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.

This paper highlights the most important results achieved at the spherical tokamak Globus-M2 with a high magnetic field. This paper also covers the most important topics of fusion research: thermal energy confinement in regimes with neutral beam injection, toroidal Alfvén eigenmode and correspondent fast ions confinement issues, L-H transition, turbulence suppression and edge-localized modes' behavior, experimental and theoretical study of regimes with nitrogen seeding that allow to significantly reduce thermal loads on the divertor plates, and experiments and simulations of lower hybrid current drive. The research results provide the basis for the next step toward a fusion neutron source—the development of the Globus-3 spherical tokamak.

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.

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 I_p,min 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.

New fusion devices are being discussed in Russia. One item is a superconducting magnet development for the demonstration hybrid facility-fusion neutron source based on a tokamak concept with the conventional aspect ratio ~3. The magnetic parameters of this device are planned to be: 5 Ton the plasma axis and about 12 Ton the high field side of the toroidal field coils. Because of a high neutron flow and a thick shielding, the space remained for the inner legs of the toroidal field coil is very tight. Therefore, a very high current density is needed for the windings. Several magnet designs have been addressed up to now with required engineering critical current density sim 1000 A/mm 2 at least. In this review, we present conceptual design of the magnet system, necessary strand parameters, and their irradiation properties. Few possible designs of the toroidal field, conductors are discussed. The R&D of efficient HTS current leads to be used for fusion machines are discussed as well.

The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion-fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet system and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k (95) = 2. The fusion power is P (FUS) = 40 MW. The toroidal magnetic field on the plasma-filament axis is B (t0) = 5 T. The plasma current is I (p) = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb3Sn, NbTi and Nb3Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron-water radiation shield (70% of stainless steel and 30% of water).

The Globus-M spherical tokamak has demonstrated practically all of the project objectives. The increasing of the magnetic field up to 1.0 T together with the plasma current up to 0.5 MA should significantly enhance plasma performance in Globus-M2 machine. Simultaneously it will entail serious rise in loads on the magnetic system in the upgraded tokamak. Thereupon a review of the design was developed. The vacuum vessel remains the same in order to reduce the project cost. Results of the complete 3-D finite element model thermal and stress analysis are presented for the novel magnetic system. The mechanical strength was enhanced significantly. Radius of the toroidal field coil outer limb was enlarged slightly in order to reduce field ripple. The central column and the toroidal field coil joints were fully redesigned. Final design of the tokamak upgrade is discussed in the report as well as current status of the work.