Search

At the Institute for Laser and Plasma Technologies of NRNU MEPhI, a compact spherical tokamak MEPhIST (MEPhI-Spherical Tokamak) has been developed and constructed for educational, demonstrational and research purposes. The creation of plasma diagnostic systems involves several stages, determined by a successive complication of the plasma research tasks, the device upgrading and the development of educational and methodological materials for the laboratory works to be performed at the tokamak. Testing of the in situ methods for analyzing the plasma-surface interaction is one of the main scientific and technological goals set for this tokamak. The diagnostic complex described in the paper provides cumulative information on the processes occurring after the plasma-surface contact; it represents a set of very informative and well-tested diagnostic tools that allow the students to obtain visual reliable information on the processes occurring in the tokamak vacuum vessel.

An array of ten built-in divertor probes was installed on the lower dome of the Globus-M spherical tokamak. It was used to measure the profiles of the floating potential, electron temperature, ion saturation current, and heat flux density to the divertor. It is demonstrated that there is a sharp change in the floating potentials of the probes located near the outer separatrix leg at the development of the peripheral instability.

The article presents the results of studying the transfer of heat and particles in the Globus-M2 spherical tokamak in discharges with neutral injection at the current ramp up. An atomic beam was injected into the tokamak plasma at a fixed toroidal magnetic field of 0.7 T. The plasma current on the plateau was varied in the range 0.2–0.3 MA. Based on the electron temperature and concentration spatial distributions measured by the Thomson scattering method, the transport of heat and particles in plasma was simulated using the ASTRA code. The energy confinement time of the plasma was determined, as well as estimates of the coefficients of thermal diffusivity and diffusion was made.

We present new information on the development of Alfven modes on the Globus-M2 spherical tokamak. Data have been obtained using a V-band Doppler reflectometer with probing radiation frequencies from 50 to 75 GHz. Аs a multifrequency source of probing radiation, a microwave harmonic synthesizer has been used. As a result of using the new reflectometer, the toroidal Alfven mode localization in a magnetic field of 0.5 T has been obtained. The spectral components of the Alfven mode with Doppler frequency shifts caused by toroidal plasma rotation have been registered. In the central regions of the discharge, so-called “Alfven cascades” have been investigated.

The Alfven oscillations were studied using the multifrequency Doppler backscattering (DBS) method on the Globus-M spherical tokamak during the neutral deuterium injection. The principles of Alfven mode registration by the DBS method are discussed. Microwave schemes of multifrequency probing are described. It is shown that the observed plasma rotation velocity oscillations are due to the oscillations of the drift velocity in crossed radial electric field of the Alfven wave and the magnetic field. The absolute values of the magnetic field perturbations of the toroidal Alfven eigenmode (TAE) in the internal regions of the tokamak discharge are determined. The obtained data are compared with the results of the magnetic field oscillations measurements obtained by a magnetic probe located outside the plasma region. The use of multifrequency probing in the DBS method made it possible to determine the radial distributions of the amplitude of magnetic field oscillations. These data were obtained for oscillations at the TAE fundamental frequency and at harmonic frequencies, which were found to correspond to different toroidal mode numbers. In general, the data obtained can be compared with the results of numerical modeling of the Alfven spectrum in the Globus-M tokamak. The prospects for the further use of the DBS method for studying Alfven oscillations in toroidal installations are discussed.

One of the most important structural elements of a tokamak is a vacuum discharge chamber, which satisfies specific requirements concerning the induction method of plasma generation and, accordingly, the presence of alternating magnetic fields additionally acting on the chamber. Moreover, the chamber should not only be able to create and confine a current plasma but also be able to diagnose its parameters through appropriate pipes. The small spherical tokamak being created at MEPhI for both educational and research purposes has an extensive program of such studies; therefore, the discharge chamber should not only allow fast evacuation and withstand alternating magnetic fields but also be convenient for measuring its parameters, using various diagnostics, as well as allow quick replacement of internal elements and introduction of additional power into the chamber. The paper describes the design of the vacuum chamber of the MEPhIST-1 tokamak, which meets these requirements, and presents calculations to justify the choice of material and the thickness of the characteristic elements of the chamber, and also presents the test results of the finished chamber.

As part of the recovery project of the National Spherical Tokamak Experiment–Upgrade (NSTX-U), the divertor plasma-facing components (PFCs) were redesigned to handle significantly higher heat fluxe...

