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Кутеев, Борис Васильевич

Организационные подразделения

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Институт лазерных и плазменных технологий

Стратегическая цель Института ЛаПлаз – стать ведущей научной школой и ядром развития инноваций по лазерным, плазменным, радиационным и ускорительным технологиям, с уникальными образовательными программами, востребованными на российском и мировом рынке образовательных услуг.

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Кутеев

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Борис Васильевич

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Теперь показываю 1 - 10 из 19

Только метаданные
Concept development and candidate technologies selection for the DEMO-FNS fuel cycle systems
(2021) Ananyev, S. S.; Ivanov, B. V.; Dnestrovskij, A. Y.; Kukushkin, A. S.; Kuteev, B. V.; Кутеев, Борис Васильевич
Within the framework of activities on a pilot industrial hybrid reactor (pilot hybrid plant-PHP) development, the DEMO-FNS fusion neutron source is being designed in Russia. Russian fusion and fission community considers it as the main facility developing prospective nuclear technologies, which should update the results of the ITER researches in the field of physics and control of burning plasma. Progress in the development of the plant and simulations of the fusion fuel cycle allows starting the engineering design of basic fuel cycle systems. In this paper, we consider the new prospective of technological solutions for the tokamak fuel cycle. The possibility of their integration and the analysis of the technology readiness level for them in the Russian Federation are addressed as well. The code simulating the hydrogen isotope flows in a tokamak was significantly upgraded for evaluation of the specific technological solutions as well as for combined modeling of the fuel circulation in the core and divertor plasmas allowing for the pumping, processing, and fuel isotope content control in the plasma. The article describes the core and divertor plasma model used for the calculations and shows the D/T fuel throughput values, which should be provided by the fuel injection systems (NBI and pellets). The dependencies of the HFS and LFS pellet injection frequency necessary for the main plasma supply and ELM control are found for different core and divertor plasma parameters. For the DEMO-FNS core plasma, an operating window has been found in terms of fueling control parameters. It was shown that D and T throughputs of 40-50 Pa center dot m(3) s(-1) should be provided by the fuel injection systems (NBI and pellets) to maintain the fusion power up to 40 MW. For the DEMO-FNS core plasma, an operating window defined by the particle confinement times and fuel injection throughputs has been found.
Только метаданные
Comparison of Lithium Divertor Options for the DEMO-FNS Tokamak
(2021) Skokov, V. G.; Sergeev, V. Y.; Anufriev, E. A.; Kuteev, B. V.; Кутеев, Борис Васильевич
© 2021, Pleiades Publishing, Ltd.Abstract: The choice of an idea for a divertor with evaporating liquid lithium that meets the requirements for removing the thermal load from the edge plasma and provides an acceptable level of change in the ionic composition of the main plasma for the DEMO-FNS tokamak being developed in Russia has been discussed. The results of numerical simulation and optimization of the design of divertors with multiple volumes sectioned by slotted diaphragms have been presented. The parameters of lithium streams flowing into the edge layer have been estimated for the temperature range of divertor chambers from 500 to 1000 K under the conditions of the gas-kinetic and free-molecular modes of lithium vapor outflow from the divertor. Analysis of the processes that reduce the outflux of lithium from the chambers and its penetration into the main volume of the plasma inside the separatrix showed that sectioning effectively reduces the outflow streams to acceptable levels of ≈1020 atom/s.
Только метаданные
THE CLOSED LITHIUM LOOP CONCEPT FOR DEMO-TIN TOKAMAK WITH SECTIONED DIVERTOR КОНЦЕПЦИЯ ЗАМКНУТОГО ЛИТИЕВОГО ЦИКЛА В УСТАНОВКЕ ДЕМО-ТИН С СЕКЦИОНИРОВАННЫМ ДИВЕРТОРОМ
(2024) Sergeev, V. Yu.; Skokov, V. G.; Kuteev, B. V.; Timokhin, V. M.; Кутеев, Борис Васильевич
Только метаданные
PHYSICAL RESEARCH PROGRAM ON THE T-15MD TOKAMAK ПРОГРАММА ФИЗИЧЕСКИХ ИССЛЕДОВАНИИ НА ТОКАМАКЕ Т-15МД
(2024) Velikhov, E. P.; Kovalchuk, M. V.; Anashkin, I. O.; Kirneva, N. A.; Kuteev, B. V.; Marenkov, E. D.; Melnikov, A. V.; Кирнева, Наталья Александровна; Кутеев, Борис Васильевич; Маренков, Евгений Дмитриевич; Мельников, Александр Владимирович
Только метаданные
DEVELOPMENT OF NEUTRON-PHYSICAL MODEL OF HYBRID REACTOR DEMO-FNS BY MEANS OF NESTOR CODE AND MONTE-CARLO METHOD РАЗВИТИЕ НЕИТРОННО-ФИЗИЧЕСКОИ МОДЕЛИ ГИБРИДНОГО РЕАКТОРА ДЕМО-ТИН С ИСПОЛЬЗОВАНИЕМ КОДА NESTOR И МЕТОДА МОНТЕ-КАРЛО
(2023) Shlenskii, M. N.; Dlougach, E. D.; Kuteev, B. V.; Кутеев, Борис Васильевич
Только метаданные
Nuclear heat loads to the first mirror unit of h-alpha diagnostic in the ITER equatorial #12 port ТЕПЛОВЫЕ НАГРУЗКИ НА УЗЕЛ ВХОДНОГО ЗЕРКАЛА ДИАГНОСТИКИ СВЛ В ЭКВАТОРИАЛЬНОМ ПОРТУ № 12 ИТЭР
