The results of 70 cycles of rigorous testing for resistance to thermal cycling of the medium are presented – temperature reduction of the supplied water δTin.30 sec = 262°C over the first 30 sec of the process for two check valves of the type S23307-0160-65, DN 65 (body material St12Kh18N10T) as well as four of the same thermal cycles for a bellows-type stop valve S.KZSA-100-00-00-E-03, DN100 (body material St20). Sealing gaskets containing thermally expanded graphite are installed in the body–cap connector of all valves. At the completion of the tests, all three types of gaskets retained their integrity and serviceability. However, if the stop valve DN100, which structurally has an axisymmetric temperature distribution of the metal in the above-flow part of the body, retained its tightness throughout all thermal cycles, then out of 140 thermal cycles the check valves DN65, which do not have such an axisymmetric temperature distribution dynamics, had 18 brief (40–50 sec) weak seal failures in the front portion of the valve connector. The average difference of the temperature of the metal above the flow portion of the valve body in the front and back sectors was estimated as ΔTav ~ 40°C in the presence of seal failure.

A model of the deformation of 12Kh18N9 steel used in fuel assemblies is verified based on the results of experimental investigations. The characteristics of the investigated steel under dynamic loading are determined by the Kolsky method using Nicholas scheme. The values of the ultimate strength and their dependence on the strain rate and temperature are found based on the actual strain diagrams obtained under different loading conditions. The parameters of the Johnson–Cook model were determined from the experimental results on deformation of 12Kh18N9 steel under static and dynamic loading. The verification process showed that the selected mathematical model gives a very accurate description of the behavior of the material under dynamic loading conditions.

The 230U yield in the direct reactions 232Th(3He, 5n)230U and 232Th(4He, 6n)230U on thick ThO2 targets irradiated by 65 and 55 MeV nuclei 3,4He was measured and compared with calculations. The experimental yield of 230U in the reaction 232Th(3He, 5n)230U is approximately 10 times higher than the value computed using the constants from TENDL-2019. The experimental yield of 230U in the reaction 232Th(4He, 6n)230U the same, to within t5he limits of error, as the computed value. Comparing the experimental and computed yield in the reaction 232Th(4He, 6n)230U with the cross section obtained by the ALICA program (USA) showed that the computed cross section is incorrect and the prognoses for 230U production based on it are unsubstantiated. It is concluded that the reactions 232Th(3He, 5n)230U and 232Th(4He, 6n)230U are inefficient for 230U production in the U-150 cyclotron at the NRC Kurchatov Institute in amounts required for clinical applications.

The efficacy of uranium and thorium waste in a blanket of a thermonuclear plant, whose tasks are breeding of tritium as fuel for a thermonuclear plant and the production of fuel for thermal reactors, are compared: 239Pu on loading raw uranium and 233U on loading thorium as well as power generation. Different blanket arrangements are considered and investigated. It is shown that the tasks can be fulfilled only within the framework of a lithium-uranium blanket, whose characteristics in the considered variants are significantly higher compared to lithium-thorium.

Single cascades of centrifuges for simultaneously concentrating three components of separable mixtures are considered. A method is proposed for optimizing the parameters of a cascade with expansion of flow from squared sections that has an additional product flow. By way of an example, the correctness of the proposed method was checked by comparing with known data the optimization of a squared cascade with an additional product stream. It is shown that on separating a three-component mixture and concentrating the intermediate mass components under the same external conditions a cascade with “expansion” of the flow turns out to be more efficient from the standpoint of the relative summary flow.

Post-reactor studies of experimental fuel pins with mixed nitride uranium-plutonium fuel and EP823-Sh, ChS68-ID kh.d., and EK164-ID kh.d. steel cladding, which have been irradiated in BOR-60 and BN-600 reactors, indicated the nitriding and carburization of the cladding inner layer, which leads to an increase in its microhardness and a reduction in ductility characteristics. In fuel pin claddings, made of ChS68-ID kh.d. and EK164-ID kh.d. austenitic steel, the processes of carburization and nitriding correlate with sample destruction during mechanical tests. The present article generalizes and analyzes the results of post-reactor studies concerning the carburization and nitriding of the fuel pin claddings made of ChS68-ID kh.d., EK164-ID kh.d., and EP823-Sh steel. The carburization and nitriding of the cladding inner surface was studied by electron microprobe analysis using the cross-sections of fuel pins.

