The present paper considers an alternative concept of expanding the nuclear power fuel base for fissile isotopes by producing 233U from 232Th in the blankets of fusion reactors. The advantages of this concept, such as a significant reduction in the risks of environmental pollution due to the low radioactivity of reprocessed fuel, are shown. Several scenarios for the production and use of a new fissile isotope in thermal neutron spectrum reactors (VVER) are considered and the possibility of closing the fuel cycle in a system of thermal reactors with a fusion neutron source for heavy nuclei is demonstrated.

In view of the increased interest in a molten salt reactor (MSR), the problem of choosing container materials compatible with molten fluoride salts containing a fuel additive and fission products at temperatures up to 750°C is considered. The results of corrosion tests of promising high-nickel alloys with a fuel salt and an intermediate coolant of MSR, including experimental results at the National Research Center Kurchatov Institute are presented.

The article considers the possibilities of a fusion neutron source with a molten-salt blanket to provide fuel for thermal fission reactors with uranium-thorium nuclear fuel cycle, and it also formulates recommendations for closing fuel nuclide balances in a nuclear power system with fission and fusion reactors.

The limited amount of economically suitable natural uranium resources forces us to improve the existing structure of the nuclear power system and its fuel cycle. With the current natural uranium reserves, it becomes practically impossible to significantly increase nuclear power plant (NPP) capacities in the existing system on the basis of thermal reactors operating in the open nuclear fuel cycle (ONFC). In order to solve this problem, the article considers the introduction of fusion-fission reactors (FFRs). FFR blankets produce 233U, which is further utilized in thermal reactors. In this article, fuel compositions of VVER-1200 reactors with the addition of 233U and thorium are studied, which make it possible to reduce the consumption of natural uranium and/or utilize already available raw materials in the form of depleted and regenerated uranium. On the basis of the scenario of the Russian nuclear power system (NPS) development within the ONFC with VVER-1200 reactors, the required 233U production volumes and the corresponding FFR fraction are determined for all fuel options. The risks of thorium introduction into the nuclear fuel cycle (NFC) are reviewed.

According to the analysis of the nuclear fuel cycle (NFC) for a two-unit NPP, based on supercritical water-cooled reactors (SCWRs), the priority option for three aggregated criteria includes the start-up of the reactor using mixed uranium-plutonium oxide fuel and processing of spent fuel assemblies (FAs) after 7 years of internal storage. The backup option involves the operation of the reactor using uranium oxide fuel for 10 years followed by a phased transition to full core loading with mixed uranium-plutonium oxide fuel. The results of the performed analysis can be used to enhance the efficiency of design solutions for SCWR NPPs, as well as to assess the NFCs of a two-component nuclear power system with SCWR NPPs, including uranium-plutonium and thorium-uranium NFCs and/or their combinations.

Herein is presented a preliminary concept of a nuclear power installation (NPI) based on a low-power multipurpose test research reactor with a supercritical pressure light water coolant (LMTRR-SCP), which can operate as a test or research reactor, and has the following capabilities: substantiation of operating conditions for a supercritical water-cooled power reactor in the fast neutron spectrum; adjustment of the design operating regimes of the reactor with supercritical coolant conditions (including reactor startup, attainment of the energy power level, switch-over between power levels, rated power operation, shutdown, etc.); reactor irradiation of promising types of nuclear fuel, absorbing and structural materials for nuclear installations with fast, intermediate, or thermal neutron spectrum; comprehensive experimental and numerical investigations to obtain information required for development and verification of numerical codes; evaluation of new types of equipment for various process systems (including elements of emergency cooling systems, steam generators, etc.), instruments, and control, monitoring, and diagnostics systems for power reactors of various types. Investigations into the closed nuclear fuel cycle (CNFC) problem; utilization of actinides and handling of long-lived fission products, including pilot demonstration closure of the fuel cycle; mastering of processes for the manufacture of radionuclide products for various applications; production of modified materials; and use of the reactor’s thermal energy for heat and power generation. A potential also exists for the development of international cooperation within the scope of joint research programs on the basis of LMTRR-SCP.

Techno-economic prognostications for the manufacture of fuel assemblies with mixed uranium-plutonium fuel for the VVER-SKD power reactor, which is now under development, show that at the current level of technological development of pressurized water-cooled reactors in our country and the world the share of the fuel component in the price of electricity generated by NPP with VVER-SKD will not exceed the level realized by modern VVER, PWR, and BWR reactors. Moreover, considering the higher efficiency of VVER-SKD compared to the customary VVER/PWR reactors, maintaining the share of the fuel component in the cost of electricity at the current level will give this innovative reactor technology significant competitive advantages in regional electricity markets.

