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This paper presents the verification of results of the stress analysis of the Globus-M2 spherical tokamak magnet system on the basis of measurements carried out in experiments. The magnet system is composed of toroidal and poloidal field coils and intercoil and supporting structures and is subjected to the impact of electromagnetic (EM) loads during an experiment. In the series of experiments with close to maximum values of the plasma current and toroidal field, the toroidal field coil displacements in the toroidal direction under EM loads with respect to its natural position were measured using a high-speed camera. The EM loads were determined using a 3D model of the Globus-M2 tokamak magnet system from the currents of the toroidal and poloidal field coils and plasma measured in experiments. These loads were used for the stress analysis of the Globus-M2 magnet system in the ANSYS software. Results of this work show that the approach consisting of two stages—the reconstruction of the EM loads and the determination of the mechanical response of the structure—was successfully verified for the EM system of the Globus-M2 and can be used for other tokamaks. [ABSTRACT FROM AUTHOR]

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 description of the zero-dimensional engineering-physical code GLOBSYS (Globus spherical tokamak system code), designed for parametric analysis of the next step of the program Globus-M, Globus-M2, is given. Within the framework of the zero-dimensional approximation, the definitions of the main scaling parameters of the plasma (poloidal beta, the fraction of bootstrap current, the energy lifetime of the plasma), as well as the specifics of calculating the inductance and resistance of the plasma in spherical tokamaks, are refined. The results of calculations of the plasma parameters by the code were compared with the experimental data of one of the Globus-M2 discharges (no. 38800) with neutral beam heating and showed good agreement. It is proposed to perform a comparison of calculations based on the code with the achieved and predicted parameters of the spherical tokamaks NSTX, NSTX-U, MAST, MAST-U, and ST40 in a separate paper. The goals of the next step (Globus-3) are formulated, the main ones of which are long pulse, high toroidal field, and powerful heating, which allow us to consider Globus-3 as a hydrogen prototype of a neutron source. The infrastructural restrictions on the Globus-3 parameters are given, which require further analysis of various versions of the electromagnetic system. Using the example of Globus-M2 discharge no. 38800, the effect of restrictions on the flow balance and heating of the elements of the electromagnetic system is shown. [ABSTRACT FROM AUTHOR]

The GLOBSYS code was developed for analysis and prediction of parameters of the Globus-M2 tokamak and its modifications. In [1], preliminary selection of correlations which connect physical and technical parameters was made. In this paper, the verification of the code using the achieved and predicted data from the installations NSTX, NSTX-U, MAST, MAST-U, and ST40 is given. As a whole, there is good agreement between simulations and plasma parameters at the discharge plateau. The best agreement is observed if ITER confinement scaling is used for energy confinement time with the enhancement factor Hy, 2 = 1–1.2. Simulations with other confinement scalings (Globus-2021, NSTX scalings) give good agreement with plasma parameters for the toroidal field Bt0 ~ 0.5 T. For increasing Bt0, more optimistic predicted plasma parameters are obtained for the Globus-2021 and NSTX scalings in comparison with the ITER confinement scaling. The condition of reaching the plasma quasistationary regime (or the time of establishment of quasistationary plasma profiles τL/R) is estimated for NSTX, NSTX-U, MAST, MAST-U and ST40 discharges. This time is compared with two technical restrictions, which are connected with the times of toroidal field coil heating and poloidal flux capacity. Verification of the GLOBSYS code using the data from the aforementioned spherical tokamaks is the basis for the prediction of parameters of the next step of Globus-M program. [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]

The engineering and technical aspects of the creation of the electromagnetic system (EMS) developed using high-temperature superconductors (HTSCs) for the TRT project (Tokamak with Reactor Technologies) are considered. Some proposals on the design of superconducting the wire and EMS windings are presented, taking into account a number of restrictions imposed by equipment and operating conditions of the TRT. A brief comparison of the TRT EMS with a number of existing projects is given. [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 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]

