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The objectives of NSTX-U research are to reinforce the advantages of STs while addressing the challenges. To extend confinement physics of low- A, high beta plasmas to lower collisionality levels, understanding of the transport mechanisms that set confinement performance and pedestal profiles is being advanced through gyrokinetic simulations, reduced model development, and comparison to NSTX experiment, as well as improved simulation of RF heating. To develop stable non-inductive scenarios needed for steady-state operation, various performance-limiting modes of instability were studied, including MHD, tearing modes, and energetic particle instabilities. Predictive tools were developed, covering disruptions, runaway electrons, equilibrium reconstruction, and control tools. To develop power and particle handling techniques to optimize plasma exhaust in high performance scenarios, innovative lithium-based solutions are being developed to handle the very high heat flux levels that the increased heating power and compact geometry of NSTX-U will produce, and will be seen in future STs. Predictive capabilities accounting for plasma phenomena, like edge harmonic oscillations, ELMs, and blobs, are being tested and improved. In these ways, NSTX-U researchers are advancing the physics understanding of ST plasmas to maximize the benefit that will be gained from further NSTX-U experiments and to increase confidence in projections to future devices. [ABSTRACT FROM AUTHOR]

This work presents a laboratory experiment on the magnetic reconnection of two self-sustained, tilt-unstable spheromaks. Experimental observations, confirmed by a developed 3D Hall-MHD model, demonstrate that magnetic reconnection of these spheromaks suppresses the tilt instability by reducing the amplitudes of disruptive low-number toroidal modes. The strong toroidal component of outflow jets generates a shear toroidal velocity, which may disrupt the coherent structure of these modes, leading to their suppression. The damping rate of toroidal magnetic perturbations was found to be exponentially related to the shear toroidal velocity. Following the end of reconnection, the growth rate of the modes is linearly proportional to the decrease in shear velocity. [ABSTRACT FROM AUTHOR]

The Spherical Tokamak for Energy Production (STEP) programme hypothesizes that a compact machine offers a route to reduced capital cost that directly tackles the barrier to entry of this potentially transformative technology. History has shown that with an unsolved, complex and highly interdependent design challenge, there is a need to balance exploration of the problem with progress. Almost all complex systems arise from the evolutionary improvement of simpler systems which is an approach the programme has adopted by working through a virtual natural selection of design families towards a single concept consistent with the initiating hypothesis. Issues are uncovered and solved more rapidly this way because the effort is focused on an end. In this current phase, STEP has had to be an agile fast-moving programme to work with what emerges as well as what was planned, to sit with uncertainty and to embrace self-organizing principles. The complex decision-making and compromises in emerging trades have led to a concept respectful of the tight aspect ratio hypothesis which carefully balances cost, performance and deliverability. It remains a high-risk and high-reward programme, but the character of the challenge is better understood building confidence and enhancing capability to advance the evolving design further. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

The programme to design plasma scenarios for the Spherical Tokamak for Energy Production (STEP), a reactor concept aiming at net electricity production, seeks to exploit the inherent advantages of the spherical tokamak (ST) while making conservative assumptions about plasma performance. This approach is motivated by the large gap between present-day STs and future burning plasmas based on this concept. It is concluded that plasma exhaust in such a device is most likely to be manageable in a double null (DN) configuration, and that high core performance is favoured by positive triangularity (PT) plasmas with an elevated central safety factor. Based on a full technical and physics assessment of external heating and current drive (CD) systems, it was decided that the external CD is provided most effectively by microwaves. Operation with active resistive wall mode (RWM) stabilization as well as high elongation is needed for the most compact solution. The gap between existing devices and STEP is most pronounced in the area of core transport, owing to high normalized plasma pressure in the latter which changes qualitatively the nature of the turbulence controlling transport. Plugging this gap will require dedicated experiments, particularly on high-performance STs, and the development of reduced models that faithfully represent turbulent transport at high normalized pressure. Plasma scenarios in STEP will also need to be such that edge localized modes (ELMs) either do not occur or are small enough to be compatible with material lifetime limits. The high current needed for a power plant-relevant plasma leads to the unavoidable generation of high runaway electron beam current during a disruption, where novel mitigation techniques may be needed. This article is part of the theme issue 'Delivering Fusion Energy – The Spherical Tokamak for Energy Production (STEP)'. [ABSTRACT FROM AUTHOR]

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 $$R$$ 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. [ABSTRACT FROM AUTHOR]

Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MW/m2. The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (≥1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (≥0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.

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. [ABSTRACT FROM AUTHOR]

In experiments with neutral beam injection on the Globus-M2 spherical tokamak, sequences of long-lasting harmonics of toroidal Alfvén modes were discovered, equidistant from each other in frequency and shifted from zero by a constant value. Using microwave Doppler backscattering diagnostics, the central localization of toroidal modes was determined. In this work, the possibility of "splitting" of toroidal harmonics due to the Doppler shift caused by the toroidal rotation of the plasma is being discussed. It is found that the unshifted frequency of the toroidal Alfvén mode obtained from the spectrum of the magnetic probe signal is in good agreement with the frequency of the mode calculated at the mode location radius, and the toroidal rotation frequency, also determined from the spectrum of the magnetic probe signal, correlates well with the rotation frequency measured using charge exchange spectroscopy diagnostics, but differs by a constant amount. Possible reasons for the discrepancies are being discussed. [ABSTRACT FROM AUTHOR]