The possibility of using the Monte Carlo method for high-speed calculations of the fast-neutron fluence on a VVER-1200 vessel during some time interval is substantiated. In the model developed, the flow attenuation behind the core is described using the MCU-PD code in a computational regime at low power with an external source set alternately in the FA at the periphery of the core. The distribution of a time-varying source of fast neutron over the core is determined, in the diffusion approximation by means of the DYN3D engineering code, for actual reactor operation. The source irregularity along the FA profile is described in the x and y directions in the P1 approximation.

A method of calculating the neutron-kinetics parameters with and without the aid of the value function implemented in the MCU program code and the computational results are presented. Calculations of nineteen critical experiments from the ICSBEP Handbook were performed in order to investigate the influence of the neutron-value function on the accuracy of the calculation of the effective delayed-neutron fraction βeff and the prompt neutron generation time Λ. Analysis showed that the neutron-kinetics parameters are determined with greater precision if the value function is taken into account.

Decontamination solutions with high chemical activity effecting the corrosion of the structural materials used in equipment are used in the radiochemical industry. This communication reports on investigations of the resistance of siliconized graphite in solutions of alkalis and acids. Preliminary results on the satisfactory stability of material in alkali are presented.

The developed model employing the KARAT code made it possible to study the dynamics of plasma flows and neutron generation in a laser-plasma diode. The depth of the skin layer, instead of the Debye radius, as a spatial discretization parameter significantly reduced the number of grid nodes. The isolation system based on permanent magnets showed almost complete suppression of the return electron flow in the diode. The calculation also showed that an increase in the voltage in the discharge gap effects an increase in the ion current but it effects an even greater increase in the neutron yield. Investigations performed on an experimental model of the generator confirmed these findings.

This article continues the debate on choosing a sequel to a strategy for the development of nuclear energy in the country. The provisions set forth in articles by the National Research Center Kurchatov Institute on the transition to a thorium-uranium nuclear fuel cycle are analyzed and compared with the Strategy for the Development of Nuclear Energy in Russia to 2050 and Prospects for the Period to 2100, which the Presidium of the Scientific and Technical Council of Rosatom State Corporation approved in 2018. The environmental impacts of nuclear power under these two approaches are compared. It is concluded that in an open nuclear fuel cycle thorium-uranium fuel has no significant advantages, including in the sphere of radiation safety.

Aspects of experimental investigations of the physical characteristics of reactor cores, performed at Aleksandrov NITI as part of tests conducted on prototype test bases for nuclear propulsion reactors, are analyzed. Aside from conventional methods, new or improved methods of measurement were used to determine the neutronics characteristics. One of the basic problems solved by the Institute’s experimental physicists for the purpose of increasing the quality of neutronics measurements is the development of wide-ranging apparatus, based on modern hardware components, for monitoring the neutron flux density and for performing reactivity calculations.

The results of mathematical modeling performed to make an assessment of the strength, the normal operation, and the resource of the critical element of the load-transfer machine. Good convergence obtains with measurements of the deflection of the central part of the chain on a real structure. The results of the modeling are used in designing a chain-type cable layer whose purpose is to bring electric cables to mobile equipment. It is shown that the considered design meets the requirements of strength under static and dynamic loading. The resource characteristics are substantiated for a specified number of loading cycles.

The results of R&D work on nuclear batteries of the capacitor-type with a liquid electrolyte are reported. A high-porosity carbon matrix containing the radioactive isotope 90Sr is used in one of the electrodes of the new type of nuclear batteries. Ionic liquids, where the charge carriers are free ions formed on dissociation of ionic pairs of the ionic liquid on account of the kinetic energy of the β-particles of 90Sr, are used as the electrolyte. The charges are separated by a double electric layer at the boundary of an electrode and liquid electrolyte similarly to supercapacitors.

