The neoclassical tearing mode (NTM) stability metric-minimum marginally stable island width w(m)(& lowast;)-was compared across 14651 inductive high-field tokamak pilot plant equilibria. Larger devices with reduced elongation and/or increased minor radius demonstrated an order-of-magnitude increase in w(m)(& lowast;), primarily due to a reduction in bootstrap drive. This work is part of an ongoing effort to ensure passive NTM-stability in the ARC tokamak, in which the technology to achieve active tearing-suppression with localised electron cyclotron current drive does not yet exist. The equilibrium scenarios in the database were Monte Carlo generated and normalised to the same >400MW fusion power, minimum pressure scenario at a range of plasma currents, before tearing analysis using the modified Rutherford equation was applied for all resonant poloidal and toroidal m, n modes up to n = 4. Single-helicity toroidal Delta ' calculations in resistive DCON set the minimum marginally stable island width, and a simple modal scaling proportional to -m(2)n(-1) was identified for high-m Delta ' values. The dominant correlates of w(m)(& lowast;) and Delta ' across the database were analysed using interpretable machine learning techniques.

Utilizing the TokaMaker code, this manuscript presents a scheme for determining the optimal locations of poloidal field coils in tokamak configurations, aimed at reducing the currents required to reproduce target equilibrium in future devices. The definition of the coil position space is described, as well as the cost function is to be minimized, which incorporates the constraints of the problem. The MANTA reactor design, featuring strong negative triangularity shaping, is considered as a test example, where preliminary results are obtained with a reduced number of degrees of freedom in order to highlight the procedure for finding local minima. Application of the optimization procedure, guided by a study of the electric current regularization parameter, results in new suggestions for the placement of each poloidal field (PF) coil. This leads to a reduction in the total current flowing through the PF coils by 37% while accurately reproducing the target plasma boundary compared to a baseline case.

In liquid breeding blanket concepts for fusion reactors, such as liquid metals and molten salts, the circulating breeder may interact directly with plasma dynamics through electromagnetic effects. As blanket designs mature and pilot-plant studies progress, simple models are needed to capture this interaction with sufficient fidelity. To address this need, we present a new axisymmetric simulation framework for coupled plasma–blanket systems. The blanket regions are treated with the time-dependent visco-resistive incompressible magneto-hydrodynamic (MHD) equations. For the plasma region, the ideal MHD equilibrium equation is solved at each timestep allowing the plasma to be simulated on the liquid blanket flow timescale. Flexible geometry definitions and support for anisotropic electrical conductivity enable exploration of a wide range of candidate blanket configurations, including electrically isolated toroidal blanket segments. We demonstrate the framework’s capability by simulating a vertical displacement event (VDE) within a toroidally conducting Pb-Li blanket, showing how the fluid treatment of the blanket alters predicted plasma evolution timescales and structural forces compared to static models. Simulations of both VDEs and midplane current quenches in toroidally conducting blankets produce peak liquid metal velocities greater than 1 m/s. These transient flows are on the same order as the steady-state background flows required for self-cooled concepts, and an order of magnitude greater than those in externally cooled designs. In toroidally insulating configurations and molten salt concepts, significantly less flow is induced, with peak fluid velocities of several cm/s. These findings underscore the importance of coupling blanket and plasma dynamics to accurately assess the viability of conducting liquid blanket designs and set engineering requirements.

