufacture of Fuel Assemblies with Mixed Uranium-Plutonium Fuel for the VVER-SKD Power Reactor

Thermogravimetric study of mixed uranium-plutonium fuel for prospective generation IV reactors

The analysis of the accidental regime of the MNIT fuel reactor facility has indicated a possible scenario of decomposition of the fuel composition with the release of metallic plutonium, which is unacceptable during operation of a nuclear reactor. The present paper studies the behaviour of mixed uranium-plutonium nitride nuclear fuel (MNIT fuel) at high temperatures (up to 2400 K) in a helium flow by thermogravimetry using an STA 449 F1 thermal analyser. The nitride fuel is shown to sublimate at a temperature of about 2000 K. X-ray microanalysis of the samples prior to and after high-temperature testing has established that mass loss in the mixed nitride is primarily due to plutonium evaporation.

Certain Aspects of Using 15N-Enriched Mixed Uranium-Plutonium Nitride Fuel in a Fast-Reactor Fuel Cycle

Certain Aspects of Using 15N-Enriched Mixed Uranium-Plutonium Nitride Fuel in a Fast-Reactor Fuel Cycle

SNF processing electrochemical operations: liquid-metal and salt medium purification

The paper investigates the process of regeneration of a liquid metal medium used in the pyroelectrochemical reprocessing of spent mixed uranium-plutonium nitride fuel produced by a fast neutron reactor. The investigation concerns the interaction of liquid cadmium with sludge formed during the anodic dissolution of ceramic nitride pellets in a 3LiCl-2KCl melt medium as well as the possibility of its purification by filtration from individual metal fission products. Anode sludge is represented by fission products of the platinum group, zirconium, molybdenum and technetium. It was determined by scanning electron microscopy that the metal product is composed of several intergrowth phases. It was found that upon contact of a polymetallic alloy simulating anode sludge with a melt, the liquid metal phase is saturated to 0.025 wt% of Pd, 0.01 wt% of Rh for 50 hours at 500 °C, while zirconium forms an insoluble dispersed intermetallic compound ZrCd3. Powders of molybdenum and technetium, which are not wetted with cadmium, can be completely removed using a filter mesh of plain weaving of the P-200 type. It is also possible to remove zirconium from anodic cadmium by filtration. The filtration efficiency of ruthenium and palladium powders did not exceed 54.3 and 13.1 wt%, respectively, due to partial dissolution and thinning of particles, which will lead to saturation of the liquid metal phase and the need to purify it by alternative methods.

Key Outcomes of Comprehensive Computational and Experimental Validation of Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel for the BN-1200 and Brest Reactors

Key Outcomes of Comprehensive Computational and Experimental Validation of Fuel Rods with Mixed Uranium-Plutonium Nitride Fuel for the BN-1200 and Brest Reactors

РАДИАЦИОННО-ГИГИЕНИЧЕСКИЕ ИССЛЕДОВАНИЯ ЭКСПЕРИМЕНТАЛЬНОГО ПРОИЗВОДСТВА СМЕШАННОГО НИТРИДНОГО УРАН-ПЛУТОНИЕВОГО ТОПЛИВА НА АО «СХК». Часть 2: Дозы и риски

Purpose: Assessment of the compliance of the radiation protection of workers at the complex experimental installations of JSC SChC with the requirements of the Russian radiation safety standards NRB-99/2009 to limit the generalized risk of potential exposure and the IAEA recommendations for not exceeding the control level of the minimum significant radiation risk. Materials and methods: The results of radiation-hygienic investigations of radiation exposure factors affecting workers involved in the manufacture of mixed uranium-plutonium nitride (MUPN) fuel at the complex experimental installations of JSC SChC are used as the input data for the preliminary assessment of radiation doses to workers. The models for assessment of radiation risk of potential exposure have been developed in accordance with the recommendations of the ICRP and the IAEA. Results: Preliminary estimates of the doses of external gamma-neutron (2.5 ± 0.5 mSv / year) and internal exposure of workers (~1 mSv/year1) are related to the current levels of exposure of workers of complex experimental installations. These levels are the result of exposure to ionizing radiation sources associated both with the development of new technologies and with residual radioactive contamination resulting from previous activities not related to the manufacture of MUPN fuel. The presented dose estimates are related to the use of raw materials that have undergone deep preliminary purification from radiogenic impurities. When irradiated nuclear materials are used as raw materials, the levels of gamma-neutron exposure to workers will be significantly higher. The maximum increase in the generalized risk of potential exposure due to annual exposure is estimated for women aged 18 years at the beginning of exposure, for which the increase in this risk is 1.45 × 10-4 year-1, which is 1.37 times lower than the limit established by the Russian radiation safety standards NRB-99/2009: 2 × 10-4 year-1. All predicted values of the lifetime attributable fraction of radiation (LARF) in mortality from malignant neoplasms are significantly less than the control level of the minimum significant risk recommended by the IAEA (LARF = 5 %), and the maximum value of LARF = 2.8 % is achieved for women aged 18 years at the beginning of exposure. Conclusion: Restrictions on the radiation risks of potential exposure, established by NRB-99/2009, as well as those recommended by the IAEA for not exceeding the reference level of the minimum significant risk, are met with a large reserve. The results obtained and the developed methods will be used to ensure the radiation safety of workers during the transition from experimental installations to pilot industrial implementation of the technology for the production of MUPN fuel.