Globus-M2—a new 1-Tesla spherical tokamak—was recently launched. The main features and research directions of this machine in scope of fusion–fission reactor development are described. Main results of the first experimental campaign with toroidal magnetic field up to 0.73 T and plasma current up to 0.33 MA are discussed. Significant improvement of the discharge parameters as compared to Globus-M was achieved. Plasma total stored energy higher than 7 kJ was obtained. Energy confinement time increase was consistent with predictions by spherical tokamak scalings. Toroidal Alfven eigenmode-induced losses decrease with increase of plasma current and toroidal magnetic field. For the first time, LHCD with the toroidal wave slowing-down was successfully used at a spherical tokamak.

Plasma turbulence plays a key, governing role in determining the spatial-temporal evolution of plasmas in astrophysical, geophysical and laboratory contexts. In particular, turbulence on disparate spatial and temporal scales limits the level of confinement achievable in magnetic confinement fusion experiments and therefore limits the viability of sustainable fusion power 1 . The TDoTP project* (Turbulent Dynamics of Tokamak Plasmas) aims to advance understanding of plasma turbulence coupling across a wide range of scale sizes. As part of this effort, a mm-wave based scattering diagnostic is being designed for the MAST-U spherical tokamak, to measure binormal oriented high-k (electron scale) turbulence in the core plasma. We present the results of Gaussian wave optics and beam-tracing calculations 2 that demonstrate the predicted spatial and wavenumber resolution of the diagnostic, as well as estimates for the sensitivity of measurement. Current specifications for the diagnostic include an operating frequency of 260GHz, a turbulence wavenumber measurement range of k ⊥ ρ e = 0.1 -> 0.5 (where k ⊥ is the poloidal turbulence wavenumber and ρ e the electron gyroradius) and a minimum spatial localisation of ~3cm along the primary beam path.

The problem of accurate and reliable plasma shape control is significant, both for modern operating tokamaks, for example for the Globus-M/M2 spherical tokamak, and for future thermonuclear tokamak-reactors using magnetic plasma confinement. The article presents the new results of design and modeling the plasma shape control system for the Globus-M/M2 tokamak with the pre-trained artificial neural network as a plasma equilibrium reconstruction algorithm, which is included in the feedback of the control system. To collect the necessary data for training the artificial neural network and to model the plasma control system the developed magnetic plasma evolutionary code was used.

New polarizer and launcher systems on a 28-GHz electron-cyclotron (EC) heating and current drive (HCD) system have been developed for non-inductive second-harmonic EC-plasma-current ramp-up in the QUEST spherical tokamak. A launcher system with two quasi-optical mirrors providing beam steering capability was designed to focus the incident beam to a small-sized waist of ∼0.05 m at the second-harmonic EC resonance layer. A relatively large focusing mirror was designed based on a Kirchhoff integral code developed to derive wave solutions. The focusing property of the launched beam was first confirmed with a 3-dimensional (3D) electromagnetic-wave simulator. Focusing characteristics were also checked at low-power test facilities, together with the steering capability. The performance of this launcher system was demonstrated to work as designed, and assembled in the QUEST device. The system was applied to the non-inductive second-harmonic EC plasma ramp-up experiments with no optimization required regarding the incident polarization. The results obtained for the non-inductive plasma ramp-up are also presented.

Results of experimental studies of toroidal Alfven eigenmodes (TAEs) in the Globus-M spherical tokamak (R = 36 cm, a = 24 cm) are reported. The experiments were carried out in a wide range of plasma parameters at a magnetic field of up to 0.5 T and plasma current of up to 250 kA. Auxiliary plasma heating was performed by tangential injection of a deuterium beam with a power of Pb= 0.75 MW and particle energy of Eb = 28 keV into deuterium plasma. The experiments have shown that the TAE-induced loss of fast particles decreases with increasing plasma current and magnetic field. Using multifrequency Doppler backscattering diagnostics, it is established that the TAEs are localized at the plasma periphery. Results of simulations of the Alfven continuum and TAE structure by means of the modified KINX and CAXE codes agree satisfactory with the experimental data on the TAE frequencies and localization.

