First Wall Research Articles

Damage evolution in Plasma Facing Materials by a sequential multiscale approach

Abstract Describing the time evolution of Plasma Facing Materials (PFMs), through quantitative evaluations of erosion, roughness, and physical properties degradation, is one of the difficult challenges to reach the goal of efficient energy production by nuclear fusion. To follow all the aging-connected physical and chemical phenomena through their characteristic dimensional scale, and to estimate the PFM microstructural transformation over time, we propose a predictive sequential multiscale methodology, consisting of two database-provided coupled codes. The first is a time-dependent, volume-averaged, plasma simulator which describes completely this system in terms of thermodynamics, composition and evaluation of the sheath potential. Plasma solutions are geometrically rearranged by adding surface reactions and 3D geometric features. To increase sensitivity, plasma information is provided to the second code as an initial condition. Such a code is a 3D kinetic Monte Carlo in-cell algorithm for the nano-scale erosion simulation describing the PFM interactions through an extendable set of physical phenomena, such as sticking, sputtering, ion enhanced removals and ion penetration. In this paper, we perform simulations for the case of study of Hydrogen (H) plasmas produced in linear devices, reaching the quasi-atomic detail of the plasma induced material modification of tungsten (W) as PFM.

The evolutionary process of W-V mixed dumbbell in tungsten crystals: A study about W-V alloy as a plasma-facing material in fusion devices

Alloying is widely used to improve the radiation resistance of plasma-facing materials. The first-principles method based on density function theory was used in this work to study the stability and mobility properties of the W-V mixed dumbbell pair. The diffusion of monovacancy and recovery of mixed dumbbell pairs were also studied. The rotations and diffusion results indicated that the W-V mixed dumbbell pair diffusions are favored in two/three-dimensional motion because of low energy barriers. And the results are also suggested the mixed dumbbell lowers the diffusion energy barriers of the monovacancy. New vacancy diffusion modes were also observed along the 〈110〉 direction during the simulation. Besides, the addition of V may also moderate the diffusion of W SIA in 〈111〉 direction.

An integral approach to plasma-wall interaction modelling for EU-DEMO

Abstract The integral approach to plasma-wall interaction (PWI) modelling for DEMO is presented, which is a part of EUROfusion Theory and Advanced Simulation Coordination (E-TASC) activities that were established to advance the understanding and predictive capabilities for modelling of existing and future fusion devices using a modern advanced computing approach. In view of DEMO design, the aim of PWI modelling activities is to assess safety-relevant information regarding erosion of plasma-facing components (PFC), including its impact on plasma contamination, dust production, fuel inventory, and material response to transient events. This is achieved using a set of powerful and validated computer codes that deal with particular PWI aspects and interact with each other by means of relevant data exchange. Steady state erosion of tungsten PFC and subsequent transport and re-deposition of eroded material are simulated with the ERO2.0 code using a DEMO plasma background produced by dedicated SOLPS-ITER simulations. Dust transport simulations in steady state plasma also rely on the respective SOLPS-ITER solution and are performed with the MIGRAINe code. On the way to improved simulations of tungsten erosion in the divertor of DEMO, relevant high density sheath models are being developed based on Particle-in-Cell (PIC) simulations with the state-of-the-art BIT code family. PIC codes of the SPICE code family, in turn, provide relevant information on multi-emissive sheath physics, such as semi-empirical scaling laws for field-assisted thermionic emission. These scalings laws are essential for simulations of material melting under transient heat loads that are performed with the recently developed MEMENTO code, the successor of MEMOS-U. Fuel retention simulations assess tritium retention in tungsten and structural materials and permeation to coolant accounting for neutron damage, in particular for divertor monoblocks of different sizes, using the FESTIM code, and for the first wall, in particular in view of tritium self-sufficiency, using the TESSIM code. Respective code-code dependencies and interactions, as well as modelling results achieved to date are discussed in the contribution.

