The challenging geopolitical landscape and the imposition of sanctions necessitate a comprehensive overhaul of economic processes within our nation. In response, the Government of the Russian Federation has delineated a list of products and technologies earmarked for substitution with domestic equivalents in the immediate future. A pivotal focus lies on the import substitution of software. Pursuant to Presidential Decree No. 166 dated March 30, 2022, titled “Measures to Ensure Technological Independence and Security of the Critical Information Infrastructure of the Russian Federation”, effective January 1, 2025, state authorities and procurement entities are prohibited from deploying foreign software on their critical information infrastructure assets. The spearhead of the IT import substitution initiative in Russia has been State Corporation “Rosatom”, which, within the framework of a unified digital strategy aimed at bolstering IT import independence, is focused on fostering collaboration with Russian small and medium enterprises. A noteworthy initiative within Rosatom’s import substitution program, involving SMEs, entails the modernization of existing GEFEST and GEFEST800 PCs. These systems are designed for operational neutronic calculations for fast reactors utilizing sodium coolant at the Beloyarsk Nuclear Power Plant, the world’s only station with industrial-scale fast neutron reactors, namely, BN-600 and BN-800. The overarching objective of this project is to establish a unified cross-platform PC, known as GEFEST-M, based on the current GEFEST and GEFEST800 models. This unified system will incorporate specially developed calculation modules and methodologies, with the future applicability envisaged for reactors like BN-1200 and other BN-type reactor units. A pivotal attribute of the GEFEST-M project is the shift away from proprietary closed software development tools. Additionally, it underscores the capability to operate seamlessly on computing equipment and operating systems developed domestically. The developer spearheading the GEFEST-M PC project is an accredited organization in the realm of information technology – LLC “ASU-LEADER”.

Sodium-cooled fast neutron reactors are one of the most environmentally friendly reactor types and play an important role in the development of Russia’s nuclear power industry and the long-term supply of nuclear fuel for the industry. The long-term successful operation of the industrial-scale fast neutron reactor BN-600 and the commissioning of a new power unit with a more powerful BN-800 reactor have made Russia a world leader in the construction and operation of fast neutron reactors. This paper describes the stages of mastering the capacity of the sodium-cooled fast neutron power reactors BN-600 and BN-800 at the Beloyarsk NPP. Based on the experience of BN-600 and BN-800 operation, a power unit with a BN-1200M reactor is being created, which can be considered as the head in a series of power units with BN-1200M reactors. The creation of the series will significantly expand the fuel base of nuclear power by reusing spent nuclear fuel from other NPPs, and minimizing radioactive waste by burning the longest-lived isotopes from spent nuclear fuel from other reactors. The main task in creating the BN-1200M is to achieve economic characteristics that ensure its competitiveness with a serial thermal neutron reactor VVER of a similar power level. According to the General Scheme for the Placement of Electric Power Facilities approved by the Government of the Russian Federation until 2042, the lead power unit with the BN-1200 reactor will be built at the Beloyarsk NPP as power unit No. 5.

The possibility of obtaining 238 Pu in the BN-1200M (about 1,200 MWe) fast neutron power reactor is considered. To obtain products with a purity of 80–85% in terms of 238 Pu, options were proposed with installation of irradiation assemblies (IA) with breeding and moderating elements in the radial blanket. Neptunium and americium oxides and their mixtures were considered as target materials of the breeding elements for producing target nuclide. Zirconium hydride was considered as a moderator. The effective density of targets, IA arrangement, target/moderator materials content ratio, and irradiation time were varied. Calculations were made to determine mass and composition of produced plutonium and IA decay heat, as well as IA effect on the global and local pin-to-pin power profile in the standard fuel assemblies of the reactor core. When using targets made of neptunium oxide, options were effective only at an effective density of approximately 1.1 g/cm³ were only effective because of self-shielding effect. Irradiation during 4–5 years ensures required 80% purity of 238 Pu with target purity in terms of 236 Pu within 2 ppm. When using Am, 236 Pu purity is also ensured, but 238 Pu fraction is somewhat lower than 80% due to the higher fraction of 242 Pu. The use of Am and Np mixture makes it possible to reduce the fraction of 236 Pu and prevent high accumulation of 242 Pu, but in this case 242 Cm accumulation should be taken into account, and appropriate time period for its decay heat reduction should be chosen.