The results of experiments and simulations of the transport of heat and particles in the Globus-M spherical tokamak are presented. Investigations were carried out in the ohmic mode in hydrogen and deuterium plasma. It is shown that in the phase of current growth under the same initial conditions, as long as the condition is met at which the minimum of the safety factor exceeds unity, two scenarios of discharge development are possible: with an electron temperature dive or with an electron density peak. After the safety factor reaches the value of 1, the differences are no longer observed. Modeling of transport processes using the ASTRA code showed that the dive of the electron temperature and density is a consequence of the decrease in the transport of heat and particles in the plasma core (i.e., the presence of internal transport barriers).

The paper treats the methodology of the design and analysis of a tokamak plasma magnetic cascade multiloop control system with the example of the Globus-M spherical tokamak (Ioffe Institute, S-Petersburg, Russia) including a set of new features. The methodology covers: plasma equilibrium reconstruction based on the experimental data, derivation of linear plasma models relative to the reconstructed plasma equilibrium taking into account plasma diagnostics methods and equipment, development of a new original model of a thyristor current inverter as an actuator included in the system feedback, original multivariable cascade multiloop structure analysis by Relative Gain Array/ μ -procedures, and design of an H ∞ , multivariable plasma shape control system with adaptation of the magnetic axis position.

The Lithium Tokamak eXperiment (LTX) is a spherical tokamak device designed to study lithium plasma facing components (PFCs). The lithium coated wall of LTX has been demonstrated to produce a plasma edge with high electron temperature (200 eV or greater). Plasma density in the outer scrape-off layer (SOL) is also found to be very low, around 2 × 1017m-3, as a result of the low recycling lithium boundary. The high temperature, low collisionality region of the plasma extends into the SOL. The recent upgrade to LTX-β includes installation of a neutral beam, which will provide further heating and fueling of the core plasma. Core and edge diagnostics will also be expanded. As part of this expansion, a Retarding Field Energy Analyzer (RFEA) has been developed for the SOL of LTX-β. Measurements of the ion temperature, ion energy distribution, and the local space potential will be performed in the SOL plasma using this RFEA. Upgraded high field side (HFS) and low field side (LFS) Langmuir probes will replace existing triple probes so that higher electron temperatures can be more reliably measured. The HFS probes are also positioned to give radial and vertical gradient measurements. The design of the RFEA will be presented, along with calibration data.Since a high temperature, low collisional edge is expected for LTX-β, with a high mirror ratio near the LCFS (around 4), the majority of particles in the SOL will be mirror-trapped. Trapped particle effects will therefore become significant in the physics of the SOL plasma, and warrant further theoretical investigations. Here we present a theoretical study of the ambipolar potential formed in the collisionless SOL via differential loss of the electrons and ions, known as the Pastukhov potential in the literature. Numerical results will also be presented. Keywords: Lithium PFC, Scrape off layer, Ambipolar potential, Retarding field energy analyzer

In this article a model for calculating energy and particles densities is presented in order to control burning plasma in the case of D3He fuel in spherical tokamak (ST). A spherical tokamak is a possible candidate for a D3He fusion reactor due to its high-beta value. In this model a one-dimensional approximation of the transport equation for energy as well as the density of deuterium-helium3 fuel ions and alpha particles is represented in cylindrical coordinates by a system of partial differential equation. By applying these simulation equations to control particles and energy densities profiles, the system is discretized in space using a finite difference method and a back-stepping design for D3He aneutronic fuel in tokamaks. Results obtained for boundary control conditions shows that initial values are approaching to equilibrium profiles and also the system achieves the stable equilibrium. It is also seen that with the controller modulation of the alpha particles, deuterium-helium3 and energy densities at the edge of the plasma, they are approaching to stable equilibrium points. The results of boundary control law show that the profiles can be successfully controlled with just one step of back-stepping.

The article presents research on neutral beam current drive in Globus-M compact spherical tokamak. The experiments were performed in the plasma current range of 0.17–0.20 MA with a 0.4- or 0.5‑T toroidal magnetic field. The injection impact parameter was 33 cm. The variable parameters included compositions of plasma and the heating beam (hydrogen, deuterium), plasma density, and vertical displacement of the plasma column. The simultaneous increase in the plasma current and drop in the loop voltage were used to determine the neutral beam current drive. The injection of a hydrogen/deuterium beam into deuterium or hydrogen plasma resulted in a significant and reproducible drop in the loop voltage (up to 0.5 V). In order to process the obtained data, the authors developed an ASTRA code-based model, which allows one to calculate the neutral beam-driven current and bootstrap current. The share of noninductive currents as a function of plasma density during the injection of a hydrogen beam (28 keV, 0.5 MW) into the deuterium plasma was calculated. The authors analyze the results of experiments on off-axis beam injection achieved by vertical displacement of the plasma column and the effect of increasing the toroidal magnetic field on the consumption of the poloidal magnetic flux in discharges with atomic beam injection.