ARC is designed to produce ${400}\, extrm {MW}$ of net electricity and prove the commercial feasibility of a fusion power plant. In order to achieve this goal ARC has to operate with optimal core performance in a stationary scenario that minimises wear on the first wall and divertor. This requires avoiding or mitigating magnetohydrodynamic (MHD) instabilities which have the potential to not only degrade the plasma core but also lead to deleterious transient heat loads on plasma facing components. Therefore, this work aims at characterising the MHD stability of the high performance ARC scenario and inform the design of error field correction coils. Firstly, simulations of vertical displacement events show that an in-vessel coil is not needed and instead the poloidal shaping coils can be used to control vertical stability. These simulations also inform the demands on the corresponding coil power supplies. Stability analysis of the ideal kink mode with or without a conducting wall and kinetic effects suggests that the ARC baseline scenario operates deeply in the stable region. Using RDCON, tearing modes at the $m/n=2/1$ and $3/2$ surfaces (with poloidal mode number $m$ , and toroidal mode number $n$ ) are shown to be linearly stable, and including thermal transport effects in the rational surfaces lead to further stabilisation. However, other transient plasma instabilities can seed neoclassical tearing modes (NTMs). The marginally stable width of NTMs in ARC strongly depends on the internal inductance and can fall below ${0.1}{\,\,\%}$ of the normalised poloidal flux. Furthermore, an empirical cross-machine model of the $n=1$ error field leading to a disruption predicts a critical error field larger than SPARC but smaller than ITER. Three-dimensional coils can be designed with the Generalised Purturbed Equilbium Code based on a simple model that calculates the maximum correctable error field that is limited by the neoclassical toroidal viscosity torque. Broad scans of different coil geometries identify a set of 2 rows of off-midplane coils to be a suitable solution. It is also determined that such a set of three-dimensional coils is capable of correcting $n=2$ error fields to some degree and creating strong enough $n=2$ or $n=3$ edge resonant perturbation fields for the suppression of edge-localised modes at reasonable coil currents. The final design of the first ARC will be further informed by results from SPARC.

A technique, volumetric power optimization, is presented for enhancing the power output of magnetic confinement fusion devices. Applied to a tokamak, this approach involves shifting the burning plasma region to a larger plasma volume while introducing minimal perturbations to the plasma boundary shape. This edge perturbation—squareness—is analogous to pinching and stretching the edge boundary. Stability calculations confirm that this edge alteration is compatible with maintaining plasma stability. This optimization method for optimizing fusion power output could improve the performance of magnetic confinement fusion power plants.

Advances in vacuum, surface, and lithium conditioning techniques throughout five years of continuous operations in LTX-beta have produced mirror-like liquid lithium surfaces and demonstrated the feasibility of highperformance tokamak discharges fully surrounded by liquid metal without significant operational problems. Improvements in conditioning techniques and procedures, including many weeks of baking and accumulation of 70 g of Li, led to reduced residual gasses and clean Li surfaces - all while still maintaining enough operational flexibility for multiple in-vacuum diagnostic upgrades and calibrations. Coatings had a visibly clean appearance, with reflective liquid metal demonstrating good wetting and surface adhesion with films that were now macroscopically thick. Solidified Li showed large crystal grains, while surface science measurements observed reduced impurities in the lithium. Steadily improved plasma performance was achieved with liquid lithium, with discharges able to match solid Li in terms of evolution of Ip and ne, including rapid density pumping indicating low recycling. There were indications of moderately increased Li impurity influx, though few significant disturbances by the large liquid surfaces on tokamak operations over hundreds of discharges. Liquid metal plasma facing components are a potential solution to the extreme heat and particle fluxes that could cause unacceptable damage to solid materials, while liquid lithium also has the potential for greatly increased confinement in the low-recycling regime. While many liquid metal approaches are possible, and numerous experiments have been conducted in test stands and small modules in fusion devices, LTX-beta is the only tokamak operated while fully surrounded by liquid metal.

Integrals to calculate generalised magnetic coordinates from an input magnetic flux function asymptotically close to the separatrix are presented, and implemented in the GPEC/DCON code suite. These integrals allow characterisation of the magnetic equilibrium of a diverted tokamak, in magnetic coordinates, arbitrarily close to the last closed flux surface, avoiding the numerical issues associated with calculating diverging field-line integrals near a magnetic x-point. These methods may assist ongoing efforts to develop robust asymptotic equilibrium behaviour for spectral 3D MHD codes at the separatrix.

The NIMROD code is used to perform simulations of MHD activity in the HBT-EP tokamak, including the effect of a resistive wall. Linear simulations are conducted to scan current and pressure profiles to establish self-consistent baseline equilibria that result in MHD instabilities with low error between experimental diagnostics and synthetic equivalents. A family of such equilibria is found with relatively peaked current and pressure profiles. Nonlinear simulations are performed using the optimal equilibrium, which is linearly unstable to a Resistive Wall Mode—stable in the ideal wall limit and with an Alfvénic growth rate in the no-wall limit. Using resistive wall boundary conditions, nonlinear Hall MHD simulations yield a rotating, saturated n=1 mode similar to the type observed experimentally in HBT-EP discharges. An n=1 mode around the q=2 surface mediates a localized axisymmetric perturbation that flattens the plasma current from a linearly unstable state to a 3D, rotating, stable state. During the saturated period, periodic changes in inter-diagnostic phase lag allow for qualitative estimations of effective, macroscopic transport coefficients used in MHD closure models and could be extended to model post-disruption MHD structure rotation after thermal quenches seen in HBT-EP.