Coupled model for liquid lithium plasma facing components

Numerical analysis provides the design choice and operating window of liquid metal Plasma Facing Components (PFC) concepts. Coupled analysis of boundary plasma together with the surrounding boundary structures is required. To achieve this goal, PPPL is developing a comprehensive multi-physics model for modeling of PFCs in fusion devices. The model includes the fluid-kinetic code SOLPS-ITER and the flow and heat transfer code CFX from ANSYS. SOLPS-ITER was augmented with a liquid metal boundary condition algorithm, allowing direct two-way coupling of the plasma analysis with the two-dimensional analytical slab flow model which includes heat convection in the liquid metal PFC. The target heat flux resulting from this coupled analysis is used as a boundary condition for detailed 3D Computational Fluid Dynamics (CFD) Magneto Hydro Dynamics (MHD) and heat transfer analysis. A new formulation of MHD equations is introduced in the numerical procedure ensuring current conservation of the discretized equations. Results of the 3D analysis are used for final validation of the coupled model. A PFC design where a porous wall is used to stabilize the liquid metal surface, while MHD drive is used to push the liquid metal flow inside the PFC, will be investigated in the regimes where vapor shielding is created for enhanced volumetric plasma heat dissipation.

Realization of a gas puff imaging system on the Wendelstein 7-X stellarator.

A system for studying the spatiotemporal dynamics of fluctuations in the boundary of the W7-X plasma using the "Gas-Puff Imaging" (GPI) technique has been designed, constructed, installed, and operated. This GPI system addresses a number of challenges specific to long-pulse superconducting devices, such as W7-X, including the long distance between the plasma and the vacuum vessel wall, the long distance between the plasma and diagnostic ports, the range of last closed flux surface (LCFS) locations for different magnetic configurations in W7-X, and management of heat loads on the system's plasma-facing components. The system features a pair of "converging-diverging" nozzles for partially collimating the gas puffed locally ≈135mm radially outboard of the plasma boundary, a pop-up turning mirror for viewing the gas puff emission from the side (which also acts as a shutter for the re-entrant vacuum window), and a high-throughput optical system that collects visible emission resulting from the interaction between the puffed gas and the plasma and directs it along a water-cooled re-entrant tube directly onto the 8 × 16 pixel detector array of the fast camera. The DEGAS 2 neutral code was used to simulate the Hα (656nm) and HeI (587nm) line emission expected from well-characterized gas-puffs of H2 and He and excited within typical edge plasma profiles in W7-X, thereby predicting line brightnesses used to reduce the risks associated with system sensitivity and placement of the field of view. Operation of GPI on W7-X shows excellent signal-to-noise ratios (>100 at 2 Mframes/s) over the field of view for minimally perturbing gas puffs. The GPI system provides detailed measurements of the two-dimensional (radial and poloidal) dynamics of plasma fluctuations in the W7-X edge and scrape-off layer and in and around the magnetic islands outside the LCFS that make up the island divertor configuration employed on W7-X.

TECXY simulations of the power exhaust in the multi-impurity plasma of DTT reactor

Reduction of the heat load to plasma-facing components is a crucial problem for future fusion reactors like Divertor Test Tokamak (DTT). Mitigation of the power load via increased plasma radiation with the use of puffed ions of impurities would be one way to mitigate power in the scrape-off layer. This paper presents a numerical investigation of the impact of seeded impurities on the radiation pattern and the power load to the divertor plates of the high-field DTT reactor in the single null (SN) configuration. The simulations have been done with the use of the TECXY code, which solves multi-species plasma transport equations for multiple impurity species simultaneously and all associated ionization stages in a two-dimensional poloidal geometry. TECXY represents the model of plasma transport in the scrape-off layer region by a classical set of transport equations of multi-species plasma derived by Braginskii. The paper aims to compare the mitigation capabilities of neon and argon impurities seeded in the plasma of the DTT device and to obtain a significant energy flux reduction to the target plate at the smallest possible impurity concentration. Performed investigations showed the effects of neon and argon impurities seeding separately for constant electron density at the separatrix. It has been found that the decrease in electron temperature on the divertor plates up to 3 eV at the outer and the inner divertor plates and the peak power load below 15 MW m−2 at the outer divertor plate can be achieved with argon seeding and much lower impurity concentration than that in the case of neon impurity seeding. Studies have shown the complexity of the effect of neon and argon impurities on the boundary plasma. It was found that the reduction of temperature and the power on both divertor plates was the most effective for the high upstream plasma densities. The results show also that diffusive perpendicular transport strongly affects impurity radiation and thus plasma condition at divertor plates.

Data streaming infra-red video bolometer (IRVB) of Korea Superconducting Tokamak Advanced Research (KSTAR).