Radiochemical and Elemental Analysis of Mixed Uranium-Plutonium Fuel Irradiated in the BN-600 Reactor

Radiochemical and Elemental Analysis of Mixed Uranium-Plutonium Fuel Irradiated in the BN-600 Reactor

Spark plasma sintering of (U,Ce)O2 as a MOx nuclear fuel surrogate

Mixed uranium plutonium oxides (MOx) are currently used as nuclear fuel in thermal fission reactors and are the leading candidate for use in solid fuelled fast reactors given a wealth of irradiation data from research and prototype systems. Spark plasma sintering (SPS) has been shown to reduce the sintering temperature and time required to manufacture UO2 fuel pellets and may benefit MOx production through the more efficient retention of volatile minor actinides (MAs) such as AmO2 in the manufacture of these fuels. In this study we have examined SPS as an alternative field assisted sintering route to UO2 and MOx fuel manufacture on surrogate (U0.93,Ce0.07)O2 prepared by a micronized master blend (MIMAS) type process. SPS has been shown to produce pellets of high density, > 95% TD at sintering temperatures of 1573 K and hold time of 5 min, comparable to equivalent pellets produced through conventional sintering at 1973 K for 5 h. The formation of Ce-rich regions has been measured showing the size distribution of Ce-rich particles in SPS Ce-MOx is similar to conventionally sintered Ce-MOx and are appropriate for MOx fuel specifications. However, a significant difference is observed in the Ce content of the Ce-rich particles which remain close to their primary blend composition, as well as having a high degree of heterogeneity within the Ce-rich regions. The effect of the shortened sintering temperatures and times on other microstructural properties are also revealed.

РАДИАЦИОННО-ГИГИЕНИЧЕСКИЕ ИССЛЕДОВАНИЯ ЭКСПЕРИМЕНТАЛЬНОГО ПРОИЗВОДСТВА СМЕШАННОГО НИТРИДНОГО УРАН-ПЛУТОНИЕВОГО ТОПЛИВА НА АО «СХК». Часть 1: Методы и результаты

Purpose: To present the methods and results of studies of the factors of radiation exposure to workers involved in the manufacture of mixed uranium-plutonium nitride (MUPN) fuel at the complex experimental installations CEI-1 and CEI-2 of JSC SChC.
 Material and Methods: Regularities of the formation of external exposure doses have been revealed based on the study of the dynamics of the ambient dose equivalent rate (ADER) of photon and neutron radiation at the CEI-1 and CEI-2 workplaces, as well as instrumental individual dosimetric control of the equivalent doses to workers. In order to assess the inhalation intake and possible doses from internal irradiation, studies of the physicochemical properties of radioactive aerosols were carried out.
 Results: It has been found that the main sources of penetrating radiation in the premises of CEI-1 are boxes where tablets are pressed, chips and rejected tablets are crushed, as well as temporary storage of products is occurred. The highest ADER values have been measured in those boxes, where the radiation exposure was due to radioactive contamination caused by past activity, and is not associated with fabrication of MUPN fuel. A significant contribution of neutron exposure to individual doses of workers was measured, which exceeded the contribution of gamma exposure at some workplaces of the CEI-1. At CEI-2, a non-functioning exhaust ventilation pipe passing over the premises was found to be a powerful source of external radiation. This pipe contained a significant amount of radioactive material. Assessment of the contribution of gamma exposure from the ventilation pipe to the external exposure of workers reached 85% at some workplaces. Studies of the physicochemical properties of radioactive aerosols have revealed a high reactivity of MUPN compounds, leading to instant oxidation of the thoracic fraction of MUPN fuel aerosols under contact with air. The complex morphological and dispersed composition of aerosol particles in combination with a complex chemical composition caused by the aging processes of aerosols, can lead to a fundamental difference in the biokinetics of MUPN aerosols, the process of dose formation and, consequently, the degree of radiological hazard compared to those adopted in the ICRP models for U and Pu.
 The results of the current radiation-hygienic research are of a preliminary nature, since the object of this research is an experimental installation, which was used to develop a new technology for the production of MUPN fuel. The instrumental and methodological approaches to assess the factors of radiation exposure to workers tested at these experimental installations, will be used in the future to conduct similar studies during the pilot industrial operation of new modules for fabricating and refurbishing of MUPN fuel.

