During the design of the secondary circuit for a nuclear power plant with the BREST-OD-300 reactor, engineers and designers faced specific challenges related to the properties of the lead coolant. The key issue was its high crystallization temperature, which necessitated fundamentally new approaches to the thermal circuit design of the power unit. Particular attention was paid to developing the feedwater heating system, as its parameters and configuration directly affect the thermal balance and operational stability of the reactor plant. Throughout the design process, the configuration of this system underwent multiple revisions, reflecting the complexity of achieving optimal technical solutions while meeting stringent reliability and safety requirements. Each modification to the feedwater heating system design requires comprehensive analysis of its impact on all key components of the power unit, including both the reactor and turbine circuits. Engineers employed an Integrated Сalculating Mathematical Model (ICMM) to evaluate the system performance in normal and emergency operating modes, incorporating interconnections between all technological processes. This approach not only helps identify potential issues at early design stages but also optimizes equipment operating parameters. This approach achieves significant improvements in the power unit’s economic efficiency while simultaneously meeting all reliability and safety requirements. The research placed special emphasis on analyzing accident scenarios, particularly the feedwater cooldown line failure. Engineers applied mathematical modeling to pinpoint critical parameters and design failure mitigation solutions. These included upgrades to backup heating systems and optimization of control algorithms. Thus, this work highlights the necessity of employing modern modeling methods when developing innovative nuclear reactors. Such approaches enable advance development of technical solutions to ensure safe and efficient operation of these advanced power units.

The article presents the experience of training foreign students of Rosatom State Corporation’s flagship universities at the MEPhI Resource Center. It is shown that over the period 2018–2024, more than 500 bachelor’s, specialist and master’s students from 33 countries studying at the five largest flagship universities of the state corporation (VSU, MEPhI, SPbPU, TPU, UrFU) underwent practical training in English and Russian at the Resource Center of Obninsk Institute for Nuclear Power Engineering (IATE MEPhI). Students were trained in six educational programs of 13 and 14 specialty groups. In the field of non-energy application of nuclear technologies, practice-oriented training at the Resource Center is carried out to a greater extent on radiation technologies in agriculture and the food industry. High results of cooperation in training foreign students in the field of agro-nuclear technologies between the IATE MEPhI and the National Research Center “Kurchatov Institute” – RIRAE were noted. Practical training of students at RIRAE is carried out on a unique gamma-installation GUR-120 for irradiation of agricultural raw materials and food products for the purposes of phytosanitary and microbiological safety, increase in shelf life, radiation stimulation. The competencies acquired by students in the field of agricultural radiation technologies are in demand by them when applying for jobs in the Centers of Nuclear Science and Technology. The interaction of the Resource Center of the IATE MEPhI and the leading research institutes of Obninsk in training foreign students is the basis for the development of the international scientific and educational center “Obninsk Tech” for training personnel for foreign businesses of the State Corporation “Rosatom” in the nuclear industry and related areas of application of nuclear technologies.

The paper presents the results of the development of modified solid-phase extractants (SPE) based on N,N,N’,N’-tetraoctyl diglycolamide (TODGA) intended for the selective extraction of americium-241 from nitric acid solutions of liquid radioactive waste. (LRW). The structural and physico-chemical properties of synthesized materials, including granulometric composition, porosity, and stability in nitric acid media, have been studied. The developed SPEs are stable in nitric acid solutions and meet the size requirements for use in industrial sorption columns. The results obtained can be used as a basis for creating technological schemes for the processing of LRW based on synthesized experimental modified samples of SPETODGA, in particular for the isolation of minor actinides, including americium-241.

In this study, a simplified computational model of the blanket mock-up is created using the SERPENT Monte Carlo Code. The nuclear data is obtained from the enriched ENDF/B.VII.1 data library to conduct this study. The model is validated, as the error percentage for 63Cu(n,2n)62Cu and 65Cu(n,2n)64Cu reactions is less than 10% when compared to experimental results. The computational model is used to calculate the tritium production rate in different lithium zones with various neutron multipliers (U, Pb) and without any multipliers. The results show that the tritium production rate with a uranium multiplier is 86% higher than with a lead multiplier and 238% higher than with no multiplier. The neutron energy spectrum shows a peak in the 0.1 MeV to 10 MeV energy range for every case. This study also examines the effects of fusion neutrons on different isotopes, providing valuable data on how materials behave under high-speed neutron exposure.

