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Loic Corso (CINaM - Aix Marseille University)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
In fusion devices like WEST, ASDEX Upgrade, EAST and ITER, tungsten (W) has been chosen as the plasma facing material for the divertor, where the heat and particles fluxes are the most intense. In particular, helium (He) irradiation leads to the formation of nano-sized bubbles in the subsurface area, which increase hydrogen isotopes retention. More dramatically, W-fuzz may form, presumably...
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Robert Kolasinski (Sandia National Laboratories)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
We examined the effects of divertor plasmas on 14 distinct tungsten and ultra-high temperature ceramic (UHTC) materials, providing insight into how combined high heat and particle fluxes affect their surface composition and structure. The experiments were carried out using the Divertor Materials Evaluation System (DiMES) at DIII-D. The test matrix included commercially available tungsten...
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Michael Simmonds (University of California San Diego)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
The ability to quantify irradiation damage and recovery of thermal transport and elastic response in tungsten is essential for understanding and accurately predicting stress evolution, heat management, and even failure modes in fusion plasma-facing components. Transient grating spectroscopy (TGS) non-destructively quantifies material properties in the first few microns [1]. Whereas previous...
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Benjamin Burazor Domazet (TU Wien)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
In fusion devices, the plasma-facing wall is primarily eroded via physical sputtering, which introduces impurities into the plasma core that degrade device performance. Aiming to gain a fundamental understanding of the sputtering processes relevant for plasma-wall interactions with machined plasma-facing components, we investigate angular distributions of sputtered particles in a controlled...
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Youpeng Wang (University of Basel)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
The Core Plasma Thomson Scattering (CPTS) diagnostic system is designed to precisely measure the electron temperature and density in the ITER core plasma region. During the fusion plasma operation, the metallic first mirror in the CPTS first mirror unit (FMU) is expected to degrade gradually due to plasma exposure, which requires radio-frequency (RF) plasma cleaning for the periodic CPTS...
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V Kotov18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
Hydrogen plasma chemistry plays crucial role for both power and particle exhaust in low temperature detached divertor plasmas as it can give rise to molecular assisted recombination via $H_2^+$, $H^-$, $H^+_3$. In the present contribution the application of the linear plasma device PSI-2 for validation of the corresponding plasma-chemical models will be discussed.
The effect on which the...
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Torben Schmitz (ZJFJ)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
Exposing a surface to an ion beam or hot plasma leads to erosion and the development of surface structures on the nanoscale. Such nanostructures have been observed on tungsten samples exposed to plasma in the PSI-2 linear plasma device and the LHD stellarator. Existing studies show that these nanostructures can have an influence on the erosion process of plasma facing components (PFCs). Better...
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yukinori hamaji (national institute for fusion science)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
The application of flowing liquid metal for plasma-facing components (PFCs) in fusion reactors has attracted interest due to its potential heat removal capabilities, self-healing surface, and the expectation of improved plasma confinement. However, research on the interaction between liquid metal and plasma in well-controlled laboratory-scale experiments is not sufficient to understand the...
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Matej Mayer (MPPL)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Since the operational period OP 2.1 the stellarator Wendelstein 7-X (W7 X) operates with a CFC divertor, graphite tiles, and a mostly stainless-steel outer wall. Boronizations are used for wall conditioning in order to reduce the amounts of oxygen and water.
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The divertor strike line is a net erosion area [1, 2]; the amount of eroded carbon depends on the concentration of oxygen in the... -
Jeremy Mateja (University of Tennessee-Knoxville)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Build-up of plasma facing component (PFC) debris, a.k.a. “PFC slag”, is a potentially serious concern for the next generation of the magnetic fusion devices. A PFC slag management experiment on DIII-D tokamak tested the ability of strike point sweeps at cleaning low-Z slag from the outer divertor shelf. Layers of enriched boron isotope B10 were deposited on the outer target with the impurity...
