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Aug 26

Pre L-H transition radial electric field and transport validations of edge and scrape-off layer gyrokinetic simulations at ASDEX upgrade

Baptiste Jimmy Frei, Clemente Angioni, Guillaume Lo-Cascio, Wladimir Zholobenko, Philipp Ulbl, Roberto Bilato, Frank Jenko

Plasma Physics and Controlled FusionAug 26, 2026Plasma & Confinement AI, Modeling & Simulation

This work presents a stepwise validation of the evolution of the radial electric field Er and transport during the pre L-H transition phase in the ASDEX Upgrade (AUG) tokamak using edge and scrape-off layer full-f gyrokinetic simulations including X-point geometry. Several L-mode time slices up to the L-H transition from a dedicated hydrogen discharge, featuring stepwise increases in ECRH input power, are selected [N. Bonanomi et al., Phys. Plasmas 31, 072302 (2024)] and simulated with the GENE-X code. As the edge boundary conditions are progressively increased between the time slices, particle and heat fluxes rise, and the radial electric field Er well deepens. A detailed validation of the Er profiles and of the Er well depth shows excellent agreement with experimental measurements at the successive time slices approaching the L-H transition. A force balance decomposition identifies turbulence-driven poloidal flows as the dominant contribution within the Er well. Edge turbulence is governed by a competition between electron drift waves and trapped-electron modes. The introduction of an edge density source, modeling neutral gas ionization, is shown to be essential to reproduce experimentally relevant density profiles, Er , and edge ion heat fluxes, which are dominated by both turbulent and diamagnetic contributions. This stepwise validation constitutes an important milestone toward predictive, first- principles gyrokinetic simulations of the L-H transition power threshold.

Max Planck Institute for Plasma Physics

Effect of electron cyclotron waves on plasma with runaway electrons

Pavel Aleynikov, Alexander Franklin Battey, Carlos Paz-Soldan, Eric Matthias Hollmann, Andrey Lvovskiy, Claudio Marini, Daisuke Shiraki, Charles Lasnier

Nuclear FusionAug 26, 2026Plasma & Confinement Control & Diagnostics Heating & Current Drive

Runaway electrons generated during tokamak disruptions are a major concern for the safe operation of future fusion devices. The interaction of runaway electrons with waves has been proposed as a potential mechanism for their mitigation. This study investigates the effect of electron-cyclotron (EC) waves on post-disruption plasmas containing runaway electrons (REs). O- and X-mode EC waves are routinely used for plasma heating and current drive. However, these modes do not interact directly with relativistic electrons and cannot be injected into plasmas with densities exceeding their respective cutoff densities. In contrast, the internal slow X-mode (sX) can resonate with relativistic electrons and may therefore provide a pathway for their mitigation. We report DIII-D experiments designed to access the internal slow X-mode through O--sX conversion during post-disruption RE plateaus. ECH increased the background-plasma density, doubled the loop voltage, and strongly enhanced the RE synchrotron signal, demonstrating substantial ECH--plasma coupling and increased effective dissipation of the RE channel. Although ECH produced a pronounced response consistent with enhanced RE dissipation, we could not isolate conclusive evidence of resonant sX--RE interaction. Strong heating, ionization, and impurity-redistribution effects appear to dominate the response and may mask a direct wave--particle signature. Nevertheless, the results demonstrate the potential of ECH for RE control through background-plasma modification, with O--sX access offering an additional possibility of direct resonant interaction.

General AtomicsSwiss Federal Institute of Technology LausanneMax Planck Institute for Plasma Physics

Aug 24

Impact of nitrogen injection on L-H transitions in JET with Be/W wall

Costanza F Maggi, Clarisse Bourdelle, Ephrem Delabie, Mathias Brix, Pedro Carvalho, Francis J Casson, Alex V Chankin, Nicholas Hawkes, Evie Litherland-Smith, Sheena Menmuir, et al.

