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The characteristics of the major disruption of plasma discharges in the Globus-M spherical tokamak are analyzed. The process of current quench is accompanied by the loss of the vertical stability of the plasma column. The plasma boundary during the disruption is reconstructed using the algorithm of movable filaments. The plasma current decay is preceded by thermal quench, during which the profiles of the temperature and electron density were measured. The data on the time of disruption, the plasma current quench rate, and the toroidal current induced in the tokamak vessel are compared for hydrogen and deuterium plasmas. It is shown that the disruption characteristics depend weakly on the ion mass and the current induced in the vessel increases with the disruption time. The decay rate of the plasma toroidal magnetic flux during the disruption is determined using diamagnetic measurements. Such a decay is a source of the poloidal current induced in the vessel; it may also cause poloidal halo currents.

A magnetic diagnostics allowing one to reliably reconstruct equilibrium plasma configurations in a tokamak over a wide range of operating parameters is developed. The accuracy of determining the geometrical parameters and thermal energy of the tokamak plasma is analyzed in detail. The experimental data obtained in the Globus-M tokamak are processed the with help of the EFIT code. The influence of the plasma configuration on the intensity of the main impurity lines is investigated.

Results are presented from experimental studies of runaway electrons in the ohmic heating regime in the Globus-M tokamak. The periodical hard X-ray bursts observed with the help of two hard X-ray spectrometers with high time resolution are attributed to MHD oscillations in the plasma core and at the periphery.

Experiments and simulations to achieve high values of plasma parameters at the Globus-M spherical tokamak (ST) at moderate absolute auxiliary heating power (up to 0.8 MW) and high specific heating power (up to 2–3 MW m−3) are described. Important distinguishing features are the low edge safety factor range, which is unusual for STs, 2.7 < q < 5 and small plasma–outer wall space (3–5 cm). High ion heating efficiency with neutral beam injection (NBI) was demonstrated. Results of numerical simulation of fast ion trajectories are described and fast ion generation during the NBI and ion cyclotron resonance heating is discussed. Investigations on their confinement and slowing down are also presented. Reasons for achievement of high IC heating efficiency are outlined. Reliable H-mode regime achievement is described. Transport ASTRA modelling demonstrated that during NB heated H-mode ion heat diffusivity remains neoclassical and the particle diffusion coefficient inside transport barrier decreases significantly. Analysis was performed of divertor tile and special probe surfaces after irradiation by plasma during a large number of shots (3000–10 000 shots). Mixed layer composition is measured and deuterium retention in different tokamak first wall areas is estimated. Plasma jet injection experiments with upgraded plasma jet are described. Jet penetration to the plasma centre with immediate increase of density and temperature drop is proved and analogy with pellet injection is outlined.