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Raphael Gurschl (TU Wien)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Plasma-facing components in fusion reactors are exposed to extreme conditions including high heat flux and bombardment by energetic particles. Tungsten, the chosen first wall material in ITER, offers favorable properties such as a high melting point and low sputter yield. To enhance plasma performance and minimize impurity influx such as oxygen, wall conditioning techniques like boronization...
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Dr Ionut Jepu (UKAEA, Culham Campus, Abingdon, OX14 3DB, UK)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Runaway electrons (REs) generated during plasma disruptions in tokamak reactors present a challenge due to their capacity to induce severe damage to the plasma facing components (PFCs) [1]. Investigations on the Joint European Torus (JET), operating with the metallic wall (formerly known as ITER Like Wall – ILW) [2] between 2010 and 2023 demonstrated the damaging effect of REs on beryllium...
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Marc Sackers (Forschungszentrum Jülich GmbH, Institute of Fusion Energy and Nuclear Waste Management - Plasma Physics, 52425 Jülich, Germany)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Re-baselining ITER to have a full tungsten wall eliminates the strong impurity gettering capabilities of beryllium. Consequently, the intrinsic oxygen level makes it challenging to start-up the plasma in limiter configuration and achieve high-performance operating conditions in divertor configuration. The selected procedure in ITER to getter oxygen is boronization [1] by depositing a thin...
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Noriyasu Ohno (Nagoya University)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Helium(He)–tungsten(W) co-deposited layers are expected to form in future magnetic fusion devices as a consequence of simultaneous sputtering and re-deposition of tungsten (W) under mixed hydrogen and He plasma exposure. Understanding their sputtering behavior is essential for predicting long-term material erosion and impurity sources in ITER and DEMO. In this study, He–W co-deposited layers...
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Mykola Ialovega (GenF)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
GenF, a spin-off of the Thales Group, is developing an accelerated pathway toward a direct-drive inertial confinement fusion (ICF) reactor. As part of the TARANIS project, GenF collaborates with French research institutions including CEA and CNRS to address the critical physics problems in ICF physics.
One of the most demanding challenges is the design of the ICF reaction chamber,...
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Samuli Saari (VTT)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Net erosion of tungsten (W) in the divertor region has been investigated by exposing small platinum (Pt) marker samples, as well as bulk molybdenum (Mo) and W samples, to a series of plasma discharges in the outer strike point (OSP) region during dedicated L- and H-mode experiments in ASDEX Upgrade (AUG). Pt was chosen as a proxy for W due to its comparable erosion characteristics, addressing...
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Gabriele Alberti (Politecnico di Milano)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
The Divertor Tokamak Test (DTT) facility, currently under design and construction at the ENEA Research Centre in Frascati (Italy), aims to assess alternative solutions to the heat and power exhaust challenge in future fusion reactors [1]. A key issue for its operation is plasma-wall interaction (PWI) [2], as material erosion can limit component lifetime while eroded impurities may migrate into...
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Fan Feng (Southwestern Institute of Physics)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
In fusion reactors, the divertor fulfills two core functions: first, expelling impurities produced by plasma-first wall interactions and helium (a fusion product); second, withstanding high heat flux from the plasma and dissipating plasma energy out of the tokamak device. During operation, the divertor endures extreme thermal loads. As one of the most promising candidate materials for future...
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clément Monet-Vidonne (Aix -Marseille Univ., CNRS, IUSTI)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
During the WEST experimental campaigns, various evolutions of the plasma-facing components (PFCs) have been observed. One of these degradations is erosion of material and redeposition on other components. These deposits have unknown but probably degraded thermal properties (low diffusivity and/or low thermal contact with the PFC). Some of these deposits created Unidentified Flying Object (UFO)...
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Chandra Prakash Dhard (Max-Planck-Institut für Plasmaphysik, Greifswald, Germany)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Tungsten (W) has emerged as a favourable material for plasma-facing components (PFCs) in nuclear fusion devices. It has been incorporated in several tokamaks, however, in the stellarator, with 3D geometry, its suitability as PFCs, in terms of erosion, redeposition, ionization, transportation and accumulation of impurity particles in the plasma core, is yet to be demonstrated. In the...