Plasma Physics and Controlled FusionAug 24, 2026Plasma & Confinement

L-H transition experiments in JET with Be/W wall (JET-ILW), with N2 injection into the divertor region, have revealed that the H-mode power threshold, PL-H, increases with nitrogen injection rate in the high-density branch and modifies the density at which the power threshold is minimum, ne,min. PL-H values approaching those obtained in JET with C wall are achieved for a sizeable increase in nitrogen concentration. At a given density in the high density branch of PL-H, both the total loss power (Ploss) as well as the power crossing the separatrix (Psep) at the L-H transition increase by a factor ~ 1.5 at high N2 levels. At a given line averaged density, while the edge density is similar in plasmas with and without N2 injection, the edge temperature at the L-H transition is higher in pulses with high nitrogen seeding levels and a larger shear of the edge radial electric field Er is measured. At low N concentrations, little change in PL-H is observed. Qualitative agreement is reported with Zeff impact on L-mode edge turbulence drive [1]. The enhanced L-mode edge turbulence drive with N2 seeding at higher Zeff implies that a higher PL-H is expected as, indeed, measured in experiment. It is assumed that the mean equilibrium Er×B shear is a key player for reduction of L-mode edge turbulence, leading to the L-H transition. In this framework, the L-H transition observations in JET-Be/W + N2 are interpreted by an increase in L-mode edge plasma turbulence with increased N concentration and the need for a larger stabilizing ErxB shear, thus higher Psep, to trigger the L-H transition. For future fusion reactors with scenarios with extrinsic low-Z impurity seeding for divertor heat load mitigation, our results suggest the need to predict Psep for H-mode access taking into account not only the high radiation fractions from the bulk plasma, but also the impact of low Z impurity concentration on Ploss and Psep at the L-H transition.

Max Planck Institute for Plasma PhysicsUnited Kingdom Atomic Energy AuthorityOak Ridge National Laboratory

Aug 21

Real-time total ECRH power control for reliable long-pulse operation and density feedback at Wendelstein 7-X

Laurent Krier, Stefan Marsen, Heike Laqua, Heinrich Laqua, Dmitry Moseev, Frank Noke, Hans Oosterbeek, Niklas Simon Polei, Sergiy Ponomarenko, Taurino Reichert, et al.

Nuclear FusionAug 21, 2026Plasma & Confinement Control & Diagnostics Heating & Current Drive

The Electron Cyclotron Resonance Heating (ECRH) system at the Wendelstein 7-X (W7-X) stellarator is equipped with eleven megawatt-class gyrotrons that operate at a frequency of 140 GHz, designed for pulse durations of up to thirty minutes. For long-pulse detached-divertor plasma experiments at W7-X, a stable total ECRH power delivered to the plasma is crucial, because the radiated power at the plasma edge is preferably close to the heating power. Consequently, the unexpected shutdown of only one gyrotron can cause a premature end of the experiment due to radiation collapse of the plasma. This paper introduces a system-level controller that dynamically redistributes power among gyrotrons to maintain total ECRH power, improving reliability and enabling feedback control. The controller maintains a specified total ECRH power output by adjusting the accelerating voltage of all gyrotrons. First implemented in the operational phase (OP) 2.2, the system contributed to two major milestones of W7-X during OP 2.3: highest long-pulse triple product for 43 s and highest energy turnover of 1.8 GJ during 360 s. In addition to reliability improvements, the controller enables the adjustment of the total ECRH power for other control objectives, such as maintaining a desired line-integrated plasma density to counteract the ECRH pump-out effect in high-performance scenarios. This paper also presents the first proof-of-principle experiments demonstrating an ECRH-based density feedback system.

Max Planck Institute for Plasma Physics

Aug 19

Magnetic equilibrium of the CN-H1 heliac: A comparative study of design,as-built, and historically built configurations

Zhengkun Gao, Axel Koenies, Jinjia Cao, Hengqian Liu, Guodong Yu, Dong Xiang, Caoxiang Zhu, Yong-Zhi Dai, Haipeng Wu, Michael Drevlak, et al.