The article considers the possibility of using the mixed (Am, Cm)-fraction extracted from minor actinides (MA) for large-scale production of powerful heat sources for space radioisotope thermoelectric generators (RTG). Such a choice of the MA fraction removes the limitation on the proportion of 236 Pu in plutonium and leaves only a limitation on the proportion of 238 Pu (above 80%). The irradiation of the (Am, Cm) mixture and the separated Am-fraction of MA are considered and compared. The irradiated material was placed in the central assembly of the VVER-1000 light-water power reactor core. It is proposed to remove only fission products (FP) from the irradiated material and to use the remaining three-component (Pu, Am, Cm) mixture as a heat source in the RTG. It has been shown that the mixed (Am, Cm)-fraction of MA is a preferable starting material compared to the Am-fraction since it provides the higher specific heat release of the produced three-component (Pu, Am, Cm) mixture at comparable rates of plutonium production and its isotopic compositions.

Cr-Mn Austenitic Stainless Steel (ASS) are utilized in nuclear reprocessing plants, fast-breeder reactors, pressurised water reactors and boiling water reactors because of their ease of fabrication and welding. In this study two fillers were used, ER308L and ER308L-16, to weld ultra-low nickel Cr-Mn ASS using gas metal arc welding (GMAW) and shielded metal arc welding (SMAW) process. The simulation of the welding processes was carried out for the evaluation of stresses induced in the welding. The heat input is employed at the weld bead for modelling and analysis. From analysis, it was found that the weld bead width is less in GMAW as compared to SMAW. The transient thermal and structural analysis was carried out for evaluation of stresses in the weld using ANSYS. The weld for both welding methods was analysed using varying welding speeds. It was revealed that as the welding speed increases there is a decrease in Von-Mises stresses in the weld.

This study applies the COMPLET nuclear reaction code to calculate excitation functions for eleven alpha-induced reactions on stable copper ( 63 Cu, 65 Cu) and antimony ( 121 Sb, 123 Sb) isotopes, aiming to predict production cross-sections for medically significant radionuclides such as 68 Ga, 67 Ga, 66 Ga, 65 Zn, 124 I, and 123 I. Reactions are simulated across an alpha energy range of 10–80 MeV to evaluate excitation functions. Fixed nuclear model parameters, such as an initial exciton number n 0 = 4 (2p+2n+0h) and level density parameters ACN/K (K = 10) expressions tied to compound nucleus mass, were used to compute theoretical cross-sections. Model outputs were systematically compared with experimental data obtained from the EXFOR database. Statistical and graphical analyses demonstrated an excellent agreement, with Pearson correlation coefficients ranging from 0.74 to 0.94. Sensitivity analyses confirmed that variations in exciton numbers and level density parameters significantly influenced the shape and peak positions of the excitation functions, highlighting the importance of accurate parameter selection. These findings validate the COMPLET code as a reliable tool for modeling alpha-induced nuclear reactions, especially when experimental data are scarce. The results contribute to improved nuclear data evaluations and provide critical support for the planning of radionuclide production in medical applications. The study also includes a detailed covariance analysis to minimize discrepancies between model predictions and experimental data, emphasizing the importance of theoretical methods in contemporary nuclear research.

The paper presents the results of the computational analysis of an accident scenario in the spent fuel storage pool of the multipurpose fast research reactor. The simultaneous failure of the spent fuel cooling system and the ventilation systems is being considered as beyond design basis accident (BDBA) conditions. The computational analysis of the BDBA scenario was performed with the computer program KUPOL-MT designed to simulate thermohydraulic processes in the room atmosphere of the reactor facility containment systems. During the computational analysis with KUPOL-MT, a nodalization scheme for the spent fuel storage pool rooms was developed. The scheduled loading of spent fuel assemblies in the storage pool was considered in the accident simulation, together with the emergency unloading of the entire reactor core. It was shown that the water temperature in the spent fuel storage pool remained below the boiling point and there was no violation of the integrity of fuel cladding for three days after the start of the accident. As the fuel cladding did not collapse, there were no radiation consequences in the considered accident scenario. The calculation results demonstrated insignificant hydrogen stratification in the room above the spent fuel storage pool. It was shown that hydrogen content did not exceed the maximum concentration limit for three days after the start of the BDBA.