The spherical tokamak (ST) operates in a steady state with a high fusion gain. The 0-dimensional power balance model, including radiation losses to determine Q value as an inductive fusion gain, and the current balance model for determining Q CD as a non-inductive fusion gain, is used to investigate the viability of D– 3 He fuel for a steady-state operation. The spherical tokamak’s geometry, including the magnetic field B t and β th as a ratio of its kinetic pressure to the magnetic pressure, is used to analyse the impact of the confinement enhancement factor H y 2 and the impurity density fraction f I on Q CD . By comparing the obtained values with the device data, plasma characteristics, such as the safety factor q I and Greenwald density limit N G are examined to determine the optimum density limit and safety factor for an assurance about Q ≈ Q CD as the aim of steady-state operation. A comparison with ARIES-III performance is also made. The overall plant power balance equation is included. Furthermore, the desirable plant thermal efficiency value η th and normalised beta value β N for producing net electric power P NET > 1 GW for the ST are achieved. Therefore, ST’s capability of having a lower aspect ratio A and higher elongation κ s than ARIES-III in generating more significant fusion power with lower H y 2 and higher energy confinement time τ E is approved. [ABSTRACT FROM AUTHOR]

Mock-up of the system for remote monitoring of the accumulation of hydrogen isotopes in the walls of the Globus-M2 tokamak was assembled and tested. The measurements were performed using the LIA-QMS laser diagnostics (laser-induced ablation with registration using quadrupole mass-spectrometry). The data were obtained on the content of hydrogen isotopes in deposits appearing after exposing tungsten collectors to the loads in the volume of the Globus M tokamak. After testing the diagnostics under laboratory conditions, it was mounted at the Globus-M2 facility. In-situ measurements of the content of hydrogen isotopes in the graphite tiles of the tokamak divertor were performed. The possibility of combining the L-IA‑QMS diagnostics with the LIBS (laser-induced emission spectroscopy) diagnostics has been confirmed, in order to obtain information on the composition of the ablated material. In addition, the LIBS method was used for obtaining the deuterium/protium isotopic ratio during measurements in the Globus-M2 facility. [ABSTRACT FROM AUTHOR]

Understanding and control of electron thermal transport is a critical point of research in magnetic fusion experiments. Previous experiments have shown that operation with reversed magnetic shear (RMS) can suppress electron thermal transport, resulting in the generation of internal transport barriers (ITBs), with the location of the ITB correlated with the location of minimum magnetic shear. The recent upgrades to NSTX—increased magnetic field up to 1 T, increased plasma current up to 2 MA, 2nd neutral beam—present an increased operating space in which to explore electron thermal transport in RMS plasmas. Utilizing TRANSP, we have developed operating scenarios by which to generate RMS in NSTX-U. The results suggest that RMS in NSTX-U can be generated through fast current ramp and early beam injection into a large plasma volume. This is very similar to the procedure that was followed in both TFTR and NSTX to generate RMS. Sustainment of RMS, disregarding non-( $q_{\mathrm{min}}$ = 1) MHD events, requires maintaining a large plasma volume, and increasing the core $T_{\mathrm{e}}$ , either via increased plasma current and/or adding heating power. Using this procedure, RMS was sustained for ∼1 s, with $q_{\mathrm{min}}$ $ \gt $ 1 for that period.

The achievement of a high toroidal magnetic field in a small spherical tokamak is challenging because of the small bore area in the central cylinder of the vacuum vessel. In this paper, we present a toroidal field coil of 0.3 T at the center of the MT-II tokamak. It has been designed, developed, and tested for installation at Pakistan Tokamak Plasma Research Institute (PTPRI). The coil is made of highly pure oxygen-free copper. It has a cross-sectional area of 10 × 15 mm2 (150 mm2) for the flow of an approximately 20-kA current to produce a 0.33 T toroidal magnetic field at the center of the tokamak. Mechanical support for the central stack of the inner legs is provided by a twisted grooved nylon cylinder to control the torque and attractive forces. The repulsive force density between the joints of the outer and inner legs is balanced by nuts and bolts along with an insulated ring of Teflon and an isolated metallic clamp from both ends. This compressive force also reduces connection resistance. The simulated currents and magnetic field are confirmed from the experimental results as well. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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 differs from the currently operating Globus-M2 tokamak in the stronger toroidal magnetic field (1.5–3.0 T) and increased duration of plasma discharge. The parametric analysis data are presented that determined the preliminary selection 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-temperature superconductors. For the first option, the concept for designing the electromagnetic system and vacuum vessel of the facility has been developed. The basic shot scenario with duration of up to 3 s at the field of 1.5 T and plasma current of 0.8 MA is presented. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The results of experiments are described, which were performed at the Globus-M2 tokamak and aimed at studying the fast ion loss at the outer tokamak wall due to fast ions interaction with the toroidal Alfvén eigenmodes. The local heating of carbon tiles was experimentally measured, and the corresponding heat flux was calculated. It was shown how simulations of the lost particle orbits can explain the characteristic features of the spatial map of wall heating. The flux of lost fast particles onto the wall was studied as a function of the instability amplitude. It has been demonstrated that the simulations predict similar dependence of the fast ion flux on the instability amplitude and also correlate its nature to the peculiarities of fast ions spatial distribution. [ABSTRACT FROM AUTHOR]

