Analysis of Dynamics of Plasma Current Quench in the Globus-M Spherical Tokamak

Data on the dynamics of the plasma current quench in the Globus-M tokamak are presented. The main current quench characteristics at different toroidal magnetic fields are compared. The distribution of the toroidal current induced in the vessel wall is determined from magnetic measurements, and the electromagnetic loads acting on the vessel wall during the current quench are calculated. By extrapolating the experimental data, the additional pressure on the vessel wall during the current quench in the upgraded Globus-M2 tokamak is estimated. It is shown that the current quench results in the appearance of bending stresses in the vessel domes. Using numerical simulations, it is shown that the best agreement between the measured and calculated plasma current dynamics during the current quench corresponds to the linear (in time) influx of the carbon impurity.     

Distinctive Features of Lower Hybrid Current Drive in Plasma of the FT-2 Tokamak

Plasma Physics Reports - Experiments on the lower hybrid current drive (LHCD) were performed in plasma of the FT-2 tokamak at the microwave pulse duration of ΔtRF ≤ 20 ms covering the...     

Study of Turbulence in the Globus-M Tokamak Plasma during the Transition to the ELM-free H-mode

Plasma Physics Reports - Here we report the results of the turbulence study in the high-confinement mode (H-mode) with and without edge localized modes (ELMs). The study was performed by the...     

Observation of Parametric Decay of a Pumping Wave in Lower Hybrid Wave Heating and Current Drive Experiments at the FT-2 Tokamak

Plasma Physics Reports - The systematization of experimental data related to development of the parametric decay instability in tokamak plasma during RF power input within the lower hybrid...     

High-Power Neutral Beam Injector with Tunable Beam Energy for Plasma Heating and Stabilization

Plasma Physics Reports - For plasma heating and stabilization in open magnetic traps, the high-power neutral beam injector with tunable beam energy was developed at Budker Institute of Nuclear...     

Scenario of the Plasma Current Ramp up in the TRT Tokamak

Plasma Physics Reports - 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...     

Nonmonotonic Spatial Distributions of Charge-Exchange Neutral Fluxes and Optical Radiation from Plasma in the DAMAVAND Tokamak

Results are presented from investigations of the nonmonotonic spatial distributions of charge-exchange neutral fluxes and optical radiation from plasma in the DAMAVAND tokamak. It is shown that, during ohmic heating of the plasma, the regions with enhanced confinement of both the background plasma particles and heavy impurity ions arise near rational magnetic surfaces with q = 1 and 2. These regions are characterized by enhanced emission of accelerated charge-exchange neutrals and optical radiation from impurity ions.     

Anomalous Fast-Ion Transport During Neutral Beam Injection on EAST

The fast-ion transport during neutral beam injection on the Experimental Advanced Superconducting Tokamak (EAST) is studied. Based on the NUBEAM and TRANSP codes, it is found that fast-ion transport is anomalous when the minimum safety factor (q_min) is about 2, while it is neoclassical when q_min is around 1. Neutral beam injection heating efficiency, plasma stored energy, and the total heating power are reduced when the fast ion transport is anomalous. The Alfvén continuum spectrum and the mode structures of toroidal Alfvén eigenmodes (AEs) are also calculated for comparison between neoclassical fast-ion transport and anomalous fast-ion transport. High-q_min discharge with anomalous fast-ion transport has more AE activity than that of lower q_min discharge with neoclassical fast-ion transport.     

Studies of runaway electrons in the Globus-M tokamak

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.     

Diagnostics of MHD instability in the Globus-M spherical tokamak

The paper describes a diagnostic system for studying MHD plasma perturbations in the Globus-M spherical tokamak (a major radius of 0.36 m, a minor radius of 0.24 m, and an aspect ratio of 1.5). The system includes a poloidal and a toroidal array consisting of 28 and 16 Mirnov probes, respectively, as well as a 32-channel proportional soft X-ray detector. Methods are described for calculating the poloidal and toroidal numbers of the dominant helical perturbations by using data from probe measurements. Results are presented of processing the experimental data from some tokamak discharges with a plasma current of 150–250 kA, an average electron density of up to 10²⁰ m^?3, and a toroidal magnetic field of 0.4 T. Specific features of MHD perturbations and their influence on the parameters of the plasma column in different stages of a discharge are briefly discussed.     