Nuclear FusionAug 19, 2026Plasma & Confinement Magnets & Superconductors AI, Modeling & Simulation

The relocation of the Australian H-1 National Facility (H-1 NF) to China and its reconstruction into the CN-H1 device necessitated a rigorous reassessment of its magnetic configuration to address systematic deviations introduced dur ing the engineering process.This paper presents a comprehensive comparative study of the magnetic equilibrium across three distinct definitions of the de vice: the idealized H-1 Design, the historically built as-built H-1 NF, and the newly reconstructed CN-H1. High-precision laser tracker metrology was em ployed to generate a high-fidelity coil model for CN-H1, incorporating the re manufactured poloidal and helical windings. Calculations of the vacuum mag netic field reveal that, compared to the H-1 NF, the CN-H1 exhibits a systematic inward radial shift of the magnetic axis, a slight increase in the rotational trans form, while maintaining a comparable level of stellarator symmetry breaking. Three-dimensional MHD equilibria were reconstructed using the VMEC code and validated against field-line tracing. Comparative analysis across represen tative configurations demonstrates that the defining characteristics of the flexi ble heliac—including high rotational transform, shear tunability, and magnetic well depth—are preserved in CN-H1. Magnetic spectra in Boozer coordinates of CN-H1 are nearly indistinguishable from the H-1 NF historically built baseline, despite coil positioning errors in the millimeter range. These results validate the engineering fidelity of the CN-H1 reconstruction and establish a precise equilib rium basis for future finite-beta plasma experiments.

Max Planck Institute for Plasma PhysicsUniversity of Science and Technology China

Tungsten limiter start-up experiments on ASDEX Upgrade and WEST in different boronization states in support of ITER

Joerg Hobirk, Richard A Pitts, Pierre Manas, Clemente Angioni, Matthias Bernert, Dominik Brida, Guido Ciraolo, Laurent Colas, Corinne Desgranges, Ralph Dux, et al.

Nuclear FusionAug 19, 2026Plasma & Confinement Materials & Plasma-Facing Components

Understanding the performance of limiter plasmas in the ITER start-up phase is important for the whole pulse and a possible challenge if performed on tungsten, as will now be the case following the switch to a W first wall in the new 2024 ITER Baseline. Experiments were performed on ASDEX Upgrade and WEST to characterise limiter plasmas using boronizations with different degree of boron surface coverage and toroidal asymmetries. Non-boronized start-up is shown to be slow and laborious, and, on both machines, was aborted in favour of a non-homogeneous boronization, performed also in support of the ITER re- baseline to study the impact of spatially non-uniform boron coatings. This allows normal start- up and short (few 100 ms, ITER will run ≈ 10s long limiter phases) limiter phases can be run without problems. Even with a full boronization, the limiter can de-condition and long limiter plasmas of several seconds suffer from high densities and radiation. Short limiter plasmas for plasma current ramp-up remain possible. The conditioning effect on limiter plasmas is documented, but also how the start-up is affected by an ageing boronization.

French Alternative Energies and Atomic Energy CommissionMax Planck Institute for Plasma PhysicsITER Organization

Aug 18

ELM suppression and confinement in negativetriangularity with stronger shaping in ASDEXUpgrade

Branka Vanovac, Joerg Hobirk, Andrew Oakleigh Nelson, Olivier Sauter, Mike G Dunne, Michael Faitsch, Thomas Pütterich, Rainer Fischer, Dirk Stieglitz, Erika Strumberger, et al.

Nuclear FusionAug 18, 2026Plasma & Confinement AI, Modeling & Simulation

Recent experiments conducted during the 2025 campaign in ASDEX Upgrade have pushed the limits of negative triangularity shaping, achieving top triangularity values of δ top ∼-0.55 while keeping bottom triangularity around 0 in diverted plasmas. Guided by ideal-MHD stability calculations using the linear MHD BALOO solver and supported by TCV results, these experiments have demonstrated a transition to an edge that is more stable against peeling-ballooning modes, leading to ELM-free plasmas. Those ELM-free plasmas also exhibit a dithering behavior similar to limit-cycle oscillations. Linear MHD modeling showed that the achieved shapes remained marginal with respect to the second stability region for ballooning modes suggesting that with minor further shaping, a robust NT edge could be achieved. The associated energy confinement is generally reduced, with clear signs of strong power degradation. In addition, a high fraction of fast-ion energy is observed, particularly in high-power/low-density discharges. An approximately 15% improvement in normalized energy confinement is obtained when seeding nitrogen. As this effect comes from the edge, it highlights a path towards pedestal tailoring for further confinement optimization and supporting negative triangularity as a potentially viable reactor- relevant operational scenario in full-metal-wall devices.

