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Boronisation was first used for wall conditioning in W7-X during the OP 1.2b operational period, which was characterized by the use of the fine-grain graphite Test Divertor Unit (TDU) and inertial cooling only. After this period, deposited layers were observed on all inner surfaces. Deposited layers were analyzed on 21 inner wall tiles using ion beam analysis methods, the deposited layers consisted mostly of boron with additional carbon and oxygen. During the operational period OP 2.1 with an actively water cooled divertor made of carbon fiber reinforced carbon, different materials were exposed during two individual boronisations using the multi-purpose manipulator. Deposited boronisation layers on the samples were analyzed using nuclear reaction analysis. The deposited layer thicknesses showed some variation depending on substrate material and surface roughness, but a systematic dependence on material and/or roughness was not observed. Under the typical boronisation conditions at W7-X, one A × h (Ampere times hour) of boronisation results in a boronisation layer with a thickness of about 30 ± 15 × 1015 B-atoms/cm2 (about 3 ± 1.5 nm) at the position of the multi-purpose manipulator. The oxygen gettering capacity of the layers is up to 0.5 – 0.9O/B.

A 13C tracer experiment was carried out to investigate impurity transport and migration in the stellarator Wendelstein 7-X (W7-X). About 4.0 × 1022 molecules of 13C marked methane were injected through the divertor gas injection system during plasma operation. After the experiment, the divertor target elements were extracted to be analyzed by post-mortem methods. In particular, the results of the Nuclear Reaction Analysis are presented, and the 13C balance and predominating contributors to migration are discussed.A two-dimensional 13C deposition pattern on 12C graphite modules was obtained with over 6000 single measurements. The post-mortem analysis results extrapolations show a high local deposition of around 25% of the 13C within a 5 cm radius around the injection holes embedded in a graphite target module of one of the ten divertor units. Additionally, a striation is detected in the toroidal direction along the outer strike line and the net carbon deposition zone along the intersecting magnetic island containing about 23% of the 13C within the same divertor unit where injections were made. The residual 52% of 13C are globally migrated across the machine. In total, ∼ 90% of the injected 13C could be extrapolated by post-mortem analysis. Thus, the 13C balance is closed. The distribution indicates a non-negligible amount of global transport and the strong influence of the magnetic topology on migration. In contrast to the strike line pathway, the transport along the magnetic island only takes place counterclockwise in toroidal direction. These results are used to provide benchmark data for ERO2.0 impurity tracing simulations and to be a reference for impurity migration processes in a stellarator.

Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified.This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx.

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.

Recently, experiments on basic plasma physics issues for solving future problems in fusion energy have been performed on a Large Helical Device. There are several problems to be solved in future devices for fusion energy. Emerging issues in burning plasma are: alpha-channeling (ion heating by alpha particles), turbulence and transport in electron dominant heating helium ash exhaust, reduction of the divertor heat load. To solve these problems, understanding the basic plasma physics of (1) wave–particle interaction through (inverse) Landau damping, (2) characteristics of electron-scale (high- k ) turbulence, (3) ion mixing and the isotope effect, and (4) turbulence spreading and detachment, is necessary. This overview discusses the experimental studies on these issues and turbulent transport in multi-ion plasma and other issues in the appendix.

The present work presents a 2D and 3D modeling of the neutral gas flow in the sub-divertor region of the W7-X. The investigations have been done using the DIVGAS code. The complex 2D and 3D geometries of the divertor components in the sub-divertor region have been considered and the Standard and High-Iota magnetic configurations have been numerically simulated. The main objective of this study is to investigate the dynamics of neutral particles in the sub-divertor region including the effects due to geometry and toroidal and poloidal leakages located at the divertor targets and baffles on the achieved pumping efficiency. A sensitivity analysis has been performed for the effect of various geometrical and flow parameters on the pumping performance, under different plasma scenarios. The considered incoming fluxes in the sub-divertor range between 10 ^20 to 10 ^22 (H _2 s ^−1 ). The main conclusions, which can be extracted from the present numerical analysis could be summarized as follows; a large fraction of incoming neutral particle flux i.e. >70% on the low iota side and >40% for the high iota side is leaked back to the main divertor region, while higher incoming neutral fluxes facilitate the increase of the pumped flux as well as the decrease of the outflux. It has been estimated that a small fraction ∼3%–4% of the incoming neutral flux is being pumped via the turbo-molecular pumps. The closure of the toroidal leakages as well as the inclination of the pumping gap panel by 9 ^o facilitate the increase of the pumped flux, but considering the all the engineering constraints, the latter option seems to be more easy to be implemented. For low incoming neutral fluxes (∼10 ^20 H _2 s ^−1 ) and for the case of AEH section, free molecular flow conditions are estimated and therefore neutral-neutral collisions could be neglected. For higher incoming neutral fluxes and for both AEH and AEP sections neutral-neutral collisions play a significant role in the flow establishment. A comparison with available experimental measurements of the neutral pressure in the sub-divertor has been performed for Standard and High-Iota plasma discharges. The 3D DIVGAS simulations predict qualitatively the experimental data with relative deviation between 25 and 63%. All the above numerical findings will actively support the optimization of the W7-X particle exhaust, in view of the experimental campaign OP2.

