Spherical Tokamak Research Articles

First Experiments on Reduction the Heat Load on the Divertor Plates of the Globus-M2 Tokamak Using Nitrogen Seeding and Their Comparison with Simulation Results

At the compact spherical Globus-M2 tokamak, a series of experiments was conducted to study theeffect of the injection of nitrogen on the discharge parameters. The experiments were carried out in dischargesin deuterium in the divertor configuration, and the auxiliary heating was performed by deuteriumneutral beam injection. During the nitrogen seeding, a substantial decrease in electron temperature near thedivertor was recorded as well as a sharp decrease of the heat flux onto the divertor plate, while the density andtemperature of the main plasma changed insignificantly. Simulations by the SOLPS-ITER showed a satisfactoryagreement with the experiment.

Free-Boundary Plasma Equilibrium Computation in Spherical Globus-M2 Tokamak by Means of the pyGSS Code

The pyGSS code constructed for computation of free-boundary plasma equilibrium in sphericalGlobus-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 inthe course of code execution in such a way that the results of reconstruction would agree with the experimentalmeasurements of the poloidal magnetic flux by means of toroidally closed loops. The results of computationof equilibrium are compared with those obtained by means of other codes and experimentally measuredthermal plasma energy, position of the separatrix outer leg, the diamagnetic-loop signal, etc.

Development of Next-Generation Spherical Tokamak Concept. The Globus-3 Tokamak

Development of Next-Generation Spherical Tokamak Concept. The Globus-3 Tokamak

Features of Plasma Disruption in the Globus-M2 Spherical Tokamak

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 quenchand in the stage of plasma current quench are measured using the diagnostics of Thomson scattering of laserradiation. The dependence of the plasma current decay time during disruption on the pre-disruption currentvalue is determined. The distribution of the toroidal current, which is induced during disruption, in the shellof the vessel is determined on the basis of magnetic measurements. Electromagnetic loads on the vessel arecalculated.

Diagnostic Complex of the Globus-M2 Spherical Tokamak

The diagnostic complex of the Globus-M2 spherical tokamak (R = 36 cm, a = 24 cm), the onlyoperating tokamak in Russia with a divertor plasma configuration, which operates in the range of subthermonucleartemperatures (Te to 1.6 keV, Ti to 4.5 keV) and densities (ne to 2 × 1020 m–3), is described. The Globus-M2 tokamak is the unique scientific facility, which is a part of the Federal Center for Collective Use ofthe Ioffe Institute, Russian Academy of Sciences “Materials Science and Diagnostics in Advanced Technologies.”This allows third parties to perform their research using it. The work contains a list of all diagnosticscurrently available on the tokamak. The description of the diagnostics is structured in such a way that thereader gets an idea of their capabilities for measuring plasma parameters with an emphasis on the limits andaccuracy of the measured values, and also spatial and time resolution. At the same time, many technicaldetails are omitted in order to save space; references are given to papers with a more detailed description ofindividual diagnostics.

Development of Next-Generation Spherical Tokamak Concept. The Globus-3 Tokamak

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 differsfrom the currently operating Globus-M2 tokamak in the stronger toroidal magnetic field (1.5–3.0 T) andincreased duration of plasma discharge. The parametric analysis data are presented that determined the preliminaryselection 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-temperaturesuperconductors. For the first option, the concept for designing the electromagnetic system and vacuum vesselof the facility has been developed. The basic shot scenario with duration of up to 3 s at the field of 1.5 Tand plasma current of 0.8 MA is presented.

Diagnostic Complex of the Globus-M2 Spherical Tokamak

Diagnostic Complex of the Globus-M2 Spherical Tokamak

Neutral Injection Complex for Globus-M2 Spherical Tokamak

The injection complex that heats plasma of Globus-M2 spherical tokamak consists of two injectorswhich supply high-energy beams of hydrogen (deuterium) atoms into the device’s plasma. Injectors areautonomous and allow for preparing and setting up the device and measuring the parameters of the atomicbeam independently of the other injector and of the tokamak. The authors provide detailed analysis of eachinjector’s configuration, substantiate their selection of the experiment layout for introducing beams into thetokamak plasma, set forth characteristics of both injectors’ atomic beams injected into the tokamak plasma,and discuss the results of experiments on additional heating of Globus-M2 tokamak plasma obtained byinjecting two atomic beams.

Features of Plasma Disruption in the Globus-M2 Spherical Tokamak

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.