The paper examines the radiation characteristics of noble platinum group metals (PGMs) extracted from spent nuclear fuel (SNF) of the VVER-1000 reactor. These are the so-called reactor-grade ruthenium, rhodium and palladium. PGMs are radioactive when extracted from SNF, but after the decay cooling of ruthenium for about 27 years, and of rhodium for about 13 years, they can be used in unlimited quantities. There is no sense in decay cooling of reactor grade palladium due to its radioactive isotope 107Pd having a half-life of 6.5 million years. As specified by regulatory documents, such palladium can be freely used only in quantities of up to 34 g. Pd is a soft beta emitter with a maximum beta particle energy of 34 keV. The calculation results show that the mean free path of beta particles from 107Pd in palladium metal is 0.8 μm, so reactor-grade palladium emits only from the surface layer, and other electrons are absorbed in the material itself. The mean free path of electrons with an energy of 34 keV in biological tissue is about 20 μm, which does not exceed the thickness of the skin epidermiscorneous layer. Calculations have shown that the equivalent dose rate (EDR) on the surface of reactor-grade palladium is 0.04 μSv/h, which is below the public EDR value. As a result, a conclusion is made that reactor-grade palladium does not pose a danger in the event of external contact.

The focus of this work is to analyse the suitability of two-group diffusion coefficients and macro constants generated from SERPENT using out-scattering approximation (OSA), transport correction (TRC) and cumulative migration methods (CMM) for fuel and non-fuel materials. For this purpose, various assembly and core models of ACP-100 SMR were designed. Assessment of these constants was conducted using COMSOL Multiphysics. For six distinct fuels, the best models were proposed with the least error margin in keff. Fuel material affects the group constants of non-fuel components except for radial reflectors. The sufficiency of two-group calculation was justified through spectrum analysis. Additional analysis revealed that MOX-RG has the hardest spectrum among all the fuels. Moreover, the effectiveness of boric acid to control excess reactivity was observed. Subcriticality was achieved for all fuel types except MOX-RG at a boric acid concentration of 4500 ppm. The influence of variation of boric acid concentrations on group constants was investigated using TRC and OSA. The reactivity difference between SERPENT and COMSOL was determined. It was found that OSA generates the most accurate results for MOX-RG with maximum 863 pcm error, while TRC produces higher accuracy with maximum error of approximately 250 pcm for other fuels.

The paper analyzes the possibility for large-scale production of highly pure plutonium suitable for use in radioisotope thermoelectrical generators in a VVER-SKD light-water power reactor with supercritical coolant parameters. Neptunium dioxide (NpO2) was used as the starting material. Neptunium can be extracted from minor actinides, i.e. from the main components of radioactive transuranium waste. Large-scale production of useful high-purity plutonium via neutron irradiation of useless radioactive waste can be estimated as one of the possible ways to convert harmful waste to a harmless form. The assembly with the starting material was placed in the central part of the reactor core with the maximum neutron flux to increase the plutonium accumulation. Numerical studies were based on a multi-physical approach with coupled analyses of neutronic and thermal-hydraulic processes in the downward and upward reactor zones. The results obtained in numerical studies have demonstrated that large quantities of high-purity plutonium with a low content of 236Pu and a high content of 238Pu could be accumulated in the central NpO2 assembly if the following conditions are satisfied. Firstly, it is desirable that the lattice of NpO2-containing rods would be sufficiently wide with a high volume fraction of light-water coolant. Secondly, the central NpO2 assembly needs to be surrounded by protective assemblies to form a barrier against high-energy fission neutrons which are capable to intensify the threshold 237Np(n,2n)236Pu reaction. The protective function can be performed by the assemblies containing rods with natural lead or radiogenic lead. The paper estimates the production scales of high-purity plutonium in the central assembly with a wide lattice of NpO2-containing rods and surrounded by a layer of protective Pb-containing assemblies.

