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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]

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]

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.

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 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.

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 $\beta_N \approx 3.8$ .

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]

In most present-day tokamaks, the majority of the heating power comes from sources such as neutral-beam injection (NBI) and other types of auxiliary heating which allow for the transfer of energy to the plasma by a small population of externally introduced fast particles. The behavior of the fast ions is important for the overall plasma dynamics, and understanding their influence is vital for the success of any future magnetic confinement devices. In the TUMAN-3M tokamak, it has been noted that the loss of fast particles during NBI can lead to dramatic changes in the rotation velocity profiles, as they are responsible for the negative radial electric field on the periphery. [ABSTRACT FROM AUTHOR]

Featured Application: The "Lite Neutral Beam" model (LNB) efficiently combines an injected beam detailed description with data processing pipeline for following up particle statistics. This approach allows one to study the issues of beam-plasma interaction in fusion devices and to perform on-the-fly optimization of beam-driven toroidal systems. The results are important for a steady-state current drive and fusion controllability in low aspect ratio and spherical tokamaks; they are also applicable to a conventional tokamak design. FNS-ST is a fusion neutron source project based on a spherical tokamak (R/a = 0.5 m/0.3 m) with a steady-state neutron generation of ~1018 n/s. Neutral beam injection (NBI) is supposed to maintain steady-state operation, non-inductive current drive and neutron production in FNS-ST plasma. In a low aspect ratio device, the toroidal magnetic field shape is not optimal for fast ions confinement in plasma, and the toroidal effects are more pronounced compared to the conventional tokamak design (with R/a > 2.5). The neutral beam production and the tokamak plasma response to NBI were efficiently modeled by a specialized beam-plasma software package BTR-BTOR, which allowed fast optimization of the neutral beam transport and evolution within the injector unit, as well as the parametric study of NBI induced effects in plasma. The "Lite neutral beam model" (LNB) implements a statistical beam description in 6-dimensional phase space (106–1010 particles), and the beam particle conversions are organized as a data flow pipeline. This parametric study of FNS-ST tokamak is focused on the beam-plasma coupling issue. The main result of the study is a method to achieve steady-state current drive and fusion controllability in beam-driven toroidal plasmas. LNB methods can be also applied to NBI design for conventional tokamaks. [ABSTRACT FROM AUTHOR]

The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in the summer of 2015. NSTX-U first plasma was subsequently achieved, diagnostic and control systems have been commissioned, the H-mode accessed, magnetic error fields identified and mitigated, and the first physics research campaign carried out. During ten run weeks of operation, NSTX-U surpassed NSTX record pulse-durations and toroidal fields (TF), and high-performance ∼1 MA H-mode plasmas comparable to the best of NSTX have been sustained near and slightly above the n = 1 no-wall stability limit and with H-mode confinement multiplier H98y,2 above 1. Transport and turbulence studies in L-mode plasmas have identified the coexistence of at least two ion-gyro-scale turbulent micro-instabilities near the same radial location but propagating in opposite (i.e. ion and electron diamagnetic) directions. These modes have the characteristics of ion-temperature gradient and micro-tearing modes, respectively, and the role of these modes in contributing to thermal transport is under active investigation. The new second more tangential neutral beam injection was observed to significantly modify the stability of two types of Alfven eigenmodes. Improvements in offline disruption forecasting were made in the areas of identification of rotating MHD modes and other macroscopic instabilities using the disruption event characterization and forecasting code. Lastly, the materials analysis and particle probe was utilized on NSTX-U for the first time and enabled assessments of the correlation between boronized wall conditions and plasma performance. These and other highlights from the first run campaign of NSTX-U are described.

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.

A new outer-off-midplane lower hybrid (LH) wave launcher was developed for improved core absorption in the TST-2 spherical tokamak. In the previously developed outer-midplane and top launch scenarios, plasma current was driven only very near the plasma edge ( $r/a \gt 0.7$ ), resulting in limited driven current density. Strong thick-target x-ray radiation has also been observed for the top launch scenario, which indicated LH wave driven fast electron losses. The outer-off-midplane launch scenario was designed to drive current at the core ( $r/a \sim 0.5$ ) at low phase velocity to suppress production of unnecessarily high energy electrons. In the non-inductive plasma start-up experiment with the new outer-off-midplane launch scenario, improved core electron heating was confirmed by the Thomson scattering diagnostic that showed the electron temperature was about twice as high as the previous two scenarios. Reduction of the LH wave driven fast electron losses was confirmed by the measured x-ray radiation intensity that was substantially lower than the top launch scenario. It was also easier to maintain the optimum level of the electron density compared to the outer-midplane launch scenario that indicated the undesired LH wave interactions with the scrape-off-layer plasma may have been reduced due to strong core absorption.

