Spherical Tokamak Research Articles

Measurement of spherical tokamak plasma compression in the PCS-16 magnetized target fusion experiment

Abstract A sequence of magnetized target fusion (MTF) devices built by General Fusion have 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 δBpol/Bpol < 20% to a high radial compression factor (CR > 8) with significant poloidal flux conservation (77% up to CR = 1.7, and ≈30% up to CR = 8.65) and a total compression time of 167 μs. 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 thermal energy was in the range of 350 ± 25 J. Plasma equilibrium was a low-β state with βtor ≈ 4% and βpol ≈ 15%. Ingress of impurities from the lithium coated aluminum wall was not a dominant effect. Neutron yield from D-D fusion increased significantly during compression. Thermodynamics during the early phase of compression (CR < 1.7) were consistent with increasing Ohmic heating of the electrons due to a geometric rise in the current density at near-constant resistivity, and with increasing ion cooling that approximately matched ion compressional heating power. Ion cooling by electrons was significant because the electrons were cooler than the ions (Te = 200 eV, Ti = 600 eV). Magnetohydrodynamic simulation was used to model the emergence of instabilities that increased 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.

Fully-relativistic electron Bernstein wave current drive simulations in the STEP spherical tokamak

Fully-relativistic electron Bernstein wave current drive simulations in the STEP spherical tokamak

Plasma self-driven current in tokamaks with magnetic islands

Abstract Magnetic island perturbations may cause a reduction in plasma self-driven current that is needed for tokamak operation.
 A novel effect on tokamak self-driven current revealed by global gyrokinetic simulations is due to magnetic-island-induced
3D electric potential structures, which have the same dominant mode numbers as that of the magnetic island,
whereas centered at both the inner and outer edge of the island. The non-resonant potential islands
are shown to drive a current through an efficient nonlinear parallel acceleration of electrons.
In large aspect ratio (large-A) tokamak devices, this new effect can result in a significant
global reduction of the electron bootstrap current when the island size is sufficiently large,
in addition to the local current loss across the island region due to the pressure profile flattening.
It is shown that there exists a critical magnetic island width for large-A tokamaks beyond which
the electron bootstrap current loss is global and increases rapidly with the island size.
As such, this process may introduce a size limit for tolerable magnetic islands in large-A tokamak
devices in the context of steady state operation. On the other hand, the current loss caused by magnetic islands in
low-A tokamaks such as spherical tokamak (ST) NSTX/U is minor.
The reduction of the axisymmetric current by magnetic islands scales with the square of island width.
However, the loss of the current is mainly local to the island region, and the pace of current
loss as the island size increases is substantially slower compared to large-A tokamaks.
In particular, the bootstrap current reduction in STs is even smaller in the reactor-relevant
high-$\beta_p$ regime where neoclassical tearing modes are more likely to develop.

Thermal hydraulic and material analysis of upgraded flat-type Graphite divertor mock-up for Pakistan Spherical Tokamak (PST)

The design of a divertor is critically dependent on managing power deposition, erosion effects, and plasma configuration. An upgraded double-null divertor configuration has been developed for the Pakistan Spherical Tokamak (PST), featuring graphite targets that are actively cooled and designed to withstand a peak heat flux of 0.3 MW/m² at a pressure of 0.1 MPa. This paper presents a comprehensive thermal-hydraulic and material analysis for the upgraded flat-type divertor mock-up system, covering aspects such as material surface heat load, peak temperature rise (∆T °C) on the mock-up, and surface temperature increase in the cooling channel. The analysis includes total deformation (mm) and equivalent strain for the inner vertical target (IVT), outer vertical target (OVT), and dome structure, along with material comparison. The recommended SST K-ω turbulence model is utilized in the pressure-based transient analysis, with an inlet velocity of 1.5 m s-1 and an inlet temperature of 16.8 °C. A comparative study of the material and thermal-hydraulic analyses was performed using CFD and RELAP5. The findings reveal that graphite is more suitable than tungsten for the PST's upgraded divertor system, demonstrating its effectiveness as the preferred surface material to address heat enhancement challenges in the PST.