This paper describes the development of a dispersion-type uranium-zirconium fuel rod. Uranium is distributed in the zirconium matrix material in the form of axis-oriented fibers. The fuel rod is designed for the conversion of the IVG.1M research reactor (Republic of Kazakhstan) from highly enriched uranium (HEU) to low enriched uranium (LEU). The need for the HEU-LEU conversion arose in connection with Kazakhstan joining the program to convert research and test reactors to fuel with reduced enrichment (RERTR 2023). The study solves the problem of deformation of a low-tech U-Zr alloy (located in a zone of low plasticity) by replacing them with a heterogeneous compound. The manufacture of fuel rod is based on metal forming processes. Initially, a fuel rod wire with a core of fiber structure is formed by triple co-extrusion of cylindrical uranium and coaxial zirconium billets. At the next stage, the wire is processed to the required diameter by drawing, then the operation of flattening, twisting and cutting into specified lengths is carried out. Upon reaching a high total degree of deformation obtained during cold work, relaxation annealing is carried out at temperatures of 550 to 600 °C, which leads to the formation of a transboundary layer of the UZr2 intermetallic compound with a thickness of 1 to 2 μm. The intermetallic layer, without having a significant effect on the strength and thermal conductivity of the compound, ensures high quality diffusion bonding of all fuel rod components. The final operations are melting of the ends of the fuel rods and sealing by electroplating with nickel. As a result, blade-profile fuel elements are obtained with a thickness of 1.5 mm, a diameter of the circumcircle of 2.8 mm and an average effective diameter of uranium fiber of 40 μm. A set of 14040 fuel rods was manufactured and loaded into an operating IVG.1M reactor. The power start-up took place in 2023. Due to the unification and wide variability in the loading of the fuel component, in size and shape of the cross section, in structure and materials of the matrix compound, the fiber fuel element design can be used in the development of fuel rods for advanced reactors for various applications.

Thermal Power Plants (ThPP), along with transport facilities, are the major sources for industrial emissions of carbon dioxide (CO2) believed to be responsible for the greenhouse effect leading to overheating of the lower atmosphere. In the opinion of many scientists, there is a threshold value of the average atmospheric temperature exceeding, which entails the potential for the development of irreversible processes threatening the existence of humankind. To avoid this danger, governments in nearly 200 countries have chosen voluntarily to reach net-zero CO2 emissions by 2050. Renewable Energy Sources (RES), including wind and solar power plants, have been selected as substitutes for ThPPs. However, energy systems based on RES only need to be multiply redundant in terms of installed capacity due to their efficiency being heavily dependent on daily, seasonal and weather factors, leave alone the scale of the required material resources (metals, polymers, concrete, glass, etc.). The major drawback of such energy systems is, however, the RES common-cause failure, e.g., in the event of a global volcanic eruption, when no energy-security requirement can be met to provide energy for satisfying the most vital needs. A need is fully evident for furnishing such energy system with another power source to be not dependent on the event that has caused the mass RES failure. With the net-zero-carbon requirement taken into account, Nuclear Power (NP) appears to be the best option in this respect. Modern NP does not however fully suits this role due to its inherent drawbacks (limited fuel resources, pending Spent Nuclear Fuel (SNF) and RadioActive Wastes (RAW) handling and nuclear-material nonproliferation issues). A potential solution to these drawbacks is a two-component NP technology in a closed nuclear-fuel cycle currently in the process of development. In Russia, where the greatest progress has been achieved in this field of development, under construction is a pilot and demonstration energy complex with the BREST-OD-300 nuclear unit expected to be started up in 2026–2027. Another promising designs to be developed are Small Modular Reactors (SMRs) / Small Nuclear Power Plants (SNPPs).

The calculation and visualization of the dose field in the decommissioning of nuclear facilities is one of the important functions of the decommissioning virtual simulation system. The dose field simulation tools can provide radiation field distribution and play an important role in determining the decommissioning plan and protecting personnel during the engineering implementation process. This article investigates the development of dose field calculation and visualization in the reactor decommissioning virtual simulation systems. A preliminary technology plan suitable for the development of the decommissioning dose field calculation and visualization display programs of the first Heavy Water Research Reactor (HWRR) in China is proposed. The applicability of the selected scheme is analyzed. The functional requirement and development direction of the HWRR reactor decommissioning dose field tool are preliminarily determined. Furthermore, the reactor vessel of HWRR reactor is modeled, the dose field distribution is calculated and visualized based on the preliminary decommissioning code. This research can provide technical support for the development of the decommissioning simulation system for the first HWRR reactor in China.