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D.L. Rudakov (University of California, San Diego)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Controlling the Plasma-Material Interactions (PMI) is one of the key issues to be resolved for the success of machines like ITER and SPARC and the following generation of the magnetic fusion devices. The Divertor Material Evaluation System (DiMES) was a workhorse of PMI research at the DIII-D tokamak for about three decades and still remains one of the leading material testing facilities in...
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Dr Andreas Kirschner (FZJ)18/05/2026, 16:10B. Material Erosion, Migration, Mixing, and Dust FormationPoster
Tungsten as high-Z material has a relatively low physical sputtering. Chemical erosion – if occurring at all – is negligibly small. In addition, tungsten has a very large melting point of about 3400°C. However, the core plasma tungsten concentration in a fusion device has to be kept at extremely low values around 3E-5 to minimize plasma dilution and in particular plasma cooling. Therefore, it...
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Christoph Kawan (FZJ)18/05/2026, 16:10C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Picosecond-Laser-Induced Ablation Quadrupole Mass Spectrometry (ps-LIA-QMS) was employed to investigate deeply located deuterium retention in polycrystalline tungsten (W). As the foreseen primary material for fusion reactors, its high heat conductivity, low sputtering yield, and good fuel retention properties make it a promising choice as first wall material. However, high-energy particles,...
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Andreas Theodorou (MPPL)18/05/2026, 16:10C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
A critical challenge in the fusion materials science lies in understanding the influence of radiation damage on hydrogen isotopes (HIs) retention, particularly concerning safety and tritium self-sufficiency. Reduced activation ferritic/martensitic (RAFM) steels are leading candidates for use as structural material in the first wall such as the breeding blanket modules. In deuterium-tritium...
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Anurag Maan (Princeton Plasma Physics Laboratory)18/05/2026, 16:10C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
We show that beam fueling of a low recycling discharge with low energy neutral beams is feasible with modest (~ 5%) lithium impurity concentrations. Low recycling discharges (R ~ 0.5) on LTX-beta have been documented with a near flat electron temperature profile. Edge temperature measurements in the SOL indicate a hot and sparse SOL with collisionality < 0.1 and values as low as 0.01. Low...
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Rongxing yi (FZJ)18/05/2026, 16:10C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Hydrogen (H) isotopes retention in plasma-facing materials is a critical issue for nuclear safety in fusion devices operating with a deuterium-tritium mixture and tungsten plasma-facing components. Its reliable detection is of great importance for both safety and material characterisation after plasma exposure. However, we observed that in vacuum conditions, the Balmer alpha spectra (Ha)...
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Mr Alexander Feichtmayer (Max Planck Institute for Plasma Physics)18/05/2026, 16:10C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
The development of suitable materials for structural and plasma-facing components is a decisive challenge on the path towards nuclear fusion as a future energy source. In particular, the transport and retention of hydrogen isotopes is of relevance as they have a significant influence on the tritium inventory. The Soret effect describes diffusion processes driven by a temperature gradient,...
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Dr Laurent COLAS (CEA, IRFM, F-13108 St-Paul-Lez-Durance, France)18/05/2026, 16:10C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
In present-day magnetic fusion devices, hydrogen serves as a minority species for Ion Cyclotron Resonant Heating (ICRH). The efficiency of the heating scenarios depends on the hydrogen fraction in the plasma core that should be controlled within a few percents. Plasma-Facing Components (PFCs) release/trap hydrogen into/from the discharge, thereby complicating this control. Hydrogen increases...
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Maria Popova (Forschungszentrum Jülich GmbH)18/05/2026, 16:10C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
Tungsten is the leading candidate for plasma-facing components in fusion devices. In future fusion reactors, neutron irradiation creates material defects that trap hydrogen isotopes, including the fusion fuel tritium and deuterium. Monitoring the hydrogen inventory is mandatory for nuclear safety and important for the efficient use of fuel. Laser-based diagnostic methods such as Laser-Induced...