The diagnostic complex of the Globus-M2 spherical tokamak (R = 36 cm, a = 24 cm), the only operating tokamak in Russia with a divertor plasma configuration, which operates in the range of subthermonuclear temperatures (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 of the 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 diagnostics currently available on the tokamak. The description of the diagnostics is structured in such a way that the reader gets an idea of their capabilities for measuring plasma parameters with an emphasis on the limits and accuracy of the measured values, and also spatial and time resolution. At the same time, many technical details are omitted in order to save space; references are given to papers with a more detailed description of individual diagnostics. [ABSTRACT FROM AUTHOR]

Analysis of helicon wave heating and current drive capability in EXL-50 spherical torus plasmas has been conducted. It is found that the driven current increases with the launched parallel refractive index n | | and peaks around n | | = 4.0 when the frequency of the helicon wave is between 300 and 380 MHz. The helicon wave current drive efficiency shows a relatively stable upward trend with increasing plasma temperature. Moreover, the driven current decreases as the plasma density increases. We also analyzed the current drive with helicon waves of 150 MHz and 170 MHz and found that the driven current at a lower frequency was lower than that at a higher frequency. A positive proportional relationship exists between the driven current and n | | . Besides, as n | | increases, the profile of the driven current becomes wider. Finally, the effect of the scrape-off layer (SOL) region on the helicon wave current drive was also investigated. [ABSTRACT FROM AUTHOR]

With an increase of magnetic field up to 0.8 T and plasma current to 400 kA, fast ion losses rate in the discharges with toroidal Alfven eigenmodes decreased in tokamak Globus-M2 comparing with Globus-M tokamak discharges. Taking into account the data on the discharges with increased magnetic field and plasma current, the regression fit of neutral particle analyzer flux drop in energy channel close to neutral beam energy on relative eigenmode magnitude, the value of magnetic field and plasma current was analyzed. The power of flux drop dependence on TAE magnitude was found to be ~0.5 and inverse proportional on the value of product of magnetic field and plasma current, which is highly likely is determined only by plasma current due to weak dependence on magnetic field. The result obtained indicates that fast ion losses in Globus-M2, stimulated by toroidal Alfven eigenmodes are mostly determined by the shift of passing orbits to the plasma edge. With the increase of plasma current and magnetic field, neutron flux drops arising in the moments of toroidal mode bursts have also decreased. [ABSTRACT FROM AUTHOR]

The paper discusses the results of the calibration of two corona neutron counters used to measure the total neutron yield from the plasma of the Globus-M2 tokamak. The calibration was carried out in the experimental hall of the Globus-M2 facility using an AmBe source. During the calibration, the source moved at a constant speed around the central solenoid in the equatorial plane of the vacuum chamber, and one of the detectors was gradually moved away from the tokamak along a line with a constant toroidal angle. The dependence of the calibration coefficient obtained depending distance of the detector from the tokamak is presented. The calibration technique made it possible to separate the contributions from the direct neutron flux emitted by the plasma and from the flux of neutrons scattered on the elements of the experimental hall in the detector signal. [ABSTRACT FROM AUTHOR]

BackgroundENN Science and Technology Development Co., Ltd. (ENN Fusion Technology R&D Center) is upgrading its compact fusion research facility EXL-50 to EXL-50U. Both devices are the conventional conductor tokamak, on which the magnet power supply system is composed by 1 TF (Toroidal Field) power supply, 1 CS (Center Solenoid) power supply and 10 PF (Poloidal Field) power supplies PF1-10. All 12 sets of power supply system are powered by 2 AC pulse generators and output DC current through thyristor-based converters.PurposeThis study aims to design EXL-50U magnet power supply for satisfying high parameter requirements of EXL-50U.MethodsPower supply capacity was the first concern for upgrading and the corresponding protection strategies under high parameter conditions was taken into account as well. The configuration of AC pulse generator was introduced at the beginning. Then transformers and converters were listed and designed in scheme. Control system and protection process were implemented respectively, followed by detail power supply system illustration and commissioning waveforms display for each power supply.ResultsThe reliability and controllability of developed power supply system are verified by the waveforms that forms plasma current under the condition of CS breakdown.ConclusionIt is proved that this power supply system can work stably, and output waveforms can be repeated no matter it works alone or under complex condition of joint debugging.

A key aspect in the design of a spherical tokamak reactor is the optimisation of the plasma equilibrium, together with a compatible divertor configuration, and the corresponding poloidal field system. This is a complex multi-disciplinary problem, integrating plasma physics and engineering in order to satisfy a multitude of often conflicting requirements and constraints. The equilibrium design process employed in this work takes into account the reference plasma operating scenario, the power exhaust solution, and the engineering limits.Managing the heat exhaust proves to be one of the most challenging issues in a compact device such as the UKAEA STEP reactor. With a smaller major radius, the available target area over which the scrape-off layer heat load must be deposited is relatively small as compared to conventional aspect ratio devices. Consequently, alternative and advanced divertor concepts need to be considered, having significant implications for the whole reactor design. Here we address the inner divertor power handling challenges. With very limited inboard space, and the small radius of the inner strike point, the associated heat loads are likely to exceed the power handling capacity of standard divertors (SD). An alternative divertor configuration approaching an X-divertor (XD), created by inducing a secondary X-point near the inner strike point, is compared with an SD configuration optimised for the maximal connection length and maximal poloidal flux expansion. The inner X-divertor, simultaneously achieving strong poloidal flux expansion, increased connection length and higher divertor volume, proved to be advantageous in reducing target heat loads and favouring detachment. Amongst a number of viable exhaust solutions considered, the inner-X divertor is indeed emerging as a promising candidate.

