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The concept of next-generation spherical tokamak is being considered: the Globus-3 project, which, in its characteristics, is compatible with the infrastructure existing at the Ioffe Institute, but differs from the currently operating Globus-M2 tokamak in the stronger toroidal magnetic field (1.5–3.0 T) and increased duration of plasma discharge. The parametric analysis data are presented that determined the preliminary selection of the facility parameters. Three options for the electromagnetic system were considered: with the warm copper coils, with the pre-cooled copper coils and with the coils made of high-temperature superconductors. For the first option, the concept for designing the electromagnetic system and vacuum vessel of the facility has been developed. The basic shot scenario with duration of up to 3 s at the field of 1.5 T and plasma current of 0.8 MA is presented. [ABSTRACT FROM AUTHOR]

Data on plasma disruption processes in the modernized Globus-M2 spherical tokamak are presented. Electron temperature and density profiles before the disruption, immediately after thermal quench and in the stage of plasma current quench are measured using the diagnostics of Thomson scattering of laser radiation. The dependence of the plasma current decay time during disruption on the pre-disruption current value is determined. The distribution of the toroidal current, which is induced during disruption, in the shell of the vessel is determined on the basis of magnetic measurements. Electromagnetic loads on the vessel are calculated. [ABSTRACT FROM AUTHOR]

The engineering part of the GLOBSYS code is presented, and the parameters of the Globus-3 facility, which is a development of the Globus program, are analyzed. The facility is primarily designed to provide a long pulse, a large toroidal magnetic field and strong heating. The concepts of searching for Globus-3 parameters under physical and engineering limitations are described. Obviously that reliable confinement and a large part of noninductive current are necessary to ensure existence of a plasma for a long time. Engineering constraints are involved in the choice of parameters in a more complex way: in some cases, it is overheating of the coils, in other cases, it is the total power supply, or the limit on the flux provided by the ohmic solenoid, or the strength of the constructions. The parameters of the Globus-3 spherical tokamak were preliminarily selected for the cases of a "warm" copper EMS (Electromagnetic system) and the EMS precooled to liquid nitrogen temperature. The exceeding of the duration of the plasma current plateau Δtplateau over the characteristic settling time of the plasma profiles τL/R was chosen as the key condition. At values of the toroidal magnetic field Bt0 = 3 T, the condition Δtplateau > τL/R cannot be attained even for precooled EMS. At Bt0 = 2 T, only options with precooled EMS can be considered acceptable, but the facility dimensions are fairly large. For the field Bt0 = 1.5 T, the options with "warm" EMS correspond to the duration of the plasma current plateau ~3 s (Δtplateau/τL/R ~ 1–1.5). In the case of precooled EMS, the duration of the plateau can increase to 12–13 s (Δtplateau/τL/R ~ 5). In the latter case, as a basis for further development of the Globus-3 facility, options with the following geometric dimensions are reasonable: R0 ~ 0.6–0.7 m, a ~ 0.35–0.4 m, А ≤ 1.7–1.8, k95 ~ 1.7–1.8. The minimum allowable value of the plasma current under the condition of effective absorption of the input power of neutral injection has been calculated. In the Globus-3 facility, Ip ≈ 0.8 MA was chosen as the base value. [ABSTRACT FROM AUTHOR]

At the present time, the construction of the T-15MD Tokamak is being completed at the National Research Center Kurchatov Institute. The magnet system of the Tokamak T-15MD should provide generation and confinement of the hot plasma in the divertor configuration. The design plasma parameters are as follows: major radius of 1.48 m, minor radius of 0.67 m, elongation of 1.7–1.9 and triangularity of 0.3–0.4, plasma current of 2 MA, toroidal magnetic field on the plasma axis of 2 T. The electromagnetic system includes toroidal and poloidal coils. The installation will be equipped with the auxiliary plasma heating and current drive in quasi-stationary discharges lasting hundreds of seconds. Scenarios of quasi-stationary plasma discharges are determined by the capabilities of the electromagnetic system to create and maintain an equilibrium magnetic configuration of the plasma column. Trouble-free operation of the power supply is provided by the protection and blocking system. This work is devoted to calculations of the maximum duration of currents in toroidal and poloidal windings in order to avoid their overheating, as well as the duration and scenarios of quasi-stationary plasma discharges. [ABSTRACT FROM AUTHOR]

The results of modeling of the plasma initiation and an inductive plasma current ramp up using the SCENPLINT, TRANSMAK, and DINA codes in the TRT (Tokamak with Reactor Technologies) are presented. The initial poloidal magnetic flux provided by the poloidal system, consumption of the poloidal flux for breakdown and the poloidal flux consumption for obtaining the plasma current of 5 MA within 30 s with a maximum field limit of 14 T in the central solenoid and less than 7 T in the other coils are calculated. At the beginning of the scenario, the plasma has a limiter configuration, the elongation of which increases from k ≈ 1 to k ≈ 1.5. The divertor configuration is formed by time of 13 s at the plasma current of Ip ≈ 2 MA. The plasma current ramp up to the design value Ip ≈ 5 MA is carried out simultaneously with the formation of an equilibrium configuration with k ≈ 2.0 at a gradual increase of plasma density and an increase of ECR heating power to 10 MW. [ABSTRACT FROM AUTHOR]

At present, the T-15MD tokamak is being built and its supporting technological systems are being modernized at the National Research Center Kurchatov Institute in the framework of the federal target-oriented program "New Generation Nuclear Power Technologies in 2010–2015 and in Prospect to 2020." The T-15MD magnet system is to ensure producing and confinement of hot plasma in the divertor configuration. The T-5MD plasma parameters are as follows: major radius R0 = 1.48 m, minor radius a = 0.67 m, elongation k95 = 1.7–1.9, triangularity δ95 = 0.3–0.4, plasma current Ip = 2 MA, and toroidal magnetic field at the plasma axis BTO = 2 T. The tokamak will be equipped with an auxiliary plasma heating and current drive system (Paux = 15–20 MW), which will enable achieving a high plasma temperature (Ti ~ Te of 5–9 keV) and a plasma density (ne ≈ 1020 m–3) during a discharge with a pulse duration of up to 30 s. The magnet system includes a toroidal winding and a poloidal magnet system. The poloidal magnet system may generate the one- or two-null divertor magnetic configuration. The power supply system provides the necessary current scenarios in the magnet system windings. The T-15MD magnet together with the vacuum vessel was pre-assembled at the NPO GKMP Ltd. plant in Bryansk. All elements of the magnet system and the vacuum vessel are delivered to the National Research Center Kurchatov Institute in Moscow, where the T-15MD tokamak will be assembled. The technological systems are to be reconstructed by the end of 2019. The physical start-up of the T-15MD is scheduled for December 2020. [ABSTRACT FROM AUTHOR]

Currently reconstruction is initiated on the T-15 tokamak at the Institute of Tokamak Physics, NRC “Kurchatov Institute”. The purpose of the reconstruction is to build the facility with a small aspect ratio A=2.2, elongated cross-section (k=1.8) and the divertor with a major radius of 1.67 m, long plasma current pulse of 2 MA up to 10 s and powerful auxiliary plasma heating (up to 15 MW). The facility is intended for demonstration of the plasma steady-state burning with high parameters for physics research to support the ITER program and program for development of volumetric neutron sources. The reconstruction involves replacement of the existing superconducting TF magnet system and cryoresistive PF system by the “warm” system. The developed magnet system provides the above-mentioned tokamak parameters. The magnet system of the new facility use of the T-10 and T-15 infrastructure, thus essentially determinig the choice of its main parameters [ABSTRACT FROM PUBLISHER]