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COUPLED SEISMIC ANALYSES OF THE ITER TOKAMAK COMPLEX BUILDING AND TOKAMAK MACHINE
World Conference of Earthquake Engineeringnull
Microturbulence Suppression by Alfvén Eigenmodes in the DIII-D Tokamak
Physical Review LettersMitigation and suppression of low-k turbulence are observed during the nonlinear evolution of toroidicity-induced Alfvén eigenmodes (TAEs) in DIII-D experiments. Turbulence mitigation begins when the dominant TAE starts to depart from the typical shear Alfv…
How turbulence sets boundaries for tokamak operation
Reviews of Modern Plasma PhysicsAbstract Various boundaries, such as the transition to high confinement, density limits, and power exhaust requirements, impose limits on the safe and efficient operation of a tokamak. Turbulent transport at the separatrix plays a decisive role in these lim…
Solution of the linear wave-particle kinetic equation for global modes of arbitrary frequency in a tokamak
Fundamental Plasma PhysicsThe linear response of a plasma to perturbations of arbitrary frequency and wavelength is derived for any axisymmetric magnetized toroidal plasma. An explicit transformation to action-angle coordinates is achieved using orthogonal magnetic coordinates and t…
Progress and innovations in the TCV tokamak research programme
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Integration of X-point radiator divertor operation with high beta hybrid core plasmas in DIII-D
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Development of mechanically jointed plasma facing component for divertor plates
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A snowflake divertor: a possible solution to the power exhaust problem for tokamaks
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Overview of the KSTAR experiments and future plan
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Investigation of Drift Effects in UEDGE Simulations of NSTX-U Edge Plasma With Lithium Divertors
WileyLithium is a low-Z material, and lithium-based plasma-facing components (PFCs) are planned for the National Spherical Torus Experiment Upgrade (NSTX-U) to explore potential benefits for divertor power exhaust and core plasma management. NSTX-U is a medium-sized spherical tokamak with up to 12 MW of auxiliary heating, capable of generating reactor-relevant plasma conditions. This work presents boundary plasma simulations for NSTX-U with lithium PFCs using the UEDGE code, incorporating full magnetic and E ×B drift physics. The simulations show that drifts strongly influence heat and particle transport: they enhance convective transport, broaden the scrape-off layer heat-flux width λ<sub>q</sub> , and reduce the anomalous heat diffusivity χ required to reproduce predicted SOL heat-flux width. E ×B drifts provide poloidal transport, while ∇B (which includes both gradB and curvature) drifts provide radial heat and particle transport. Lithium transport is also affected by drifts, with lithium ions migrating from the outer divertor to the inner divertor through the private flux region (PFR) following the E ×B drifts flow, lowering upstream impurity lithium densities. UEDGE is self-consistently coupled with the Wall-Li model to study plasma lithium PFC interactions depending on the local lithium sourcing based on local plasma conditions and lithium surface temperature. In these simulations, lithium evaporation shows a vapor-shielding effect that reduces divertor heat flux and increases radiative losses once surface temperatures exceed 450°C. This research work provides a first step toward self-consistent modeling of lithium PFCs in NSTX-U, demonstrating the impact of drift-driven plasma transport in SOL and divertor regions.
ANALYTIC EXPRESSIONS FOR EXTRACTING FAR SCRAPE-OFF LAYER TRANSPORT COEFFICIENTS FROM THE PLASMA-WALL INTERACTION OF DIVERTOR TOKAMAKS
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Maintenance strategy, structural design, and site layout of the ST-E1 fusion power plant
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Turbulent heat transfer for coolant water flow in plasma facing component
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Power and particle exhaust in the ST-E1 fusion power plant
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Overview of the MAST Upgrade physics programme: testing novel concepts at low aspect ratio to inform future devices
IOP Science - IAEAThe research programme performed on the Mega Amp Spherical Tokamak (MAST) Upgrade experiment has made significant advances in developing the physics understanding of low aspect ratio tokamaks in support of the operation of ITER and design of fusion powerplants. High performance plasma scenarios have been developed to facilitate a broad programme of experiments, in which confinement is constrained by the presence of m/n = 2/1 modes that cause substantial losses of fast ions. The onset of these modes coincides with the q = 2 surface residing in a local minimum in the toroidal current density profile. The maximum electron temperature at the pedestal top, T<sub>e,ped</sub> is limited with gas fuelling to ∼350 eV to maintain regular ELMs; higher T<sub>e,ped</sub> results in a transition to a non-stationary ELM-free regime. The operational space of spherical tokamaks has been expanded into small and ELM-free regimes. Strong shaping of the last closed flux surface can induce a transition from large to small ELMs, and ELM suppression with resonant magnetic perturbations has been observed for the first time in a low aspect ratio tokamak. Negative triangularity shaping has induced a transition from ELMy H-mode to a high-performance L-mode regime for the first time in a low aspect ratio tokamak. In studies of fast ion confinement, losses of fast particles due to Global Alfvén Eigenmodes have been identified. Interactions between fast ions generated by off-axis neutral beam injection and thermal neutrals can result in significant losses of fast ions. Experiments with on- and off-axis neutral beam injection exhibit a flux pumping mechanism, where the central safety factor is held to ∼1 in the absence of sawteeth. In studies of pedestal physics, it has been found that elevated main chamber neutral pressures result in an increase in the electron density and reduction in the temperature at the pedestal top. Advances in understanding plasma exhaust include the integration of a high-performance plasma core with detached outer divertors in the X-point target configuration. A newly commissioned lower divertor cryopump reduces the lower divertor neutral pressure by up to 50%, with minimal effect on the main chamber or upper divertor. New measurements and SOLPS-ITER simulations emphasise the importance of plasma–neutral interactions on divertor detachment in the conditions accessible in experiments. Real-time control of the ionisation front location in both divertor chambers independently has been demonstrated in double null experiments, enabled by the tightly baffled divertor chambers.
