The cross sections of the reactions 186W(4He, p3n)186Re, 186W(4He, pn)188Re, and 186W(4He, p)189Re were measured by the activation method in the energy range of 4He nuclei 24–63 MeV. For all three reactions the cross sections in the TENDL-2019 nuclear data library differ from experiment by more than a factor of 2 in a wide energy range.

The chromium-nickel alloy 42KhNM developed at VNIINM is used for the cladding on the fuel rods in nuclear icebreakers as well as the absorbing rods for the control and protection of VVER reactors. The alloy is promising for use in the cladding of dispersion-type fuel rods for low-power nuclear power plants (RITM-200 and Shelf-M installations) and floating power units as well as VVER container-type

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

Calculations of the personnel irradiation dose rate were performed to assess the radiation consequences of the rupturing of a tube in the sampling line in a complex preparing samples for automated chemical monitoring of the primary loop of NPP with VVER-1000. The calculations were performed by the Monte Carlo method using the MCU-PD high-precision code based on data on the activity of fission products

In connection with the use of the SPZO-M modernized system for prestressing the NPP containment, it became necessary to determine the effect of loads inside the containment wall on the stress-strain state of the containment. Because of the particularities of the technology used to install the reinforcing elements, where the ducts are filled with grout under pressure up to 1 MPa, there is a possibility

An economic and mathematical model has been developed to determine the competitiveness of REMIX-A and -C uranium-plutonium fuel obtained by reprocessing spent fuel from thermal reactors. The presented model makes it possible to calculate the cost metrics of nuclear fuel components and the fuel component of the cost of NPP electricity as a function of the cost metrics of key process changeovers. It

Mathematical models of the corrosion destruction of protective barriers of nuclear and radiation hazardous objects in the marine environment and the software tools developed for predicting the time of destruction and possible consequences of depressurization with pollution of the marine environment are analyzed. The models are based on the idea of an assemblage of protective barriers in the form of

Matters concerning the improvement of safety guidelines related to the use of risk-informative methods in determining the safety class of elements of nuclear power plants in the process of design and assessment of malfunctions are considered. Clarifi cations of the terms and defi nitions used are proposed. Considerable attention is given to the interpretation and calculation of the conditional probability

The exportation of fast reactors with a closed nuclear fuel cycle to countries that do not have nuclear weapons could commence within the next few decades, which will require strengthening the non-proliferation regime and improving the effectiveness of IAEA safeguards. Both requirements can be achieved by improving the system of accounting and management of nuclear materials and the reliability of

The results of experimental studies of the hydrodynamics of the model coolant flow at the entrance to the VVER-1000 core support grid are presented. A 1:17 scale model of the in-reactor pressure path was built, which allowed physical modeling of the conditions for coolant flow entry into the core by installing support risers. A method was developed for measuring the flow velocity distribution at the

The project Proryv [Breakthrough] now being implemented in our country is aimed at achieving a new quality of large-scale nuclear power, development, origination, and industrial implementation of a closed nuclear fuel cycle with thermal and fast reactors. Here the current status of the technologies for reprocessing spent nuclear fuel and recycling plutonium fuel is analyzed, and it is shown that the

ETVS fuel assemblies containing nitride fuel rods with different structural implementation and cladding made of different materials are tested in the cores of the BN-600 and BOR-60 reactors in order to substantiate the efficiency of fuel rods with mixed uranium-plutonium nitride fuel used in sodium-cooled (BN-1200) and lead-cooled (BREST-OD-300) reactors. We report here the results of post-reactor

The results of a study of the effect of the homogeneous addition of 237Np, 241Am, 243Am into the fuel on the neutronics characteristics of a lead-cooled fast reactor are reported. The reasons for the change in the delayed- neutron fraction and the effective delayed-neutron fraction, breeding ratio, density effect of reactivity, and residual energy release of the fuel unloaded from the core are analyzed

Experimental data were obtained on the kinematic viscosity of 0.75LiF–0.2BeF2–0.05ThF4 melt in the temperature range 920–1155 K in order to substantiate the concept of a reactor with circulating fuel based on molten salts of lithium, beryllium, and thorium metal fluorides with added fissile materials. It is shown that the obtained viscosity is described by an exponential dependence on the reciprocal

Currently, the typical risk for the global nuclear energy industry is the risk of violating the planned NPP construction schedule with the concomitant increase in cost. The results of an (accounting) settlements comparison of the realization of such risk situations during the construction of large, small, and medium- size NPPs are discussed. Changes in project efficiency metrics, such as the payback

The designs of gas-filled ionization chambers KNK-2-7M and KNK-2-8M are described. The main metrological characteristics of the detectors are presented. A method is given for determining the fluence of >0.1 MeV neutrons in the process of testing radio electronic equipment for radiation resistance in reactors with a metal core, aperiodic solution reactors, the BIGR reactor, NPP reactors, and hydrogen-containing