The MAST-U fusion plasma research device, an upgrade to the Mega Amp Spherical Tokamak, has recently completed its first campaign of physics operation. MAST-U operated with Ohmic, or one or two neutral beams for heating, at 400–800 kA plasma current, in conventional or 'SuperX' divertor configurations. Equilibrium reconstructions provide key plasma physics parameters vs. time for each discharge, and diagrams are produced which show where the prevalence of operation occurs as well as the limits in various operational spaces. When compared to stability limits, the operation of MAST-U so far has generally stayed out of the low q, low density instability region, and below the high density Greenwald limit, high beta global stability limit, and high elongation vertical stability limit. MAST-U still has the potential to reach higher elongation, which could benefit the plasma performance. Despite the majority of operations happening below established stability limits, disruptions do occur in the flat-top phase of MAST-U plasmas. The reasons for these disruptions are highlighted, and possible strategies to avoid them and to extend the operational space of MAST-U in future campaigns are discussed. [ABSTRACT FROM AUTHOR]

NBI-assisted plasma heating with one or two injectors of fast neutral atoms was studied at the Globus-M2 spherical tokamak at the toroidal magnetic fields of 0.8–0.9 T and plasma currents of 0.35–0.4 MA. Measurements of the spatial temperature and electron density distributions, performed using the Thomson scattering diagnostics, showed a twofold increase in heating of plasma electrons during the injection of neutral particles with energies of up to 45 keV at the beam power of 0.75 MW, as compared to the ohmic heating regime. Switching on the second additional beam with the particle energy of up to 30 keV and power of up to 0.5 MW resulted in obtaining the hot ion mode in the range of mean plasma densities of (1.6–10) × 1019 m−3. According to the data of active spectroscopy and neutral particle analyzer diagnostics, in the hot zone, the ion temperature reached 4 keV at the plasma density of 8 × 1019 m−3, which is more than 2.5 times higher than the electron temperature. [ABSTRACT FROM AUTHOR]

The application of a neutral particle analyzer (NPA) diagnostic at the Globus-M/M2 spherical tokamaks is discussed. Physical principles of the diagnostic are reviewed. Two general approaches—active and passive measurements—are described. Examples of NPA application for the ion temperature and isotope composition measurements are presented. NPA-aided studies of the energetic ions in the MHD-free discharges, as well as in the experiments with sawtooth oscillations and toroidal Alfvén eigenmodes, are considered. [ABSTRACT FROM AUTHOR]

A sequence of magnetized target fusion devices built by General Fusion has compressed magnetically confined deuterium plasmas inside imploding aluminum liners. Here we describe the best-performing compression experiment, PCS-16, which was the fifth of the most recent experiments that compressed a spherical tokamak plasma configuration. In PCS-16, the plasma remained axisymmetric with $\delta B_\textrm{pol}/B_\textrm{pol} \lt 20\%$ to a high radial compression factor ( $C_\mathrm R \gt 8$ ) with significant poloidal flux conservation (77% up to $C_\mathrm R = $ 1.7, and ${\approx}30\%$ up to $C_\mathrm R = 8.65$ ) and a total compression time of 167 $\mu\mathrm{s}$ . Magnetic energy of the plasma increased from 0.96 kJ poloidal and 17 kJ toroidal to a peak of 1.14 kJ poloidal and 29.9 kJ toroidal during the compression, while the thermal energy was in the range of 350 ± 25 J. Plasma equilibrium was a low- β state with $\beta_\textrm{tor} \approx 4\%$ and $\beta_\textrm{pol} \approx 15\%$ . Ingress of impurities from the lithium-coated aluminum wall was not the dominant effect. Neutron yield from D-D fusion increased significantly during compression. Thermodynamics during the early phase of compression ( $C_\mathrm R \lt 1.7$ ) were consistent with increasing Ohmic heating of the electrons due to a geometric increase in the current density at near-constant resistivity, and with increasing ion cooling that approximately matched ion compression heating power. Ion cooling by electrons was significant because the electrons were much cooler than the ions ( $T_\mathrm e = 200\,\mathrm{eV}, T_\mathrm i = 600\,\mathrm{eV}$ ). Magnetohydrodynamic simulations were used to model the emergence of instabilities that increase electron thermal transport in the final phase of compression. Conditions for ideal stability were actively maintained during compression through a current ramp applied to the central shaft and, after this current ramp reached its peak two-thirds of the way through compression, we measured a transition in plasma behavior across multiple diagnostics.

The objectives of NSTX-U research are to reinforce the advantages of STs while addressing the challenges. To extend confinement physics of low- A , high beta plasmas to lower collisionality levels, understanding of the transport mechanisms that set confinement performance and pedestal profiles is being advanced through gyrokinetic simulations, reduced model development, and comparison to NSTX experiment, as well as improved simulation of RF heating. To develop stable non-inductive scenarios needed for steady-state operation, various performance-limiting modes of instability were studied, including MHD, tearing modes, and energetic particle instabilities. Predictive tools were developed, covering disruptions, runaway electrons, equilibrium reconstruction, and control tools. To develop power and particle handling techniques to optimize plasma exhaust in high performance scenarios, innovative lithium-based solutions are being developed to handle the very high heat flux levels that the increased heating power and compact geometry of NSTX-U will produce, and will be seen in future STs. Predictive capabilities accounting for plasma phenomena, like edge harmonic oscillations, ELMs, and blobs, are being tested and improved. In these ways, NSTX-U researchers are advancing the physics understanding of ST plasmas to maximize the benefit that will be gained from further NSTX-U experiments and to increase confidence in projections to future devices.