FIREFLY: heat load and particle exhaust approximations for rapid evaluation of divertor designs
DOE OSTIThe divertor in a magnetic confinement fusion reactor is an essential component for power dissipation and particle removal. The FIREFLY package for rapid evaluation of divertor designs is presented as an extension of the FLARE code for field line reconstruction from a flux tube mesh. First, divertor loads are approximated with a simplified heat transport model. Neutralized particles are then sampled from the resulting load distribution, and the EIRENE code is used to track molecules and atoms in a plasma background while accounting for dissociation, charge exchange and ionization. Particles are removed on pumping surfaces in order to estimate the exhaust efficiency for a given divertor geometry. Optimization of the divertor geometry for more efficient particle exhaust is explored by using W7-X as an example, and the sensitivity to model parameters for the plasma background in the proxy calculations is evaluated.
Overview of Wendelstein 7-X high-performance operation
IOP Science - IAEAThe Wendelstein 7-X (W7-X) stellarator has completed two consecutive experimental campaigns OP 2.2 (Sep.-Dec. 2024) and OP 2.3 (Feb.-May 2025) under a new operational strategy enabling more than one year of uninterrupted device availability. This approach, supported by exceptionally high subsystem reliability, allowed sustained high-efficiency plasma operations with up to 80–100 discharges per day across a broad range of magnetic configurations. Several key technical upgrades-most notably the first operation of a 1.5 MW class steady-state gyrotron, a new steady-state pellet injector, and advanced real-time feedback control systems significantly enhanced heating, fueling, and plasma control capabilities. Together, these improvements enabled major advances in long-pulse performance, high-β operation, and confinement optimization. Long-pulse discharges achieved 1.8 GJ of injected energy under fully detached divertor conditions, while reduced-field scenarios facilitated record volume-averaged β values approaching 3%. High-performance plasmas with centrally peaked density profiles, created via neutral beam injection (NBI) or sustained pellet fueling, demonstrated strongly reduced turbulent transport and stellarator-record fusion triple products. Complementary studies of power exhaust and divertor heat loads revealed the role of scrape-off-layer drift physics in shaping strike-line patterns under attached conditions. Together, the results from OP 2.2 and OP 2.3 significantly expand the operational space of W7-X and strengthen its role as a leading platform for steady-state stellarator research and reactor-relevant plasma scenarios.