Due to the increasing demands for active plasma control operations, in situ diagnostics are highly sought after. Tungsten plasma-facing components have been utilized in the Korea Superconducting Tokamak Advanced Research (KSTAR) lower divertor since the 2023 campaign. Plasma radiation is a key parameter for plasma control, especially in radiation front control experiments. Therefore, the KSTAR infra-red video bolometer (IRVB) needs to be reconfigured into an in situ data streaming diagnostic. This requires comprehensive changes in both infra-red camera control and data analysis compared to the previous system. To ensure the stability of the reconfigured system, functional parts are grouped and separated into individual processes to protect camera acquisitions from errors in other processes. In addition, to enhance the speed of data streaming analysis, the analysis code has been optimized and converted into graphics processing unit (GPU) code. Besides the data streaming analysis, the system is also designed to support post-shot analysis with the entire frame data for the same shot to address frame drop issues encountered during data streaming. Radiation front control experiments with N2 gas seeding are successful results of the data streaming IRVB for its commissioning. This paper focuses on the development of the data streaming IRVB system.

Numerical simulation of deuterium retention in tungsten under ELM-like conditions

Retention of hydrogen isotopes (HI) in plasma-facing components (PFCs) is a crucial process influencing the operation of a fusion device. The dynamics of HI retention is mainly determined by irradiation conditions of PFCs. These conditions can change due to the onset of fast transient events, such as edge-localised modes (ELMs). The development of ELMs results in repetitive short-term plasma bursts on PFCs, affecting the exposure regime. In this work, the effect of ELM-like loads on the fuel retention is numerically analysed using a one-dimensional diffusion model, implemented in the FESTIM code. As a representative simulation case, the deuterium (D) retention in tungsten (W) is considered with the geometry and inter-ELM exposure conditions relevant for large fusion devices. Temporal dependencies of heat and particle fluxes during intra-ELM stages are accounted for by applying the free-streaming model for an ELM filament transport. The simulations were conducted for three inter-ELM exposure regimes with various properties of transient loads, D trapping sites, and D recombination rates. Compared to the ELM-free irradiation, the onset of transients is shown to mainly reduce the D retention rate because of significant material heating induced by the arrival of energetic ELMy particles. This relative difference can decrease if the material is characterised by strong trapping sites or limited desorption from a surface. The findings also demonstrate that additional heating during transients provides conditions for a faster D migration into the bulk. In addition, we discuss the possibility of using ELM-average loads to obtain quick assessments of the D content. The approach is shown to be applicable for the case of small ELMs and is used to derive the analytical expressions for the D distribution in W within the steady-state approximation.

Microstructural evolution and hardening of W-ZrC, W-1Re-ZrC and W-3Re-ZrC alloys induced by He ions irradiation

Bulk W-0.5 wt%ZrC (W-ZrC), W-1 wt%Re-0.5 wt%ZrC (W-1Re-ZrC), and W-3 wt%Re-0.5 wt%ZrC (W-3Re-ZrC) alloys were prepared and exposed to He2+ ion irradiation with an incident energy of 500 keV at 400 °C. Post-irradiation analysis revealed the formation of He bubbles, with consistent number density and average size across all alloys. However, the W-3Re-ZrC alloy exhibited a notably higher density of dislocation loops within grain interiors compared to W-ZrC and W-1Re-ZrC. This led to a significant increase in the hardening rate of W-3Re-ZrC to 61.5 %, nearly twice that of the other alloys. By considering the binding energies of He-vacancy, He-He, and Re-He, the formation and growth behavior of irradiation-induced defects, including He bubbles and dislocation loops, were also discussed. This work provides valuable insight into the relationships among Re alloying, He-induced defects, and irradiation hardening in Re-alloyed W materials, thus offering valuable guidance for the design of W-based plasma-facing materials.

Interaction of iron melt with tungsten and WFe composite structure evolution

Development of new plasma facing components (PFC) draw attention of numerous scientific groups in fusion energy field. Tungsten as a main PFC material main have poor machinability, thus various designs were proposed to overcome this problem. However, such technologies as functional graded layers, sintering and additive technologies are limited in production size and have low cost-efficiency. This research considers the alternative approach of steel melt injection or melt infiltration of tungsten mesh. The results obtained establish the mechanism of phase evolution during interaction of solid W with liquid steel and iron. Energy dispersive spectroscopy, X-ray diffraction and Thermocalc calculations used to analyze the phase composition of tungsten wetted with various steels and pure iron. Results show that interaction rate significantly depends on melt temperature and overheating above its liquidus. Then the overheating exceeds 150 °C erosion of tungsten occur.