A more complete non-perturbative magnetohydrodynamic (MHD)-kinetic hybrid formulation is developed by including the perturbed electrostatic potential δφ in the particle Lagrangian. The fluid-like counter-parts of the hybrid equations, in the Chew-Goldberger-Low high-frequency limit, are also derived and utilized to test the new toroidal implementation in the MARS-K code. Application of the updated non-perturbative hybrid model for a high- β spherical tokamak plasma in MAST finds that the perturbed electrostatic potential generally plays a minor role in the n = 1 ( n is the toroidal mode number) resistive wall mode instability. The effect of δφ is largely destabilizing, with the growth rate of the instability increased by several (up to 20) percent as compared to the case without including δφ . A similar relative change is also obtained for the kinetic-induced resonant field amplification effect at high- β in the MAST plasma considered. The updated capability of the MARS-K code allows quantitative exploration of drift kinetic effects on various MHD instabilities and the antenna-driven plasma response where the electrostatic perturbation, coupled to magnetic perturbations, may play important roles.

The first pedestal stability and structure analysis on the new MAST Upgrade (MAST-U) spherical tokamak H-mode plasmas is presented. Our results indicate that MAST-U pedestals are close to the low toroidal mode number ( n ) peeling branch of the peeling-ballooning instability, in contrast with MAST H-mode pedestals which were deeply in the high- n ballooning branch. This offers the possibility of reaching the ELM-free quiescent H-mode (Burrell et al 2005 Plasma Phys. Control. Fusion 47 B37–B52) or high-performance super H-mode (Snyder et al 2015 Nucl. Fusion 55 083026; Snyder et al 2019 Nucl. Fusion 59 086017) regimes. In addition, the coupling between the peeling and ballooning branches is weak in MAST-U, suggesting that a path to very high pedestal pressure gradient at high density may exist with sufficient heating power. A possible explanation for the differences between MAST and MAST-U pedestal stability is given in terms of plasma shaping parameters, in particular squareness and elongation, as well as the pedestal top temperature and collisionality.

High-power ion heating of merging spherical tokamak (ST) plasma has been investigated using TS-3U, TS-4, and UTST at the University of Tokyo for future direct access to burning high-beta ST plasma without using any additional heating. We developed a two-fluid/kinetic interpretation of the promising scaling of ion heating energy that increases with the square of reconnecting magnetic field B _rec ∼ poloidal magnetic field B _p . We find that reconnection heating creates interesting high-beta ST plasmas with hollow currents and broad/hollow T _i profiles. These high-beta ST plasmas often have reversed-shear or absolute minimum-B profiles, depending on their reconnection heating power and q-values.

Here we report the ion heating/transport characteristics of the merging startup scenario in the TS-6 spherical tokamak. In addition to the previously investigated impulsive heating process during magnetic reconnection, here we also focus on a longer time scale response of the ion temperature profile both during and after merging, including the semi-steady plasma confinement phase. During magnetic reconnection, (i) the ion temperature profile forms a poloidally asymmetric profile around the X-point in the initiation phase and (ii) radially asymmetric higher deposition is obtained at the high field side. After merging, (iii) the radially asymmetric double-peak structure is affected by parallel heat conduction and is aligned with field lines, but it does not simply become a flux function on a microsecond time scale—inboard/outboard asymmetry lasts even in the semi-steady confinement phase. (iv) Under the influence of the low-aspect-ratio configuration, there is a two to three times higher toroidal field on the high-field side on the same closed flux surface: characteristic asymmetry of inboard/outboard ion temperature has been found experimentally for the first time.

Establishment of an efficient central solenoid (CS) free tokamak plasma start-up method may lead to an economical fusion reactor. CS-free start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Plasma current of about a quarter of CS-driven discharges has been obtained fully non-inductively using the outer-midplane and top LH launchers. Recently, an outer-off-midplane LH launcher was developed to achieve higher plasma current by optimizing for core absorption and minimal fast electron losses. Using the (outer-)off-midplane launcher, fully non-inductive plasma current start-up up to about 8 kA was achieved. Coupled ray-tracing and Fokker–Planck simulation was performed on equilibria reconstructed with an extended MHD model. It was found that the experimentally observed plasma current was in reasonable agreement with the numerical simulation. The simulation predicted appreciable orbit losses for the off-midplane launcher driven discharge at the present parameters, which was consistent with the experimentally observed x-ray radiation characteristics. The simulation showed that the current density was saturated for the present off-midplane launcher discharges and higher density and higher LH power was necessary to achieve higher plasma current.

This paper introduces an innovative method for reconstructing 2D magnetic flux contours and plasma parameters of dynamically moving tokamak plasmas. While conventional methods like EFIT, based on the Grad–Shafranov equation, are suitable for plasma equilibria with a single magnetic axis, our approach utilizes the MHD equations and shows promise for tokamak plasmas in motion or containing multiple magnetic axes, which may not strictly adhere to plasma equilibria. By utilizing limited edge magnetic probe measurements, our developed model successfully reconstructs the time evolution of two merging plasma toroids in the TS-6 experiment. A comparison with direct 2D magnetic probe measurements in a low β regime reveals a reconstruction error of approximately 3%.