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Tom Wauters (ITER Organization (IO))18/05/2026, 16:10C. Plasma Fueling, Particle Exhaust and Control, Tritium RetentionPoster
While ITER is a full-tungsten (W) tokamak, a substantial surface area of water-cooled stainless steel remains exposed to charge-exchange neutrals (CXN) originating from the plasma. Hydrogen isotopes implanted in steel can permeate rapidly and subsequently be released into the cooling water, where inventory limits are imposed by regulatory requirements. This work presents a sensitivity study...
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Dr Tomi Vuoriheimo (University of Helsinki)18/05/2026, 16:10D. Wall Conditioning and Tritium Removal TechniquesPoster
The replacement of the originally planned beryllium first wall with tungsten in the current ITER baseline makes boronization an important strategy for reducing impurities such as oxygen in the plasma fuel. Consequently, the effects of boron on plasma-material interactions during and after plasma operations are of high importance. Although hydrogen isotope retention in tungsten has been widely...
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Anne Houben (Forschungszentrum Jülich GmbH, Institute of Fusion Energy and Nuclear Waste Management - Plasma Physics, 52425 Jülich, Germany)18/05/2026, 16:10D. Wall Conditioning and Tritium Removal TechniquesPoster
Due to the change of first wall material in ITER from Be to W, a glow discharge boronization (GDB) system is included in the re-baseline in order to guarantee efficient plasma operation (https://doi.org/10.1016/j.nme.2024.101854). Even though the GDB is used in many fusion devices for decades, its conditioning effect and possibility for recovering fuel from the B layers was not studied in...
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Ralph Dux (MPI for Plasmaphysics, Garching, Germany)18/05/2026, 16:10D. Wall Conditioning and Tritium Removal TechniquesPoster
During boronisation glow discharges, the spectral radiance on several lines-of-sight (LOS) through the vessel was measured for the complete visible spectrum in the range from 350-730 nm. The spectroscopic setup was absolutely calibrated by illuminating the respective optical heads inside the vessel with an integrating sphere of known spectral radiance. The glow discharges were run at the...
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Fotios Maragkos (MPPL)18/05/2026, 16:10D. Wall Conditioning and Tritium Removal TechniquesPoster
Wall conditioning improves plasma performance in fusion devices by reducing the amount of impurities, especially carbon- and oxygen-based impurities, in the plasma [1,2]. Standard wall-conditioning procedures include baking and glow-discharge cleaning (GDC) of the inner walls using hydrogen or helium plasmas for removing impurities contained at the plasma facing components (PFCs). The amount...
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Sebastijan Brezinsek (ZJFJ)18/05/2026, 16:10D. Wall Conditioning and Tritium Removal TechniquesPoster
Boronization is an established techniques to condition the first wall of fusion devices. Boronization systems are designed to deposit a ~10-100 nm thin boron (B) layer on PFCs by injecting diborane (B₂H₆/B₂D₆) mixed with a carrier gas (He, H₂, D₂) into a cleaning Glow Discharge (GD). For optimal boronization performance, thus, achieving a homogeneous layer is desirable...
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Alessandro Bortolon (Princeton Plasma Physics National Laboratory)18/05/2026, 16:10D. Wall Conditioning and Tritium Removal TechniquesPoster
Boronization, a process involving coating of the plasma facing components (PFCs) with boron (B) either by glow discharge (GDB) or solid boron injection (SBI), is an established way of improving operation of tokamaks and stellarators due to medium and high-Z impurity level reduction. However, a quantitative prediction of the amount of B required on next-step fusion devices such as ITER or a...
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Frederik Henke (MPPL)18/05/2026, 16:10E. Impurity Sources, Transport and ControlPoster
The island divertor concept implemented at Wendelstein 7-X (W7-X) is one of the most extensively investigated solutions for power and particle exhaust in future quasi-isodynamic stellarator power plants. In the latest experimental campaign (OP2.3), W7-X demonstrated successful feedback control of radiative detachment via impurity seeding, based on real-time bolometric measurements of the total...