It is proposed to use a direct-flow plasma centrifuge with a double skimmer to isolate intermediate-mass components of spent nuclear fuel. The separation of a three-component mixture of ions with masses 90, 140, and 240 amu is considered. It is shown that efficient isolation of the intermediate-mass component on a separate collector is possible.

On using neutron generators for geophysical research the efficiency of logging equipment depends on the amplitude and time characteristics of neutron fluxes, which are influenced by the magnetic field inside the ion source, the composition and pressure of the gas, the frequency of anode voltage pulses, and other factors. The amplitude-time characteristics of the generated neutron pulses depend on the discharge burn regime in the ion source. The key parameters determining the conditions of ignition of the discharge and the stability of the operation of the ion source are the distribution and strength of the magnetic field in the discharge cell. This article reports the results of experimental investigations of the amplitude–time characteristics of the drawn current of small-sized Penning ion sources on changing the induction and configuration of the magnetic field in a low-frequency operating regime of a pulsed power supply system.

As an alternative approach to the transmutation of Np, Am, and Cm in fast power reactors, the use of a specialized facility is considered, which includes a molten-salt burner reactor and module for fuel reprocessing and fabrication. Based on the requirements for the consumer properties of the installation, a comparative analysis of transmutation for LiF–NaF–KF and LiF–BeF2 salts was performed. Calculations of the transmutation characteristics were performed for model systems. The computational results show the advantages of a reactor based on the LiF–NaF–KF salt on account of the higher solubility of actinide fluorides. For a given transmutation capacity, a burner reactor based on LiF–NaF–KF consumes several-fold less power-grade plutonium, right up to its absence in the equilibrium regime, as well as lower thermal power and lower production of fission products. Of note is a particular aspect of the molten-salt incinerator reactor that is associated with the accumulation of even-numbered plutonium isotopes during its operation whereby the plutonium extracted from the facility is actually converted into long-lived radioactive waste.

Project Proryv (Breakthrough) is leading an effort to develop fast reactors with mixed uranium-plutonium nitride fuel as well as the installations and technologies of a closed nuclear fuel cycle (NFC) as a base for a new type of nuclear energy. The RTM-2 complex of computer programs, which performs combined modeling of the reactor core and NFC stages, was developed for performing complex investigation of NFC closure and optimizing the loads for the reactor installation. This article discusses the current state of the RTM-2 software complex and the development of an RTM-2-based virtual digital model for NFC closure in the context of the on-site power complex. In particular, it is within the purview of such a model to integrate with a verified and certified next-generation computational code EVKLID/V1 for modeling thermohydraulic processes in fast reactors as well as with the digital information model of the reactor core, which is a tool for processing and analyzing the controlled characteristics of the core and its components and fulfilling the nuclear and radiation safety requirements throughout the entire life cycle of the installation.

A neural network of a model point reactor with instantaneous feedback on power was built in order to investigate and illustrate the possibilities of neutral networks in application to problems of reactor physics. A neural network is tested for its ability to predict a reactor’s capacity in the presence of uncontrolled perturbations of the reactor’s reactivity and errors of power measurements as well as for its use outside its training range. The obtained results show that in terms of accuracy a neural network model under the same conditions is highly competitive with or can appreciably surpass simple analytical assessments.

An appraisal was made of the influence of the edge effect, which is induced when branch-pipes are welded to the nozzles of a multiple forced circulation loop of power units with RBMK-1000, on the stress-strain state of the welds between the branch-pipes and the pipelines attached to them during the process of validating the concept “leak before break” (LBB). On the basis of finite-element modeling and an analytical solution of a problem of theoretical elasticity – the pinching of the edge of a cylindrical shell, it is shown that the edge generates local stresses, whose distribution is skew-symmetric with respect to the equidistant surface of the nozzle. These stresses do not contribute to the ductile failure of structural elements. In addition, on account of the rapid damping the response from the effect does not affect the stressed state of the welds between branch-pipes and pipelines. It is shown based on the findings that the approaches to the computational validation of the LBB concept for such welds are identical to those applied to the welds on straight sections of pipelines and do not need additional correction.