A multi-frequency electron cyclotron resonance heating (ECRH) and current drive (CD) system is currently under construction at the ST40 spherical tokamak. The system employs 2 GYCOM gyrotrons with a maximum output power of 1 MW each and a pulse length of 2s. Both gyrotrons can be tuned to operate either at 105 GHz or 140 GHz. The system is designed to study a non-inductive plasma start-up, current ramp-up and sustainment.

QUEST (Q-shu university experiment with steady state spherical tokamak) is a midsize spherical tokamak capable of steady-state operation, comprising all-metal plasma-facing walls and a hot wall (HW) to address issues pertaining to fuel particle balance. The HW was installed summer 2014. Quantitative analysis pertaining to the HW at 373 K is carried out, and clarify the quantitative impact of shot history that obviously appears in wall stored hydrogen just before the discharge at the wall temperature. The model indicates the plasma-induced deposition layer play an essential role in fuel particle balance. A clear temperature dependence of fuel recycling was observed using outgassing just after plasma termination and played an essential role in regulation of particle balance. Consequently, long duration discharges lasting more than 1 h has been obtained at wall temperature, TW

A spectroscopic system with multiple viewing chords was developed for QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) to measure the spatial distribution of ion toroidal velocities in discharges sustained by electron cyclotron resonance heating (ECH). Twenty-four viewing chords were aligned in the midplane and C III emission line spectra were measured for three types of ECH discharge under different magnetic field configurations. By applying an inversion method to the measured spectra, we evaluated the radial distributions of C2+ ion emissivity, temperature, and toroidal velocity. The error in the evaluated velocity was estimated to be less than 5 km/s. It was found that the velocity depends on the magnetic field configuration.

As a new spherical tokamak designed to simplify the engineering requirements of a possible future fusion power source, the EXL-50 experiment features a low aspect ratio vacuum vessel (VV), encircling a central post assembly containing the toroidal field coil conductors without a central solenoid. Multiple electron cyclotron resonance heating (ECRH) resonances are located within the VV to improve current drive effectiveness. Copious energetic electrons are produced and measured with hard x-ray detectors, carry the bulk of the plasma current ranging from 50–150 kA, which is maintained for more than 1 s duration. It is observed that over one ampere current can be sustained per watt of ECRH power issued from the 28 GHz gyrotrons. The plasma current reaches I p > 80 kA for high density (>5 × 1018 m−2) discharge with 150 kW ECRH. An analysis was carried out combining reconstructed multi-fluid equilibrium, guiding-center orbits of energetic electrons, and resonant heating mechanisms. It is verified that in EXL-50 a broadly distributed current of energetic electrons creates a smaller closed magnetic-flux surface of low aspect ratio that in turn confines the thermal plasma electrons and ions and participate in maintaining the equilibrium force balance.

The spatial distribution of the hydrogen atom density was evaluated in a spherical tokamak (ST) plasma sustained only with 28 GHz electron cyclotron heating (ECH). The radially resolved H&delta, emissivity was measured using multiple viewing chord spectroscopy and Abel inversion. A collisional-radiative (CR) model analysis of the emissivity resulted in a ground-state hydrogen atom density of 1015&ndash, 1016 m&minus, 3 and an ionization degree of 1&ndash, 0.85 in the plasma.

Plasma shape reconstruction is an important part of tokamak data analysis. An off-line plasma shape reconstruction, based on the EFIT code in Linux system, can be carried out in Q-shu Univ. Experiment in Steady-State Spherical Tokamak (QUEST). In this paper, the algorithm about equilibrium reconstruction is described in detail. Magnetic reconstruction in QUEST, which is calculated by the EFIT code according to the external magnetic measurements, is introduced and the results are presented in detail. It is known that the eddy current effect is large in QUEST experiment, the eddy current effect is first considered into equilibrium reconstruction. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. To test the accuracy of the result, a comparison with the result of Cauchy condition surface method has made, which suggests the reliability of the result. A visualization interface is made, which is convenient for physicists and operators to analyze the behavior of plasma. At the same time, the operation space of elongated discharges is discussed based on the QUEST database. The real time EFIT will be applied to the plasma control system of QUEST in the future.