Status on performance study of mixed nitride fuel pins of BREST reactor type

Within the framework of the “PRORYV” project the “The comprehensive program for the calculation and experimental justification of fuel pins with mixed uranium-plutonium nitride fuel of BN-1200 and BREST reactors for the period up to 2020” has been developed Tpoянoв et al., 2015 [1]. The purpose of the program is to judge fuel pins performance, to ensure stability and required quality of the experimental technology for manufacturing of mixed nitride, fuel pins and experimental fuel assemblies (EFA) for testing in BOR-60 and BN-600 reactors, to optimize the design and technology of fuel pins with mixed nitride for the developed BN-1200 and BREST-OD-300 reactors based on the results of reactor testing. Currently, the program aims at justifying the starting operation stages of BREST-OD-300 and BN-1200 reactors. The following results were obtained to judge the fuel performance at the starting operation stage of BREST-OD-300 reactor (the maximum mixed nitride burn-up of 6 at%, the maximum damage dose of 89 dpa): 1. A set of out-of-pile investigations has been performed in order to obtain the necessary data on the properties of fuel and cladding materials, which are required for the development of calculation models of mixed nitride behavior under irradiation and design justification of their performance, including in design accidents. 2. To March 2021 the irradiation of eight BN-600 EFAs with EP823-Sh cladding was finished. In EFA-11 the maximum burn-up 9.0 at% and the maximum damage dose 107.6 dpa have been reached. PIE of fuel pins of two OU irradiated in BOR-60 and four EFAs irradiated in BN-600 with maximum burn-up of 5 at% and damage dose of 74.7 dpa are completed. Experimental data required for verification of calculation results of temperature and stress-strain state of mixed nitride pins with EP823-Sh steel cladding, as well as for evaluation of fuel pin performance and establishment of allowable lifetime of the BREST-type fuel pins with mixed nitride have been obtained. 3. The KORAT, DRAKON and BERKUT-U codes were modified and verified. Certification of the DRAKON code has been completed and the BERKUT-U code has been submitted for certification. Obtaining the permission to extend the irradiation of EFA beyond the parameters planned in the Program was possible due to the fact that independent concurrent calculations with the use of KORAT, DRAKON and BERKUTU codes were performed during the development of technical designs of fuel pins with mixed nitride in justification of the possibility to extend the irradiation of EFA-11, EFA 14, EFA-15 in BN-600 reactor. All codes confirmed this possibility. Positive results of irradiation of BREST-type fuel pins in BOR-60 reactor were also the most important argument for obtaining permission. In order to improve fuel pins with (U, Pu)N and further burn-up increase, the technologies of their production were developed and optimized, the design of the pellets was changed (chamfer) and also fuel pins with liquid-metal sub-layer have been considered.

Results of post-irradiation examinations of mixed nitride pins with gas and liquid metal sub-layers