Electrical insulation properties in a cold plasma of alumina coating for the in-vessel stabilizing shell of the RFX-mod2 fusion device

This paper presents the results of experimental tests on samples made of copper coated with alumina layer, performed to assess the reliability of its dielectric properties for applications in the low temperature plasma at the edge of a fusion device. The cue of the study was related to the plasma facing components of the RFX-mod2 fusion device (Marrelli et al., 2019, Peruzzo et al., 2023, Peruzzo et al., 2019), devoted to the experimental study of the magnetic confinement of fusion plasmas in a variety of configurations, including the reversed-field pinch and the tokamak. In RFX-mod2 an in-vacuum copper shell for the passive stabilization of MHD modes will surround the plasma. To avoid potentially harmful electrical discharges, which could be induced by rapid transients of the plasma current, this structure must be covered with an electrically insulating layer. For RFX-mod2 an alumina coating was chosen, whose dielectric properties have been tested both in air and in the presence of weakly ionized plasma. Electrical tests, conducted on copper samples with alumina deposits of about 100μm thickness, revealed that the ceramic layer has a high electrical resistance value in air (>1GΩ), but electrical discharges can occur in presence of a weakly ionized plasma, depending on compactness and porosity of the alumina layer, causing local melting of the alumina and expulsion of copper droplets from the substrate. Scanning Electron Microscope (SEM) analyses revealed that in the failed samples the ceramic layer was irregular and rough, with interconnected cavities and cracks, which could reduce its effective thickness and explain the dielectric breakdown at relatively low voltages (<400V). The analyses also showed that samples with a more compact layer present a higher dielectric strength in the presence of the plasma, highlighting that compactness and porosity play crucial roles in ensuring good insulation for materials in a plasma. This study led the definition of the requirements for the insulating coating of the plasma facing components of the RFX-mod2 fusion machine, however the results can be useful for other fusion and non-fusion plasma applications requiring electrical insulation, which can span from industrial devices to spacecrafts.

Exploring the suppression methods of helium-induced damage in tungsten by investigating the interaction between beryllium and helium

Abstract Helium (He)-induced damage is a sensitive concern for the performance of tungsten plasma facing materials (W-PFMs). Recent experiments have revealed that trace impurities in He plasma can effectively prevent the formation of He bubbles and fuzz on W surfaces. To explore its plausibility and underlying mechanism, we performed a multiscale computational study that combines density functional theory calculations and object kinetic Monte Carlo simulations to investigate the effects of a small quantity of beryllium (Be) on the evolution of He bubbles. It is found that there is a strong attractive interaction between He and Be, which can be attributed to the decrease in electron density and the lattice distortion induced by embedded Be atoms. Therefore, the co-implantation of Be continuously introduces trapping centers for He. Due to the low implantation depth and high migration energy of Be, the Be atoms are located close to the surface, leading to the trapping of the majority of He within the near-surface region and the development of a shielding layer for He permeation. The presence of Be facilitates the dispersion of the trapped He, skewing the He clusters into smaller sizes. More importantly, the Be trapping centers bring the He clusters closer to the surface, significantly increasing the probability of bubble bursting and the release of He back to the vacuum. This ultimately leads to a lower retention of He in the case of He + Be co-irradiation, compared with the case of He-only irradiation. Consequently, our findings elucidate the suppressive effect of a low flux of Be atoms on the growth of He bubbles, highlighting the need to focus on synergetic effects between plasma species.

Managing the heat: In-Vessel Components.