Max Planck Institute for Plasma PhysicsSwiss Federal Institute of Technology LausanneMassachusetts Institute of Technology

Aug 17

Comparing HSX Reflectometer Measurements with a Full-Wave Synthetic Diagnostic

Henrique Oliveira Miller Hillebrecht, Michael Gerard, Michael Richardson, Gavin W Held, Gavin McCabe Weir, Benjamin J Faber, M J Pueschel, Xiang Han, Benedikt Geiger

Plasma Physics and Controlled FusionAug 17, 2026Plasma & Confinement Control & Diagnostics AI, Modeling & Simulation

A synthetic reflectometer diagnostic has been developed to model the reflectometer used at the Helically Symmetric Experiment (HSX) stellarator. The main purpose of the HSX reflectometer is to measure density fluctuations in order to study turbulence and coherent modes in HSX. A full-wave 2-D synthetic diagnostic modeling the HSX reflectometer has been used to investigate its radial localization, characterize its sensitivities, and relate gyrokinetic simulations to physical measurements. The synthetic diagnostic results show that signal localization peaks slightly outboard of the cutoff surface, with the main localization lobe spanning between 15% and 30% of the minor radius. The poloidal and radial wavenumber sensitivities are k θ < 1.7 cm -1 and k r < 6 cm -1 , respectively, with minimal variation in k r sensitivity across probing regions and density profiles. Conversely, k θ sensitivity increases to > 2 cm -1 in the core of the device, in accordance with an increase in local flux-surface curvature. Synthetic reflectometer signals scale linearly with density fluctuation amplitude over a large range of fluctuation levels. Multiple density fluctuation level estimators are tested to identify the optimal choice for use in HSX, with an equation from G.D. Conway Plasma Phys. Control. Fusion 1999 consistently performing best overall. Synthetic measurements of rescaled density fluctuations from a gyrokinetic simulation are then compared against experimental measurements, with coherent mode activity at 13 -15 kHz and 22 kHz observed in both experimental and synthetic frequency spectra. Synthetic spectra overestimated experimental spectra at low frequencies and underestimated at high frequency.

Max Planck Institute for Plasma PhysicsUniversity of Wisconsin Madison

Coupling time-dependent equilibrium evolution to 1D current diffusion equation towards integrated modeling of stellarators

Emiliano Fable, Fabian Solfronk, Elisa Buglione-Ceresa, David Kulla

Plasma Physics and Controlled FusionAug 17, 2026Plasma & Confinement AI, Modeling & Simulation

It is presented a framework in which time-dependent simulations of a 3D toroidal magnetized plasma (stellarator device) are self-consistently taking into account the variation of the magnetic equilibrium together with the plasma current and a background varying pressure inside a given plasma boundary shape (prescribed boundary simulations). The key result is to solve the issue of prescribing the degree of freedom present when performing prescribed boundary simulations with a 3D equilibrium solver, that is the anchor vacuum field not usually considered in this kind of framework.

Max Planck Institute for Plasma Physics

Uncertainties of magnetic equilibrium reconstructions

Rainer Fischer, Michael G Dunne, Joerg Hobirk, Tilmann Lunt, Wolfgang Suttrop

Plasma Physics and Controlled FusionAug 17, 2026Plasma & Confinement Control & Diagnostics AI, Modeling & Simulation

Uncertainties of quantities of magnetic equilibrium reconstructions are essential for the validation and quantification of estimated plasma parameters and their uncertainties. The uncertainties of a few equilibrium quantities can be evaluated analytically. A Monte-Carlo method is proposed to estimate the uncertainty of any equilibrium quantity. The Monte-Carlo method was verified using analytic formulas for the uncertainty of some equilibrium quantities and applied to various scalar, profile, flux-surface averaged and integral plasma quantities. Additionally, the uncertainty of the category of negative triangularity plasmas and alternative divertor configuration plasmas were evaluated.