Two primary systems planned for disruption mitigation in the SPARC tokamak, massive gas injection (MGI), and the runaway electron mitigation coil (REMC), are modeled with the 3D MHD code NIMROD. MGI is modeled in four configurations, and each of the 6-valve and 4-valve configurations considered is predicted to produce a high radiation fraction (>95%) and low toroidal peaking factor [<1.5 during the thermal quench (TQ)]. The MGI-induced TQ is also modeled in conjunction with the REMC in two different scenarios: one in which the REMC is assumed to be a closed-circuit, and another in which it can carry current only during the CQ. The closed-circuit coil has some effect on the timing of the TQ onset but not on the overall TQ dynamics or subsequent REMC effectiveness. The REMC is able to maintain stochastic fields in the plasma so long as the safety factor at the magnetic axis remains below 2 (q0<2), but the safety factor evolution, in turn, depends on a number of factors, including the treatment of the region outside of the SPARC limiter surrounding the plasma.

Magnetic equilibrium reconstruction in the Lithium Tokamak Experiment-Beta (LTX-beta) is complicated by strong eddy currents and toroidal asymmetries arising from its segmented, close-fitting conducting shell. In the earlier experiment LTX, these three-dimensional (3D) conductor effects significantly distorted diagnostic signals and challenged conventional axisymmetric reconstruction methods. In this work, we demonstrate that accurate plasma equilibria can be recovered on LTX-beta by incorporating a small number of dominant eddy current modes derived from realistic conductor models. Using the open-source TokaMaker Grad-Shafranov solver, we reconstruct equilibria across a systematically selected set of LTX-beta discharges and validate them against the legacy PSI-Tri hybrid 2D-3D code. Our results show that fully 2D TokaMaker reconstructions achieve significantly improved agreement with flux loop and Mirnov probe measurements, reducing total chi-squared fitting errors, especially during startup. Finally, we develop new hybrid 2D-3D TokaMaker reconstructions by integrating the ThinCurr 3D eddy current model, which we optimize for significant further reduction of chi-squared in most scenarios. These findings underscore the importance of realistic wall-current modeling in short-pulse tokamaks, and establish a physics-based reconstruction framework that is extensible to devices with complex passive structures and 3D wall interactions.

Negative triangularity (NT) tokamak configurations have several key benefits including sufficient core confinement, improved power handling, and reduced edge pressure gradients that allow for edge-localized mode (ELM) free operation. We present the design of a compact NT device for testing sophisticated simulation and control software, with the aim of demonstrating NT controllability and informing power plant operation. The TokaMaker code is used to develop the basic electromagnetic system of the R-0 = 1m, a = 0.27m, B-t = 3T, T-p = 0.75 MA tokamak. The proposed design utilizes eight poloidal field coils with maximum currents of 1 MA to achieve a wide range of plasma geometries with and . Scenarios with strong negative triangularity and high elongation are particularly susceptible to vertical instability, necessitating the inclusion of high-field side and/or low-field side passive stabilizing plates which together reduce vertical instability growth rates by 75%. Upper limits for the forces on poloidal and toroidal field coils are predicted and mechanical loads on passive structures during current quench events are assessed. The 3 T on-axis toroidal field is achieved with 16 demountable copper toroidal field coils, allowing for easy maintenance of the vacuum vessel and poloidal field coils. This pre-conceptual design study demonstrates that the key capabilities required of a dedicated NT tokamak experiment can be realized with existing copper magnet technologies.

Active feedback control in magnetic confinement fusion devices is desirable to mitigate plasma instabilities and enable robust operation. Optical high-speed cameras provide a powerful, non-invasive diagnostic and can be suitable for these applications. In this study, we process high-speed camera data, at rates exceeding 100 kfps, on in situ field-programmable gate array (FPGA) hardware to track magnetohydrodynamic (MHD) mode evolution and generate control signals in real time. Our system utilizes a convolutional neural network (CNN) model, which predicts the n = 1 MHD mode amplitude and phase using camera images with better accuracy than other tested non-deep-learning-based methods. By implementing this model directly within the standard FPGA readout hardware of the high-speed camera diagnostic, our mode tracking system achieves a total trigger-to-output latency of 17.6 μs and a throughput of up to 120 kfps. This study at the High Beta Tokamak-Extended Pulse (HBT-EP) experiment demonstrates an FPGA-based high-speed camera data acquisition and processing system, enabling application in real-time machine-learning-based tokamak diagnostic and control as well as potential applications in other scientific domains.