Detailed in situ studies of magnetic reconnection and particle acceleration, which play a crucial role in the release and redistribution of energy in solar flares, can be performed in tokamak plasmas under conditions resembling those of the flaring solar corona. Recent measurements and modelling of fast particle production during reconnection events in the Mega-Amp Spherical Tokamak (MAST) are described. Specifically, observations in this device of electron acceleration during edge localised modes, and of both ion and electron acceleration during merging-compression plasma start-up, are presented, and possible implications of these studies for particle acceleration in flares are discussed. The results from MAST lend weight to the conjecture that large numbers of ions are accelerated to sub-MeV energies in flares.

The paper describes a new hierarchical tokamak plasma shape control method implemented in the system with switching robust controllers and matching state vectors. A system robustness was achieved via the H∞ loop-shaping design. A novel modeling approach was used for the feedback system simulation: tokamak plasma equilibrium was calculated by the reconstruction code receiving the sum of experimental scenario signals and the output signals of a linear parameter varying (LPV) model, and sending the output to the multivariable plasma shape controller to close the loop. The LPV plant model is based on plasma equilibria reconstructed from the experimental data of the spherical tokamak Globus-M and is used to simulate plasma responses to small deviations from tokamak discharge scenario. The control system components were designed and combined to achieve high-performance operation during short pulses of plasma with total duration of about 200 ms and the divertor phase time length of 20–25 ms. The non-iterative Improved Moving Filaments method was developed for the plasma equilibrium reconstruction with the simultaneous determination of coordinates and currents of filaments. The plasma shape is controlled via the magnetic field at X-point and poloidal fluxes on a plasma separatrix, without calculation of the gaps between the plasma separatrix and the first wall. The hierarchical feedback control method was used in simulations of the entire plasma discharge with the transition from the limiter to the divertor plasma magnetic configuration. The control system was discretized and simulated on a high-speed computer to be prepared for application in Globus-M experiments by a real time test bed.

Публикация посвящена применению методов вычислительной геометрии, интервального анализа и линейного программирования к задачам физики управляемого термоядерного синтеза. Рассмотрены геометрические аспекты проблемы, получены проекции светимостей различных объемов сферического токамака на плоскость матричного детектора, изучены изображения предполагаемых макроскопических структур и микроскопических включений. Для набора модельных распределений светимости объема токамака поставлена задача восстановления сигнала. Решение получено с использованием задач линейного программирования., The problems of reconstruction of plasma luminosity are important for physics and technology of power plants-tokamaks. The Globus-M research tokamak obtained a large amount of data using a matrix detector in pinhole camera geometry. From the mathematical point of view, finding the luminosity for different regions of the plasma volume according to the matrix detector is an inverse problem related to the field of integral geometry. An essential feature of the particular task is the use of a single fixed camera with a small viewing angle. In this regard, application of methods of harmonic analysis of data is not enough. The paper investigates the geometric aspects of the problem. In the general view, a threedimensional object is projected onto a two-dimensional plane through a diaphragm. Under the assumption of azimuthal symmetry, there is a central projection of the luminosity of the body of rotation onto a flat matrix detector. The initial information for the calculation is the plasma boundary obtained from magnetic sensors. There is no reliable information about the internal structure of the plasma, so its division into regions of the equal luminosity is not unambiguous. The paper presents an algorithm for finding the projections of the luminosity of plasma volumes on the plane of the matrix detector. A set of model direct problems for the construction of algorithms for their recognition according to the detector data was investigated. Images of supposed macroscopic structures and microscopic inclusions were obtained. The methodological basis of the work is the use of interval analysis methods for solving geometric and algebraic problems. This approach allows obtaining qualitative and quantitative results that takes into account the uncertainty of the input data with the minimum amount of computational costs. Algebraic solvability is investigated in the interval formulation using response functionality. Solutions for a set of test problems are obtained, which demonstrate the availability of successful reconstruction for real data. An important result of the study is an information about the presence of uncertainties in geometric data and related calculations by obtaining results about the luminosity of the plasma by solving linear programming problems., №25 (2020)

