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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
Integrated modelling of core and divertor plasmas for the DEMO Fusion Neutron Source hybrid facility
(2019) Dnestrovskiy, A. Y.; Sergeev, V. V.; Kukushkin, A. S.; Kuteev, B. V.; Кутеев, Борис Васильевич
A steady state regime for the tokamak-based DEMO Fusion Neutron Source (DEMO-FNS) with parameters R/a = 3.2 m/1.0 m, B = 5 T, I-pl = 4-5 MA, P-NBI = 30 MW and P-ECR = 6 MW is studied using coupled simulations of the central and divertor plasma. In our analysis, the divertor plasma state is determined by the values of the heat flux P-SOL, the pressure of the neutrals in the divertor p(n) and the total number of neon particles N-Ne outside the separatrix. As the boundary conditions for the core plasma, we use the values at the separatrix of the electron density, n(e_sep), and temperatures of the ions, T-i_sep, and electrons, T-e_sep the concentration of the neon impurity normalized to the electron density at the separatrix, C-Ne = Sigma n(Ne)/n(e_sep) and the hydrogen neutral influx into the core plasma, Gamma(0sep). In the divertor region, all the values are calculated using the SOLPS4.3 code for a number of operating points (similar to 150 in our case) with different values of P-SOL, p(n )and N-Ne. Then the calculation results are approximated by analytical formulas. Power balance in the core plasma is calculated using the ASTRA and NUBEAM codes. The hydrogen (deuterium and tritium) density is modelled taking into account the sources generated by the neutrals originating from the divertor region, as well as by injection of fast atoms and pellet injection. The neon density and radiation in the main plasma are simulated using the STRAHL code. As the result of the simulations, the operational regime of DEMO-FNS is determined, in which the heat loading onto the divertor targets remains at the acceptable level below 10 MW m(-2), the divertor plasma does not transit into the 'full detachment' mode and the plasma in a double-null separatrix configuration is kept up-down symmetric. Variations of these conditions versus the impurity level and confinement parameters are investigated and discussed.
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Closed Lithium Cycle Concept in the DEMO-FNS Tokamak with a Sectioned Divertor
(2024) Sergeev, V. Y.; Skokov, V. G.; Kuteev, B. V.; Timokhin, V. M.; Кутеев, Борис Васильевич
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
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
Только метаданные
LIGHT 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].
Только метаданные
Disruption mitigation in tokamak reactor via reducing the seed electrons of avalanche
(2021) Sergeev, V. Y.; Kuteev, B. V.; Кутеев, Борис Васильевич
Disruption 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.
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