A new three-dimensional electromagnetic modeling tool ThinCurr has been developed using the existing PSI-Tet finite-element code in support of conducting structure design work for both the SPARC and DIII-D tokamaks. Within this framework a 3D conducting structure model was created for both the SPARC and DIII-D tokamaks in the thin-wall limit. This model includes accurate details of the vacuum vessel and other conducting structural elements with realistic material resistivities. This model was leveraged to support the design of a passive runaway electron mitigation coil (REMC), studying the effect of various design parameters, including coil resistivity, current quench duration, and plasma vertical position, on the effectiveness of the coil. The REMC is a non-axisymmetric coil designed to passively drive large non-axisymmetric fields during the plasma disruption thereby destroying flux surfaces and deconfining RE seed populations. These studies indicate that current designs should apply substantial 3D fields at the plasma surface during future plasma current disruptions as well as highlight the importance of having the REMC conductors away from the machine midplane in order to ensure they are robust to off-normal disruption scenarios.

In this paper, we present a new static and time -dependent MagnetoHydroDynamic (MHD) equilibrium code, TokaMaker, for axisymmetric configurations of magnetized plasmas, based on the well-known Grad-Shafranov equation. This code utilizes finite element methods on an unstructured triangular grid to enable capturing accurate machine geometry and simple mesh generation from engineering -like descriptions of present and future devices. The new code is designed for ease of use without sacrificing capability and speed through a combination of Python, Fortran, and C/C++ components. A detailed description of the numerical methods of the code, including a novel formulation of the boundary conditions for free -boundary equilibria, and validation of the implementation of those methods using both analytic test cases and cross -code validation is shown. Results show expected convergence across tested polynomial degree for analytic and cross -code test cases.

A Large Ion Collider Experiment (ALICE) has been conceived and constructed as a heavy-ion experiment at the LHC. During LHC Runs 1 and 2, it has produced a wide range of physics results using all collision systems available at the LHC. In order to best exploit new physics opportunities opening up with the upgraded LHC and new detector technologies, the experiment has undergone a major upgrade during the LHC Long Shutdown 2 (2019–2022). This comprises the move to continuous readout, the complete overhaul of core detectors, as well as a new online event processing farm with a redesigned online-offline software framework. These improvements will allow to record Pb-Pb collisions at rates up to 50 kHz, while ensuring sensitivity for signals without a triggerable signature.

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant beta phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent 'pedestal-to-divertor' integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T-e and radiation which need control. High-beta(P) scenarios with q(min) > 2, q(95)-7.9, beta(N) > 4, beta(T)-3.3% and H-98y2 > 1.5 are sustained with high density ((n) over bar = 7E19 m(-3), f(G)-1) for 6 tau(E), improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

We report the first observation of global recycling coefficient R near 0.5 in the Lithium Tokamak eXperiment-β (LTX-β), significantly below the minimum R previously reported in other devices. In a series of experiments with varied Li wall conditioning, estimates of the recycling coefficient have been made using a Lyman-α array and DEGAS2 modeling. A progressive reduction in Lyman-α emission with increased lithium and an increase in edge electron temperature are observed. It is also observed that with increasing Li coating thickness, the effective particle confinement time τp* is reduced and approaches TRANSP calculated energy confinement time (τE), with τp* near τE,TRANSP for the lowest recycling coefficients. Edge temperatures approaching core plasma temperatures, first reported in LTX, can now be directly connected to estimates of the recycling coefficient and qualitatively agree with previous UEDGE simulations. The particle flux to the limiting surfaces appears to be significantly reduced in comparison with fluid scrape-off layer (SOL) models, indicating that a large fraction of the SOL ions are mirror trapped. SOL collisionality drops more than an order of magnitude below the banana regime boundary, indicating the importance of kinetic effects. Full-f 1x2v gyrokinetic simulations of SOL field lines with the GKEYLL code indicate that the fraction of ions trapped along field lines increases as collisionality drops, as a result of increased lithium evaporation.