Assessment of a space and energy resolved diagnostic based on GEM technology on MAST-U

Abstract A Gas Electron Multiplier (GEM)-based detector was utilized for the first
time on a spherical tokamak, MAST-U, during the 2023 campaign to investigate Soft
X-Ray (SXR) radiation (1-20 keV) emitted from the plasma. GEM detectors, chosen
for their resilience to harsh fusion environments and their ability to provide energy-
resolved SXR emission images with sub-millisecond time resolution, are a relatively
new diagnostic compared to standard semiconductor diodes. In this study, the GEM
detector features a pinhole geometry outside the vacuum chamber and observes the
plasma through a beryllium window. Filled with an ArCO2 mixture, the detector
consists of an Aluminized Mylar cathode, three Aluminum-coated GEM foils, and an
anode made of a 16x16 matrix of 6mm2 pads for 2D readout. It employs custom
GEMINI ASICs (Application Specific Integrated Circuits) for signal readout, enabling
photon-counting techniques with Time over Threshold (ToT) analysis on each detector
channel, reaching a total maximum rate of 256 MHz.
Preliminary findings from the 2023 campaign demonstrate the effectiveness of the
GEM detector in complementing existing SXR camera data in terms of spatial and
temporal resolution. Case studies include the identification of Magnetohydrodynamic
(MHD) instabilities, such as the Snake instability, and the exploitation of energy
capabilities for plasma event energy characterization, as shown in the case of Internal
Reconnection Events. Additionally, the GEM detector allows for estimation of the
Electron Temperature of Maxwellian plasmas. These results underscore the potential
of the GEM-based diagnostic system in advancing tokamak research for comprehensive
plasma characterization and control.

Effects of Injected Current Streams on MHD Equilibrium Reconstruction of Local Helicity Injection Plasmas in a Spherical Tokamak

Open field line currents are intrinsic to DC helicity injection plasma startup and pose a challenge for inferring the plasma equilibrium with standard reconstruction analysis. Local helicity injection (LHI) is a type of DC helicity injection which uses small, modular current sources to drive force-free current along helical field lines to produce tokamak plasmas. MHD modeling and magnetic measurements during LHI indicate the injected current streams remain coherent as helical structures on the outboard edge of a core toroidal plasma that is tokamak-like in a toroidally averaged sense. To extract core plasma equilibrium properties, external magnetic diagnostics corrected for contributions from the injected current streams are fitted by a standard Grad-Shafranov equilibrium code. An iterative approach for estimating and subtracting the stream contributions from the diagnostic signals is described and applied to a model equilibrium database to reduce systematic errors introduced by the streams. Convergence is usually attained with 2 to 4 iterations, with derived equilibrium parameters matching the prescribed axisymmetric core values to within estimated experimental uncertainties. Accurate recovery of core parameters occurs when the ratio of the net toroidal windup current from the streams to the core plasma current is less than 0.2, which is typically satisfied in most experiments.

The optimisation of the STEP electron cyclotron current drive concept

Abstract 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.

Heating characteristic of electron Bernstein wave using high-field side X-mode injection in QUEST

Abstract A comparative investigation was conducted to assess the impact of high-field side (HFS) extraordinary (X) mode (HFS X) injection and low-field side (LFS) ordinary (O) mode (LFS O) injection of 8.2 GHz radio frequency (RF) power in the Q-shu University experimental steady-state spherical tokamak. In the case of the HFS X injection, the ratio of electron plasma frequency and RF frequency f pe / f RF was 1.3, indicating an overdense plasma, whereas in the LFS O injection, this ratio was 0.9. Distinctive features emerged in the electron temperature and density profiles between the two injection scenarios. In the HFS X injection, the profiles peaked between the fundamental electron cyclotron resonance (ECR) layer and the upper hybrid resonance layer. Conversely, in the LFS O injection, the peaks were situated near the 1st ECR layer. This implies effective excitation of the electron Bernstein wave (EBW) in the case of the HFS X injection, with the wave power being absorbed through the doppler-shifted ECR of the EBW. Density and temperature oscillations were observed only in the start-up phase of the HFS X injection, which may correlate with the presence of the left-hand cutoff of the X-mode. These findings will be contributed to understand the distinctive characteristics associated with plasma heating through EBW excitation in the HFS X injection.