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Prof. Vladimir Rozhansky (SPbPU)18/05/2026, 16:10E. Impurity Sources, Transport and ControlOral
Impurity transport inside the separatrix of a tokamak is considered to combine neoclassical and anomalous properties. If $L_{n_i}$, $L_{T_i}$ are main plasma density and ion temperature radial characteristic lengths and $L_{T_i}$<1/2 $L_{n_i}$ the impurity radial convective flux is outward [1]. However, for strong gradients in the edge transport barrier it’s not applicable [4] and the...
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Curtis Johnson (ORNL)18/05/2026, 16:10E. Impurity Sources, Transport and ControlPoster
Prompt redeposition of tungsten plays a central role in impurity transport, main-chamber erosion, and net wall evolution in full-metal fusion devices. Yet experimental quantification of prompt redeposition remains limited due to the diagnostic challenges of ultraviolet (UV) tungsten spectroscopy and the need for reliable atomic physics data. This work presents spectroscopic measurements,...
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Alex GROSJEAN (University of Tennessee - Knoxville (UTK))18/05/2026, 16:10E. Impurity Sources, Transport and ControlPoster
A series of experiments with plasma shaping scans were conducted in WEST and KSTAR illuminating the respective role of lower divertor impurity sourcing (high impurity production and screening) compared to the non-divertor regions (lower impurity production and screening) and the non-intuitive resulting impact on W core contamination. Since the new ITER baseline was presented in 2024, W has...
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Tyler Abrams (General Atomics)18/05/2026, 16:10E. Impurity Sources, Transport and ControlPoster
The DIII-D National Fusion Facility is preparing for a transition from graphite to tungsten (W) plasma-facing components to enable reactor-relevant studies of plasma scenarios, core-edge integration, and plasma–material interaction for next-step fusion devices.
Leveraging DIII-D’s unique capability for short-pulse, high-performance operation with flexible shaping, advanced actuator...
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Dr Volodymyr Bobkov (Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching, Germany)18/05/2026, 16:10E. Impurity Sources, Transport and ControlPoster
Following the re-baselining of ITER aimed at operation with full tungsten wall, a renewed interest in studies of the impurity sources during operation of Ion Cyclotron Range of Frequencies (ICRF) antennas emphasizes the importance of validation of the modelling tools simulating the sheath-rectified DC potentials $V_{DC}$. We use the SSWICH-SW code coupled with RAPLICASOL or TOPICA, in order to...
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Nicolas RIVALS (CEA)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
Among the potential reactor-relevant divertor regimes of operation, the “X-Point Radiator” (XPR) regime [1] which features a radiating MARFE at the X-Point and strongly reduced target heat loads, is a strong contender. This regime is investigated extensively in the WEST tokamak on the actively cooled ITER-grade tungsten divertor [2], over a wide range of operational parameters (density, power,...
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Dick Majeski (PPPL)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsOral
An overview of the effects of low recycling lithium walls, plasma-limiting surfaces, and divertor targets on the scrape-off layer (SOL) and core confined plasmas is presented. The discussion will primarily reference the tokamak, although consequences for alternative magnetic confinement configurations (primarily stellarators and mirrors) will also be discussed. For the SOL plasma, reduction of...
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Valeria Perseo (Max Planck Institute for Plasma Physics - Greifswald)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
The plasma exhaust concept of the Wendelstein 7-X (W7-X) stellarator is based on the island divertor configuration, which exploits the interaction of magnetic islands with ten discrete water-cooled targets. This translates into a scrape-off layer (SOL) with long connection lengths and small field line pitch angles, affecting the balance of the different transport channels, favouring the...
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Jae-Sun Park (Oak Ridge National Laboratory)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
Due to its high magnetic field SPARC is expected to have a narrow heat-flux width (λq~0.5mm) in the similar range to ITER, making power exhaust a central challenge. Previous studies explored SPARC’s operational space through density and impurity scans [1, 2], but the diffusivity values (D, χ) used to reproduce λq are not uniquely constrained. Recent experiments reported upstream decay lengths...
