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The concept of a fast neutron reactor with a sodium coolant with metallic U–Pu fuel in the core and metallic uranium and thorium in the blankets is considered. The achievement of the required system characteristics for the initial loading and surplus fuel breeding is demonstrated. Several scenarios of the development of a two-component nuclear power system in Russia based on the VVER-type thermal reactors and fast reactors with metallic fuel involving the thorium resources are considered.

The strategic line of development of a nuclear power system based on fission and fusion reactors which ensures electricity generation on a specified scale, solves the fuel problem for a long-term outlook, and secures the lowest risk of environmental contamination is presented. A contemporary view of prospects of developing the nuclear power industry on the basis of replacement of thermal reactors in the future by fast reactors, owing to a long duration of this process and the necessity of additional resources of natural uranium, forces us to consider the possibility of implementation of this strategy as unlikely. In addition, the fuel cycle of fast reactors requires the quick reprocessing of the highly active spent fuel, and because of this, the fuel cycle will have a high risk of a negative radioactive effect on the environment. The transition of the nuclear power industry to fast reactors will lead to a full change of the infrastructure related to reactor construction and operation. In the development of a nuclear power system with fusion and fission reactors, the demands for natural uranium will correspond to current estimates of economically effective stockpiles, the risk of radioactive contamination of the environment associated with the spent fuel reprocessing will be the lowest, and the contemporary infrastructure of the nuclear power industry will be maintained, i.e., the prevalence of generating capacities based on thermal neutron reactors. Thus, the integration of nuclear power production by fusion and fission reactions into a unified system creates a significant synergetic effect, in which the deficiencies of each technology are compensated by another technology of nuclear power production.

An option of high-density neutron flux generation locally, at a distance from the source, is considered in this paper. The energy dependence of the cross sections of neutron scattering by nuclei of some isotopes displays resonance behavior. Below the resonance level by energy, there is a deep drop of the scattering cross section dictated by the effect of resonance and potential scattering interference. When a slowing-down neutron gets the energy corresponding to a very small scattering cross section, it is capable of flying for several meters without interaction with nuclei of the medium. The possibilities of creating conditions for selection of high-energy neutrons emitted by a reactor or other source and moving in the required direction are investigated in the paper. The research described in the paper is focused on various factors (material composition, geometry) influencing the neutron radiation density and achievable neutron flux limits.

This article presents a comparison of results of simulation of VVER-440 reactor fuel burnup by means of ISTAR set of tools using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries. Code verification is required on the basis of both experimental data and computational uncertainty analysis. A VVER-440 fuel rod burnup simulation using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries was performed as preliminary step before comparison of nuclear data uncertainty analysis and calculation/experiment. Fuel irradiation conditions and SNF nuclide composition experimental data were obtained from radiochemical assays of irradiated VVER-440 fuel report. Fuel burnup simulation was carried out by means of ISTAR set of tools using Monte-Carlo code for spectrum averaged reaction rates evaluation required for solution of Bateman's equations. As a result, Kinf and nuclide composition across burnup were estimated for the sample considered. On the basis of two nuclear data libraries’ implementation for depletion simulation it was shown that Kinf uncertainties rise across burnup from 0.4% at the beginning of life up to 0.7% at the end of life. Heavy nuclides composition uncertainties run up to 9.5% for Am-241 at the end of the 4th life time; however for main uranium and plutonium isotopes the uncertainties do not exceed 1.2%. Most fission products composition uncertainties stay below 7.3% except Sm-150 nuclide (15.1%). The next step will be a comparison with experimental spent nuclear fuel isotopic composition data considering the above mentioned uncertainties.

The paper presents the 207Bi production cross-sections measured by the direct gamma-spectrometry technique in the samples of lead enriched with isotopes 208, 207 and 206, as well as in the samples of natural lead and bismuth, irradiated by protons of 11 energies in the range from 0.04 to 2.6 GeV. The obtained experimental results are compared with the previous measurements, with the TENDL-2019 data-library evaluations and the simulated data by means of the high-energy transport codes MCNP6.1 (CEM03.03), PHITS (INCL4.6/GEM), and Geant4 (INCL++/ABLA). The observed discrepancies between model predictions and experimental data are discussed.

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 233U from 232Th. 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.

This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.