Today the investigations have been completed on helium-bonded fuel pins with mixed uranium-plutonium nitride of BN-600/BN-800, BN-1200 and BREST reactors types after irradiation as a part of ten EFAs of BN-600 reactor up to maximum burn-up of 7.5; 6.0 and 4.5at%, respectively. Also the PIE of mixed nitride pins of BREST type with helium and lead sub-layers after intermediate tests to maximum burn-up of 4,8 and 3,9at% as part of dismountable EFAs of BOR-60 reactor are over. As a result of postirradiation examinations, the main intra-fuel pin processes that affect on materials properties change and fuel pins state, their relationship with the initial nitride state were identified. The regularities and quantitative characteristics of nitride swelling, the behavior of fission products, cladding corrosion and mechanical properties change are revealed. The average for the irradiation time rate of fuel swelling in cross sections near the core midplane in gas-bonded pins of different EFAs irradiated in BN-600 reactor is equal to (1.6 - 2.0) %/at%. According to the results of irradiation in BOR-60 reactor, the fuel swelling rate in lead-bonded pins is lower than in helium-bonded pins: 1.4 ± 0.2 and 1.7 ± 0.2 %/at%, respectively. The presence of carbon and oxygen impurities in the fuel can lead to local areas of claddings carburization and oxidation, randomly distributed along the height and perimeter of its inner surface. The key characteristics of the fuel that determine the cladding carburization and oxidation, and ways to prevent them, are determined. Cladding corrosion in lead-bonded pin is caused by selective dissolution of cladding steel components. The maximum depth of the corrosion zone is located in the upper part of the fuel column and is equal to 40 microns. The results of cladding mechanical tests obtained using longitudinal segment samples and tubular samples loaded by internal gas pressure showed the significant margin of strength and ductility of cladding materials along the entire height of fuel pins both at nominal operating temperatures and at room temperature. High values of strength characteristics indicate a weak influence of corrosion on the strength of studied claddings materials. The mechanical cladding properties of a lead – bonded pins have the same pattern of change along the pin height depending on irradiation and testing temperatures, as of fuel pins with helium sublayer, keeping enough ductility in the area of low-temperature irradiation embrittlement. So, for example, at irradiation and test temperatures of 350-380 °C, the values of the strength limit from 1070 to 1200 MPa were obtained with the values of the total elongation of not less than 2 %. Thus, the results of post-irradiation examinations confirmed the performance of mixed nitride pins at the achieved test parameters and the possibility of their irradiation prolongation to higher fuel burn-up.

Unification of the BN-1200 reactor core designs with MOX and MNUP fuel

Two types of mixed uranium-plutonium fuel are considered for BN-1200 reactor (Matveev and Khomyakov, 2012) within the framework of Proryv Project: conventional, well-mastered MOX fuel and advanced nitride fuel (MNUP fuel) having higher density. Decision about the expediency of use of this or that type of fuel in the first BN-1200 reactors at the initial stage of their operation will be made taking into account the results of MNUP fuel pins tests carried out in BOR-60 and BN-600 reactors. In this report the information on basic engineering decisions taken in the design of BN-1200 reactor core (Vasiliev et al., 2010; Shepelev et al., 2013; Vasiliev et al., 2017) is given, features of cores with MOX and MNUP fuels are analyzed, possible variants of cores with unified design of fuel pins and FSA are presented, estimates of core characteristic change with increased fuel burnup are given, problems of transition from MOX fuel to MNUP fuel are considered.

Burnup and neutronic analysis for UOX and MOX fuel assemblies in VVER reactor

The present work examined the feasibility of utilizing Mixed Uranium–Plutonium Oxide (MOX) fuel in the VVER-1000 reactor core. In this study, about 30% of Uranium Oxide (UOX) fuel assemblies of the original core are replaced by MOX fuel assemblies. The effect of MOX fuel on the safety parameters is studied by using the MCNP6 code and the nuclear data library (ENDF/B)-VII.1. The safety parameters examined included the effective multiplication factor, temperature coefficients, void coefficient, boron coefficient, shutdown margin and the kinetic parameters. The excess reactivity, neutron flux and power density distributions and power peaking factor are also examined. All estimated values of reactivity coefficients analyzed for the core contains 30% MOX fuel confirmed to be negative, which guarantees the reactor safety and power stability in a normal state. Also, the effective delayed neutron fraction was lower than that of the full UOX core as well as the neutron lifetime was shorter, this shows that the MOX core responds faster than the UOX core. The maximum power factor for the MOX core was equal to the design limit. The shutdown margin satisfied the design requirement. The results confirmed that the VVER-1000 is able to meet the safety requirement for core designs with MOX loadings of up to 30%.

