[en] Creation fast critical molten salt reactor for burning-out minor actinides and separate long-living fission products in the closed nuclear fuel cycle is the most perspective and actual direction. The reactor on melts salts - molten salt homogeneous reactor with the circulating fuel, working as burner and transmuter long-living radioactive nuclides in closed nuclear fuel cycle, can serve as an effective ecological cordon from contamination of the nature long-living radiotoxic nuclides. High-flux fast critical molten-salt nuclear reactors in structure of the closed nuclear fuel cycle of the future nuclear power can effectively burning-out / transmute dangerous long-living radioactive nuclides, make radioisotopes, partially utilize plutonium and produce thermal and electric energy. Such reactor allows solving the problems constraining development of large-scale nuclear power, including fueling, minimization of radioactive waste and non-proliferation. Burning minor actinides in molten salt reactor is capable to facilitate work solid fuel power reactors in system NP with the closed nuclear fuel cycle and to reduce transient losses at processing and fabrications fuel pins. At substantiation MSR-transmuter/burner as solvents fuel nuclides for molten-salt reactors various salts were examined, for example: LiF - BeF2; NaF - LiF - BeF2; NaF-LiF ; NaF-ZrF4 ; LiF-NaF -KF; NaCl. RRC 'Kurchatov institute' together with other employees have developed the basic design reactor installations with molten salt reactor - burner long-living nuclides for fluoride fuel composition with the limited solubility minor actinides (MAF3 < 2 mol %) and have estimated its basic characteristics. On the basis of these data employees RRC KI and VNIPIET carry out conceptual binding reactor installations with molten salt reactor - burner to the project of a factory on processing 500 tons spent fuel of reactors of type WWER-1000 in a year. During a settlement-experimental research in RRC KI it is shown, that fluoride fuel composition with high solubility minor actinides (MAF3 > 10 mol %) allows to develop in some times more effective molten salt reactor with fast neutron spectrum - burner/ transmuter of the long-living radioactive waste. In high-flux fast reactors on melts salts within a year it is possible to burn ∼300 kg minor actinides per 1 GW thermal power of reactor. The technical and economic estimation given power-technological complex shows on economic efficiency of use such burner/transmuter. After separation from spent fuel power reactors minor actinides go on burning out in molten salt reactor. The offered concept power-technological complex with high-flux fast reactor on melts salts, intended for burning out and transmutation long-living radiotoxic nuclides, at practical realization will allow minimizing quantity of the long-living radioactive waste in system of a nuclear power. Accommodation of such reactors at the enterprises of a fuel cycle will provide with their energy and will facilitate the decision of a problem of radioactive waste management with the minimal losses. Small share MSR (5-7) % from full electric power in structure of the future nuclear power provides practically full burning of all minor actinide (Authors)

[en] It is shown that three-component Nuclear Power (NP) system consisting of thermal reactors (TR), fast reactors (FR) and molten salt reactors-burners (MSR) can operate in mode when actinides in system are not stored up proportionally to energy generated but are practically at steady level proportionally to system power. In the paper there was considered the influence of nuclear fuel cycle (NFC) duration and level of irretrievable reprocessing losses on NP structure and fuel consumption efficiency. U-Pu and U-Th variants of NP systems were considered with various levels of irretrievable reprocessing losses for all actinides - 0%, 0.1% and 1%, and various duration of cooling and reprocessing time of TR fuel cycle (TRR time) - 3 years, 6 years, 9 years, 20 years. Steady state calculations were performed for three-component NP systems and there were obtained steady state amounts and radioactivity of actinides for main components of NFC closed by actinides. Also, there were obtained rates and radioactivity of accumulation of actinide irretrievable reprocessing losses. (authors)

[en] The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: "2"3"5U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or "2"3"3U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle

[en] It is shown that reprocessed uranium can be enriched with simultaneous dilution of ^232,234,236U isotopes in a cascade of gas centrifuges that has three feed flows (depleted uranium, low-enriched uranium, reprocessed uranium). Computational experiments are carried out for different ²³⁵U content of the lowenriched uranium. It is demonstrated that the chosen combination of diluents can simultaneously reduce the cost of the separation procedure and the consumption of natural uranium, which in turn ensures cost reduction relative to not only the earlier used multiflow cascades but also the standard cascade for enrichment of natural uranium.

[en] The possibility of the recovered uranium enrichment in a cascade of gas centrifuges with three feed flows (depleted uranium, low-enriched uranium, recovered uranium) with simultaneous dilution of U-232,234,236 isotopes was shown. A series of numerical experiments were performed for different content of U-235 in low-enriched uranium. It has been demonstrated that the selected combination of diluents can simultaneously reduce the cost of separative work and the consumption of natural uranium, not only with respect to the previously used multi-flow cascade schemes, but also in comparison to the standard cascade for uranium enrichment. (paper)

[en] A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the ²³⁶U isotope is compensated by re-enrichment in the ²³⁵U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the ²³²U isotope are obtained.

[en] The effect of the uncertainties of the isotopic composition of the reprocessed uranium on its enrichment process in gas centrifuge cascades while diluting it by adding low-enriched uranium (LEU) and waste uranium. It is shown that changing the content of ²³²U and ²³⁶U isotopes in the initial reprocessed uranium within 15% (rel.) can significantly change natural uranium consumption and separative work (up to 2-3%). However, even in case of increase of these parameters is possible to find the ratio of diluents, where the cascade with three feed flows (depleted uranium, LEU and reprocessed uranium) will be more effective than ordinary separation cascade with one feed point for producing LEU from natural uranium. (paper)

[en] A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing "2"3"3U from "2"3"2Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated

[en] The use of fissionable materials accumulated during the open fuel cycle faces both challenges and opportunities for modern nuclear power engineering. A growing disproportion between the production and consumption of natural uranium draws attention to the problem of used nuclear fuel. The paper examines the current situation in the field of reuse (single or multiple) of reprocessed uranium (recovered from spent nuclear fuel), which is used for the production of low-enriched fuel (LEU) to fuel the fleet of light water reactors (LWR). The world experience in handling this material has been analysed. This study also gives an overview of the currently proposed methods for enriching the ²³⁵U content in reprocessed uranium to the required level by means of gas centrifugation process, while simultaneously meeting the limitations on the presence of ^232,234,236U in commercial LEU. Savings of natural uranium was estimated for repeated recycling of VVER spent fuel. It was supposed that re-enrichment of reprocessed fuel would be done by arranging the most promising for such purposes cascade schemes. The obtained results can be used as a basis for further scientific, technical, and feasibility studies on the large-scale utilization of reusable materials in the different fuel cycles of LWR. (paper)