Max Planck Institute for Plasma Physics

Identifying non-performing or Dud Plasmas for Burning Plasma Control: Insights from JET and TFTR Deuterium-Tritium Campaigns

Lidia Piron, Nicolò Ferron, Eric Fredrickson, Morten Lennholm, Alessandro Pau, Timo Ravensbergen, Olivier Sauter, Fulvio Auriemma, Matteo Baruzzo, Krassimir K. K Kirov, et al.

Plasma Physics and Controlled FusionAug 17, 2026Plasma & Confinement Control & Diagnostics

Among the burning plasma controllers for research fusion reactors, the dud detector will be of primary importance as it determines whether the plasma is performing well or if it is a dud. In the latter case, the discharge needs to be terminated to remain within tritium and neutron activation limits. To this scope, monitors which track the plasma performance will be integrated in the plasma control system. In this work, we present a novel dud detector that has been empirically identified based on Deuterium-Tritium campaigns carried out at JET and TFTR. This controller estimates a proxy of the neutron rate using a combination of the diamagnetic energy and the density peakedness. If the predicted neutron rate deviates from the real-time measurement, then the dud detector will trigger an alarm leading to a safe plasma termination if plasma recovery is not expected or, to actuator requests when adjustment is possible. This monitoring function can also be coupled with equilibrium solver and control-oriented models, such as RAPDENS, as proposed in the 15 MA plasma current, 5.3 T toroidal magnetic field baseline Deuterium-Tritium ITER scenario.

United Kingdom Atomic Energy AuthorityMax Planck Institute for Plasma PhysicsCulham Centre for Fusion EnergyConsorzio RFXPrinceton Plasma Physics Laboratory

Aug 14

The physics of ballooning-limited ELM-free regimes in EUROfusion tokamaks

Mike G Dunne, Michael Faitsch, Olivier Sauter, Eleonora Viezzer, Benoit Labit, Athina Kappatou, David Keeling, Branka Vanovac, Itziar Balboa, Petra Bilkova, et al.

Nuclear FusionAug 14, 2026Plasma & Confinement

The development of operational scenarios without large Type-I ELMs is of utmost importance for the stable operation and longevity of future tokamaks. The EUROfusion tokamak exploitation program has therefore made the understanding of ELM-free regimes a major topic of exploration across all its contributing devices (ASDEX Upgrade, JET, MAST-Upgrade, TCV, and WEST). An integrated program to investigate a range of Type-I ELM-free regimes has been developed covering the enhanced D-alpha (EDA), magnetic perturbations (MP), negative triangularity (NT), quasi-continuous exhaust (QCE), quiescent H-mode (QH), the baseline small ELMs (SE), I-mode, and X-point radiator (XPR) regimes. This contribution focuses on the development and understanding of the NT and QCE regimes on ASDEX Upgrade, JET, and TCV. The importance of transport via ballooning modes in both regimes is highlighted, as well as the progress in developing access models based on ideal-MHD. In the case of the QCE, this can also be expressed as a minimum separatrix density, which corresponds well to experimentally measured separatrix densities. Particular focus is paid to the performance of the QCE in terms of the achieved pedestal top values, which, when appropriately normalised, do not differ significantly from ELMy H-mode plasmas. This, combined with the predicted minimum separatrix density for the 15~MA ITER baseline plasma, highlight the relevance of the QCE as a potential operational scenario for both ITER and future reactors.

Culham Centre for Fusion EnergyMax Planck Institute for Plasma PhysicsSwiss Federal Institute of Technology LausanneCzech Academy of Sciences

First campaign with alternative divertor configurations in ASDEX Upgrade

Tilmann Lunt, Felix Albrecht, Matthias Bernert, Dominik Brida, Ralph Dux, Michael Faitsch, Tabea Gleiter, Sebastian Josef Hörmann, Joey Kalis, Bernd Kurzan, et al.

