Персона: Кутеев, Борис Васильевич
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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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Борис Васильевич
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- ПубликацияТолько метаданныеLIGHT BEAM MODEL FOR NEUTRAL BEAM INJECTION OPTIMIZATION ЛУЧЕВАЯ МОДЕЛЬ ПУЧКА ДЛЯ ОПТИМИЗАЦИИ ПАРАМЕТРОВ НЕИТРАЛЬНОИ ИНЖЕКЦИИ(2021) Dlougach, E. D.; Ananyev, S. S.; Kuteev, B. V.; Кутеев, Борис Васильевич© 2021 National Research Center Kurchatov Institute. All rights reserved.A simple 3-dimensional model LNB (Light Neutral Beam) is proposed for fast ions current evaluation in fusion devices. The model implements analytical expressions for plasma description, fast ion generation and slowing down. LNB model allows one to evaluate the beam ionization in plasma volume, fast ions thermalization [1], current drive, and phase distributions. The technique can be compared with the analytical approach adopted in system codes, but LNB takes into account the particular details of beam-plasma shaping; thus all the beam-driven quantities in plasma are obtained regardless to the empirical scaling laws. The model is especially effective for compact and spherical devices optimization, as for those the injected beam box shape and the beam internal angular distribution can play an important role in the result power deposition and current drive. LNB provides a simple and efficient tool for fast verification, sensitivity analysis, and beam-plasma optimization. These are to be requisite in operation windows selection for fusion neutron sources (FNSs) design [2, 3].
- ПубликацияТолько метаданныеDisruption mitigation in tokamak reactor via reducing the seed electrons of avalanche(2021) Sergeev, V. Y.; Kuteev, B. V.; Кутеев, Борис ВасильевичDisruption mitigation technology remains the key factor for ensuring safe and reliable operation in future large tokamaks including ITER. A novel approach analyzed in this paper aims at mitigating the runaway avalanche (RA) development by reducing the population of runaway seeds using the tungsten projectile injection into the ITER plasma just after the thermal quench. As opposed to the scenario currently discussed by the ITER team, the approach does not involve injecting a large mass of deuterium into the tokamak vacuum vessel for collisional suppression of the RA current generated. This should significantly reduce the operational problems of the ITER technological systems dealing with fueling and pumping, isotope separation, plasma heating, etc. Collecting seeds and runaways by the projectile injected might reduce the avalanche current in ITER below the level of 1 MA and its magnetic energy by more than two orders of magnitude. A tungsten rod being 8 mm inside of the square section and 80 mm in length that crosses the plasma volume with the velocity of 0.8 km s(-1) radially or tangentially seems appropriate for ITER disruption conditions. Simulations of the projectile interaction with the ITER post TQ plasma revealed that its surface temperature remains below the tungsten melting point. It is also found that a railgun using the toroidal magnetic field of the tokamak is the optimal way to accelerate such a projectile.
- ПубликацияТолько метаданныеCalculations of Consistent Parameters of FNS-ST Plasma Using Ion Transport Equations and Simulations of Tritium Fuel Cycle Using FC-FNS Code(2026) Ananyev, S.; Kuteev, B.; Gorkunov, S.; Ананьев, Сергей Станиславович; Кутеев, Борис Васильевич; Горкунов, Сергей Владимирович
- ПубликацияТолько метаданныеEutectic 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
- ПубликацияТолько метаданныеTHE 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
- ПубликацияТолько метаданныеConcept 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 ВасильевичWithin 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.
- ПубликацияТолько метаданныеIntegration 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.
- ПубликацияТолько метаданныеChoice of Gas Isotope Composition for Neutral Beam Injectors of the FNS-ST Compact Fusion Neutron Source(2023) Ananyev, S.; Dnestrovskij, A.; Kukushkin, A.; Ivanov, B.; Kuteev, B.; Кукушкин, Александр Борисович; Кутеев, Борис ВасильевичThe dependence of the neutron yield of the FNS-ST (spherical tokamak) fusion neutron source on the fraction of tritium in the core D+T plasma is analyzed for the case of using tritium neutral beam injectors with 200-keV energy and 6-MW power. The FNS-ST operating regimes are explored using the SOLPS4.3 and ASTRA codes for different values of core plasma density ne, T fraction in the plasma, and particle diffusivity. The FC-FNS code is used to estimate the fluxes of the fuel components in the fuel cycle (FC), which are produced by different injection systems: gas puffing, pellet injection, and neutral beam (T) injection. It is shown that in the case of the Т beam injection, in the operating range of parameters, the neutron yield can reach 6.0 × 1017 s−1, which is the value comparable to that obtained for the scenario of D-beam injection into the balanced D+T plasma. In the case of the T-beam injection, in the range of parameters, for which the neutron yield is close to its maximum, the amount of tritium in the FC is lower than in the case of the D-beam injection. The neutron yield can be increased to 6.5 × 1017 n/s if full separation of the D and T is introduced for the gas pumped out from the divertor and puffed back into the torus. With this approach, in the case of the tritium beam, the amount of tritium in the FC is Tinv of ~170 g. If this approach is used in the case of the deuterium beam, the neutron yield can reach 7.0 × 1017 n/s. However, in this case, the amount of tritium contained in the FC increases to 215 g. The results of the analysis performed are used for optimizing the FC of the FNS-C (compact) fusion neutron source, which is planned for construction in the framework of the comprehensive program of the State Corporation Rosatom “Development of Engineering, Technology and Scientific Research in the Field of Using Atomic Energy in the Russian Federation for the Time Period up to 2030.”. © 2022 American Nuclear Society.
- ПубликацияТолько метаданныеDEVELOPMENT 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
- ПубликацияТолько метаданныеNuclear 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.