The paper discusses the prospects of an alternative way to close the fuel cycle, without recycling highly radioactive spent fuel. The essence is the potential production of the fissile isotope 233U from the isotope 232Th in a hybrid fusion reactor (HFR). The feasibility of using a hybrid fusion plant as a fuel accumulator in the nuclear fuel cycle, and the potential need of the system for hybrid fusion reactors necessary for its closing are estimated. A system in which two technologies, fission and fusion, working together, avoid the difficulties that arise as a result of the independent implementation of each of the technologies under consideration is simulated.

An implementation assessment of PSKD-600 type power reactors was made. It was established that the readiness level of all PSKD-600 nuclear fuel cycle technologies is no lower than TRL5, and most correspond to readiness level no lower than TRL7, i.e., the basic technological solutions were also demonstrated under actual industrial conditions. The relevant R&D should be aimed at large-scale industrial implementation of technologies. The analysis showed that there are no fundamental constraints on the integration into the NFC infrastructure of a complex of interconnected technological objects that secure the operation of PSKD-600. Thus, the PSKD-600 reactor stock can be unrolled on the existing technological base.

The objectives of the development of reactor technologies for transitioning to two-component nuclear power with thermal and fast reactors as well as the issues arising in the analysis of development scenarios, including the uncertainty of the resource base and technical-economic performance, are discussed. It is shown that attempts to detail the structure of the nuclear power system in the second half of the 21st century do not yield unambiguous results, giving compelling reasons for strategic decisions. It is suggested that future work on strategic development be focused on the horizon periods of key decisions concerning the adoption of new reactor technologies.

The use of modular design and arrangement to increase the operating life and simplify decommissioning of power units is considered, as well as other advantages aimed at reducing the integral risk of NPPs throughout the project life cycle.

Decades of research into the damage of structural materials of nuclear power plants have not yet led to a clear and consistent understanding of the mechanisms of a significant number of phenomena and processes occurring in metals and alloys under the influence of high-energy neutron irradiation. Creation of construction materials resistant to radiation exposure requires very significant financial and time resources. In this paper, we consider several problems that were actualized in the transition to the development of nuclear power as a system under uranium resource constraints.

The aim of the present work is to examine the possibilities of the MSR-I molten-salt nuclear actinide-incinerator reactor with a cavity-type core as possessing internal and passive safety characteristics and meeting the requirements imposed on next-generation reactors. Starting from the premises that the basic principles of nuclear safety must not differ from all other types of reactors the MSR-I reactor was analyzed in transient processes without activation of the accident protection system in situations where fuel could not be discharged into subcritical drainage tanks because of operator error or failure of the emergency discharge system. A computational study showed that the basic types of emergency situations can be overcome without destroying the protective barriers or actuating accident protection systems.

A view on Russia’s nuclear power strategy at the present stage – including the formation of a two-component system including fast reactors and the possibility of fuel cycle closure on the basis of thermal reactors – as well as on the prospects for using thorium and including thermonuclear neutron sources in nuclear power in the distant future is discussed. The key problems of the impending state of implementation of the country’s nuclear power strategy are examined.

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q similar to 1, fusion power of similar to 40MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for U-233 conversion from Th-232. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF - NaF - KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

A concept is offered for tokamak and molten-salt thorium blanket with liquid-metal mass-exchanger which allows an uninterrupted extraction of protactinium from the blanket and its accumulation in a cascade salt trap separately from fission products. At uninterrupted extraction of protactinium from neutron field with the same rate, such a facility can become attractive for industrial production of nuclear fuel ( 233 U) from thorium. For this, it is offered to use a reduction extraction of radio-nuclides into a liquid-metal carrier (directly contacting with the molten salt) by managing RedOx potential (Fermi level) of the salt composition. Establishing Fermi level in the first cascade of molten-salt trap only for oxidizing the lanthanides allows extracting only them from the liquid-metal carrier. In the second cascade of this trap, one can extract protactinium by shifting down Fermi level at higher oxidation potential. For correct operation of the trap cascades, the lanthanides portion in the second cascade will be less than 0.01% of the first one and the portion of protactinium in the first cascade will be four orders less than in the second one.