A mathematical model of the technological process of the carbothermal synthesis of mixed uranium-plutonium nitride fuel is presented. The model was developed for use in the KOD TP software complex – a code for optimization and diagnostics of technological processes. The applied studies performed in this work are aimed at determining the optimal technological modes and increasing the efficiency of the

This article contains information on the structure and number of medical radiological imaging procedures by type and organ group and the collective and average radiation doses in 2018 for patients served by the FMBA of Russia. The radiation dose from fluorography, radiography, fluoroscopy, tomography, and other procedures in different regions of the country was analyzed. Compared to the preceding period

149Tb is currently considered to be a promising α-emitting isotope radioimmunotherapy. It is shown that it can be obtained without the use of a carrier by irradiating a target consisting of Eu2O3 with 4He with energy 63 MeV. After irradiation, radiochemical separation of terbium radioisotopes was performed in a time not exceeding the half-life of 149Tb (4.12 h). It was based on reductive co-precipitation

It was clear from the outset of work on nuclear power that liquid metals are attractive coolants (small neutron absorption cross section, high thermal conductivity, etc.). The measurements of temperature fields in flows of liquid metals form the basis for obtaining reliable calculated dependences on the intensity of heat transfer. This article calls attention to the necessity of preserving this data

The results of the dismantling of the active drain system and the decontamination of technological enclosures of the loop installations of the MR reactor, which is being decommissioned at the NRC Kurchatov Institute, are presented. The radioactive contamination of the pipelines of the system and the soil had a complex nuclide composition. Methods for assessing the volume of radioactive waste being

The authors’ point of view, as expressed in publications, on the evolution of models taking into account the irregularity of irradiation in time, which are contained in the ICRP Recommendations and are used in radiation risk assessment, is generalized. The formal nature and imperfection of these models is noted. The need to build models based on biological concepts of radiation carcinogenesis is justified

The operational features determining innovative approaches to the design of space nuclear energy installations are analyzed. It is shown that there is no alternative to space nuclear energy for solving potential problems and the main directions of its use are determined. The main results obtained at NIKIET on the development of a nuclear rocket motor and space nuclear energy installations with turbomachine

Tests of the fuel rods of a high-temperature gas-cooled reactor were performed in the loop installation of the MIR.M1 reactor. General information about the objectives of the experiment and the modernized loop installation is presented. The designs of the loop channel, experimental fuel assembly, and fuel rods, including the thermometric versions, are described. The main results of post-reactor studies

A mathematical model of remote diagnostics of the condition of sealed cable glands in NPP was developed based on an analysis of methods and techniques used in on-line monitoring of sealed cable glands. The availability of a mathematical model of the condition of sealed cable glands is important for practical purposes and necessary for their reliability assessment under NPP operating conditions, especially

Experimental fuel rods with mixed uranium-plutonium nitride fuel are being tested in the BOR-60 reactor as part of collapsible irradiation devices, which are prototypes of fuel rods of the BREST-OD-300 reactor being designed. Currently, reactor tests of fuel rods with a helium sublayer as part of OU-2, as well as lead alloyed with components of EP823-Sh steel as part of OU-4, are continuing. In 2018

A two-stage scheme is proposed to reduce the concentration of 232,234,236U in regenerated uranium hexafluoride. In the first cascade, waste uranium hexafluoride, whose contaminated stream is taken in the additional product, is directed into the additional feed. The main product stream of the first stage, enriched in 235U, is directed to the feed of the second stage. Its waste purified from 232,234U

The possibility of using analog simulation of neutron evolution in conjunction with a non-stationary thermohydraulic calculation in reactor dynamics problems is considered. The coordination of neutron and thermohydraulic calculations is accomplished within the framework of an asynchronous approach. Numerical examples illustrating the applied approach on models of typical operating modes of nuclear

A method of deep burial of long-lived radioactive waste and spent nuclear fuel by melting of the geological mass due to the residual heat release of radionuclides confined in a container is analyzed. A variant is proposed for regulating the power of heat release by adding relatively short-lived highly active radionuclides inside a capsule, which also accumulate during the burnout of nuclear fuel in

A reliability evaluation of the unique equipment in the nuclear fuel fabrication/refabrication module of a pilot-demonstration energy complex during process operations is a pressing problem that affects both the economic efficiency and the safety of the entire complex. The developed approaches to the qualitative and quantitative reliability analysis are aimed at accounting for features of the fabr

The aim of this work is to make an assessment of the dynamics of the radioecological conditions in the 30-km zone of influence of the Beloyarsk NPP after the commissioning of the BN-800 power-generating unit. Based on the established network of terrestrial ecosystems, it is shown that the operation of BN-800 did not increase the content of 90Sr, 137Cs, and 239+240Pu in the soil and vegetation. The