Nuclear FusionAug 14, 2026Plasma & Confinement Materials & Plasma-Facing Components

After a major hardware extension ASDEX Upgrade has now established and characterized a variety of alternative divertor configurations (ADCs) with heating powers reaching up to 20 MW and plasma currents up to 1 MA. These high performance conditions in configurations with arbitrarily small field line incidence angles were made possible due to the high precision in the tile alignment of 300 μm as well as a careful optimization of the error fields caused by the current feeds. The formation of an X-point radiator (XPR) was observed in a low-field side snowflake minus (LFS SF - ) configuration at a very low impurity concentration determined by the intrinsic sources only, i.e. without additional seeding of impurities. Compared to a typical lower single-null (SN) configuration the ELMs were found to be substantially smaller in size and higher in frequency and therefore barely detectable. While the primary strike line is fully detached the secondary one in the far-SOL shows heat fluxes of a similar magnitude as the primary strike line in a previous upper SN phase. These fluxes might be reduced by installing configuration-optimized baffles and/or by increasing the plasma current and thereby reducing λ q . In fact significantly smaller far-SOL heat fluxes were found when increasing the plasma current from 800 to 1000 kA. According to the divertor Langmuir probes the peak heat flux is then by a factor of two smaller than the ones in the SN reference.

Max Planck Institute for Plasma Physics

Aug 13

Direct comparison of 3D non-linear JOREK simulations of shattered pellet injection with ASDEX Upgrade experiments

Weikang Tang, Matthias Hoelzl, Paul Heinrich, Di Hu, F J Artola, Pascal de Marne, Mathias Dibon, Mike G Dunne, Ondřej Ficker, Peter Halldestam, et al.

Nuclear FusionAug 13, 2026Plasma & Confinement Control & Diagnostics AI, Modeling & Simulation

Shattered pellet injection (SPI) as primary mitigation method for major disruptions in ITER has a large parameter space available for optimization including the total amount of injected material, the size of the individual pellet fragments, the material composition, and the timing of multiple injections. This flexibility needs to be exploited to simultaneously minimize thermal heat loads, electromagnetic vessel forces, and formation of relativistic electrons and their impacts on plasma facing components. In this article, we apply 3D non-linear magnetohydrodynamic modelling to SPI experiments in the ASDEX Upgrade tokamak, going beyond our previous work [Tang et al Nucl. Fusion 65 116003 (2025)] by resolving some discrepancies between simulations and experiment and carrying out direct qualitative and quantitative comparisons to experimental measurements. The key element that enables the transition is the incorporation of the parallel heat-flux limit, which is done here in a simplified form. The work increases the confidence of reproducing key processes of disruption mitigation in direct 3D non-linear simulations in view of future predictive studies for ITER.

French Alternative Energies and Atomic Energy CommissionITER OrganizationMax Planck Institute for Plasma PhysicsInstitute of Plasma Physics of the Czech Academy of Sciences

Aug 11

Sheared poloidal plasma flows in the island divertor scrape-off layer of Wendelstein 7-X

Sean Bozkurt Ballinger, Seung Gyou Baek, Olaf Grulke, Carsten Killer, Floris Scharmer, Jim L Terry, Adrian von Stechow

Plasma Physics and Controlled FusionAug 11, 2026Plasma & Confinement Control & Diagnostics