MAST-U is equipped with on-axis and off-axis neutral beam injectors (NBI), and these external sources of super-Alfvénic deuterium fast-ions provide opportunities for studying a wide range of phenomena relevant to the physics of alpha-particles in burning plasmas. The MeV range D-D fusion product ions are also produced but are not confined. Simulations with the ASCOT code show that up to 20% of fast ions produced by NBI can be lost due to charge exchange (CX) with edge neutrals. Dedicated experiments employing low field side (LFS) gas fuelling show a significant drop in the measured neutron fluxes resulting from beam-plasma reactions, providing additional evidence of CX-induced fast-ion losses, similar to the ASCOT findings. Clear evidence of fast-ion redistribution and loss due to sawteeth (ST), fishbones (FB), long-lived modes (LLM), Toroidal Alfvén Eigenmodes (TAE), Edge Localised Modes (ELM) and neoclassical tearing modes (NTM) has been found in measurements with a Neutron Camera (NCU), a scintillator-based Fast-Ion Loss Detector (FILD), a Solid-State Neutral Particle Analyser (SSNPA) and a Fast-Ion Deuterium- α (FIDA) spectrometer. Unprecedented FILD measurements in the range of 1–2 MHz indicate that fast-ion losses can be also induced by the beam ion cyclotron resonance interaction with compressional or global Alfvén eigenmodes (CAEs or GAEs). These results show the wide variety of scenarios and the unique conditions in which fast ions can be studied in MAST-U, under conditions that are relevant for future devices like STEP or ITER.

Axial merging of two torus plasmas is utilized as a center-solenoid free start-up scheme for a high-beta spherical tokamak (ST) plasma, in which magnetic reconnection under a strong guide field plays dominant roles in energy conversion and equilibrium formation. The ion heating source in magnetic reconnection is the plasma outflow with $E \times B$ drift velocity in the downstream region where the reconnected field lines flow out. Since the inductive reconnection electric field is almost parallel to the magnetic field, particularly in the inboard-side downstream region of magnetic reconnection under a strong guide field, a large electrostatic field in the poloidal plane is spontaneously formed to sustain steady plasma outflow motion in the downstream region. In ST plasma merging experiments, the self-generated electrostatic field in the downstream region does not always balance with the inductive electric field to make the total electric field strictly perpendicular to the total magnetic field. The excess electrostatic field will provide an even faster outflow plasma velocity than the magnetic field line motion and a quick reversal of the toroidal plasma current to form convex flux surfaces.

The elimination of the need for an Ohmic heating solenoid may be the most impactful design driver for the realization of economical compact fusion tokamak reactor systems. However, this would require fully non-inductive start-up and current ramp-up from zero plasma current and low electron temperature of sub-keV to the full plasma current of ∼10–15 MA at 20–30 keV electron temperature. To address this challenge, an efficient solenoid-free start-up and ramp-up scenario utilizing a low-field-side-launched extraordinary mode at the fundamental electron cyclotron harmonic frequency (X–I) is proposed, which has more than two orders of magnitude higher electron cyclotron current drive (ECCD) efficiency than the conventional ECCD for the sub-keV start-up regime. A time dependent model was developed to simulate the start-up scenarios. For the Spherical Tokamak Advanced Reactor (STAR) (Menard et al 2023 Next-Step Low-Aspect-Ratio Tokamak Design Studies (IAEA)), it was found that to fully non-inductively ramp-up to 15 MA, it would take about 25 MW of EC power at 170 GHz. Because of the relatively large plasma volume of STAR, radiation losses must be considered. It is important to make sure that high Z impurities are kept sufficiently low during the early current start-up phase where the temperature is sub-keV range. Since the initial current ramp up takes place at a factor of ten lower density compared to the sustained regimes, it is important to transition into a higher bootstrap fraction discharge at lower density to minimize the ECCD power requirement during the densification. For the sustainment phase an array of eight gyrotron launchers with a total of about 60 MW of fundamental O-mode was found to be sufficient to provide the required axis-peaked external current drive. High efficiencies between 19–57 kA MW ^−1 were found with optimal aiming, and these were resilient to small changes in aiming angles and density and temperature profiles.

In 2019 the UK launched the Spherical Tokamak for Energy Production (STEP) programme to design and build a prototype electricity producing nuclear fusion power plant, aiming to start operation around 2040. The plant should lay the foundation for the development of commercial nuclear fusion power plants. The design is based on the spherical tokamak principle, which opens a route to high pressure, steady state, operation. While facilitating steady state operation, the spherical design introduces some specific plasma control challenges: (i) All plasma current during the burn phase should to be generated through non-inductive means, dominated by bootstrap current. This leads to operation at high normalised plasma pressure ${\beta _{\text{N}}}$ with high plasma elongation, which in turn imposes effective active stabilisation of the vertical plasma position. (ii) The tight aspect ratio means very limited space for a central solenoid, imposing that even the current ramp up must be non-inductively generated. (iii) The compact design leads to extreme heat loads on plasma facing components. A double null design has been chosen to spread this load, putting strict demands on the control of the unstable vertical plasma position. (iv) The heat pulses associated with unmitigated ELMs are unlikely to be acceptable imposing ELM free operation or active ELM control. (v) To reduce and spread heat loads, core and divertor radiation and momentum loss has to be controlled, aiming to operate with simultaneously detached upper and lower divertors. (vi) High pressure operation is likely to require active resistive wall mode (RWM) stabilisation. (vii) The conductivity distribution in structures near the plasma must be carefully selected to reduce the growth rates for the vertical instability and the RWM without damping the penetration of the of magnetic fields from active control coils too much. This article describes the initial work carried out to develop a STEP plasma control system.

