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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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Bonding of Tungsten and Graphite Using Spark Plasma Sintering for Divertor Component in LHD
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FIREFLY: heat load and particle exhaust approximations for rapid evaluation of divertor designs
IOP PublishingAbstract The divertor in a magnetic confinement fusion reactor is an essential&#xD;component for power dissipation and particle removal. This article introduces the&#xD;FIREFLY package for rapid evaluation of divertor designs based on an extension of&#xD;the FLARE code for field line reconstruction from an unstructured flux tube mesh. First, divertor heat&#xD;loads are approximated with a simplified heat transport model. Neutralized particles&#xD;are then sampled from the resulting distribution, and the EIRENE code is used to&#xD;track molecules and atoms in a plasma background while accounting for dissociation,&#xD;charge exchange and ionization. Particles are removed on pumping surfaces in order&#xD;to estimate the exhaust efficiency for a given divertor geometry. Optimization of the&#xD;divertor geometry for more efficient particle exhaust is explored by using W7-X as an&#xD;example, and the sensitivity to model parameters for the plasma background in the&#xD;proxy calculations is evaluated.
Heat Transfer Performance of EVAPORON-3 Developed for an Enlarged Heat Transfer Surface of Divertor
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6.7 Plasma Facing Component (Divertor)
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Heat pipe technology based divertor plasma facing component concept for European DEMO
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Impact of the plasma geometry on the divertor power exhaust in a magnetic fusion reactor
Agence Bibliographique de l'Enseignement SupérieurImpact de la géométrie du plasma sur l'extraction de puissance au divertor d'un réacteur à fusion magnétique Une compréhension profonde du transport du plasma au bord d'un réacteur à fusion par confinement magnétique est obligatoire pour gérer l'extraction de puissance. Dans les dispositifs de fusion de nouvelle génération, des limites technologiques contraignent le flux de chaleur maximal au divertor. Pour une puissance d'échappement donnée le flux de chaleur maximal est déterminé par l'amplitude de l'empreinte du plasma au mur. Les profils de flux de chaleur au divertor peuvent être paramétrés par deux échelles de longueur du transport. Nous remettons en question l'interprétation actuelle de ces deux échelles de longueur en étudiant l'impact de la géométrie du divertor sur l'échappement. En particulier, un élargissement des profils de flux de chaleur avec la longueur de la jambe du divertor externe est diagnostiqué. Des efforts de modélisation ont montré que les simulations diffusives reproduisent les profils expérimentaux de flux de chaleur pour les plasmas à jambes courtes. Inversement, l'étalement du flux de chaleur pour une longe jambe du divertor est reproduit par un modèle turbulent, soulignant l'importance de la turbulence aussi dans le divertor. Ces résultats remettent en question l'interprétation de la largeur du flux de chaleur comme grandeur liée a la main SOL uniquement. Les configurations magnétiques avec une longe jambe du divertor mettent en évidence l'importance du transport asymétrique dans le divertor. Nous concluons que le transport dans la main SOL et celui dans le divertor ne sont pas à découpler et nous soulignons l'importance de la géométrie magnétique sur le transport turbulent avec l'avantage potentiel d'un inattendu étalement du dépôt de puissance.
Physics basis for the reference flat-top plasma scenario in the ST–E1 fusion power plant
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Tokamak Energy’s pre-concept design for a fusion power plant: an overview of ST-E1
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Analytic 1D approximation of the divertor broadening S in the divertor region for conductive heat transport
IOP PublishingAbstract An analytic 1D approximation for the divertor broadening S is introduced, depending only on the electron temperature between X-point and target. It is compared to simulations solving the 2D heat diffusion equation, in order to describe the divertor broadening along a field line solely by the ratio of the perpendicular to the parallel diffusivities. By assuming the temperature dependence of these two diffusivities an integral form of S is derived for the area along the separatrix between X-point and target. Integration along the separatrix results in an approximation for S , being in agreement with the 2D simulations. This approximation is furthermore compared to recent studies, which find a power law with negative exponent to describe S in terms of target temperature. This dependence is not reproduced in a pure conductive description, which instead shows a finite S for zero target temperature. This points to other mechanisms changing the shape of the heat flux profile—by additional widening or radiation losses—not included in the presented reduced approximation.