Characteristics of major plasma discharge disruption in the Globus-M spherical tokamak

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.     

Noninductive plasma generation and current drive in the Globus-M spherical tokamak

Experimental results on the generation and maintenance of the toroidal current in the Globus-M spherical tokamak by using waves in the lower hybrid frequency range without applying an inductive vortex electric field are presented. For this purpose, the original ridge guide antennas forming a field distribution similar to that produced by multiwaveguide grills were used. The high-frequency field (900 MHz) was used for both plasma generation and current drive. The magnitude of the generated current reached 21 kA, and its direction depended on the direction of the vertical magnetic field. Analysis of the experimental results indicates that the major fraction of the current is carried by the suprathermal electron beam.     

Particle diagnostics of ion cyclotron resonance plasma heating in the Globus-M tokamak

Basic experimental results on cyclotron heating of the ion plasma component in the Globus-M spherical tokamak obtained by means of the ACORD-12 charge-exchange ion analyzer are presented. A procedure for determining the maximum energy of fast ions confined in the plasma is described. The procedure was applied to estimate the limiting energy of hydrogen minority ions accelerated during ion cyclotron heating in the Globus-M tokamak. The experimental evaluation of the maximum hydrogen ion energy is confirmed by simulations of ion orbits. Recommendations for optimizing experiments on ion cyclotron heating in the Globus-M tokamak are formulated.     

First results obtained with a sweeping pulsed radar reflectometer in the Globus-M tokamak

A weeping pulsed radar reflectometer designed for measuring the spatial electron density distribution in the Globus-M spherical tokamak with a minor plasma radius of a=24 cm, a major radius of R=36 cm, a toroidal field of B _T=0.5 T, a plasma current of I _p=200 kA, and an average density of n=(3–10)×10¹³ cm^?3 is described. The reflectometer operation is based on the reflection of microwaves with a carrier frequency f from a plasma layer with the critical density n=(0.0111f)², where n is the electron density in units of 10¹⁴ cm^?3 and f is the microwave frequency in GHz. By simultaneously probing the plasma at different frequencies, it is possible to recover the electron density profile. Microwave pulses with different frequencies are obtained by frequency sweeping. To increase the range of measured densities, channels with fixed frequencies are also used; as a result, the instrument has eleven frequency channels: a 19.5-GHz channel, eight channels in the 26-to 40-GHz frequency range, a 51.5-GHz channel, and a 60-GHz channel, which corresponds to eleven points in the density profile: 0.47×10¹³ cm^?3, eight points in the (0.8–1.95)×10¹³-cm^?3 range, 3.27×10¹³ cm^?3, and 4.5×10¹³ cm^?3. The reflectometer allows detailed measurements of the density profile with a time resolution of several tens of microseconds, which can be useful, in particular, in studying the processes related to the formation of an internal transport barrier in plasma. The first results obtained using this reflectometer in the Globus-M tokamak under various operating conditions are discussed.     

Study of plasma heating in discharges with neutral beam injection in the Globus-M spherical tokamak

Results from experimental studies on the injection of high-energy neutral hydrogen beams into the plasma of the Globus-M spherical tokamak are reviewed. In the Introduction, the importance of these studies for implementing the controlled fusion research program and constructing the ITER tokamak is proved. Some problems related to the use of neutral beam injection in small and low-aspect-ratio tokamaks is analyzed. Results are presented from numerical simulations of the experiment by using the ASTRA transport code. It is shown that the use of neutral beam injection in the Globus-M tokamak ensures efficient ion heating and increases the plasma stored energy. The greater part of the review is devoted to the survey of experiments on the injection of 22-to 30-keV hydrogen and deuterium beams with a power of 0.4–0.8 MW into the plasma of the Globus-M spherical tokamak in a wide range of plasma currents and densities. The experimental results are analyzed and compared with the results of numerical simulations. The achievement of top plasma parameters is highlighted.     