ST40 is a compact, high-field ( $B_{\mathrm{T0}}\unicode{x2A7D} 2.1\,\,\,\textrm{T}$ ) spherical tokamak (ST) with a mission to expand the physics and technology basis for the ST route to commercial fusion. The ST40 research programme covers confinement and stability; solenoid-free start-up; high-performance operating scenarios; and plasma exhaust. In 2022, ST40 obtained central deuterium ion temperatures of $9.6 \pm 0.4\ \textrm{keV}$ , demonstrating for the first time that pilot plant relevant ion temperatures can be reached in a compact, high-field ST. Analysis of these high-ion temperature plasmas is presented, including a summary of confinement, transport and microstability characteristics, and energetic particle instabilities. Recent scenario development activities have focused on establishing diverted H-mode plasmas across a range of toroidal fields and plasma currents, along with scenarios with high non-inductive current fractions. In future operations, beginning in 2025, a 1 MW dual frequency (104/137 GHz) electron cyclotron (EC) system will be installed to enable the study of EC and electron Bernstein wave plasma start-up and current drive. Predictive modelling of the potential performance of these systems is presented.

The plasma current start-up experiment is conducted through electron cyclotron (EC) heating in the QUEST spherical tokamak. During the EC heating, the application of a toroidal electric field in the opposite direction to the plasma current effectively inhibits the growth of energetic electrons. Observations show rapid increases in plasma current and hard x-ray count immediately following the cancellation of the retarding electric field. When a compact tokamak configuration maintains equilibrium on the high field side, along with the retarding field, it leads to effective bulk electron heating. This heating achieved an electron temperature of T _e ≈ 1 keV at electron density n _e > 1.0 × 10 ^18 m ^−3 . Ray tracing of the EC wave verifies that more power absorption into plasma through a single-pass occurs around the second resonance layer with higher values of electron density and temperature.

STEP is a spherical tokamak prototype power plant that is being designed to demonstrate net electric power. The design phase involves the exploitation of plasma models to optimise fusion performance subject to satisfying various physics and engineering constraints. A modelling workflow, including integrated core plasma modelling, MHD stability analysis, SOL and pedestal modelling, coil set and free boundary equilibrium solvers, and whole plant design, has been developed to specify the design parameters and to develop viable scenarios. The integrated core plasma model JETTO is used to develop individual flat-top operating points that satisfy imposed criteria for fusion power performance within operational constraints. Key plasma parameters such as normalised beta, Greenwald density fraction, auxiliary power and radiated power have been scanned to scope the operational space and to derive a collection of candidate non-inductive flat-top points. The assumed auxiliary heating and current drive is either from electron cyclotron (EC) systems only or a combination of EC and electron Bernstein waves. At present stages of transport modelling, there is a large uncertainty in overall confinement for relevant parameter regimes. For each of the two auxiliary heating and current drive systems scenarios, two candidate flat-top points have been developed based on different confinement assumptions, totalling to four operating points. A lower confinement assumption generally suggests operating points in high-density, high auxiliary power regimes, whereas higher confinement would allow access to a broader parameter regime in density and power while maintaining target fusion power performance.

Generation of Runaway Electrons (REs) during plasma disruptions is of great concern for ITER and future reactors based on the tokamak concept. Unmitigated RE generation in the current STEP (Spherical Tokamak for Energy Production) concept design is modelled using the code DREAM, with hot-tail generation found to be the dominant primary generation mechanism and avalanche multiplication of REs found to be extremely high. Varying assumptions for the prescribed thermal quench (TQ) phase (duration, final electron temperature) as well as the wall time, the plasma-wall distance, and shaping effects, all STEP full-power and full-current unmitigated disruptions generate large RE beams (from 10 MA up to full conversion). RE mitigation is first studied by modelling idealised mixed impurity injections, with ad-hoc particle transport arising from the stochasticity of the magnetic field during the TQ, but no combination of argon and deuterium quantities allows runaways to be avoided while respecting the other constraints of disruption mitigation. Initial concept of STEP disruption mitigation system is then tested with DREAM, assuming two-stage shattered pellet injections (SPI) of pure $\mathrm D_2$ followed by Ar+ $\mathrm D_2$ . Such a scheme is found to reduce the generation of REs by the hot-tail mechanism, but still generates a RE beam of about 13 MA. Options for further optimising the SPI scheme, for mitigating a large RE beam in STEP (benign termination scheme), as well as estimations of required RE losses during the current quench (from a potential passive RE mitigation coil) will also be discussed.

A fusion reactor based on the spherical tokamak is very likely to be completely non-inductive for the majority of the plasma ramp-up and steady-state phases, due to the limitations imposed on the central coil assemblies by the compact design. Efficiency gains from solenoid-driven current cannot be relied upon. It is also critical that an electricity-producing plant maximises the wall-plug efficiency of its heating and current drive (HCD) system, this being one of the largest consumers of recirculating power. It is therefore essential that the HCD system is well-optimised for current drive efficiency in order to meet the goal of net electricity production. The UK’s Spherical Tokamak for Energy Production (STEP) reactor design program has recently taken the decision to use exclusively microwave-based heating and current drive actuators for its reactor concepts. We present the optimisation of an electron cyclotron current drive scheme for a spherical tokamak reactor, based around the STEP concept, arriving at a solution which overcomes the limitations imposed by the spherical tokamak geometry in terms of microwave access and high trapped particle fraction. The solution uses high-field side absorption and a mix of fundamental and 2nd harmonic O mode, with overall power requirements reducing with increasing number of frequencies used. An additional fundamental frequency is also added to further boost the efficiency during non-inductive plasma ramp.