The 2/1 tearing mode instability during the plateau phase of the plasma current time evolution has been observed since the start of the Globus-M spherical tokamak operation. Because of a small amount of shots in a database, the poloidal beta dependence of the instability growth rate had been investigated with an insufficient accuracy. In this work, a more detailed analysis has been produced that proves the fact that the observed modes can be treated as neoclassical. The saturated island width as a function of poloidal beta has been obtained; a time evolution of the magnetic island has been calculated from the Mirnov magnetic coil signals and compared to the modified Rutherford theory, and the critical magnetic island width has been found for the neoclassical tearing mode (NTM) of the Globus-M tokamak.

With respect to a particle fueling method in a spherical tokamak reactor with an advanced fuel, we investigate the availability of a merging process following the translation of a small torus plasma through which a plasma current flows in the same direction as the main plasma. A three-dimensional resistive MHD model is used to examine the particle fueling effect by reproducing the translation process and simultaneously performing an ion trajectory calculation. The simulation shows the merging process between the translated plasma and the main plasma and exhibits potential as a particle fueling method. The results of the study indicate that up to 30% of the ions in the secondary plasma are supplied to the main plasma.

In this study a hybrid simulation environment to investigate the lower-hybrid-wave-driven tokamak plasmas is presented, and its application to the spherical tokamak TST-2 is described. These plasma are formed and driven by radio-frequency waves without the use of the central solenoid, and are characterized by low density and low magnetic field. A hybrid simulation environment which is divided into two groups, one using magneto-hydrodynamic (MHD) as well as particle-in-cell (PIC) approaches, and the second group using ray-tracing and Fokker–Planck solvers, is applied to describe the behavior of energetic electrons, bulk plasma, wave propagation, and the wave-particle interaction. Both groups of solvers can be coupled via the energetic-particle velocity distribution function and the equilibrium conditions of magnetic field, pressure, and density profiles to obtain a self-consistent solution. First results show the impact of a self-consistent equilibrium on ray trajectories and current density profiles. Therefore, new insights in lower-hybrid-wave-driven plasmas of TST-2 can be obtained using the proposed hybrid simulation environment.

With good confinement and relatively high beta (up to 40%), the spherical tokamak could be a suitable magnetized target. To illustrate this potential, the theoretical fusion yield from compressing a spherical tokamak adiabatically to fusion conditions is calculated. While it is found that the large amount of energy stored in the compressed toroidal field constrains the potential energy gain, a system can be designed to produce significant useful output by directly recovering a large fraction of the compression energy. With 80% energy recovery, an example system using 140 MJ of pressurized gas to compress the spherical tokamak plasma within a 2 m radius liquid metal flux conserver can produce a fusion yield of 140 MJ and a net output of 40 MJ per pulse. At ~ 1 Hz this would produce 40 MWelectric.

Experiments intended to study the dependence of the ion temperature on the plasma current were carried out on the Globus-M spherical tokamak. The ion temperature was measured at a few spatial points with the help of charge exchange spectroscopy and a charge exchange atom analyzer. On the basis of the experimental data, modeling with the use of the ASTRA numerical code was carried out and the thermal diffusivities for the ion plasma component were determined. The calculation results show that the ion behavior is similar to neoclassical.

In tokamaks, the temperature of the plasma-facing wall is an important parameter for achieving particle balance and therefore steady-state operation. QUEST, which is a middle-sized spherical tokamak, has hot walls that act as plasma-facing walls. They can be actively heated with sheath heaters and actively cooled with water. To control the wall temperature, heating and cooling systems have been developed. These systems adjust the power of the sheath heaters and the motor valves of the cooling system, respectively. The two systems communicate via Ethernet through UDP and control the hot-wall temperature cooperatively. The plasma control system (PCS) in QUEST has also been modified, especially with respect to gas fueling, in order to enable long-duration plasma sustainment. A feedback controller has been installed in the PCS, together with a mass flow controller, allowing Hα emission from the plasma which is used as a reference signal, to be well controlled. Plasma density calculations using a field-programmable gate array are proposed for the feedback control system.

