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Alexander Huber (Institute of Fusion Energy and Nuclear Waste Management–Plasma Physics, Forschungszentrum Jülich GmbH)18/05/2026, 10:40A. Physics Processes at the Plasma Material InterfaceInvited
Tungsten (W), chosen as the plasma-facing material for ITER, offers high thermal robustness and low tritium retention, but its erosion and penetration into the plasma core can degrade the confinement through dilution and radiation. The net W content depends on the interplay of erosion, divertor screening and transport, which vary with plasma species. To resolve these dependencies, we...
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Athina Kappatou (Max-Planck-Institut für Plasmaphysik, Garching, Germany)18/05/2026, 11:50E. Impurity Sources, Transport and ControlInvited
High-Z impurities such as tungsten must be kept out of the plasma, to avoid excessive radiation or plasma collapse. Previous assessments of the neoclassical impurity transport expected at the edge of ITER plasmas has indicated favorable outward convection of tungsten, due to the expected high pedestal ion temperatures and lower density gradients [R. Dux et al, PPCF 2014; NME 2017] leading to...
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Richard Pitts (ITER Organization)18/05/2026, 13:20J. Plasma Exhaust and Plasma Material Interactions for Fusion ReactorsInvited
The ITER 2024 re-baseline defines a revised approach to reach the main burning plasma objective. A key component of this strategy is the switch from Be to W main wall armour, making ITER a full-W device from the beginning of operations. In fact, the new wall becomes two walls, with a Temporary First Wall (TFW) in place for the “Start of Research Operations” campaign (SRO) in the revised ITER...
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Jörg Hobirk (MPPL)18/05/2026, 14:50E. Impurity Sources, Transport and ControlInvited
The ITER Research Plan re-baseline [1], especially the change of first wall (FW) material from beryllium to tungsten (W) has required the re-validation of the initial ramp-up phase in limiter configuration. The direct plasma contact on W surfaces, without the screening of eroded atoms provided by diverted operation, is expected to lead to high radiative ractions of about f rad = 70-80% as...
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Dr Hui Wang (Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences)18/05/2026, 15:20E. Impurity Sources, Transport and ControlInvited
Kinetic effects are found to significantly diminish the thermal force on W relative to conventional fluid modeling, underscoring the necessity of kinetic approaches for modeling W transport in boundary plasmas. A recently developed kinetic impurity transport model in DIVIMP has been used to investigate kinetic effects on W transport and screening across various divertor conditions. Results...
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Jonathan Gaspar (Aix Marseille Univ., CNRS, IUSTI)19/05/2026, 08:40G. Power Exhaust, Plasma Detachment and Heat Load ControlInvited
The plasma facing components of next step fusion devices will handle unprecedented heat flux and particle fluence. The WEST tokamak, equipped with an actively cooled tungsten ITER grade divertor, aims to assess the divertor performance under tokamak conditions. A first high fluence campaign was performed in WEST, in 2023, on the new actively cooled tungsten divertor based on the ITER-grade...
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Marta Pedrini (GNOI)19/05/2026, 10:20I. Plasma Edge and First Wall DiagnosticsInvited
We present a systematic quantification of the energy balance during controlled runaway electron (RE) beam terminations in the TCV tokamak.
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Disruptions in tokamaks can generate relativistic REs capable of carrying a large fraction of the plasma current at multi-MeV energies. A sudden loss of confinement of the REs can deposit extreme energy densities onto plasma-facing components (PFCs),... -
Srikanth Sureshkumar (CEA)19/05/2026, 11:50H. Far SOL Transport and Plasma Wall Interaction in Main ChamberInvited
The new ITER baseline in which tungsten (W) replaces beryllium as the chosen first wall (FW) material reinforces the need for plasma backgrounds for the analysis of W sources and transport. Crucial in this context are the FW particle/heat fluxes, which depend on the far scrape-off layer (SOL) radial transport profile. This is still very much uncertain, but there is strong evidence from current...