Important progress in the development of high-temperature superconductors (HTSC) of the second group made it possible to design the quasi-stationary tokamak with reactor technologies (TRT) with the high magnetic field (Bt0 = 8 T). The high magnetic field will ensure the achievement of plasma fusion regimes in the tokamak with the fusion energy gain Q > 1 at the considerably reduced size of the facility (R0 = 2.15 m, a = 0.57 m), and, consequently, at its reduced cost. TRT will be capable of operating in the quasi-stationary regimes (≥100 s) with hydrogen, helium, and deuterium plasmas (with the densities ne of up to 2 × 1020 m–3) and in the regimes with short (duration Δt < 10 s) deuterium–tritium plasma shots with the fusion energy gain Q > 1 limited by the radiation heating of toroidal coils. TRT is being designed as a plasma prototype for both the pure fusion reactor and the fusion neutron source for the hybrid (fusion–fission) reactor. The TRT missions are the development of the key fusion technologies and their integration in one facility. These technologies are as follows: the HTSC electromagnetic system operating at the extremely high magnetic fields; the metal and liquid-metal (lithium) first wall and innovative divertor; the unique advanced systems for the auxiliary plasma heating and non-inductive current drive, including the systems for atomic beam injection with energy of 0.5 MeV and power of several tens of megawatts, the electron cyclotron heating system based on the megawatt-power gyrotrons with a frequency of 230 GHz and a total power of ~10 MW, and the ion cyclotron heating system at frequencies of 60–80 MHz with a power of several megawatts; the tritium fuel cycle; the remote control technologies; the technologies for diagnostics capable of operating under the fusion reactor conditions; the technologies for maintaining quasi-stationary plasma discharges; and the technologies for the tokamak operation in the fusion ignition regime, in which the heating by alpha particles is the dominant heating mechanism at the axis of the plasma column, in the deuterium–tritium experiments limited by the radiation heating of the toroidal coils. The results are presented from the conceptual design of the basic TRT components, as well as the expected characteristics of its operation. It is shown that TRT has a wide window of working parameters suitable for studying the reactor operating regimes. The high magnetic field provides the necessary margins of the pressure, MHD stability, and plasma controllability variation. Implementation of the advanced divertor and first wall concepts, including those using the liquid-metal technologies, will provide the optimum choice of design options in order to reliably control the heat and particle fluxes under the reactor conditions. The advanced systems for the auxiliary heating and current drive will make it possible to implement both the pulsed and stationary regimes of the reactor operation. Calculations of the TRT discharge scenarios show that, for the DT mixture with equal content of components, the long discharges (with duration exceeding 100 s) can be realized with a neutron flux of more than 0.5 MW/m2 onto the wall, as well as the stationary discharges with a flux of approximately 0.2 MW/m2. Thus, TRT can be a real prototype of the fusion neutron source for the hybrid reactor. [ABSTRACT FROM AUTHOR]

The targeted plasma parameters of the compact spherical tokamak (ST) Globus-M have basically been achieved. The reasons that prevent further extension of the operating space are discussed. The operational limits of Globus-M together with an understanding of the limiting reasons form the basis for defining the design requirements for the next-step, Globus-M2. The recent experimental and theoretical results achieved with Globus-M are discussed, the operational problems and the research programme are summarized and finally, the targeted Globus-M2 parameters are presented. The magnetic field and plasma current in Globus-M2 will be increased to 1 T and 0.5 MA, respectively. The plasma dimensions will remain unchanged. With auxiliary heating at a high average plasma density, the temperatures will be in the keV range and the collisionality parameter with ν* ≪ 1 will define the operational conditions. Noninductive current drive will be a major element of the programme. The engineering design issues of Globus-M2 tokamak are discussed and the technical tokamak parameters are confirmed by thermal load and stress analysis simulations. The experimental results obtained on Globus-M2 and the limits of its performance should clarify the feasibility of an ST-based super compact neutron source. [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]

The objective of the present paper is to investigate the vertical stability of the plasma column in the project of the T-15M tokamak. It has been shown that a passive stabilizing structure consisting of a vacuum vessel and two (upper and lower) passive coils is capable of providing vertical stability of the plasma column. [ABSTRACT FROM AUTHOR]