The Spatial Variation of Membrane Potential Near a Small Source of Current in a Spherical Cell

A theoretical analysis is presented of the change in membrane potential produced by current supplied by a microelectrode inserted just under the membrane of a spherical cell. The results of the analysis are presented in tabular and graphic form for three wave forms of current: steady, step function, and sinusoidal. As expected from physical reasoning, we find that the membrane potential is nonuniform, that there is a steep rise in membrane potential near the current microelectrode, and that this rise is of particular importance when the membrane resistance is low, or the membrane potential is changing rapidly. The effect of this steep rise in potential on the interpretation of voltage measurements from spherical cells is discussed and practical suggestions for minimizing these effects are made: in particular, it is pointed out that if the current and voltage electrodes are separated by 60°, the change in membrane potential produced by application of current is close to that which would occur if there were no spatial variation of potential. We thus suggest that investigations of the electrical properties of spherical cells using two microelectrodes can best be made when the electrodes are separated by 60°.     

Kinetic Simulations of EC Plasma Heating and Current Drive in the L-2M Stellarator

The ECHLAB code, intended for a self-consistent numerical analysis of the evolution of the electron distribution function and the spatial structure of the electromagnetic field during EC plasma heating in a stellarator, is described. The results from calculations of plasma heating and current drive under conditions corresponding to experiments on EC plasma heating by an X2-mode in the L-2M stellarator are presented. It is shown that, at the existing level of microwave power, the energy deposition region displaces only slightly during heating. The energy is mainly absorbed by relatively fast passing electrons. The influence of locally trapped electrons on the efficiency of current drive is insignificant.     

Correction of the axial asymmetry of the poloidal magnetic field in the Globus-M spherical tokamak

The toroidal inhomogeneity of the poloidal magnetic field—the so-called error fields that arise due to imperfections in manufacturing and assembling of the electromagnetic system-was measured in the Globus-M spherical tokamak. A substantial inhomogeneity corresponding to the n = 1 mode, which gave rise to a locked mode and led to discharge disruption, was revealed. After compensation of this inhomogeneity with the help of special correction coils, the discharge duration increased and the global plasma parameters improved substantially. A technique for determining and compensating the n = 1 mode inhomogeneity is described, the measured dependences of the penetration threshold of the m = 2/n = 1 mode on the plasma parameters are given, and results of experiments in which record parameters for the Globus-M tokamak were achieved after correction of the poloidal magnetic field are presented.     

Plans for Liquid Metal Divertor in Tokamak Compass

The COMPASS tokamak (R = 0.56 m, a = 0.2 m, B_T = 1.3 T, I_p ~ 300 kA, pulse duration 0.4 s) operates in ITER-like plasma shape in H-mode with Type-I ELMs. In 2019, we plan to install into the divertor a test target based on capillary porous system filled with liquid lithium/tin. This single target will be inclined toroidally in order to be exposed to ITER-relevant surface heat flux (20 MW/m²). Based on precisely measured actual heat fluxes, our simulations predict (for 45° inclination, without accounting for the lithium vapor shielding) the surface temperature rises up to 700°C within 120 ms of the standard ELMy H-mode heat flux with ELM filaments reaching hundreds MW/m². Significant lithium vaporization is expected. The target surface will be observed by spectroscopy, fast visible and infrared cameras. The scientific program will be focused on operational issues (redeposition of the evaporated metal, ejection of droplets, if any) as well as on the effect on the plasma physics (improvement of plasma confinement, L–H power threshold, Z_eff, etc.). After 2024, a closed liquid divertor may be installed into the planned COMPASS Upgrade tokamak (R = 0.84 m, a = 0.3 m, B_T = 5 T, I_p = 2 MA, P_in = 8 MW, pulse duration ~2 s) with ITER-relevant heat fluxes loading the entire toroidal divertor.