Development of the plasma facing components for the dome-liner component of the ITER divertor
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Power and particle exhaust for the ARC fusion power plant
Cambridge University PressTo successfully show that fusion is an attractive energy source, the ARCTM fusion power plant will need to operate with a robust, integrated power and particle exhaust solution. To maximise ARC’s fusion power output while avoiding excessive erosion of the plasma-facing components, we will need to radiatively dissipate most of the power crossing the last-closed flux surface, injecting radiating impurities such as argon or neon to access divertor detachment. Divertor detachment will need to be integrated with a high-performance core plasma, and with efficient impurity pumping to prevent the accumulation of helium ash in the core. To access and control detachment in high-performance plasmas, we have designed ARC with up–down-symmetric divertors, with secondary X-points in long, tightly baffled outer legs. Using a core-edge modelling workflow, we predict that with this divertor design, ARC will access detachment with modest argon seeding in the divertor (c<sub>Ar,div</sub> ∼0.9%), which should have minimal impact on the core ( Z<sub>eff,core</sub> <0.5) for reasonable argon enrichment (c<sub>Ar,div</sub>/c<sub>Ar,core</sub> =6). Due to the high predicted divertor neutral pressure (p<sub>div </sub>∼20 Pa), we predict that ARC will sufficiently pump helium to limit ash accumulation in the core (c<sub>He,core</sub> <2%) for a helium enrichment of c<sub>He,div</sub>/c<sub>He,core </sub>=0.4. ARC’s divertor design is expected to increase the stability of a detachment front in the outer divertor leg, helping to prevent divertor reattachment during smaller heat-flux transients such as scrape-off-layer filaments associated with the quasi-continuous exhaust regime. However, this buffering will not be sufficient to prevent divertor reattachment during large type-I edge-localised modes (ELMs), and as such these will need to be avoided on ARC. Experiments on SPARC will be used to select an integrated scenario which avoids or mitigates type-I-ELMs while maintaining access to detachment, good core fusion performance and sufficient impurity exhaust. SPARC experiments will also be used to finalise ARC’s divertor design, by studying the impact of magnetic and first-wall geometry on detachment stability, impurity enrichment and neutral baffling under conditions similar to those expected for ARC. In conclusion, our analysis finds that ARC will have a viable power and particle exhaust solution which is compatible with high-power operations, and this solution will be validated in experiments on SPARC.
Effects of transient heating events on tungsten plasma-facing materials in a steady-state divertor-plasma environment*
IOP PublishingNearly all heat-pulse tests of plasma-facing materials (PFMs) have been completed in vacuum environments without the presence of a background plasma. Thus, combined effects of thermal transients on materials undergoing plasma exposure need to be explored. Heat-pulse experiments have been conducted in the PISCES-A device using a pulsed laser in a divertor-like plasma background. The results indicate that the erosion of PFMs is enhanced as compared with transient-only or plasma-only experiments, and the threshold energy for material removal by a transient heat pulse in a steady-state plasma background is reduced. There appears to be a minimum ion fluence to the surface needed to cause these effects. Initial experiments with pre-loaded material samples exposed to plasmas and heat pulsed in the lower divertor of the DIII-D tokamak using the Divertor Material Evaluation System indicate that similar effects occur in confinement devices.
Fracture mechanical analysis of a tungsten monoblock-type plasma-facing component without macroscopic interlayer for high-heat-flux divertor target
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Divertor detachment and heat exhaust mitigation control in KSTAR with tungsten divertor
IOP PublishingAbstract KSTAR has recently undergone an upgrade to use a new tungsten divertor to run experiments in ITER-relevant scenarios. Even with a high melting point of tungsten, it is important to control the heat flux impinging on tungsten divertor targets to minimize sputtering and contamination of the core plasma. Heat flux on the divertor is often controlled by increasing the degree of detachment of scrape-off layer plasma from the target plates. In this work, we have demonstrated successful divertor detachment and heat exhaust dissipation control experiments using two different methods. The first method uses attachment fraction as a control variable which is estimated using ion saturation current measurements from embedded Langmuir probes in the divertor. The second method uses a novel machine-learning-based surrogate model of 2D UEDGE simulation database, DivControlNN. We demonstrated running inference operation of DivControlNN in realtime to estimate heat flux at the divertor and use it as the control variable in a feedback loop with impurity gas flow. We present interesting insights from these experiments including a systematic approach to tuning controllers and discuss future improvements in the control infrastructure and control variables for future burning plasma experiments.
Overview of the physics basis for the ARC fusion power plant
Cambridge University PressCommonwealth Fusion Systems plans to build ARC as the first fusion power plant at a site in Chesterfield County, Virginia, USA by the early 2030s. We present an overview of analysis comprising the physics basis of the ARC V3A design, a high-magnetic-field tokamak with B<sub>0</sub> =11.4 T, I<sub>p</sub> =12.0 MA, R<sub>0</sub> =4.62 m, a =1.18 m. ARC V3A is designed to produce P<sub>fus</sub> ≈1.13 GW DT fusion power and deliver ⩾400 MW net electric power to the grid. This overview includes quantitative analysis of fundamental issues for design of and operational plasma scenarios for a tokamak power plant, and lays out the design targets and strategic choices for ARC, including empirical fusion performance projections, assessment of H-mode access, ion cyclotron resonance heating simulations, alpha particle physics and time-dependent full-pulse simulations. This is complemented by topical papers on fusion performance and transport, disruption physics, boundary physics and magnetohydrodynamic stability. Critically, these studies identify key model uncertainties and physics risks to be retired through SPARC operation. Due to the modular nature of ARC, early results from SPARC can be incorporated into the design of the first ARC as well as subsequent replacements of the ARC vacuum vessel.
Stress analysis of divertor plasma-facing component designs using tungsten particle-reinforced copper composite heat sink
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