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Adriano Stagni (Consorzio RFX)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
This contribution presents first experimental observations on the effect of nitrogen (N2) seeding on the scrape-off layer (SOL) properties of type-I ELMy H-mode plasmas in the TCV device. This work is motivated by the significant uncertainties still persisting in the physical understanding of SOL transport processes in seeded regimes, which represents a crucial aspect for reliable...
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Andreas Holm (Lawrence Livermore National Laboratory)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
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Scrape-off layer (SOL) transport simulations including magnetic and ExB drifts flows, performed with the multi-fluid edge-plasma code UEDGE, predict passive stabilization of the detachment front along the low-field side (LFS) divertor leg when main-SOL pumping upstream of the LFS divertor target is applied. With the ion $\mathbf{B}\times\nabla\mathrm{B}$ drift directed into the divertor,... -
Yanjie Zhang (Nanyang Technological University)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
Double-Null (DN) configurations are a leading candidate for future tokamak power exhaust management, yet the precise impact of the magnetic configuration on divertor asymmetry remains a critical challenge. This work employs SOLPS-ITER simulations with full drifts to investigate the interplay between triangularity ($\delta$) and transport in TCV L-mode DN discharges, focusing on the physics of...
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Marco Cavedon (University of MIlano-Bicocca)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
Understanding the mechanisms that govern heat and particle transport in the divertor region is critical for the design and operation of future fusion reactors. Turbulent cross-field transport plays a key role in determining the heat flux distribution at divertor targets, affecting both the peak heat load and the overall power exhaust scenario. A key metric for characterizing heat flux...
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Tilmann Lunt (Max Planck Institute for Plasma Physics)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
Toroidal symmetry is one of the basic design principles of a Tokamak. ASDEX Upgrade (AUG) has recently undergone a major hardware upgrade [Herrmann FED 2017,Herrmann FED 2019,Teschke FED 2019] in order to study alternative divertor configurations [Lunt NME 2017,Pan PPCF 2018]. Apart from a pair of in-vessel coils, a new inner and outer upper target and a new cryo-pump a large set of new edge...
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Patrick Tamain (CEA)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
Mean-field transport coefficients are the main work-horse for the modelling of the edge plasma of magnetic fusion devices and are heavily used in the frame of predictive studies for the design and preparation of operation of future devices. These codes offer a high-fidelity description of plasma-neutrals and multi-species interactions at the cost of a simplified description of transverse...
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Michele Lambresa (Aix-Marseille Université)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
The magnetic field topology has been observed to systematically affect the power threshold for the L–H transition in several tokamaks. In particular, a lower (higher) power threshold is found when the ion-$\nabla B$ drift points toward (away from) the X-point, indicating a "favorable" ("unfavorable") configuration for the transition. Early explanations for this behavior were proposed by...
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Dr Dieter Boeyaert (University of Wisconsin-Madison)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
Divertors are important for particle and heat removal in future stellarator power plants. Currently, three types of stellarator divertors are studied: the island divertor, the helical divertor, and the non-resonant divertor [1]. Where the island divertor and the helical divertor have been studied experimentally in W7-X and LHD, respectively, the non-resonant divertor has mainly been studied...
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Carsten Killer (MPPL)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
In the island divertor of Wendelstein 7-X, modular divertor targets intersect a chain of resonant magnetic islands that act as the scrape-off layer (SOL). The island SOL is characterized by long parallel connection lengths of several 100m, resulting in a high efficiency of perpendicular (radial and bi-normal) transport compared to parallel transport. The two main perpendicular transport...
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Vesa-Pekka Rikala (Aalto University)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
The agreement in the EDGE2D-EIRENE predicted and the measured Balmer-$\alpha$ line emission and the molecular Fulcher band emission for a set of deuterium Ohmic plasmas in JET-ILW validates the neutral and the target recycling models in EDGE2D-EIRENE. Since the plasma-neutral interactions are the strongest in the divertor, a validated EDGE2D-EIRENE neutral model provides higher fidelity to...