Gas puff imaging (GPI) has been routinely operated since the installation of the water-cooled high heat flux divertor in the Wendelstein 7-X (W7-X) stellarator. GPI provides the first systematic, two-dimensional experimental characterization of perpendicular plasma flows in the magnetic island divertor scrape-off layer (SOL) on W7-X. Using spatiotemporal analysis of turbulent fluctuations, we find that, across magnetic configurations and plasma scenarios, poloidal flows ubiquitously dominate over radial flows, with typical poloidal velocities of order km/s. In contrast to tokamaks, radially propagating blob filaments are not observed, consistent with the predominantly normal fluctuation statistics measured in the island SOL. Net radial motion is detected only in the vicinity of an X-point of the magnetic island, which is accessed in a low-iota magnetic configuration by the present GPI field-of-view. In general, the poloidal flow profiles in the island SOL exhibit a complex structure, with multiple counter-propagating flow channels as narrow as 1 cm. The experimental parameters that order the observed poloidal flow structure near island O-points are investigated. Both the magnitude of the poloidal velocity and the number of shear layers, as well as their radial location, depend sensitively on the size and position of the magnetic islands, as well as on the toroidal current, plasma density, and toroidal field direction. These results demonstrate that cross-field transport in the W7-X island divertor SOL is correlated primarily with topology-driven ExB sheared flows, rather than by radial filamentary transport, and provide the first experimental documentation of the complex, shear-layer-dominated flow dynamics in the island divertor over a wide range of plasma conditions.

Massachusetts Institute of TechnologyTechnical University of DenmarkMax Planck Institute for Plasma Physics

Aug 10

Challenges of Locked Mode Avoidance in the Low-ν* ITER Baseline Scenario Plasmas in ASDEX Upgrade

Lidia Piron, Thomas Pütterich, Tomas Markovic, Olivier Sauter, Wolfgang Suttrop, Paolo Bettini, M Bonotto, Athina Kappatou, Marc Maraschek, Leonardo Pigatto, et al.

Plasma Physics and Controlled FusionAug 10, 2026Plasma & Confinement Control & Diagnostics AI, Modeling & Simulation

In AUG, the requirement to explore the low-ν* regime within the same ITER baseline scenario parameters, i.e. a Greenwald fraction of ~0.85, q₉₅ ≈ 3, βN ≈ 1.8, and H98 ≈ 1, has been addressed by inducing density pump-out through n=2 magnetic field perturbations. This approach required additionally relaxing the high-δ plasma shape to δ ≈ 0.1 to enable sufficiently strong density pump-out. However, at low density also the generation of locked modes was observed which deteriorates the plasma performance. Several strategies were therefore tested, including model-based error field correction combined with different heating schemes, as well as a core–edge decoupling metric derived from the GPEC code. Nevertheless, these attempts were unsuccessful, largely because the gap between the onset of density pump-out and the excitation of the core mode is very small.

Max Planck Institute for Plasma PhysicsSwiss Federal Institute of Technology Lausanne

Aug 5

FIREFLY: heat load and particle exhaust approximations for rapid evaluation of divertor designs

Heinke Frerichs, Dieter Boeyaert, Yuhe Feng, Detlev Reiter

Plasma Physics and Controlled FusionAug 5, 2026Plasma & Confinement Tritium & Fuel Cycle AI, Modeling & Simulation

The divertor in a magnetic confinement fusion reactor is an essential component for power dissipation and particle removal. This article introduces the FIREFLY package for rapid evaluation of divertor designs based on an extension of the FLARE code for field line reconstruction from an unstructured flux tube mesh. First, divertor heat loads are approximated with a simplified heat transport model. Neutralized particles are then sampled from the resulting 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.

University of Wisconsin MadisonMax Planck Institute for Plasma Physics

Aug 3

Development of tungsten actively cooled divertor target plasma facing components for W7-X and JT-60SA fusion devices

Marianne Richou, Mehdi Firdaouss, Joris Fellinger, Thierry BAFFIE, Bernd Böswirth, Mariano Di Bartolomeo, Diogo Dias Alexio, Daniel Dickes, Daniel Dorow-Gerspach, Martin Draksler, et al.

Nuclear FusionAug 3, 2026Materials & Plasma-Facing Components Manufacturing & Supply Chain