Results from the successful solenoid-free plasma startup using the method of transient coaxial helicity injection (transient CHI) in the QUEST spherical tokamak (ST) are reported. Unlike previous applications of CHI on HIT-II and on NSTX which required two toroidal insulating breaks to the vacuum vessel, QUEST uses a first of its kind, floating single biased electrode configuration, which does not use such a vacuum break. Instead, the CHI electrode is simply insulated from the outer lower divertor plate support structure. This configuration is much more suitable for implementation in a fusion reactor than the previous configurations. Transient CHI generated toroidal currents of 135 kA were obtained. The toroidal current during the formation of a closed flux configuration was over 50 kA. These results bode well for the application of transient CHI in a new generation of compact high-field STs and tokamaks in which the space for the central solenoid is very restricted.

Ion temperatures of over 100 million degrees Kelvin (8.6 keV) have been produced in the ST40 compact high-field spherical tokamak (ST). Ion temperatures in excess of 5 keV have not previously been reached in any ST and have only been obtained in much larger devices with substantially more plasma heating power. The corresponding fusion triple product is calculated to be ${n_{i0}}{T_{i0}}{\tau _E} \approx 6 \pm 2 \times {10^{18}}{{\text{m}}^{ - 3}}{\text{keVs}}$ . These results demonstrate for the first time that ion temperatures relevant for commercial magnetic confinement fusion can be obtained in a compact high-field ST and bode well for fusion power plants based on the high-field ST.

In this work, the MARS-F/K codes (Liu et al 2000 Phys. Plasmas 7 3681; Liu et al 2008 Phys. Plasmas 15 112503) are utilized to model the passive and active control of the n = 1 ( n is the toroidal mode number) resistive wall mode (RWM) in a spherical tokamak (aspect ratio A = 1.66). It is found that passive stabilization of the RWM gives a relatively small increase in normalized beta above the no-wall limit, relying on toroidal plasma flow and drift kinetic resonance damping from both thermal and energetic particles. Results of active control show that with the flux-to-voltage control scheme, which is the basic choice, a proportional controller alone does not yield complete stabilization of the mode. Adding a modest derivative action, and assuming an ideal situation without any noise in the closed-loop, the RWM can be fully stabilized with the axial plasma flow at 5% of the Alfvén speed. In the presence of sensor signal noise, success rates exceeding 90% are achieved, and generally increase with the proportional feedback gain. On the other hand, the required control coil voltage also increases with feedback gain and with the sensor signal noise.

In EXL-50 plasma currents over 100 kA are non-inductively generated and maintained solely by electron cyclotron heating (ECH) power with an efficiency of ∼1 A W ^−1 . These currents are carried by energetic electrons (EEs) in the energy range from several tens of keV up to several hundreds of keV which also account for almost all pressure in plasma. This EE component can be viewed as a large number collection of various periodic orbits of energetic particles. Based on this picture we have developed a method for particle orbit description of the EE component in a typical plasma at $I_{\textrm P}$ = 121 kA as analysis target. We use a fluid description as a bridge to describe successfully the EE component as a collection of various passing and trapped orbits in the approximation of monochromatic particle energy with good matching to the flux loop signals. The description has revealed characteristics of passing and trapped particles. Passing particles carry almost all toroidal current, while they account for only 20 $\%$ of total particle number of the EE component. While net current carried by trapped particles is a very small fraction, they account for a major fraction in number and carry a large positive current outside the last closed flux surface (LCFS) and a large negative current inside. As a result, trapped particles redistribute the current from inside of the LCFS to outside both radially and vertically, generating a large vertically elongated cross section in current as well as number density profiles. There is a ridge-like structure along the LCFS in the current density profile, with no such structure in the number density profile. The results suggest that forward passing particles are more advantageous in confinement than backward passing particles. This advantage increases with particle energy and contributes to the current generation observed in EXL-50 experiments.

Doppler backscattering (DBS) is a microwave diagnostics method typically used to study the plasma rotation velocity. Apart from conventional techniques, more advanced forms of DBS implementation were suggested on Globus-M. More specifically the study of a variety of oscillating processes was performed using DBS. In this review we present a detailed description of all of the methods and techniques employed in Globus-M alongside results obtained using DBS in all the years up until the shutdown of the tokamak. These include research similar to that done on other devices into the properties of such phenomena like geodesic acoustic modes or limit cycle oscillations, along with innovative works regarding the detection and investigation of Alfven eigenmodes and filaments that were the first of their kind and that provided important and novel results. Apart from that, the specific aspects of DBS application on a spherical tokamak are discussed. An in-depth look into the gradual change and improvement of the DBS diagnostics on Globus-M is also presented in this paper. [ABSTRACT FROM AUTHOR]