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.

The as-built stiffness in the ITER central solenoid (CS) modules (CSM1 thorough to CSM4 are currently manufactured) determines the range of vertical compression forces that can be tolerated by the CS modules during ITER operation. Since the as-built stiffness of the CS modules manufactured (∼32 GPa and ∼34 GPa for CSM1 and CSM2, respectively and similar for the other modules) has been reduced from the design value (53 GPa), the CS axial (vertical) force criteria have been updated assuming a conservative stiffness (25 GPa) with margins for all six CS modules. Initial analysis using the updated CS force criteria has revealed that this reduction affects only the plasma initiation with fully charged CS in the ITER 15 MA Baseline DT scenario, resulting in a slight reduction of poloidal magnetic flux, from 117.5 Wb to 116.2 Wb at initial CS magnetization. Therefore, the 15 MA Baseline scenario has been re-developed with an updated plasma start-up, and then the entire evolution of the CS and poloidal field coil parameters has been validated against all the coil currents, fields and forces criteria. To explore potential risks and opportunities for further optimization of scenarios, the equilibrium operational space (the plasma internal inductance versus the poloidal magnetic flux produced by the coils) at flat-top burn has been analyzed using the CORSICA and DINA codes. The three major ITER reference DT operation scenarios, 15 MA Q = 10 Baseline, 12.5 MA Q > 5 Hybrid and 10 MA Q ∼ 5 Steady-State, satisfy all the coil criteria including the CS force updated reflecting the as-built stiffness. The evolution of the plasma discharge parameters within the equilibrium operational spaces provided a guidance for potential optimization with margins.

A set of magnetic coils used to correct the error fields at the Globus-M2 spherical tokamak, which appear due to the imperfections of the production and assembly of the tokamak magnetic system, is described. The magnetic sensors that are used to monitor the locked helical MHD modes are also described. The results of experiments on detecting the locked modes in the discharges with plasma heating by neutral beam injection are presented. A correlation is found between the appearance of the locked modes accompanied by the loss of fast ions and the confinement of the main plasma.

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 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.

In ITER, tritium retention primarily occurs through co-deposition with beryllium. To avoid exceeding the strict tritium inventory limit, efficient tritium recovery techniques are essential. Baking is the ITER baseline for tritium recovery, but its effectiveness in removing tritium from thick beryllium layers is limited. A raised strike point scenario is considered an alternative method for removing tritium from the ITER inner vertical divertor target by heating components via plasma flux. This paper presents SOLPS-ITER code simulations conducted under various conditions, assessing the divertor performance and tritium outgassing of the raised strike point scenario. As the strike point is raised, recycled neutrals are not efficiently baffled by the dome and scrape-off layer, significantly changing the neutral trajectory and ionization source distribution. This improves detachment accessibility but worsens core-edge compatibility compared to the baseline scenario. However, in the partially detached condition, the impact of raising the strike point, perpendicular transport, and q (95) on target heat flux is not significant, as it primarily scales with the input power. Target heat flux is translated to target surface temperature using a simplified heat transfer model that considers the 3D target monoblock geometry and active cooling condition, excluding Be layer thermal properties. For partially detached divertor conditions, the bulk tungsten monoblock surface temperature remains below the baking temperature, which is insufficient for efficient tritium outgassing under the actively cooled ITER divertor condition. However, considering the potential thermal contact resistance between the beryllium and tungsten layers, which may significantly impact temperature distribution, the temperature of the beryllium layer can be raised to a level sufficient for efficient tritium outgassing. Therefore, the raised strike point scenario can be considered as an alternative in-vessel tritium removal technique.

The main results of numerical simulation of transient electromagnetic processes in the modules of the first and second rows of TRT first wall panels are presented. The study was carried out using the TYPHOON software designed for numerical simulation of quasi-stationary eddy currents in conducting shells of complex shape randomly located in space, taking into account their multicoupling and branching. The basic design of the first wall panel and its five modifications are considered. It is determined that the electromagnetic forces and moments for the basic design reach extreme values, and the edge elements of the panel are the most loaded. Options with inserting electrical insulation at the places of contact of the plasma facing elements and the base of the panel, as well as incorporating a shunt between adjacent panels, make it possible to achieve the strongest reduction in loads. These options are accepted as basic for further modification of the panel design and conduction of additional EM analysis.

The pyGSS code constructed for computation of free-boundary plasma equilibrium in spherical Globus-M2 tokamak is described. Currents in the coils of the electromagnetic system, their coordinates, plasma current, positions of the limiter and current-conducting wall, etc., are used as the input parameters. Free parameters determining spatial distribution of equilibrium pressure and current density are selected in the course of code execution in such a way that the results of reconstruction would agree with the experimental measurements of the poloidal magnetic flux by means of toroidally closed loops. The results of computation of equilibrium are compared with those obtained by means of other codes and experimentally measured thermal plasma energy, position of the separatrix outer leg, the diamagnetic-loop signal, etc.

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.

Long Pulse Scenarios (LPS) in ITER foreseen during the Pre-Fusion Power Operation (PFPO) phase of the ITER Research Plan (IRP) are assessed using 1.5D transport simulations within the ASTRA framework. Such assessment is required to predict the operational space for LPS operation in PFPO, as well as to evaluate which physics processes for LPS operation during Fusion Power Operation (FPO) could be studied during PFPO. An important aspect in the development of LPSs in PFPO is to minimize lifetime consumption of the Central Solenoid (CS) for these scenarios. The maximum pulse length achievable for LPSs in PFPO with no consumption of CS lifetime (currents in CS coils ⩽30 kA per turn) has been assessed for a range of heating schemes and heating mixes, confinement regimes (L-mode and H-mode) and for helium and hydrogen plasmas. The operational space of LPS and pulse length has been explored through density scans with the Heating and Current Drive mix required for the FPO Q ⩾ 5 steady-state plasma scenario (namely Neutral Beam Injection and Electron Cyclotron Heating) including acceptable shine through losses on the first wall for both helium and hydrogen plasmas. Fast particle physics aspects that are common between FPO plasmas and LPS PFPO H-mode plasmas at low densities are studied including MHD stability analysis with the KINX code and non-perturbative critical gradient model based on high-n Toroidal Alfven Eigenmodes (TAE) stability kinetic ballooning code HINST calculations.

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.

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.