Neutral Injection Complex for Globus-M2 Spherical Tokamak

Neutral Injection Complex for Globus-M2 Spherical Tokamak

Generation of energetic electrons by an electron cyclotron wave through stochastic heating in a spherical tokamak

This study presents novel findings on stochastic electron heating via a random electron cyclotron wave (ECW) in a spherical tokamak. Hard x ray measurements demonstrate the time evolution of hard x ray counts at different energy bands, consistent with predictions from the stochastic heating model. The ECW heating rate shows a positive correlation with applied power, confirming the effectiveness of stochastic heating. Remarkably, the ECW-driven plasma current remains insensitive to ECW incidence angle, consistent with model predictions. The observed stochastic heating of electrons offers potential for exploring innovative non-inductive current drive modes in spherical tokamaks. This research contributes to the understanding of plasma behaviour and motivates the development of new models for non-inductive current drive in fusion devices.

Electromagnetic gyrokinetic instabilities in STEP

We present herein the results of a linear gyrokinetic analysis of electromagnetic microinstabilites in the conceptual high reactor-scale, tight-aspect-ratio tokamak Spherical Tokamak for Energy Production, https://step.ukaea.uk. We examine a range of flux surfaces between the deep core and the pedestal top for two candidate flat-top operating points of the prototype device. Local linear gyrokinetic analysis is performed to determine the type of microinstabilities that arise under these reactor-relevant conditions. We find that the equilibria are dominated at ion binormal scales by a hybrid version of the kinetic ballooning mode (KBM) instability that has significant linear drive contributions from the ion temperature gradient and from trapped electrons, while collisional microtearing modes (MTMs) are sub-dominantly also unstable at similar binormal scales. The hybrid-KBM and MTM exhibit very different radial scales. We study the sensitivity of these instabilities to physics parameters, and discuss potential mechanisms for mitigating them. The results of this investigation are compared to a small set of similar conceptual reactor designs in the literature. A detailed benchmark of the linear results is performed using three gyrokinetic codes; alongside extensive resolution testing and sensitivity to numerical parameters providing confidence in the results of our calculations, and paving the way for detailed nonlinear studies in a companion article.

Line integrated density measurements on the Versatile Experiment Spherical Torus (VEST) using frequency sweep interferometer

A frequency sweep interferometer (FSI) operating in the frequency range of 50–75 GHz is installed in the versatile experiment spherical torus spherical tokamak to measure the line integrated density (LID). FSI measures the time derivative of phase to calculate the group delay, which is proportional to the LID under the condition that the microwave frequency is much higher than the plasma frequency. Since the group delay is calculated from the time derivative of phase and the frequency sweep rate, FSI is very sensitive to the measurement noise. In the view point of signal processing, derivative exaggerates the measurement noise. Therefore, sophisticated techniques for phase measurement and frequency linearization are required to obtain meaningful results with FSI. The detailed techniques and the hardware setup are explained in the paper. The LID measured by FSI is benchmarked with the LID measured by a conventional 94 GHz heterodyne interferometer. The two measurements agree well. A conventional interferometer can no longer provide LID when severe phase errors occur. This is because phase errors propagate to subsequent measurements. However, FSI provides LID during the entire discharge time successfully regardless of frequent measurement failure because the LID is obtained in FSI from the time derivative of phase rather than the phase. In this sense, FSI is suitable as a diagnostics for steady state plasmas. The main cause for the phase errors is identified as the beam path displacement due to the refraction of the plasma.

Erosion estimates for the divertor and main wall components from STEP

The tungsten erosion within Spherical Tokamak for Energy Production (STEP) assuming tungsten main wall and tungsten divertor has been estimated with ERO at the inner and outer divertor, at the inner and outer midplane and at the outboard baffle entrance. Plasma parameters are based on SOLPS simulations applying argon puffing for edge cooling. The plasma parameter range covers peak electron temperatures T e between 3 and 25 eV in the divertor. At the inner midplane T e ∼ 13 eV, at the outer midplane ∼7 eV and at the outboard baffle entrance between 1 eV and 4 eV. The modelled peak gross erosion is highest in the divertor with up to 1E19 W m−2 s−1 within the inner and 7E19 W m−2 s−1 in the outer one for the plasma parameter range studied. At the main wall the gross erosion is about 2E18 W m−2 s−1 at the inner midplane and 1.3E17 W m−2 s−1 at the outer one. However, tungsten deposition within the divertor is much larger with amounts between 88% and 98% and only between 10% and 60% at the midplane. At all locations studied, tungsten erosion due to deuterium ions is negligibly small compared to the erosion by argon ions. Erosion due to deuterium atoms has been studied for the outer midplane and is there at least four times smaller than the erosion due to argon ions. The simulations have been performed considering singly ionised Ar. However, according to the SOLPS runs the mean charge of Ar impinging the surfaces is about two at the locations of largest erosion, which leads to an increase of the gross erosion by a factor between 1.5 and 5 with the largest increase occurring at the outer divertor target.