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Sebastian Hörmann (Max-Planck-Institut für Plasma Physik (Garching))18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
Understanding and quantifying particle and energy transport at the boundary, detachment and its stability in the divertor region, is crucial for magnetic confinement fusion, as this determines both plasma performance and target loads. For this reason, several fast helium beam systems, based on the ASDEX Upgrade (AUG) midplane system [1], have been installed in the divertors of AUG [2] and...
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Erin Joy Tinacba (ORNL)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
The electron temperature (T$_{e,sep}$) and density (n$_{e,sep}$) at the separatrix at the outboard midplane (OMP) are the key parameters in mapping the operational space for power exhaust handling in tokamak fusion devices. Here, three methods of determining the Te,sep and ne,sep in ST40 are compared. First is through the reverse two-point modeling [1], wherein the electron temperature...
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Sergio Garcia Herreros (Swiss Plasma Centre - EPFL)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
While the conventional lower single-null (LSN) divertor will be tested in ITER, its extrapolation to power-plant conditions remains uncertain, motivating the exploration of further optimised solutions. Three geometrical parameters can be changed independently in the LSN configuration: poloidal length of the divertor legs, poloidal flux expansion at the target and radial position of the strike...
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Andrei Pshenov (ITER Organization)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The ITER divertor design has been guided by extensive scoping studies focused on baseline burning conditions at $Q_{DT} = 10$. They were conducted with the SOLPS-4.3 plasma boundary code without drifts and currents, assuming fuel injection from the top of the machine and pumping directly underneath the dome umbrella. The resulting simulation database was used to optimize the divertor...
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Ulrich Stroth (MPPL)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The X-point radiator (XPR) plasma regime displays favorable properties with regard to power exhaust in tokamaks: An H-mode-like confinement quality, a detached divertor, and the suppression of type-I ELMs are achieved simultaneously [1]. XPR scenarios may also pave the way for more compact and cheaper divertor solutions,as demonstrated on ASDEX Upgrade [2]. The parameter to control XPR...
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Andrei Khodak (Princeton Plasma Physics Laboratory)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
A new divertor concept, referred to as the Diffusion Pump Divertor [McComas et al. US Patent Application #63/919,661], is presented. This concept adapts vacuum-pump technology to enable controlled delivery and removal of vapor within fusion devices. This approach addresses longstanding limitations in present lithium vapor delivery schemes, which rely on evaporation and provide only limited...
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Vsevolod Soukhanovskii (LLNL)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The snowflake (SF) divertor is studied in the MAST-U tokamak as an alternative concept for next-step compact fusion devices and is planned for high-power NSTX-U tokamak experiments. The studies focus on the SF plasma transport mechanisms and their scaling with plasma current in the range 0.4-1.0 MA. The SF divertor features divertor geometry, radiation and transport enhancements (cf. standard...
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Dr Yu Gao (MPPL)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
Power crossing the last closed flux surface is guided toward the divertor plates, where it deposits on a small area (wetted area), which in a reactor-scale device would result in an enormous heat flux. For given technical constraints, the maximum wetted area achievable in axisymmetric tokamaks is determined by the power decay length ($\lambda_{q}$) [1]. For Wendelstein 7-X (W7-X), the divertor...
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Anastasia Poletaeva (Peter the Great St. Petersburg Polytechnic University)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The X-point radiator (XPR) regime (see the Ref. [1] and the references therein) is now commonly explored in most modern tokamaks using a variety of impurity seeding gases. Given its usual association with almost complete divertor detachment and benign or no ELM activity, it is an extremely attractive potential option for future reactor-scale machines.