Poloidal magnetic field reconstruction by laser-driven ion-beam trace probe in spherical tokamak

The poloidal magnetic field ( plays a critical role in plasma equilibrium, confinement and transport of magnetic confinement devices. Multiple diagnostic methods are needed to complement each other to obtain a more accurate profile. Recently, the laser-driven ion-beam trace probe (LITP) has been proposed as a promising tool for diagnosing and radial electric field ( ) profiles in tokamaks [Yang X Y et al 2014 Rev. Sci. Instrum. 85 11E429]. The spherical tokamak (ST) is a promising compact device with high plasma beta and naturally large elongation. However, when applying LITP to diagnosing in STs, the larger invalidates the linear reconstruction relationship for conventional tokamaks, necessitating the development of a nonlinear reconstruction principle tailored to STs. This novel approach employs an iterative reconstruction method based on Newton’s method to solve the nonlinear equation. Subsequently, a simulation model to reconstruct the profile of STs is developed and the experimental setup of LITP is designed for EXL-50, a middle-sized ST. Simulation results of the reconstruction show that the relative errors of reconstruction are mostly below 5%. Moreover, even with 5 mm measurement error on beam traces or 1 cm flux surface shape error, the average relative error of reconstruction remains below 15%, initially demonstrating the robustness of LITP in diagnosing profiles in STs.

Inferring the scrape-off layer heat flux width in a divertor with a low degree of axisymmetry

Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MWm−2. 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.

Experimental observations of bifurcated power decay lengths in the near Scrape-Off Layer of ST40 High Field Spherical Tokamak

The scrape-off layer parallel heat flux decay lengths measured at ST40, a high field, low aspect ratio spherical tokamak, have been observed to bifurcate into two groups. The wide group follows established H-mode scalings (ranging between 2 to 8 mm) while the narrow group falls up to 10 times below these scalings (between 0.2 and 0.8 mm), being comparable to the ion total Larmor radius rather than the ion poloidal Larmor radius. The heat flux profiles of the latter group can only be described by a multi-exponential function, rather than the single exponential function convoluted with a Gaussian. The onset of the narrow scrape-off layer width is observed to be associated with suppressed magnetic fluctuations, suggesting reduced electromagnetic turbulence levels in the SOL.

Soft x-ray tomography on the high field spherical tokamak ST40.

As part of its roadmap to developing commercial fusion plants, Tokamak Energy Ltd. operates the high field spherical tokamak ST40. Studies on this device will help to expand the high field spherical tokamak physics basis by characterizing confinement and the fusion triple product. In support of this, bolometers and broadband and x-ray sensitive diodes can provide information on key energy loss mechanisms of the plasma. These mechanisms include core magnetohydrodynamic activity that deteriorates confinement, such as sawtooth crashes that can be used to characterize relaxations in the q-profile. In addition, combinations of these diagnostics can be used to infer the total radiated power losses and plasma composition. Here, we present results from a new, midplane, tangential, Be-filtered diode with 16 channels spanning the radial extent of the plasma. The system is shown to resolve magnetohydrodynamic instabilities (up to 100kHz) and be able to provide radiation profiles through tomography. The tomographic inversion routine is compared against other diagnostics on ST40 and provides emissivity measurements across a variety of operating scenarios. Finally, we look ahead to implementing multiple soft x-ray cameras on ST40 and the improvements this will have on the diagnostic capabilities.

Safety Analysis of Metallic Tokamak-I Against Electromagnetic and Structural Loads

During tokamak operation, the structural integrity of the vacuum vessel (VV) of Metallic Tokamak-I (MT-I), a small spherical tokamak, was evaluated. This evaluation involved simulating real experimental data of electromagnetic (EM) and structural loads using the ANSYS platform. Internal heat generation, induced currents, and inertial and pressure loads in the VV were analyzed to determine their effects on the VV. This analysis was conducted on a 180-deg sector model over a 10-ms-event period. To create multiple checkpoint events, the plasma current was assumed to be formed at variable positions of the VV, hence inducing variable current for each event. The events are divided into four cases based on the radial and vertical displacements of plasma. The response of the VV structure was calculated using coupling of EM and structural modules of ANSYS. It is observed from the numerical results that the maximum stress on the VV is in a safe range and that the temperature rise on the vessel can be reduced by natural convection only if the event is ended in 10 ms. A prolonged event can result in permanent deformation in the VV structure. A disruption event on the limiter region is also studied.