The concept of the small spherical tokamak project is described, the main objectives of the project are to increase competencies at the University in the training of personnel in the field of physics and technologies of controlled thermonuclear fusion, as well as attracting young people to this area. The implementation of the project also implies giving impetus and new opportunities for improving the methods of plasma diagnostics developed at the University, studies on the plasma surface interactions and computer simulation of processes in the plasma and on the plasma facing surfaces. The installation has a large radius of 25 cm, an aspect ratio of less than 2, and a vertical elongation of ∼ 3, which allows, in principle, for small sizes and costs of the installation, taking into account the subsequent increase in the toroidal field to ∼ 2 T, to carry out important studies for progress in plasma performance in tokamaks. Namely, research on physics of plasma confinement, current drive with RF power and plasma interaction with materials. The project includes a phased implementation with a multiple increase in the magnetic field and plasma current, as well as the possibility of quick and convenient access to the internal elements of the discharge chamber and the simultaneous use of a large number of diagnostics.

A microwave pre-ionization source is designed, fabricated and installed on Glass Spherical Tokamak (GLAST) to reduce the high-loop voltage requirement during the plasma startup. In GLAST II, limited space is available at the in-bore side, which restricts the size and heating capability of central solenoid. Therefore, the central solenoid alone cannot produce the required loop voltage for plasma breakdown. Electron Cyclotron Resonance Heating (ECRH) can be very useful to overcome this limitation. In this work, an economical microwave source is fabricated using magnetron of 2.45 ± 0.02 GHz, obtained from a household microwave oven. Pulsed-mode operation of the magnetron is achieved through certain modifications in the half-wave circuit of domestic microwave oven. A new technique is proposed to enhance the microwave power, where an additional electromagnet is introduced around the magnetron cavity that confines the fast moving electrons. This modified microwave source is sufficient to achieve the breakdown in GLAST II with improved plasma current of 5kA.

Presence of a large population of fast particles may qualitatively modify the tokamak equilibrium from the one that can be described by the standard magneto-hydrodynamics (MHD). The kinetic modification of the MHD equilibrium was studied for a lower-hybrid wave driven plasma on the TST-2 spherical tokamak. The analysis was performed using an equilibrium reconstruction method based on an extended MHD model that considers a two-component plasma of bulk MHD fluid and kinetic fast electrons. Scrape-off-layer current appeared naturally in the extended MHD analysis because of the finite electron orbit excursion from the flux surfaces. This reduced the bulk pressure contribution to the toroidal current substantially. The resulting poloidal flux profile was more consistent with that inferred from the Thomson scattering measurement.

Local helicity injection (LHI) is a non-solenoidal startup technique that utilizes electron current injectors at the plasma edge to initiate tokamak discharges. Viable non-solenoidal startup techniques require high central Te to combat resistive losses and enhance coupling to auxiliary methods of current drive/heating. Thomson scattering measurements of LHI discharges in Pegasus showed peaked Te profiles at Ip∼0.15 MA and Bt∼0.15 T with Te,0∼100– 150 eV. These results are similar to Te profiles observed with Ohmic induction. At lower levels of Bt, LHI Te profiles were hollow with Te,0∼40 eV and Te,max≤120 eV depending upon the helicity input. Regardless of the Bt level and helicity input, the electron pressure profiles were flat/peaked with hollow JR profiles. Equilibrium reconstructions and measurements of core absolute extreme ultraviolet radiation suggest the hollow Te profiles are a result of very low resistive heating power in the core due to the edge-localized nature of LHI and low-Z line radiation losses. Estimates of Zeff from the plasma conductivity indicate averaged values of ∼1 or ∼3 assuming neoclassical or Spitzer conductivity, respectively. When auxiliary heating power from magnetic reconnection is considered, this observed LHI performance is comparable to expectations from a linear Ohmic confinement scaling estimate and a collisional stochastic confinement scaling estimate of the core plasma region.

The method of Doppler backscattering (DBS) has been applied for the study of Alfven oscillations in the tokamak plasma for the first time. The oscillations of the plasma rotation velocity at the Alfven frequency have been able to be registered using DBS during the studies on the Globus-M spherical tokamak. The area of the development of Alfven instability has been determined. The measurement of the amplitude of rotation velocity oscillations has allowed the estimation of the absolute values of the amplitudes of the oscillations of the electric and magnetic fields of Alfven oscillations in the region of their existence in the plasma-discharge volume in tokamak.