Achievement of a high-density, high-confinement, and high-beta tokamak plasma regime in DIII-D, and implications for a lower-current path for ITER and FPP
IOP PublishingExperiments on DIII-D have demonstrated a density-confinement synergy that enables sustainment of high performance in a previously unattained parameter regime of simultaneous very high energy confinement quality (H<sub>98y2</sub> ≥ 1.5), very high line-average density Greenwald fraction (ƒ<sub>Gr</sub> = πa<sup>2</sup>< n >/I<sub>P</sub> ≥ 1.4), and high toroidal beta (β<sub>T</sub> ≥ 3%). Tokamak operation in this regime is essential for a compact steady-state FPP, as well as for Q=10 with 500 MW of fusion power in ITER at I<sub>P</sub> << 15 MA. These experiments leveraged the knowledge that, in the high-poloidal-beta (β<sub>P</sub>) regime, impurity and density gradients can enhance turbulence stabilization caused by high α<sub>MHD</sub> (α<sub>MHD</sub>~(dβ<sub>P</sub>)⁄dr). This was described by theoretical predictions and gyrokinetic transport simulations [M.T. Kotschenreuther et al, 2024 Nucl. Fusion, 64 076033], and later confirmed by experiments on DIII-D [S. Ding et al, 2024 Nature 629 555]. To increase both β<sub>P</sub> and β<sub>T</sub>, the new experiments increased the ideal-wall stability β<sub>N</sub>-limit by using a smaller plasma-outer wall distance and higher triangularity in the plasma cross section (top/bottom average δ~0.9), enabled by the recent “shape & volume rise” (SVR) modification to the DIII-D divertor. The higher triangularity also contributed to achieving higher ƒ<sub>Gr</sub> by enabling higher pedestal density. At high density, the pedestal is ballooning limited and exhibits small and frequent ELMs, while the divertor is near detachment even without any impurity seeding. High plasma performance was attained and sustained reproducibly, with the eventual terminations brought about by an MHD mode destabilized as the current profile slowly continued to evolve. A path to stationary fully noninductive operation might include ECH injection to reduce both core impurity accumulation and the electron collisionality, thus increasing the bootstrap current. These experiments provide the first experimental demonstration of the ƒ<sub>Gr</sub>, H<sub>98y2</sub>, and β<sub>T</sub> values required simultaneously for ITER Q = 10 at I<sub>P</sub> < 10 MA, pointing to practical ways to improve the energy confinement in a fusion reactor.
Thermal stress intensity factor of interfacial cracks of a plasma facing component under high heat flux loading
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Recent advances in plasma control and physics research in the Large Helical Device
IOP Science - IAEAThe Large Helical Device (LHD), the largest superconducting helical system in the world, is equipped with advanced heating and diagnostic tools, facilitating plasma control and physics research. Data assimilation was employed for electron temperature control using a real-time Thomson scattering system and real time prediction code. A virtual LHD environment enabled visualization of escaping high-energy tritium ions and demonstrated that these ions impact the rear side of the divertor plate. Pioneering results crucial to plasma control have also been achieved. Real-time wall conditioning using Lithium granule dropping improved bulk ion energy and particle transport while simultaneously enhancing the heavy impurity transport. Progress has also been made in the investigation of turbulence-driven transport. At the confinement bifurcation, ion-scale turbulence decreased, while electron-scale turbulence increased. A change in the anisotropy of turbulent eddies was also observed at the confinement bifurcation. Coexistence of local and non-local turbulence was identified in electron-scale turbulence. Non-local turbulence exhibited the rapid spatial propagation of perturbations throughout the plasma, while local turbulence followed the temperature gradient. A transition between drift-wave turbulence and magnetohydrodynamics (MHD) turbulence was observed with the turbulence minimized at the transition condition. Machine learning analysis was employed to evaluate the temperate and density conditions of this turbulence transition. Then, real-time control of fueling and heating was applied to maintain the turbulence transition condition, improving the energy confinement enhancement factor by 20%. In addition, evidence was obtained for collisionless ion heating by energetic-ion-driven geodesic acoustic modes and MHD bursts. These achievements represent unique contributions to the development of fusion reactors.
Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups
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Material ejection and surface morphology changes during transient heat loading of tungsten as plasma-facing component in fusion devices
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