Accuracy of kinetic equilibrium reconstruction of NSTX and NSTX-U plasmas and its impact on the transport and stability analysis

Abstract An accurate magnetohydrodynamic (MHD) equilibrium reconstruction is an essential starting point for stability and transport plasma analysis. This work describes an approach for obtaining kinetic equilibrium reconstructions using the OMFIT framework, which has been applied for the first time to spherical tokamak data from NSTX and NSTX-U. The EFIT equilibrium solver is integrated with experimental data analysis procedures and subsequent TRANSP transport simulations to enhance the accuracy of the reconstruction, in particular at the edge region, by adding constraints on the total pressure and current density profiles, based on the transport code solution. The accuracy of the equilibrium depends on the uncertainty and number of constraints, as well as the choice of basis functions to represent the pressure and current density profiles. An improved fidelity of the equilibrium reconstruction is demonstrated by reducing the variability of the magnetic axis and boundary locations from several centimeters for reconstructions based on magnetic and experimental pressure constraints to only several millimeters for kinetic reconstructions based on transport code constraints, when different representations for basis functions were tested. Variability of the safety factor on-axis was reduced ten times in the same sensitivity study. The accuracy of the equilibrium reconstruction and subsequent mapping of experimental kinetic profile data have a significant impact on TGLF and linear CGYRO turbulence simulations, which predict different spectra of unstable modes and turbulent fluxes for cases with different numbers of constraints in the equilibrium reconstruction. Conversely, the stability analysis performed with the GATO code shows plasmas stable to n = 1 MHD modes in both equilibria using magnetic and experimental pressure constraints as well as the transport code constrained equilibrium. However, a scan of parameters away from these conditions shows considerable deviation in the threshold of unstable modes between these reconstructions. Therefore, for reliable plasma analysis and use in turbulence and stability calculations, a high fidelity equilibrium reconstruction with accurate kinetic constraints based on transport code solutions is necessary.

Evaluating the Viability and Optimization of Plasma Pilot Plants

Nuclear fusion stands as a promising avenue for achieving a sustainable and abundant energy source. Central to this endeavor is the development of effective pilot plants capable of demonstrating the feasibility of fusion power on a commercial scale. This paper provides a comprehensive evaluation of three leading magnetic confinement fusion configurations: the Advanced Tokamak (AT), Spherical Tokamak (ST), and Compact Stellarator (CS).The Tokamak, characterized by its toroidal shape and strong magnetic fields, has been the most researched and developed fusion device. The analysis focuses on the challenges related to maintaining stability and minimizing disruptions. Furthermore, the optimization strategies involving advanced materials, superconducting magnets, and innovative plasma control techniques are discussed to enhance the Tokamak's viability. In contrast, the Spherical Tokamak, a variant with a more compact and spherical design, promises improved current advancements including the handling of higher heat loads and magnetic field configurations. The Stellarator, with its complex, twisted magnetic field structure, eliminates the need for continuous external current drive, addressing some intrinsic issues of the Tokamak. By comparing these configurations, the paper identifies the relative strengths and weaknesses of each approach in terms of confinement efficiency, operational stability, and engineering feasibility. The evaluation is supported by recent experimental data, simulation results, and technological advancements. Finally, the paper proposes a roadmap for the future development of fusion pilot plants, highlighting the need for an integrated approach that leverages the strengths of each configuration while addressing their individual challenges. The synthesis of this evaluation underscores the importance of continued research, cross-collaboration, and investment in advanced technologies to realize the goal of practical and economically viable fusion energy. This comparative analysis aims to provide a strategic framework for policymakers, researchers, and engineers in the fusion community, fostering informed decisions and prioritizing research efforts towards the most promising fusion energy configurations.

The formation of an radial edge electric field due to finite ion orbit width effects is the possible root cause of the H-mode edge

Abstract Full orbit-following simulations of thermal ions show that finite ion-orbit width effects create charge separation near the last closed flux surface (LCFS) which generates a localized radial electric field. Experimentally, edge electric fields are observed in H-mode plasmas and they are necessary for the edge turbulence suppression via the E × B flow shear mechanism. Confined trapped (and to a lesser extent co-passing) ions near the plasma edge form a positive charge distribution outside the LCFS, while thermal electrons are tied more tightly to field lines owing to their small mass and are poorly confined outside the LCFS, hence charge neutrality is violated outside the LCFS. A large number of reported observations from spherical and conventional tokamaks support the results from the simulations although the simulations were not performed fully self consistently. The results suggest ways to lower the H-mode power threshold and optimize the H-mode plasma edge.