ПРОБЛЕМЫ ОЦЕНКИ СОСТОЯНИЯ ЗДОРОВЬЯ ПЕРСОНАЛА, РАБОТАЮЩЕГО В УСЛОВИЯХ НОВЫХ ТЕХНОЛОГИЙ ПРОИЗВОДСТВА ЯДЕРНОГО ТОПЛИВА

Purpose: To identify the problem of assessing the health status of personnel working under the conditions of new technologies for the production of nuclear fuel.
 Material and method: The object of the research was the general morbidity of workers in the production of mixed nitride uranium-plutonium fuel (MNUP-fuel). The material for the study was the data presented in the «Health Passports». The paper used the method of comparative analysis of the overall morbidity of workers in the production of MNUP-fuel and workers in enterprises dealing with nuclear fuel.
 Results and analysis: At present, in our country, within the framework of the «Breakthrough» project, new technologies are being developed for the fabrication and refurbishment of mixed uranium-plutonium (MNUP) fuel. In the absence of radiation and hygienic standards for the content of fuel products in working rooms, in order to assess the influence of production factors, along with the radiation dose, the incidence of personnel is studied as an integral indicator of health. A study of the incidence of 50 workers in the production of MNUP fuel revealed: Relatively high incidence of general morbidity – 1122 diseases per 100 people or an average of 93.5 diseases per 100 people per year, regardless of the length of service. The leading diseases in the overall morbidity structure are diseases of the respiratory system – 26.0 % 
 (1st place), eyes – 13.4 % (2nd place), musculoskeletal system – 11.4 % (3rd place), circulatory system – 10,9 % (4th place), injuries and poisoning – 8.4 % (5th place), digestive organs and genitourinary system – 7.7 % and 7.0 %, respectively (6th place), which make up 84.7 % of the total morbidity. Obviously, the effective dose of 4.6 mSv/year cannot be the only reason for the high morbidity in workers in complex radiochemical production, but characterizes only the influence of one of the many nonspecific factors of production. The existing system for assessing the health of personnel working in radiochemical production, in addition to analyzing the risks of deterministic and stochastic effects, should include an assessment of the overall morbidity of personnel.

Mathematical Model of the Technological Process of Carbothermal Synthesis of Mixed Uranium-Plutonium Nitride Fuel

Mathematical Model of the Technological Process of Carbothermal Synthesis of Mixed Uranium-Plutonium Nitride Fuel

ПРЕИМУЩЕСТВА БЫСТРОГО РЕАКТОРА С ФИЗИЧЕСКИ ТОЛСТЫМ ОТРАЖАТЕЛЕМ ИЗ СВИНЦА-208

The paper analyzes the effects produced by the use of lead isotope 208Pb instead of natural lead as a coolant and neutron reflector on neutron-physical parameters of fast reactor. A possibility is demonstrated to reach substantial (up to three orders of magnitude) elongation of mean prompt neutron lifetime in the fast reactor with physically thick 208Pb-reflector. So large elongation of prompt neutron lifetime opens an opportunity for remarkable enhancement of the reactor safety under conditions of reactivity-induced accidents. Numerical analyses have shown that it became possible to form high flux of slowed down and epithermal neutrons in 208Pb-reflector even at large distances from the reactor core. Spacious area of high neutron flux out of the reactor core may be used to transmute long-lived fission products and to produce significant quantities of necessary isotopes. Weak neutron absorption in 208Pb makes it possible to reach remarkable (up to 20 %) reduction of plutonium fraction in mixed uranium-plutonium fuel composition. It is proposed to place a layer with neutron absorber in 208Pb-reflector above the reactor core for the reactivity control with no variations in neutron-physical properties of the reactor core. If reactivity of the control rods in the reactor core is lower than effective fraction of delayed neutrons, then the reactor is obviously subcritical on prompt neutrons of the reactor core. In this case the danger of the reactivity-induced accident on prompt neutrons of the reactor core disappears. If the reactivity-induced accident is caused by joint action of the neutrons produced in the reactor core and in 208Pb-reflector, then the accident propagation will be substantially slowed down because of the longer lifetime of neutrons in 208Pb-reflector.

Possible ways to solve the problems of using MOX fuel in thermal reactors

Modern nuclear power has a number of crucial problems, including the accumulation of spent nuclear fuel (SNF) and the projected shortage of nuclear resources. The conversion to a closed fuel cycle is the most promising solution to these problems. Mixed uranium-plutonium fuel (MOX fuel) has been successfully tested and commissioned under this concept by several European countries. However, its operation in reactors designed for uranium fuel has a number of problems, most of which have been resolved. Nevertheless, the spent MOX fuel has neutronic characteristics that do not allow its reuse in thermal reactors and solve the problem of SNF accumulation. This is due to the accumulation of heavy (even) plutonium isotopes, which are not fissionable in the thermal spectrum. In fast reactors, it is possible to create conditions under which the isotopic composition of plutonium from the spent MOX fuel from thermal reactors changes in such a way that it becomes suitable for reuse in thermal reactors. This process is called improving. The concept for the development of two-component nuclear energy based on the joint operation of thermal and fast neutron reactors and the creation of a closed fuel cycle was adopted in Russia to implement this process and solve crucial problems of nuclear power.