We present a method to geometrically quantify the three magnetic island chains with the poloidal mode numbers m = 4, 5, and 6 (referred to in this paper as high-iota, standard, and low-iota islands, respectively), on which the W7-X divertor relies. The focus is on a comparative study of their detachment performance using a series of models of different physical and geometrical complexity, ranging from one- to three-dimensional (1D to 3D). In particular, it aims to identify the key physical elements behind the correlation between impurity radiation and island geometry and the associated detachment stability. Assuming intrinsic carbon as a radiator, we scan the three island chains with the EMC3-Eirene code based on otherwise identical code inputs. We find that the three islands behave differently in the radiation distribution, in the development of the radiation zones during detachment, and in the ‘radiation costs’, defined as the product of impurity and electron density near the last closed flux surface. While the radiation costs for the iota = 5/4 and 5/5 island chains linearly increase with the total radiation, the low-iota island with iota = 5/6 shows a bifurcation behavior in the sense that the radiation costs initially increase and then decrease when the total radiation exceeds a critical level. Consistent with the numerical trends, stable detachment, which is experimentally easy and robust to achieve with the standard iota = 5/5 island chain, remains an experimental challenge with the low-iota configuration. Dedicated numerical experiments show that the recycling neutrals and the ratio of parallel to perpendicular heat transport, which depends closely on the field line pitch, play a significant role in the formation and evolution of the radiation layer. A deeper understanding of the underlying physics relies on simpler models that explain why and how flux expansion can reduce the radiation costs. From these insights, we derive the conditions in which detached plasmas can benefit from the expansion of flux surfaces around the X-point. We show and explain why the current divertor design limits the actual capability of the high-iota configuration and propose solutions. The work is presented within a theoretical/numerical framework but cites relevant experimental evidence to emphasize its practical significance.

The net carbon erosion rate at the divertor strike line of W7-X was determined experimentally during the operational phases OP 1.2a and OP 1.2b [M. Mayer et al. Phys. Scr. T171 (2020) 014035; M. Mayer et al., Nuclear Fusion 62 (2022) 126049]. OP 1.2a was characterized by high concentrations of oxygen and a very high net carbon erosion rate. The oxygen concentration decreased by 1–2 orders of magnitude in OP 1.2b due to boronizations, and the experimentally observed erosion rate decreased by a factor of 5–6. The carbon erosion rate at the W7-X divertor is calculated using an analytical erosion model taking physical and chemical erosion by hydrogen, carbon and oxygen impact into account. Experimentally determined plasma parameters from several discharges and experimental surface roughnesses were used for the simulations. The calculated erosion rates for the selected discharges are in reasonable agreement with the experimental values. According to the calculations, during OP 1.2a carbon erosion at the strike line was dominated by chemical erosion by oxygen. The decrease of the oxygen impurity concentration in OP 1.2b decreased the carbon erosion rate by a factor of about five, which is in good agreement with the experimental values. During OP 1.2b carbon erosion was dominated by sputtering by incident hydrogen and carbon ions. The surface morphology has a profound influence on the net carbon erosion rate and develops during a campaign by erosion/deposition phenomena. Plasma-exposed surfaces get smoother with plasma exposure time, which can result in an increase of the erosion yield.

Hot spots and bright patterns on plasma-facing components (PFC) in visible light are observed in electron cyclotron resonance heated (ECRH) plasmas at low densities in the stellarator Wendelstein 7-X. The events are often located far outside of any convective plasma loads and led in some cases to damages of diagnostic components, where the interaction zones qualitatively agree with fast particle loss simulations. Reciprocating electric probes indicate that this phenomenon is related to a non-thermal electron population that can have a beam like character, being directed in one parallel direction. Further, the electrons can be trapped on rational flux surfaces and be used to map magnetic islands.

In the published paper titled ‘Gas exhaust in the Wendelstein 7-X stellarator during the first divertor operation’ (2022 Nucl. Fusion https://doi.org/10.1088/1741-4326/ac7ac4 ), figure 8 was incorrect. This corrigendum provides the correct figure 8 .

The main objective of the Wendelstein 7-X (W7-X) stellarator is to demonstrate the integrated reactor potential of the optimized stellarator line. An important element of this mission is the achievement of high heating-power and high confinement in steady-state operation. Such an integrated plasma operation has not yet been demonstrated and represents the major scientific goal of W7-X. The way towards this goal is staged. In the first phase, called OP 1.1, December 2015-March 2016, a limiter configuration was used. In this paper, the preparation of the first operation phase as well as lessons learned during the first commissioning and the operation phase are discussed, while the physics results from OP 1.1 are reported elsewhere (Wolf et al 2017 Nucl. Fusion57 102020). [ABSTRACT FROM AUTHOR]