(2020) Afanasenko, R. S.; Alekseev, A. G.; Morozov, A. A.; Vukolov, D. K.; Kuteev, B. V.; Кутеев, Борис Васильевич
© 2020 National Research Center Kurchatov Institute. All rights reserved.3D modeling of neutron associated processes is performed to assess radiation loads in the area of the first mirror unit (FMU) of H-Alpha and Visible Spectroscopy Diagnostic located in ITER Equatorial port #12, including neutron/gamma and total heating, and absorbed dose rate in the FMU components. A temperature analysis of the structure elements is done being based on the nuclear heating data. Calculations are based on the latest ITER 40ºneutronic C-Model and highly detailed FMU model developed using code SuperMC. The highest nuclear heat load 0.8 W/cm3 is derived for the front-end wall of the First Mirror Unit. Thermal analysis shows low temperature gradient around 50 °C in FMU housing area. Maximum temperature obtained does not exceed 460 °C on the front-end wall of the FMU.
Только метаданные
Radiation loads on the first mirror unit of h-alpha diagnostic in the iter equatorial port No 12
(2020) Alekseev, A. G.; Morozov, A. A.; Vukolov, D. K.; Afanasenko, R. S.; Kuteev, B. V.; Кутеев, Борис Васильевич
© 2020 National Research Center Kurchatov Institute. All rights reserved.3D profiles of radiation loads were derived for the first mirror unit (FMU) of H-alpha and visible spectroscopy diagnostic located in ITER equatorial port no 12, including neutron/gamma fluxes, and radiation damage levels in the FMU components, using the latest ITER neutronic C-Model and highly detailed FMU models. Good conformity between the results obtained by MCNP and FISPACT-II codes had been demonstrated with the discrepancy within 10%. The highest radiation loads and ~0.22 dpa radiation damage are derived for the front wall of the FMU (at 0.54 FPY-by the end of the ITER operation).
Только метаданные
Integration of coupled modeling of the core and divertor plasmas into “FC-FNS” code and application to DEMO-FNS project
(2020) Ananyev, S. S.; Dnestrovskij, A. Y.; Spitsyn, A. V.; Kukushkin, A. S.; Kuteev, B. V.; Кутеев, Борис Васильевич
© 2020The “FC-FNS” (Fusion Cycle for Fusion Neutron Source) fuel cycle model is used to calculate the hydrogen isotope flows in the fuel systems of a tokamak-based fusion neutron source (FNS). The present development incorporates into the FNS model the coupled modeling of the core and divertor plasmas that provides the relation between the plasma parameters and particle fluxes through the vacuum chamber in steady state operation. The upgraded “FC-FNS” code is applied to the analysis of the DEMO-FNS project based on a tokamak with the parameters R/a = 3.2 m/1 m, B = 5 T, Ipl = 4–5 MА, PNBI =30 MW, PECR =6 MW and DT fusion power Pfus =40 MW. The possibility of reducing the hydrogen fluxes in the pumping and injection systems when using an admixture of Ne in a diverter is substantiated. The operation mode of the DEMO-FNS fueling systems depending on the main plasma parameters is analyzed and the gas flows for different D2/T2/DT pellet injection modes are evaluated.
Только метаданные
Discharge duration limits of contemporary tokamaks and stellarators
(2020) Sergeev, V. Y.; Kuteev, B. V.; Кутеев, Борис Васильевич
© 2020 IAEA, Vienna.The realization of long-time discharges in magnetic confinement devices (MCDs), namely, in tokamaks and stellarators, is a key issue in the development of controlled fusion energy. Experiments demonstrate that the discharge duration achieved in contemporary MCDs is limited and the normal operation is terminated after 0.3 to 3600 s depending on plasma, discharge and device parameters. This paper is devoted to the analysis of physical mechanisms which may be responsible for the discharge duration limit in MCD operation. The impact of heat transfers to plasma facing components from the plasma, coolants and thermal radiation is evaluated and compared with the available experimental database. The critical temperature T cr ≅ 2300 K of plasma facing components is considered as the key parameter that limits the discharge duration. The regimes of the first wall temperature growth governed by both the heat conductivity and heat capacity are identified experimentally and analytically. Heat removal from wetted areas by means of thermal surface radiation and linear heat transfer to the coolant is identified as the key physical mechanism that determines the boundary of time-limited discharges in MCDs. The principal role of localized wetted areas with a size of ≅0.2-0.6 m2 is revealed for operation of contemporary devices. This means that in further development of fusion reactors major attention should be devoted to the organization of a more uniform and distributed heat exhaust. It is shown that the proposed semi-analytical approach explains the experimentally discovered trends in the MCD operation and may be used for the evaluation of the discharge duration limit of new facilities designed to obtain steady-state discharges.
Только метаданные
Eutectic Lead–Bismuth Alloy as a Possible Coolant in the Fusion Reactor Cooling System
(2024) Deryabina, N. A.; Kuteev, B. V.; Pashkov, A. Yu.; Shpanskiy, Yu. S.; Кутеев, Борис Васильевич