Fast reactors can be used to adjust the isotopic composition of plutonium from spent mixed uranium-plutonium fuel from PWR and make it recyclable. The efficacy of such an adjustment is directly related to the scope for breeding in fast reactors. This can be done in the operating BN-800 reactor by multiple recycling of plutonium in the fuel of the fast reactor as well as by using 235U or in specialized

The role and place of NIKIET, as one of the leading design organizations in the nuclear industry, in the joint implementation of the large-scale international thermonuclear reactor project ITER – unparalleled in world practice – are reflected in this article. The issues involved in the establishment of a quality management system at NIKIET in line with international specifications and a production

An analysis of existing and possible designs of high-temperature gas-cooled reactor installations has shed light on the critical elements determining the reactor’s efficiency and reliability; it required determination of approaches to the development and theoretical and experimental substantiation of solutions for such critical units. Among the problems requiring a separate investigation in the development

Different stages of the development of new construction materials at NIKIET for innovative designs of reactor installations are analyzed. Justification is given for the use of large-sized intermediate products made of a nickel-based alloy (KhN55MVTs (ChS57) grade) for vessel elements and intermediate products made of molybdenum alloy of the TSM-7 type for the core elements of a high-temperature reactor

The results of development work on and safety validation of high-density, low-enriched fuel for research nuclear reactors of Russian design are presented. Attention is focused mainly on the development of fuel assemblies with fuel based on UO2 + Al, U3Si2 + Al, and U9%Mo + Al for the LVR-15 research reactor (Czech Republic). A design-basis validation is given for IRT-4M fuel assemblies with fuel composition

A concept is presented for a pressurized, water-cooled, pool-type research nuclear reactor IVV-3 under consideration as a replacement capacity for the operating IVV-2M reactor. The expanded experimental capabilities of the IVV-3 reactor, which are associated with an increase in the thermal power and neutron flux density in experimental bodies, will make it possible not only to keep the existing research

In the context of the life-extension program for the No. 3 unit of the Smolensk NPP, the life-exhausted RBMK-1000 control and protection system, the safety control systems of the power unit, and the analog-relay control apparatus of the technological equipment of the block control board (reactor and turbine departments) have been replaced with systems and software-hardware suites (SHS) developed by

The IBR-2 (JINR) pulsed reactor is used in the field of solid state physics for experiments on extracted beams. It is predicted that the IBR-2 reactor will reach resource limits in 2032–2037. To keep and develop beam research at JINR, the development of a new pulsed neutron source has begun. According to estimates, the reactor will have average thermal power 12–15 MW, extracted by sodium coolant. At

The main directions of development of small-scale energy generation are shown and a list of developed small-scale energy projects (SNPP) is given. The main results of work performed in 2019–2020 on the conceptual design of SNPP based on a developed reactor installation are presented. The key directions of modernization of a prototype reactor installation are indicated, and a plant concept and individual

The basic requirements of regeneration of spent fuel from BREST reactors in order to reach radiation equivalence in the nuclear energy system are presented. It is shown on the basis of computational modeling of scenarios for the development and long-term operation of the nuclear energy system, including nuclear power plants with thermal and fast reactors, that radiation equivalence of waste and natural

The results of modeling the form change and cracking of graphite blocks under exposure to heat and radiation at the stage of their secondary swelling within the scope of scientific and technical support for the restoration of the lifespan characteristics of the graphite stack of power units with RBMK-1000 in order to predict the deformed state of graphite at the final stages of operation of channel

The work performed on the creation of the BREST innovative reactor technology based on a lead-cooled fast reactor, mixed uranium-plutonium nitride fuel, operating in a closed nuclear fuel cycle is reviewed. It is shown that a safe and competitive large-scale nuclear energy with practically unlimited fuel resources can be developed based on the BREST reactor technology. The choice of mixed nitride fuel

Power units with an RBMK reactor have reached the stage of termination of operation and preparation for decommissioning. Having exceeded the assigned life of 30 years, as a result of aging of the graphite stack some state parameters are close to their maximum permissible value. Further operation required restoration work, which was made possible by the creation and implementation of special robotics

The purification of VVER-1000 uranium-plutonium extract from fission products by liquid chromatography was tested. The process was conducted in a liquid chromatography column filled with silica gel whose pores were impregnated with nitric acid at concentration 1 mol/L. The purification of the uranium-plutonium extract was performed in two stages. The highest purification factors were obtained from