The SMall Aspect Ratio Tokamak (SMART) is currently being commissioned at the University of Seville and will be able to compare the performance of positive and negative triangularity plasmas at low aspect ratio. Predictive simulations have been performed for different machine scenarios and heating schemes using the TRANSP code. The objectives of these simulations are to predict the parameters expected in positive triangularity plasmas, to guide diagnostic development, and to validate transport models. Several reduced turbulence models have been used to predict electron and ion temperatures for the operational phase 2. All models provide similar results from approximately mid-radius to the separatrix but important discrepancies are found in the core region. These positive triangularity results are compared with experiments from a similar size machine like GLOBUS-M2. The multi-mode model (MMM) shows the best agreement. Simulations with different boundary conditions have been performed and no strong differences have been observed between them. The impact of neutral beam injection (NBI) on the predicted profiles has also been addressed. Rotation reduces turbulence levels so higher temperatures are achieved when included in the simulations. Studying the different contributions to the thermal diffusivities, it is observed that electron temperature gradient (ETG) turbulence dominates at the plasma core while micro-tearing modes (MTM) dominate at the edge in the electron channel. In the ion channel, the neoclassical contribution is dominant at the core and at the very edge while the Weiland component, which includes ion temperature gradient mode (ITG), trapped electron mode (TEM), kinetic ballooning mode (KBM), peeling mode (PM) and collisionless and collision dominated magnetohydrodynamic (MHD) modes governs the mid-radius region. For phase 3, two plasmas with different electron densities have been studied. The case with lower density matches well a specific discharge of GLOBUS-M2. The higher density plasma shows high performance with $\beta_N \approx 3.8$ .

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 Hy, 2 = 1–1.2. Simulations with other confinement scalings (Globus-2021, NSTX scalings) give good agreement with plasma parameters for the toroidal field Bt0 ~ 0.5 T. For increasing Bt0, 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 τL/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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

An important part of high-temperature plasma study is the determination of the electron temperature dynamics in the tokamak plasma. At spherical tokamaks, one can use Thomson scattering diagnostics as well as soft X-ray emission diagnostics (SXR). The capabilities of electron temperature measurement by the first diagnostics are limited by the repetition rate of laser pulses and their number in one tokamak discharge. Data of the second diagnostics are continuous in time and are determined by the time resolution of the detectors; however, obtaining the electron temperature using these data encounters a number of difficulties considered in this study. A method of combined processing of results of these diagnostics using machine learning algorithms was developed for overcoming these difficulties and applying the adVoprosy Atomnoi Nauki i Tekhniki, Seriya: Termoyadernyi Sintezages of both diagnostics. Training data include soft X-ray diagnostic data, hard X-ray diagnostic data, and CIII line emissivity diagnostic data. Thomson local scattering measurements were used as labels for supervised machine learning. The developed technique provides significant extension of the possibilities of determining the electron temperature at the Globus-M2 tokamak. [ABSTRACT FROM AUTHOR]

Nowadays, there is much extensive research investigating nuclear fusion reaction with D-³He fuel as one of the most essential advanced fusion fuels. One of the most important quantities in fusion is the reactivity. In this work, with consideration of different temperatures for ion and electron (hot ion mode), we intend to study the effects of two different reactivities (Hively and Bosch-Hale) on D-³He fusion reaction in spherical tokamak. Accordingly, by writing the system of particle and energy balance equations for this reaction in hot ion mode, we will investigate the effects of reactivities on plasma parameters in spherical tokamak. [ABSTRACT FROM AUTHOR]

The engineering part of the GLOBSYS code is presented, and the parameters of the Globus-3 facility, which is a development of the Globus program, are analyzed. The facility is primarily designed to provide a long pulse, a large toroidal magnetic field and strong heating. The concepts of searching for Globus-3 parameters under physical and engineering limitations are described. Obviously that reliable confinement and a large part of noninductive current are necessary to ensure existence of a plasma for a long time. Engineering constraints are involved in the choice of parameters in a more complex way: in some cases, it is overheating of the coils, in other cases, it is the total power supply, or the limit on the flux provided by the ohmic solenoid, or the strength of the constructions. The parameters of the Globus-3 spherical tokamak were preliminarily selected for the cases of a "warm" copper EMS (Electromagnetic system) and the EMS precooled to liquid nitrogen temperature. The exceeding of the duration of the plasma current plateau Δtplateau over the characteristic settling time of the plasma profiles τL/R was chosen as the key condition. At values of the toroidal magnetic field Bt0 = 3 T, the condition Δtplateau > τL/R cannot be attained even for precooled EMS. At Bt0 = 2 T, only options with precooled EMS can be considered acceptable, but the facility dimensions are fairly large. For the field Bt0 = 1.5 T, the options with "warm" EMS correspond to the duration of the plasma current plateau ~3 s (Δtplateau/τL/R ~ 1–1.5). In the case of precooled EMS, the duration of the plateau can increase to 12–13 s (Δtplateau/τL/R ~ 5). In the latter case, as a basis for further development of the Globus-3 facility, options with the following geometric dimensions are reasonable: R0 ~ 0.6–0.7 m, a ~ 0.35–0.4 m, А ≤ 1.7–1.8, k95 ~ 1.7–1.8. The minimum allowable value of the plasma current under the condition of effective absorption of the input power of neutral injection has been calculated. In the Globus-3 facility, Ip ≈ 0.8 MA was chosen as the base value. [ABSTRACT FROM AUTHOR]