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Ray Chandra (Aalto University, Espoo, Finland)19/05/2026, 13:20G. Power Exhaust, Plasma Detachment and Heat Load ControlInvited
The inclusion of hydrogenic Lyman line radiation absorption in drift-enabled SOLPS-ITER [1] modeling of JET-ILW L-mode hydrogen and deuterium discharges [2,3] resulted in improved predictions of the divertor target particle fluxes by 30% and higher heat fluxes at the onset of detachment. Lyman line absorption changes the plasma density distribution in the divertor volume in detachment and is...
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Hyungho Lee (KFE)19/05/2026, 13:50G. Power Exhaust, Plasma Detachment and Heat Load ControlInvited
Since the installation of the actively cooled lower tungsten (W) divertor in 2023, KSTAR has conducted four experimental campaigns (>7,200 discharges; avg. 6.5–7 MW, ~10 s) addressing critical plasma-surface interactions and optimizing divertor performance. Previously, KSTAR operated with carbon wall for 15 years (~32,800 discharges). This overview synthesizes key physics and engineering...
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Rui Ding (Institute of Plasma Physics, Chinese Academy of Sciences)19/05/2026, 15:00F. Edge and Divertor Plasma PhysicsInvited
Dedicated EAST experiments and modeling have been performed to investigate the influence of deuterium (D) and helium (He) plasmas on divertor detachment and tungsten (W) impurity transport. As fusion reactor will be operated in the D-T plasma with He naturally existing as the product of D-T reaction, the influence of different ion species on divertor detachment remains unclear. Density ramp-up...
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Dr Hao YANG (M2P2 Aix-Marseille Univ, CNRS, Centrale Méditerranée, 13013 Marseille, France)19/05/2026, 15:30G. Power Exhaust, Plasma Detachment and Heat Load ControlInvited
Recent experiments on the WEST tokamak show that, with sufficient nitrogen seeding, a stable X-point radiator (XPR) can be sustained above the X-point for about 70s. This regime helps to reduce divertor heat loads by about 90% and tungsten divertor sources by up to 98% while enhancing core energy confinement by about 25%, highlighting the XPR’s potential for future fusion power plants.
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Gen Motojima (National Institute for Fusion Science)20/05/2026, 08:40C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionInvited
The Large Helical Device (LHD) experiments will conclude in December 2025, marking nearly three decades of divertor research. This milestone provides a unique opportunity to summarize the achievements and lessons learned from the development of the helical divertor concept and its role in steady-state stellarator operation.$\newline$
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Early LHD experiments demonstrated that the open helical... -
Sophie Gorno (ORNL)20/05/2026, 10:20C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionInvited
Experiments and modeling in DIII-D highlight the key parameters determining the dynamic change in divertor conditions with pellet fueling. In future reactors, pellet injection will be necessary for effective core fueling[Kukushkin2003NF] since a high-performance scenario typically necessitates a scrape-off layer (SOL) that is opaque to neutrals. Due to increasing alpha heating with density,...
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Yuki Hayashi (Graduate school of Frontier Sciences, The University of Tokyo)20/05/2026, 11:10G. Power Exhaust, Plasma Detachment and Heat Load ControlInvited
Edge localized modes (ELMs) deliver bursts of energy and particles that can exceed the survivability limits of plasma-facing components. While plasma detachment provides a promising scenario for reducing steady-state heat fluxes, its compatibility with transient events remains poorly understood. It is unclear whether detached plasmas can sustain their protective role under ELM-like pulses or...
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Stepan Krat (NRNU MEPhI)21/05/2026, 09:00B. Material Erosion, Migration, Mixing, and Dust FormationInvited
Tritium accumulation is going to be a serious safety issue in future fusion devices. Accumulation in co-deposited layers is one of the main channels of hydrogen isotope accumulation in tokamaks, at least for low Z-materials [1]. Even W layers can contain up to 5-10 at.% of hydrogen, with some works reporting up to 20 at.% D content [2]. Their properties vary greatly with deposition conditions,...