Modeling of ion cyclotron resonance frequency heating of proton-boron plasmas in EHL-2 spherical tokamak

Ion cyclotron resonance heating (ICRH) stands out as a widely utilized and cost-effective auxiliary method for plasma heating, bearing significant importance in achieving high-performance discharges in p-11B plasmas. In light of the specific context of p-11B plasma in the EHL-2 device, we conducted a comprehensive scan of the fundamental physical parameters of the antenna using the full-wave simulation program TORIC. Our preliminary result indicated that for p-11B plasma, optimal ion heating parameters include a frequency of 40 MHz, with a high toroidal mode number like to heat the majority H ions. In addition, we discussed the impact of concentration of minority ion species on ion cyclotron resonance heating when 11B serves as the heavy minority species. The significant difference in charge-to-mass ratio between boron and hydrogen ions results in a considerable distance between the hybrid resonance layer and the tow inverted cyclotron resonance layer, necessitating a quite low boron ion concentration to achieve effective minority heating. We also considered another method of direct heating of hydrogen ions in the presence of boron ion minority. It is found that at appropriate boron ion concentrations ( ), the position of the hybrid resonance layer approaches that of the hydrogen ion cyclotron resonance layer, thereby altering the polarization at this position and significantly enhancing hydrogen ion fundamental absorption.

Integrated modeling of anisotropic neutron yields of classical and spherical tokamaks

Estimations of counting rates of neutron spectrometers in experiments on controlled fusion with magnetic confinement, as well as calculations of energy resolved flux densities of fusion neutrons from plasma to the walls of a reactor require spatial integration of the local, usually anisotropic function of the neutron source. The integrated modeling consists of three main stages. First, sources of fast particles in beam- or wave-heated plasma are calculated. The next stage deals with spatial, energetic, and angular velocity distributions of plasma ions. Finally, double differential rate coefficients of nuclear fusion reactions are computed. This article describes calculations of spatial distributions of nuclear fusion reaction rates in classical and spherical tokamaks and the anisotropy of the neutron yield and spectra. The results are based on analytical formulas for energetic and angular distributions of the local source of fusion products in plasma. Examples of energetic spectral densities of neutron fluxes on first walls are presented, as well as energy resolved counting rates of collimated neutron spectrometers for perpendicular and tangential lines of sight.

Design of a Thomson scattering diagnostic for the SMall Aspect Ratio Tokamak (SMART).

We describe the design of a Thomson scattering (TS) diagnostic to be used on the SMall Aspect Ratio Tokamak (SMART). SMART is a spherical tokamak being commissioned in Spain that aims to explore positive triangularity and negative triangularity plasma scenarios at a low aspect ratio. The SMART TS diagnostic is designed to operate at high spatial resolution, 6mm scattering length in the low-field side and 9mm in the high-field side regions, and a wide dynamic range, electron temperature from 1eV to 1keV and density from 5×1018m-3 to 1×1020m-3, to resolve large gradients formed at the plasma edge and in the scrape-off layer (SOL) under different triangularities and low aspect ratios. A 2J @1064 nm laser will be used that is capable of operating in the burst mode at 1, 2, and 4kHz to investigate fast phenomena and at 30Hz to study 1s (or more) long discharges. The scattered light will be collected over an angular range of 60° - 120° from 28 spatial points in the midplane covering the entire plasma width and the outer midplane SOL. Each scattering signal will be spectrally resolved on five wavelength channels of a polychromator to obtain the electron temperature measurement. We will also present a method to monitor in situ laser alignment in the core during calibrations and plasma operations.

Toroidal Alfvén mode instability driven by plasma current in low-density Ohmic plasmas of the spherical tori

Due to intrinsically low magnetic fields, at low density the drift speed of the current-carrying electrons in spherical tokamaks can exceed fraction of the Alfvén speed sufficient for the excitation of the Alfvén gap modes. A particular case of the toroidal mode, observed during minor disruptions in Ohmic shots on SUNIST [Liu et al., Phys. Plasmas 23, 120706 (2016)], is considered in the present communication. Due to the negligible effect of the electron pressure gradient, the growth rate scales linearly with the drift speed, with slope inversely proportional to electron thermal velocity. In the absence of continuum damping, the threshold value of the drift speed for TAE excitation is independent of electron temperature.