Journal Issue: Nuclear Energy and Technology (NUCET)
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Volume
2025-11
Number
2
Issue Date
Journal Title
Nuclear Energy and Technology (NUCET)
Journal ISSN
2452-3038
Том журнала
Том журнала
Nuclear Energy and Technology (NUCET)
Nuclear Energy and Technology (NUCET) (2025-11)
Статьи
Публикация
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Modeling of a device for conducting experiments under low reactor neutron fluence rate conditions
(НИЯУ МИФИ, 2025) Veretennikov, D. G.; Abdelmeguid Abouellail; Ahmed H. Ali; Bedenko, S. V.
The study focuses on modeling a device designed for conducting experiments under low reactor neutron fluence rate conditions. The relevance of the work is driven by the challenges of accurately measuring low reactor neutron fluence rates, which are critical for the safety and efficiency of nuclear reactors and for conducting research in various scientific fields. To create a directed neutron flux, a horizontal collimation channel of the Prizm-AN irradiation device with a radioisotope capsule neutron source of the internal bounded neutron (IBN) type is being investigated. The paper presents the results of particle transport modeling using the PHITS and SOURCES-4C codes, which allowed for the evaluation of the generated neutron fluence rate parameters and the effective dose rate of radiation. The obtained results confirm the functionality of the proposed device and its compliance with radiation safety requirements, making it suitable for conducting experiments similar to those performed on reactor installations.
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Semi-empirical predictive calculation of the thermal-physical properties of potassium-sodium melts based on their component data
(НИЯУ МИФИ, 2025) Terekhov, S. V.
Use of liquid-metal coolants in nuclear power plants has been the cause of unfailing interest in thermal-physical properties of metals and their alloys from both experimentalists and theorists. Power polynomials are used by experimentalists to approximate temperature changes in heat capacity, the coefficient of thermal linear (volume) expansion and other quantities. These polynomials have different form in different temperature intervals and need to be joined at the interval ends. This approach creates a number of difficulties in developing a unified methodology for calculating not only the thermal functions of metals, but also for predicting the behavior of their melts and alloys. The following was used in the paper to solve the problem at hand: the author’s model of a two-phase local-equilibrium region (with different order parameters) and the modified rule of component mixing, taking into account the coordinated arrays of experimental data on the initial metals for calculating predictively the thermal-physical performance of potassium and sodium melts. It has been shown that using the model of a two-phase local-equilibrium region, new approximating functions and empirical formulas lead to a sufficiently adequate estimation of heat capacities, thermal linear expansion coefficients, densities, thermal conductivities and thermal diffusivities of melts. A discrepancy has been found between the mathematical description of the thermal conductivity of the K56Na44 melt and its experimental values.
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Boron-containing radiator coatings of ionization chambers of neutron flux detection sensors
(НИЯУ МИФИ, 2025) Baskov, P. B.; Salamakha, B. S.; Glazyuk, Ya. V.; Namakshinas, A. A.; Bondarenko, S. A.; Mushin, I. M.; Khudin, A. S.
The proposed solution for the production of new boron-containing radiator materials – composite coatings with neutron conversion material (10B isotope) – for ionization chambers electrodes is based on the technology of chemical and structural modification of the surface using heat-resistant oxide materials (silicon and boron oxides). The paper presents the results of the development of a neutron-sensitive radiator material based on an oxide micro-dimensional composite coating consisting of an intermediate adhesive layer of silicon dioxide (SiO2) and a neutron-sensitive functional layer of boric anhydride (B2O3). The technological basis consists of the sequential processes of thermal destruction of polyorganosiloxane of the PES-5 brand and pyrolytic decomposition of boric acid (H3BO3). Studies using infrared and fiber-optic interferometric spectroscopy have shown that during the formation of a silicon dioxide layer, an amorphous silicate of a linear chain structure with a developed surface is formed, which contributes to the subsequent formation of a functional boric anhydride layer. The specific neutron sensitivity of boron-containing radiator coatings was measured by alpha spectrometry. It was found that at a surface density of boric anhydride of 2.5 mg/cm2, the specific neutron sensitivity is of the order of 3∙10-18Coulomb/neutron. It is shown that the boron-containing radiator coating retains its integrity and nuclear physical properties during thermocyclic tests (4 cycles at 600 °C). The boron-containing radiator coating is characterized by high adhesion properties to the metallic surface of the electrodes (grade 321 steel) of the ionization chamber. The composite coating is resistant to vibration when exposed to high-frequency (200 Hz) and low-frequency (6–35 Hz) vibration loads.
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Synthesis, investigation of the structure and physico-chemical properties of modified solid-phase extractants (SPE) based on N,N,N’,N’-tetraoctyldiglycolamide (TODGA)
(НИЯУ МИФИ, 2025) Savelev, A. A.; Rachkov, V. I.; Рачков, Валерий Иванович; Савельев, Александр Александрович
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 SPE TODGA, in particular for the isolation of minor actinides, including americium-241.
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Computational study into the experimental capabilities of the MTRR SCW reactor
(НИЯУ МИФИ, 2025) Lapin, A. S.; Blandinsky, V. Yu.; Nevinitsa, V. A.; Pustovalov, S. B.; Sedov, A. A.; Subbotin, S. A.; Fomichenko, P. A.
Two stages of the MTRR-SCW reactor operation are planned: a test stage and a research stage. This paper considers the research stage of the MTRR-SCW experimental reactor operation, the purpose of which is to investigate current and advanced light-water reactors. The MTRR-SCW driver-type core provides a fast neutron spectrum with the possibility for the local warmup in ampoule devices and independent loop channels. Irradiation channels will be installed in the core center and periphery, as well as instead of the reactor’s changeable reflector cartridges. The MTRR-SCW irradiation channels and independent loops will provide ample opportunities both for undertaking a research on effects from neutron irradiation of different materials, and for testing a variety of fuel assembly designs and operating conditions (temperature, pressure, neutron spectrum), as well as for investigating transients and emergency processes. The MTRR-SCW channels can be used to irradiate different types of fuel, and structural and absorbing materials with different coolant inlet temperatures (from 250 to 450 °C) and, consequently, its inlet density (from 800 to 100 kg/m3 respectively), providing different neutron spectrum options for the experimental fuel assembly in a range from thermal to fast spectrum. The MTRR-SCW allows experiments to increase power and simulate emergency processes, including reactivity accidents (RIA). The strong primary and safeguard vessels of the independent loop channels also make it possible to simulate loss-of-pressure emergencies of the LB LOCA and SB LOCA types. The peripheral independent loop channel will allow undertaking experiments for simulation of alternative reactor concepts with reactors with SKD coolant parameters, such as a single-circuit concept with the pseudophase transition in the core (VVER-SKD-1700), and with natural coolant circulation in the core (SKDI). In addition, the peripheral channel allows accelerated irradiation of fuel rods used in current VVER reactors, taking into account the reproduction of the ratio between damaging dose and burnup rates.