The MANTA (Modular Adjustable Negative Triangularity ARC-class) design study investigated how negative-triangularity (NT) may be leveraged in a compact, fusion pilot plant (FPP) to take a 'power-handling first' approach. The result is a pulsed, radiative, ELM-free tokamak that satisfies and exceeds the FPP requirements described in the 2021 National Academies of Sciences, Engineering, and Medicine (NASEM) report 'Bringing Fusion to the U.S. Grid' (2021 Bringing Fusion to the U.S. Grid). A self-consistent integrated modeling workflow predicts a fusion power of 450 MW and a plasma gain of 11.5 with only 23.5 MW of power to the scrape-off layer (SOL). This low PSOL together with impurity seeding and high density at the separatrix results in a peak heat flux of just 2.8 MW m-2. MANTA's high aspect ratio provides space for a large central solenoid (CS), resulting in similar to 15 minute inductive pulses. In spite of the high B fields on the CS and the other REBCO-based magnets, the electromagnetic stresses remain below structural and critical current density limits. Iterative optimization of neutron shielding and tritium breeding blanket yield tritium self-sufficiency with a breeding ratio of 1.15, a blanket power multiplication factor of 1.11, toroidal field coil lifetimes of 3100 +/- 400 MW yr, and poloidal field coil lifetimes of at least 890 +/- 40 MW yr. Following balance of plant modeling, MANTA is projected to generate 90 MW of net electricity at an electricity gain factor of similar to 2.4. Systems-level economic analysis estimates an overnight cost of US$3.4 billion, meeting the NASEM FPP requirement that this first-of-a-kind be less than US$5 billion. The toroidal field coil cost and replacement time are the most critical upfront and lifetime cost drivers, respectively.

Negative triangularity (NT) tokamak configurations may be more susceptible to magneto-hydrodynamic instability, posing challenges for recent reactor designs centered around their favorable properties, such as improved confinement and operation free of edge-localized modes. In this work, we assess the vertical stability of plasmas with NT shaping and develop potential reactor solutions. When coupled with a conformal wall, NT equilibria are confirmed to be less vertically stable than equivalent positive triangularity (PT) configurations. Unlike PT, their vertical stability is degraded at higher poloidal beta. Furthermore, improvements in vertical stability at low aspect ratio do not translate to the NT geometry. NT equilibria are stabilized in PT vacuum vessels due to the increased proximity of the plasma and the wall on the outboard side, but this scenario is found to be undesirable due to reduced vertical gaps which give less spatial margin for control recovery. Instead, we demonstrate that informed positioning of passively conducting plates can lead to improved vertical stability in NT configurations on par with stability metrics expected in PT scenarios. An optimal setup for passive plates in highly elongated NT devices is presented, where plates on the outboard side of the device reduce vertical instability growth rates to 16% of their baseline value. For lower target elongations, integration of passive stabilizers with divertor concepts can lead to significant improvements in vertical stability. Plates on the inboard side of the device are also uniquely enabled in NT geometries, providing opportunity for spatial separation of vertical stability coils and passive stabilizers.

Extended-MHD modeling of planned Runaway Electron Mitigation Coils (REMC) for SPARC and DIII-D is performed with the NIMROD code. A coil has been designed for each machine, with the two differing in shape and location, but both having n = 1 symmetry (with n the toroidal mode number). Compared to previous modeling efforts, three improvements are made to the simulations boundary conditions. First a resistive wall model is used in place of an ideal wall. Second, the ThinCurr code is used to compute the time-dependent 3D fields used as magnetic boundary conditions for the simulations. Third, the simulation boundary is moved from the first-wall location to the Vacuum Vessel (VV), which extends the boundary past the location of the internal REMC. To remove the 3D coil from the simulation domain, an equivalent set of 3D fields is calculated at the VV boundary that produce approximately the same field distribution at the last closed flux surface assuming vacuum between the two. Each of these three boundary condition improvements leads to an improvement in the predicted performance of the REMC for both machines. The resistive wall alone primarily effects the resonance of the coil with the plasma after the TQ, affecting the q-profile evolution in the SPARC modeling, and allowing the applied spectrum to be modified in response to the plasma in the DIII-D modeling. The movement of the simulation boundary has the most significant effect on the RE confinement overall, including in the early stages, particularly for a DIII-D inner wall limited equilibrium, where the RE loss fraction increases from 90% to > 99%, with SPARC RE losses also occurring much earlier when the boundary is placed at the VV.