Research reactors (RRs) are widely used for research in fundamental physics, materials science, and for radioisotope production. One of the problems in the field of mathematical modeling of research reactors is the verification of computational codes used for the calculation of neutron-physical parameters. To pass the verification procedure, it is necessary to justify the methods of mathematical modeling and the accuracy of the RR geometric model. Currently, precision programs allow creating a RR geometric model of any degree of detail. However, the degree of this detail has not been formulated. The paper considers the justification of the parameters of spatio-temporal discretization of IVV-2M-type fuel assemblies when calculating the burnup distribution. Determining the frequency of dividing the fuel assembly model into layers by the height of its active part significantly affects such parameters as the effective neutron multiplication factor (Keff), energy release at the maximum stress point. Moreover, the degree of detail and the set of calculation statistics affect the calculation duration, which can be reduced by applying a certain approach to modeling. It should also be noted that the issue of the need for a fuel pin (or more detailed, in the horizontal plane) degree of burnup calculation was not considered due to the lack of control over the orientation of the IVV-2M fuel assemblies in the horizontal plane during the operation of the IVV-2M research reactor.

Nucleon-induced nuclear reactions are a significant field in nuclear physics with numerous applications like as in the production of medically important radioisotopes. The primary objective of this study is to analyze the excitation function of nucleon-induced nuclear reactions on the arsenic-75 isotope across projectile energies from 10 MeV to 100 MeV using COMPLET code. The excitation functions of the seven reaction channels: 75As(p, 3n)73Se, 75As(p, pn)74As, 75As(p, p5n)70As, 75As(p, p2p)73As, 75As(p, n)75Se, 75As(n, 2n)74As, and 75As(n, p)75mGe were investigated, analyzed and compared with experimental data within the energies from 10 MeV to 100 MeV. The calculated excitation functions showed strong agreement with experimental data obtained from the EXFOR data base, as assessed using Pearson’s correlation coefficient. Both pre-equilibrium and equilibrium nuclear excitation functions for all nucleon-induced reaction channels displayed a strong correlation with experimental results, except for the neutron-induced reaction channel, 75As(n, p)75mGe, which exhibited a moderate correlation. Studies have indicated that the pre-equilibrium reaction mechanism primarily governs the high-energy segment of the excitation function, whereas the low-energy segment is dominated by the equilibrium reaction mechanism for both neutron and proton-induced nuclear reactions on arsenic-75. Thus, utilization of the COMPLET code and the EXFOR data base has facilitated a detailed analysis of induced nuclear reactions in producing radionuclides with diverse applications.

The paper presents the results of determining the possibility of using the horizontal experimental channels of the MBIR reactor for neutron capture therapy studies. The collimator configuration for the neutron beam extraction with specified properties was justified computationally. The peculiarities of the reactor give grounds for a positive assessment of this prospect, primarily the hard spectrum and the uniquely high intensity of the beams. The paper considers the capability of channel No.5 as the most suitable for neutron capture therapy due to a combination of characteristics. The simplest possible axisymmetric collimator was selected for the calculations to assess the key functionalities of neutron capture therapy. The configuration and material composition of the collimator are defined by the experience of calculations. Two fundamental characteristics were analyzed to assess the capabilities of the neutron beam of MBIR’s channel No.5 for neutron capture therapy. These are the dose in the target (soft tumor tissue) containing 65 ppm of 10B, and the dose in healthy tissue containing 18 ppm of 10B. The task in the series of calculations was as follows: to determine the dynamics of the key values for neutron capture therapy with a variable thickness of the moderator in the collimator channel – the time for gaining a fixed “therapeutic” dose in the target (tumor) and the time for gaining the maximum “tolerance” dose in healthy tissue when the target moves along the depth of the phantom. The distribution of these characteristics through the depth of the tissue allows us to conclude that the beam extraction configuration under consideration is effective. The obtained results of the spectral neutron distribution at the outlet of channel No.5 and the estimated dose characteristics in healthy tissue and in the tumor confirm that it is technically possible to use this channel for neutron capture therapy.