The Spherical Tokamak for Energy Production (STEP) programme aims to deliver a first-of-a-kind fusion prototype powerplant (SPP). The SPP plasma places extreme heat, particle and structural loads onto the plasma-facing components (PFCs) of the divertor, limiters and inboard and outboard sections of the first wall. The PFCs must manage the heat and particle loads and wider powerplant requirements relating to safety, net power generation, tritium breeding and plant availability. To enable STEP PFC concepts to be identified that satisfy these wide-ranging requirements, an iterative design ('Decide & Iterate') methodology has been used to synchronize a prioritized set of decisions, within the fast-paced, iterative, whole plant concept design schedule. This paper details the 'Decide and Iterate' methodology and explains how it has enabled the identification of the SPP PFC concepts. These include innovative PFC solutions such as a helium-cooled discrete and panel limiter design to increase tritium breeding while providing sufficient coverage and enabling individual limiter replacement; the integration of the outboard first wall with the breeding zone to enhance fuel self-sufficiency and power generation; and the use of heavy water (D2O) within the inboard first wall and divertor PFCs to increase tritium breeding within the outboard breeding zone. This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

Controlling a new plasma regime.

Success of the UK's Spherical Tokamak for Energy Production (STEP) programme requires a robust plasma control system. This system has to guide the plasma from initiation to the burning phase, maintain it there, produce the desired fusion power for the desired duration and then terminate the plasma safely. This has to be done in a challenging environment with limited sensors and without overloading plasma-facing components. The plasma parameters and the operational regime in the STEP prototype will be very different from tokamaks, which are presently in operation. During fusion burn, the plasma regime in STEP will be self-organizing, adding further complications to the plasma control system design. This article describes the work to date on the design of individual controllers for plasma shape and position, magneto hydrodynamic instabilities, heat load and fusion power. Having studied 'normal' operation, the article discusses the philosophy of how the system will handle exceptions, when things do not go exactly as planned. This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

Boron powder injection experiments in WEST with a fully actively cooled, ITER grade, tungsten divertor

Reactor relevant fusion devices will use tungsten (W) for their plasma facing components (PFCs) due to its thermomechanical properties and low tritium retention. However, W introduces high-Z impurities into the plasma, degrading its performance. Different wall conditioning methods have been developed to address this issue, including coating of W PFCs with layers of low-Z material. Wall conditioning by boron (B) powder injection using an impurity powder dropper (IPD) is being studied in WEST. Two series of experiments were conducted since the installation of the new ITER grade full W divertor. During the first series in 2023 ∼ 1 g of B powder was injected in total at a maximum rate of ∼ 58 mg/s, both of which are three times greater than respective values in the initial WEST powder injection experiments. The second series of experiments included injection of B and BN powders for comparison of their effects on plasma performance. The presence of an instantaneous conditioning effect is suggested by visible spectroscopy measurements of low-Z impurity lines and a rollover of total radiated power past an injection rate of ∼ 20 mg/s was observed. Presence of B coating layer formation is supported by the evolution of the average radiance of visible lines of B, W and oxygen (O). To understand B transport, an interpretative modeling workflow is employed, utilizing the SOLEDGE-EIRENE fluid boundary plasma code and the Dust Injection Simulator (DIS) code. Parameters like B perpendicular diffusivity and recycling coefficients are varied to match experimental results to see if the initial assumption of B sticking to the PFCs immediately after the contact with the wall is adequate for correctly modelling its distribution on the PFCs.

A High–Throughput Molecular Dynamics Study for the Modeling of Cryogenic Solid Formation

To predict the favorable thermodynamical conditions and characterize cryogenic pellet formations for applications in nuclear fusion reactors, a high–throughput molecular dynamics study based on a unified framework to simulate the growth process of cryogenic solids (molecular deuterium, neon, argon) under gas pressure have been designed. These elements are used in fusion nuclear plants as fuel materials and to reduce the damage risks for the plasma-facing components in case of a plasma disruption. The unified framework is based on the use of workflows that permit management in HPC facilities, the submission of a massive number of molecular dynamics simulations, and handle huge amounts of data. This simplifies a variety of operations for the user, allowing for significant time savings and efficient organization of the generated data. This approach permits the use of large-scale parallel simulations on supercomputers to reproduce the solid–gas equilibrium curves of cryogenic solids like molecular deuterium, neon, and argon, and to analyze and characterize the reconstructed solid phase in terms of the separation between initial and reconstructed solid slabs, the smoothness of the free surfaces and type of the crystal structure. These properties represent good indicators for the quality of the final materials and provide effective indications regarding the optimal thermodynamical conditions of the growing process.