The present study describes the ongoing developments for plasma-facing components (PFCs) intended for future fusion devices, requiring PFCs to handle high heat loads of at least 10 MW/m² in the divertor region. tungsten (W) is chosen as the armor material due to its resistance to plasma-wall interactions and high heat fluxes (HHF). Significant efforts are underway on various fusion devices (WEST, EAST, KSTAR…) to test and optimize water-cooled W-based PFCs. For W7-X, new PFCs are being developed, using W armor material placed at the divertor target region. The design aims to simplify manufacturing, inspection, and installation processes compared to the currently operated actively cooled carbon fiber composite based divertor, while meeting thermal and mechanical constraints. The developed concept and related manufacturing choices are promising since, we show that it is able to sustain the required heat loads. Also, the JT-60SA tokamak plans a transition to metallic PFCs after 2029. The W monoblock concept, successfully used in WEST, EAST and KSTAR and planned for ITER, is being adapted for JT-60SA, and is currently the reference concept. However, some advanced designs exploring cost-saving measures and enhanced heat transfer capabilities are also under development. Manufacturing processes, such as laser powder bed fusion (LPBF) and hot isostatic pressing (HIP), are optimized to ensure reliable cooling channels for the development of the proposed enhanced designs. For these developments, thermo-hydraulic analyses and HHF tests have shown promising results, with surface temperatures being in agreement with models and material temperature limits. These developments lead to show a successful use of combined advanced manufacturing processes (additive manufacturing…), design hypotheses and adapted modelling tools to propose components with relevant performances.

Max Planck Institute for Plasma PhysicsJozef Stefan InstituteFrench Alternative Energies and Atomic Energy Commission

Hybrid Kinetic-MHD Simulations of Drift-Orbit Effects on the Stability and Non-Linear Dynamics of Runaway Electron Beams

Shi-Jie Liu, Hao-Wei Zhang, Hannes Bergstroem, Matthias Hoelzl, the JOREK Team

arXiv (physics.plasm-ph)Aug 3, 2026Plasma & Confinement AI, Modeling & Simulation

During tokamak disruptions, the Ohmic current may be replaced by a non-inductive runaway electron (RE) current, affecting resistive stability. Previous studies suggest that, in the linear phase, the presence of REs acts destabilizing for tearing modes (TM) compared to a scenario with Ohmic current. In the non-linear regime, this translates to larger saturation amplitudes. These results are based on the assumption of zero drift-orbit deviation from the magnetic flux surfaces corresponding to the low-energy limit. This work investigates the importance of this kinetic effect by studying the linear and non-linear TM dynamics in RE beams with different RE energies, providing a clear picture of finite-orbit-width (FOW) effects. We use a hybrid fluid-kinetic model in the 3D non-linear magnetohydrodynamic (MHD) code JOREK, treating REs kinetically with a full-f Monte Carlo approach in self-consistent interaction with the MHD mode dynamics. The study shows that the presence of REs modifies the characteristics of the instability in several ways. First, we find that the major-radial displacement of drift orbits from flux surfaces induces an $m=1$ perturbation to the equilibrium current, introducing additional mode coupling between $(m,n)$ instabilities and the $(m\pm1,n)$ sidebands. Second, we find that increasing RE energy has a stabilizing effect on the MHD modes because REs cannot support narrow current sheets on rational flux surfaces owing to the drift-orbit displacement. This counteracts the destabilizing effect that REs have on TMs in the low-energy limit. For the scenario investigated, the stabilizing effect dominates over the additional mode coupling, reducing the development of stochastic magnetic regions with increasing RE energy and thereby lowering radial particle transport. Overall, we find that FOW effects can substantially alter the MHD stability and non-linear dynamics of RE beams.

Max Planck Institute for Plasma Physics

Geometric numerical discretisation of electromagnetic quasineutral models

Nishant Narechania, Emil Poulsen, Eric Sonnendrücker

Journal of Plasma PhysicsAug 3, 2026AI, Modeling & Simulation

In this work, the geometric electromagnetic particle-in-cell (PIC) framework, GEMPICX , is extended to solve the quasineutral, fully kinetic Vlasov–Maxwell equations on dual grids using mimetic finite differences. The discrete action principle is derived, taking into account the duality between the grids. The temporal derivative of the electric field does not directly appear in the dynamical system for the quasineutral model. Hence, a discretised curl–curl equation is used to implicitly obtain the electric field at every time-step. This also circumvents the need to obtain electric potentials. A Lagrange multiplier is used to maintain the discretised divergence of the current density at machine zero.

Max Planck Institute for Plasma Physics