Resistive instabilities in toroidal anisotropic plasmas

Resistive singular layer equations are developed by applying the magnetohydrodynamic (MHD) model to toroidal anisotropic plasmas. This work extends the previous ideal MHD theory [Shi et al. Phys. Plasmas 23, 082121 (2016)] to the resistive case. These layer equations can be used to investigate resistive localized MHD instabilities, such as tearing instability and resistive interchange instability. Compared to existing resistive theory [Johnson and Hastie, Phys. Fluids 31, 1609 (1988)], our model includes plasma compressibility, allowing for a study of the coupling between parallel motion to perpendicular one, which is known as the apparent mass effect. In addition, these obtained equations are valid for low n modes, where n is the toroidal mode number. The dispersion relation is derived in a reduced model. We find that the anisotropic pressure effect (when p⊥ > p‖) not only increases the stable threshold of the resistive interchange mode but also raises the critical value Δc of the tearing mode stability index Δ′, which represents the logarithmic jump of the radial magnetic field perturbation across the rational surface. This discovery holds significant practical implications for mitigating neoclassical tearing modes in high confinement plasmas, particularly those characterized by a low aspect ratio (such as spherical tokamaks) or low magnetic shear (as designed in ITER hybrid scenarios). However, it enhances the growth rate of the tearing mode in a low growth rate region, where p‖ and p⊥ denote the pressure components parallel and perpendicular to the magnetic fields, respectively.

Analysis of Toroidal Alfven Eigenmode-Induced Fast Ion Losses in Globus-M2 Spherical Tokamak

Analysis of Toroidal Alfven Eigenmode-Induced Fast Ion Losses in Globus-M2 Spherical Tokamak

First data and preliminary experimental results from a new Doppler backscattering system on the MAST-U spherical tokamak

A new Doppler backscattering (DBS) system, consisting of Q-band and V-band, has been installed and achieved its first data on the MAST-U spherical tokamak. The Q-band and V-band have separate microwave source systems, but share the same optical front-end components. The Q-band and V-band sources simultaneously generate eight (34, 36, 38, 40, 42, 44, 46 and 48 GHz) and seven (52.5, 55, 57.5, 60, 62.5, 65 and 67.5 GHz) fixed frequency probe beams, respectively. These frequencies provide a large range of radial positions from the low-field-side edge plasma to the core, and possibly to the high-field-side edge, depending on the plasma conditions. The quasi-optical system consists of a remotely-tunable polarizer, a focusing lens and a remotely-steerable mirror. By steering the mirror, the system provides remote control of the probed density fluctuation wavenumber, and allow the launch angle to match the magnetic field. The range of accessible turbulence wavenumbers (kθ ) is reasonably large with normalized wavenumber kθρs ranging from <0.5 to 9. The first data acquired by this DBS system is validated by comparing with the data from the other DBS system on MAST-U (introduced in ref. [21]). An example of measuring the velocity profile spanning from the edge to the center in a high-density plasma is presented, indicating the robust capabilities of the integrated Q-band and V-band DBS systems.

Renovation of the IPM-2 Source Ion-Optical System for the Globus-M2 Spherical Tokamak Neutral Beam Injector

Renovation of the IPM-2 Source Ion-Optical System for the Globus-M2 Spherical Tokamak Neutral Beam Injector

Full-Wave 2D Modeling of Helicons Propagation and Absorption in the Spherical Tokamak Globus-M2

Full-Wave 2D Modeling of Helicons Propagation and Absorption in the Spherical Tokamak Globus-M2

Parameter Space Constraints for Compact Spherical Tokamak Fusion Reactors

Parameter space for spherical tokamak reactors is explored quantitatively to elucidate the main constraints for spherical tokamak design choices. Using a constant plasma current Ip search constraint, a set of four Ip scenarios (5, 10, 15, and 20 MA) is first explored in a wide parameter space. Considering modest but gradually increasing auxiliary power, a set of four machine configurations (major radius R = 1.25, 1.75, 2.25, and 3.5 m) is explored next, optimizing the Ip and the bootstrap fraction. Constraints that narrow down the vast parameter space are elaborated along with critical assumptions, such as current drive efficiency, H-mode enhancement factor, nuclear shielding efficiency, and confinement scaling. Limits on the current density of the center post and how it affects the shielding are quantitatively indicated, thereby setting a lower limit on the aspect ratio.