A method of changing the composition of trace impurities in the crystallization-recrystallization regime is proposed for obtaining ammonium tetrafluoroberyllate crystals by introducing a complexing agent (sodium ion in the form of sodium carbonate), which provides a more than three-fold reduction of the sum of limited impurities in crystals. Purer crystals yield correspondingly purer metallic beryllium

Variants of Np, Am, Cm incineration using a portion of the plutonium from VVER-1000, -1200 spent fuel in a high-power molten-salt nuclear reactor (MSR) with a cavity-type core and fuel salts with the compositions Li,Be,An/F and Li,Na,K,An/F are discussed. The high nickel alloy KhN80MTYu was chosen as the structural material for the fuel loop. To obtain the highest possible transmutation efficiency

The need to reduce operating costs of commercial uranium hexafluoride production motivates the present study. With this in mind, a variant of coolant flow reduction in the tubing of the desublimation apparatus is proposed. Mathematical calculations showed that the coolant consumption can be reduced by a factor of 2. Field experiments performed at an operating production facility have confirmed the

Approaches to modeling gas-phase transport in heavy liquid-metal coolant flow in the HYDRA-IBRAE/LM code, which is actual for safety validation of nuclear reactors with a leak in the steam-generator tubes, are examined. A hierarchy of models for determining the dispersed-particle size that are implemented in the code for bubble flow is presented: empirical correlation, transport equation for the density

A method was developed for measuring the specific activity of 90Sr in the field and a portable apparatus with the following characteristics was built: range of measurement of the activity of analytical soil samples 0.1–3000 kBq/kg and range for determining the overall activity of thin samples, for example, aerosol filters 0.05–100 kBq. A brief description of the method of radiometric measurement of

This article concludes a series of analytical publications on the largest nuclear accident in the history of the nuclear fleet, which occurred on August 10, 1985 on a submarine in the Chazhma Bay situated in the Primorsky Krai. The damaging factors of a thermal explosion of the nuclear reactor and their impact on people, structures, and watercraft are considered. Of note is a special feature stemming

The purpose of tests is experimental validation of VVER-1200 power unit operation in a daily load schedule and to work out a method of controlling and checking the operation of the equipment. A method of controlling the reactor based on the principle of localization of xenon processes in the top half of the core in combination with soft temperature control was tested. The method allows control of reactivity

The applicability of a high-pressure xenon based spectrometer for determining the parameters of radioactive contamination of soil or building structures by 137Cs resulting from a radiation accident was studied. For soil contaminated by 137Cs with an exponential depth distribution, the Monte Carlo method was used to calculate the characteristics of the radiation field at the detection height and the

The corrosion characteristics of the nickel-molybdenum alloys Hastelloy-N mod. (USA), GH3535 (China), KhN80MTYu (Russia), and KhN80MTB (Russia) were studied in the melt 0.465LiF–0.115NaF–0.42KF at temperature 750°C with added metallic tellurium, a fission simulator, and ytterbium to lower the redox potential. The corrosion rate of the alloys is in the range 100 ± 15 μm/yr and is due to the high rate

The Rosatom State Corporation activity in the nuclear fuel market requires the development of new designs and modifications meeting the current requirements of safety and operational reliability. The history of the development of a competitive fuel meeting heightened requirements of economic efficiency, safety, and operational reliability for NPP with VVER and PWR is presented. Ways to improve the

Technologies for bringing the Semipalatinsk test site into a safe state as implemented by VNIITF jointly with the VNIIEF, Radium Institute, NIKIET, and the National Nuclear Center of the Republic of Kazakhstan are shown. It is shown that as a result the Russian Federation, the Republic of Kazakhstan, and the United States jointly eliminated threats to the nonproliferation regime and expensive nuclear

The results of an investigation of fission gas release from mixed nitride fuel after irradiation to maximum burnup 7.5% h. a. in 10 experimental FA in BN-600 are reported. It is shown that gas release commences at burnup >2% h.a. and increases with increasing burnup. The burnup dependence of gas release for mixed and uranium nitride fuel coincide quite well. The fission gas beneath the fuel-rod cladding

The refueling system of sodium-cooled fast reactors is described. The history of the development of systems and their principles of construction are described. The sequence of operations performed in setting up the refueling equipment at the reactor and the main results of the research and development work performed in the sphere of refueling equipment are presented.

The history of the development and adoption of reliability analysis of equipment and systems as well as probabilistic safety analysis in the practice of design and operation of nuclear installations is presented. The basic characteristics of the developed software systems for performing probabilistic safety analysis of installations during the design process, risk monitoring during operation of NPP

Testing and experimental substantiation are a significant and integral part at all phases of the life cycle of reactor equipment. The Research and Testing Complex (RTC) operates at OKBM Afrikantov to ensure that these tasks are successfully fulfilled. The RTC provides the opportunity of performing a wide range of research and testing during the creation of reactor installations as well as the development