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On the SOLPS-ITER numerical simulation... -
Michael Faitsch (Max Planck Institute for Plasma Physics)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
The quasi-continuous exhaust (QCE) regime is naturally type-I ELM free. It combines the high density at the plasma edge needed for power exhaust with the high normalised energy confinement typical for H-mode operation. In the QCE regime large-scale type-I ELMs are replaced by high-frequency, low-amplitude filaments leading to the quasi-continuous edge transport of particles and energy...
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Nicola Lonigro (UKAEA)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
Reactors will operate in the detached divertor regime, with the hot ionizing plasma away from the target, to avoid excessive heat fluxes to the walls. A reduced sensitivity of this detachment front location is advantageous as a passive stabilization measure against variations in the upstream parameters. During transients, it can avoid the hot plasma reaching the target, or the cold neutrals...
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Auna Moser18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
Recent experiments on DIII-D measure broadened heat flux profiles in the divertor; broadening increases with auxiliary heating at both low and high plasma current, reaching up to 3 times broadening relative to the ITPA multi-machine heat flux scaling regression[1]. These discharges push to midplane parallel heat flux $q_{||}\sim1$ GW/m$^{-2}$ to study profile broadening at previously...
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David Rees (Aalto University, Espoo, Finland)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
A density scan of two heating powers, $P_{\rm{NBI}}=1$ MW and 5 MW, was performed in JET ITER-like wall (JET-ILW) experiments and SOLPS-ITER simulations of NBI-heated low-confinement mode (L-mode) helium (He) plasmas. In high-recycling conditions, ion flux to the low-field side (LFS) divertor, $I_{\rm{div,LFS}}$, is 70% lower in He then deuterium (D). SOLPS-ITER underpredicts...
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Sarah Elmore (UKAEA)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
MAST Upgrade is designed and built to have a tightly fitting baffle to allow the use of either a conventional or long legged closed divertor configuration. This design allows the main chamber plasma to be decoupled from the divertor plasma, giving greater power dissipation before the divertor target in the closed divertor chamber and better access to detachment, whilst maintaining high...
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Ivan Paradela Perez (ORNL)18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
MAST Upgrade experimental infrared thermography and Langmuir probe data have been used to study the upper to lower outer targets asymmetries, especially the ratio of the peak of the heat flux densities, for a variety of experimental discharges. SOLPS-ITER simulations with a wide range of actuator conditions and magnetic configurations (Conventional, Elongated, and Super-X; combined with...
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Dr Hang Si (National Institutes for Quantum Science and Technology(QST))18/05/2026, 16:10G. Power Exhaust, Plasma Detachment and Heat Load ControlPoster
One of the critical challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, which necessitates access to divertor detachment at relatively low main plasma density. Now Japan focuses on the design of Japanese DEMO (JA DEMO) with ITER size. The V-shaped divertor geometry formed by the vertical...
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Daniel Primetzhofer (Uppsala University)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
Tungsten (W) is the main candidate for plasma-facing materials in tokamaks, as it features, among other properties, low sputter yield and low retention of hydrogen isotopes. However, W lacks intrinsic gettering properties for mid-Z impurities, which are necessary to reduce the presence of impurities that are otherwise capable of degrading the plasma. ITER plans to use boronization as a wall...
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Dr Davis Easley (ORNL)18/05/2026, 16:10F. Edge and Divertor Plasma PhysicsPoster
We present a novel edge-SOL coupling scheme linking SOLPS-ITER SOL transport simulations with EPED pedestal predictions and validation efforts through MAST-U dedicated experiments. This scheme employs a flux-gradient-driven density pedestal model [1], taking as inputs the separatrix conditions from a converged SOLPS solution and predicting the density pedestal top through empirically informed...
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Gabrielle Koknat (Sandia National Laboratories, USA)18/05/2026, 16:10A. Physics Processes at the Plasma Material InterfacePoster
The resilience of plasma-facing materials (PFMs) under intense radiation is a significant challenge for feasible fusion reactors. Computational predictions of cumulative radiation damage are needed to explain experimental observations but are obscured due to the complex ultrafast dynamics that occur after energetic collisions with the lattice. As a result, scientific understanding of PFM...
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