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.

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 results of measuring the electron temperature and density spatial distributions in plasma of the Globus-M2 tokamak using the Thomson scattering diagnostics are presented. The diagnostics provides measurements throughout the entire tokamak discharge, starting from time of gas breakdown. The Thomson scattering data were analyzed in order to determine the positions of the last closed flux surface, the plasma magnetic axis, and the radius of inversion during the saw-tooth oscillations. The results of measurements performed during the internal reconnection of magnetic field lines are presents, as well as the dynamics of spatial distributions of electron temperature, density and pressure during the plasma transition to the H-mode. The results of measuring the electron temperature distribution in the scrape-off layer using the Thomson scattering diagnostics are also presented for distances up to 4 cm outside the last closed flux surface.

The thermal energy stored in plasma Wp, normalized internal plasma inductance li and currentbeta bi are calculated via the free-boundary equilibrium PET code. The equilibrium reconstruction algorithmis iterative method of minimizing two parameters, the distance between the reconstructed plasmaboundary and that simulated by the PET code, as well as the difference between the plasma diamagnetic fluxfrom PET and the experimental one. The discharges from the Globus-M2 tokamak with a toroidal magneticfield up to 0.9 T and a plasma current of 0.3–0.4 MA in a mode with auxiliary heating by two atomic injectorsare analyzed. The possibility of using approximate formulas for estimating and is considered. The measureddiamagnetic flux is used to determine bdia and then to calculate Wdia . The normalized internal plasmainductance is additionally determined from the measured vertical magnetic field under the assumption that bp=bdia.

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.

A cross-machine comparison of global parameters that determine the runaway electron (RE) generation and loss process during tokamak start-up was carried out with the aim to extrapolate these to ITER. The study found that all considered discharges, also those that do not show signs of RE, are non-thermal at the start, i.e. have a streaming parameter larger than 0.1. During the current ramp-up the electric field, E, remains above the critical value, E (c), that allows RE in the plasma. The distinction to be made is not if RE can form but, if sufficient RE can form fast enough such that they are detected or start to dominate the dynamics of the tokamak discharge. The dynamics of the value of E, density and temperature during tokamak are key to the formation of RE. It was found that larger devices operate with E closer to E (c), due to their higher temperatures, hence the RE generation is relatively slower. The slower time scales for the formation of RE, estimated to be of the order of 100s of ms in ITER simplifies the development of avoidance schemes. The RE confinement time is also an important determinant of the entire process and is found to increase with the device size. The study also revealed that drift orbit losses, a mechanism often attributed as the main RE loss mechanism during the early tokamak discharge, are actually more difficult to achieve. RE losses might be more likely attributed to RE diffusion due to magnetic turbulence.

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 H y , 2 = 1–1.2. Simulations with other confinement scalings (Globus-2021, NSTX scalings) give good agreement with plasma parameters for the toroidal field B t0 ~ 0.5 T. For increasing B t0 , 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 τ L / 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.

Understanding the formation of start-up runaway electrons (REs) is essential to ensure successful plasma initiation in ITER. The design of ITER start-up scenarios requires not only predictive simulations but also a validation of assumptions. The objective of this study is to strengthen the physical background required for predictive simulations aimed at ITER plasma start-up design, by validating the model assumptions. Through kinetic simulations, this study examines the validity of steady-state models for Dreicer generation under slowly-varying time scales relevant to plasma start-up and investigates the finite energy effect, commonly neglected, on the runaway avalanche growth rate. The research findings provide insights into situations where kinetic simulations are necessary. To secure a margin-of-control scheme without kinetic simulation, we suggest a strategy of scanning the Coulomb logarithm in fluid simulations as an alternative to predict runaway current takeover and avoid RE dominant scenarios. Ultimately, this paper seeks to offer a robust physical background, practically supporting the successful design of ITER start-up scenarios.

The thermal energy stored in plasma W_p , normalized internal plasma inductance l_i and current beta β_I are calculated via the free-boundary equilibrium PET code. The equilibrium reconstruction algorithm is iterative method of minimizing two parameters, the distance between the reconstructed plasma boundary and that simulated by the PET code, as well as the difference between the plasma diamagnetic flux from PET and the experimental one. The discharges from the Globus-M2 tokamak with a toroidal magnetic field up to 0.9 T and a plasma current of 0.3–0.4 MA in a mode with auxiliary heating by two atomic injectors are analyzed. The possibility of using approximate formulas for estimating W_p and l_i is considered. The measured diamagnetic flux is used to determine β_dia and then to calculate W_dia . The normalized internal plasma inductance l_i is additionally determined from the measured vertical magnetic field under the assumption that β_p = β_dia .

Absorbed power of the neutral-injection beam in spherical tokamaks Globus-M/M2 is estimated numerically. Deceleration of fast particles is simulated by means of the NUBEAM code. The signal of analyzer of charge-exchange atoms is simulated by means of the FIDASIM code using the distribution function of fast ions calculated by means of the NUBEAM code. Comparison of calculated and experimental signals allowed determining the degree of influence of instabilities on confinement of fast particles along with absorbed beam power.

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.

Исследован нагрев плазмы одним и двумя инжекторами быстрых нейтральных атомов в сферическом токамаке Глобус-М2 при тороидальном магнитном поле 0.8–0.9 Тл и токе плазмы 0.35–0.4 МА. Измерение пространственных распределений температуры и концентрации электронов, выполненные диагностикой томсоновского рассеяния лазерного излучения, показало двукратный нагрев электронов плазмы при инжекции нейтральных частиц с энергией до 45 кэВ при мощности пучка 0.75 МВт по сравнению с омическим режимом. Дополнительное включение второго пучка с энергией частиц до 30 кэВ и мощностью до 0.5 МВт позволило получить режим с горячими ионами в диапазоне значений средней плотности плазмы 1.6–10 × 10 19 м –3 . По данным активной спектроскопии и корпускулярной диагностики температура ионов достигла величины 4 кэВ при плотности плазмы 8 × 10 19 м –3 в горячей зоне, превысив температуру электронов более чем в 2.5 раза.

NBI-assisted plasma heating with one or two injectors of fast neutral atoms was studied at the Globus-M2 spherical tokamak at the toroidal magnetic fields of 0.8–0.9 T and plasma currents of 0.35–0.4 MA. Measurements of the spatial temperature and electron density distributions, performed using the Thomson scattering diagnostics, showed a twofold increase in heating of plasma electrons during the injection of neutral particles with energies of up to 45 keV at the beam power of 0.75 MW, as compared to the ohmic heating regime. Switching on the second additional beam with the particle energy of up to 30 keV and power of up to 0.5 MW resulted in obtaining the hot ion mode in the range of mean plasma densities of (1.6–10) × 10 19 m −3 . According to the data of active spectroscopy and neutral particle analyzer diagnostics, in the hot zone, the ion temperature reached 4 keV at the plasma density of 8 × 10 19 m −3 , which is more than 2.5 times higher than the electron temperature.

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.

Hot ion mode was recently achieved at the Globus-M2 spherical tokamak when the toroidal magnetic field was as high as 0.9 T and plasma current 0.35 MA. The injection of two neutral beams with a total power of 0.7 MW and an energy of 30 keV into a plasma with major radius 0.36 m and minor radius 0.22 m (aspect ratio 1.6) made it possible to heat the plasma ions up to 4 keV, while electron density in the plasma center was as high as 8 × 1019 m−3. The achievement of high temperatures became possible due to the good thermal insulation of the plasma.