Technique of Reactor Experiments of Tin-Lithium Alloy Interaction with Hydrogen Isotopes Under Neutron Irradiation Conditions

The implementation of the ITER and DEMO projects currently includes the investigation of the structural and functional material properties of fusion reactors (FRs). Research to support the use of liquid metals and alloys as plasma-facing materials (PFMs) is a crucial area of work during the development of new FRs. Recent studies indicate the prospects of the tin-lithium (Sn-Li) alloy as a new liquid metal for protecting the in-vessel elements of a FR from the energy flows and high-density particles. Sn-Li alloy has been widely explored for utilization as PFM; however, there is a shortage of investigations being performed at nuclear reactors. The utilization of Sn-Li alloy as PFM in a FR must be fully justified by validated experimental results on tests under extremely high heat, plasma, and radiation loads. The paper presents the methodology of in-pile experiments performed at the IVG.1M research reactor (Kurchatov, Kazakhstan) to study the interaction of hydrogen isotopes with Sn-Li alloy under neutron irradiation conditions. A Sn-Li sample with 73 at. % tin and 27 at. % lithium was manufactured. A unique experimental ampoule device (AD) with a Sn-Li sample had been developed and manufactured for in-pile tests. The results of neutron-physical and thermophysical calculations of designs of the experimental device with Sn-Li alloy under irradiation conditions of the IVG.1M reactor were performed to justify the AD design. Methodical experiments were performed to determine the temperature dependence of the change in the composition of the gas phase in the chamber with Sn-Li alloy. The time dependence of the partial pressure of hydrogen, tritium, and tritium-containing molecules in the AD volume with the Sn-Li alloy on its temperature under reactor irradiation conditions at a power of 3 MW has been studied. Key findings include the successful measurement of tritium release, the determination of temperature conditions for tritium generation and release, and the validation of our experimental AD for conducting such studies.

Inside high-temperature ablation phenomenon and analysis of mechanism of a flat-type W–ZrC/Cu high heat flux mock-up

As a plasma-facing material (PFM), W–ZrC is a promising candidate for the divertor in future fusion reactors. To verify the feasibility of W–ZrC, a high heat flux (HHF) test was conducted on a comprehensive experimental platform. A rare internal ablation failure mode was observed during the experiment. Considering the internal ablation phenomenon during the test, the measured data of the test platform were analyzed to predict an ablation time of 12 s. Based on the shape and size of the ablation pit, the cause and mechanism of ablation were determined using thermal simulation analysis. It can be concluded that a sudden and significant increase in heat flux combined with reduced water flow can cause thermal ablation to occur within the heat sink. Furthermore, dimensions of the ablation pit are related to the area in which the heat flux is imposed. In sum, internal ablation can have serious consequences for the divertor.

Ejection of molten tin in the presence of a hydrogen plasma

One of the significant obstacles left if the development of a fusion power plant is the development of Plasma Facing Components (PFCs) that can withstand the large heat and particle flux incident on the first wall and divertor region. As solid PFCs struggle with microstructure growth, sputtering and melting, liquid metals have been a popular potential replacement. Liquid metal's use as a PFC has been increasing due to its ability to self-repair damage as well as pump lost fuel and waste particles to create a low recycling edge. Currently, tin, lithium and lithium eutectics are the commonly considered liquid metals for use as a fusion PFC. It is therefore important to research the Plasma Material Interaction (PMI) between a hydrogen plasma and the liquid metal PFC candidates. This work, conducted at the Center for Plasma Material Interaction (CPMI), investigated how a molten tin surface reacts when exposed to a hydrogen plasma, building off observations by ASML of particles being ejected from a molten tin surface in the presence of a hydrogen plasma. With the ejection of tin affecting ASML's systems as well, since ejected tin can travel around their systems and cause contamination of equipment or wafers that can be destructive to the lithography process. So, for this work, molten tin was exposed to a hydrogen plasma, of varying electron densities and temperatures, and any ejected particles were collected on a witness plate to determine the particle sizes, angular distribution and mass flux. This work found the ejected tin particles range in size from 10′s ofnms to 10′s of microns and that the mass flux of tin from the molten surface is in the 10′s of mgm2*s, increasing with increased atomic hydrogen flux to the molten tin surface. Showing macroscopic amounts of molten tin are ejected in the presence of a hydrogen plasma, therefore showing tin may not be a suitable fusion PFC material.

Overview of the first Wendelstein 7-X long pulse campaign with fully water-cooled plasma facing components

After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the confinement.