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Filippo Scotti (Lawrence Livermore National Laboratory)21/05/2026, 10:20G. Power Exhaust, Plasma Detachment and Heat Load ControlInvited
Stable X-point radiation (XPR) was obtained via impurity seeding in DIII-D experiments ($I_p$=0.8-1.3MA, $P_{inj}$=6-12MW) for characterization of XPR access, validation of radiation stability models, and assessment of impact of XPR on H-mode pedestal. XPR regimes [1] have gathered interest for future devices thanks to the simultaneous elimination of steady-state and transient heat fluxes...
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Robert Lunsford (Princeton Plasma Physics Laboratory)21/05/2026, 13:10D. Wall Conditioning and Tritium Removal TechniquesInvited
The PPPL Impurity Powder Dropper (IPD) represents a novel capability for simultaneously controlling main ion recycling, modifying edge plasma conditions, and managing material surfaces in reactor-relevant environments. Coordinated experimental campaigns across multiple plasma confinement devices over a range of magnetic configurations have demonstrated that controlled low-Z particulate...
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Hanna Schamis (Princeton Plasma Physics Laboratory)21/05/2026, 13:40D. Wall Conditioning and Tritium Removal TechniquesInvited
Boron powder injected into KSTAR lower single null, H-mode discharges with a W monoblock divertor reduced impurity influx from the plasma-facing components (PFCs). The wall conditioning effect provided by the B powder injection resulted in a ~40% reduction in radiated power, ~30% reduction in core electron density, and ~20% reduction in $Z_{eff}$. W-I and O-II line emission brightness was also...
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Sehoon An (Max-Planck-Institut für Plasmaphysik, Garching, Germany)21/05/2026, 14:30D. Wall Conditioning and Tritium Removal TechniquesInvited
Glow-discharge boronisation (GDB) is a standard wall-conditioning technique in fusion devices to ensure reliable plasma operation. With ITER’s re-baseline decision to adopt tungsten (W) as the main-chamber material, GDB is foreseen as a key method to achieve favourable plasma–wall conditions in its full-W environment. To enable reliable extrapolation to ITER, dedicated studies on existing...
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Dr Jiansheng hu (Institute of plasma physics, HIPS, Chinese academy of sciences, 230031, Hefei, China)21/05/2026, 15:00D. Wall Conditioning and Tritium Removal TechniquesInvited
Full metal walls are a high priority choice for future fusion reactors like ITER and CFETR, but they present significant challenges in plasma-wall interaction and high-Z impurity control. Low-Z material coatings (e.g., lithium, boron) are commonly employed to improve plasma confinement and achieve H-mode, but they add considerable complexity to reactor operation and maintenance due to tritium...
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Massimo Carpita (SPC-EPFL)22/05/2026, 09:00G. Power Exhaust, Plasma Detachment and Heat Load ControlInvited
This work investigates stationary power exhaust in the X-Point Target (XPT) divertor, combining experiments in the TCV tokamak with SOLPS-ITER simulations. Power exhaust is a major challenge for magnetic confinement fusion: future reactors will face intense heat fluxes channelled through a narrow scrape-off layer onto divertor targets, exceeding material tolerances if unmitigated. Detached...
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Wladimir Zholobenko (Max Planck Institute for Plasma Physics)22/05/2026, 13:20F. Edge and Divertor Plasma PhysicsInvited
Divertor detachment is mandatory for fusion reactors and must be reconciled with good confinement. While the H-mode remains attractive for its high confinement, heat and particle loads on plasma facing components must be mitigated, which result from ELM bursts and the narrow inter-ELM SOL width.
We present novel GRILLIX simulations of the Quasi-Continuous Exhaust (QCE) and detached X-Point...
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Peter John Ryan (UKAEA)22/05/2026, 14:30G. Power Exhaust, Plasma Detachment and Heat Load ControlInvited
In the near scrape-off layer (SOL), heat transport is dominated by parallel electron conduction, resulting in a radial heat flux decay length ($\lambda_q$) that is short relative to the machine size. The cross-field extent of the power entering the divertor region is set by $\lambda_q$, which strongly affects divertor performance, including the peak heat flux and access to detachment. In the...
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