Divertor Unit Research Articles

Properties of boron layers deposited during boronisations in W7-X

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 OP2.1 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 × 1015B-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.

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.

Validation of the ERO2.0 code using W7-X and JET experiments and predictions for ITER operation

The paper provides an overview of recent modelling of global material erosion and deposition in the fusion devices Wendelstein 7-X (W7-X), JET and ITER using the Monte-Carlo code ERO2.0. For validating the modelling tool in a three-dimensional environment, W7-X simulations are performed to describe carbon erosion from the graphite test divertor units, which were equipped in operational phase OP 1.2 and analysed post-mortem. Synthetic spectroscopy of carbon line emission is compared with experimental results from the divertor spectrometer measurement system, showing a good agreement in the e-folding lengths in the radial intensity profiles of carbon. In the case of metallic wall materials, earlier modelling of the Be/W environment in JET and ITER is revisited and extended with an updated set of sputtering and reflection data, as well as including the mixing model for describing the Be/W dynamics in the divertor. Motivated by recent H/D/T isotope experiments in JET, limited and diverted configuration pulses are modelled, showing the expected trend of both Be and W erosion increasing with isotope mass. For the JET diverted configuration pulses, it is shown that Be migrates predominantly to the upper part of the inner divertor where it initially leads to strong W erosion. With longer exposure time, the growth of a Be deposited layer leads to a reduction of W erosion in that region. A similar trend is observed in simulations of the ITER baseline Q = 10 scenario, however with a more symmetric Be migration pattern leading to deposition also on the outer divertor.

Migration and balance of 13C in the stellarator Wendelstein 7-X after 13CH[formula omitted] local injection into the island divertor

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.

Multi-scale lumped modeling of micro-channels cooling structure for W7-X divertor unit target module

A novel concept for the cooling system of the W7-X divertor unit target module has been proposed, which involves tiles equipped with parallel arrays connected by hundreds of sub-millimeter rectangular micro-channels (MCs), obtained using Additive Manufacturing techniques. The structural material proposed for the heat sink substrate is galvanized copper, while the plasma facing material is tungsten. To reduce the high computational cost of thermal-hydraulic simulations of the tiles, a multi-scale lumped modeling approach has been built in previous studies. This involves replacing a group of hydraulic parallel MCs with a porous strip (PS), suitably calibrated to reproduce similar thermal-hydraulic behavior of the MCs. The aim of the current work is to verify that the PS-model is also suitable for the thermo-mechanical stress evaluation, comparing the results from arrays with MCs and PS. Since the heat sink and plasma-facing tiles are bonded, the interfacial delamination and shear stresses are analyzed, being crucial for structural integrity. The analyses, carried out in the elastic regime, show that the PS model returns conservative results when compared to the MCs model, with an overestimation of the delamination and shear stresses at the free edge. Results from both models reveal nearly identical interfacial stress predictions at the free surface edge. Moreover, it is found that both MCs and PS blocks, located near the bond interface, contribute to high-stress fluctuations that could lead to the delamination of the bond interface, suggesting that the distance of the microchannel from the interface should be increased. The PS model can reliably be used to design and verify the most convenient series/parallel connection of the tiles in the divertor unit target module, taking operational constraints into consideration.

Signal-distribution-based crack detection for divertor monoblock inspection using eddy current testing

This study evaluated the applicability of eddy current testing to the surface inspection of a divertor plasma-facing unit in a Tokamak fusion reactor. Artificial slits were introduced on the surfaces of some samples. The eddy current testing was performed to detect these slits. A quantitative detection evaluation method based on evaluating the distribution differences is proposed. The results reveal that the signals of slits near the edge are easily buried by the edge signals and difficult to detect directly from the signal amplitude image. With the aid of the proposed method, not only the slit in the middle of the sample surface but also the slits near the sample edges were successfully detected by eddy current testing.

Integral thermo-mechanical performance analysis of helium-cooled T-tube divertor

To verify the mechanical properties of the helium-cooled T-tube divertor, based on the thermal hydraulic optimization of the T-tube divertor, a preliminary analysis was conducted through a combination of elastic and elasto-plastic analysis.The results show that for the M-type damage mode, the T-tube divertor exhibits a small-scale plastic yielding phenomenon, but this has no effect on the overall divertor. The T-tube divertor has local necking deformation but does not fail due to immediate local fracture, and the T-tube divertor has no creep rupture. For the C-type damage mode, the plastic strain of the T-tube divertor is not accumulated under cyclic loading, but in the fatigue analysis, the life of the divertor does not meet the design standards and further optimization is needed.The conclusion shows that the optimized helium-cooled T-tube divertor has good thermo-mechanical properties, and the safety of the helium-cooled T-tube divertor unit can be guaranteed.

Enhancements of the Fast Interlock System for Wendelstein 7-X Operational Phase OP 2.1

An extended modification phase of the superconducting fusion experiment Wendelstein 7- X was completed late 2021 after installing actively cooled divertor units and the integration or modification of new technical components and diagnostics. Currently, the commissioning of the W7-X experiment is underway. First technical plasma operation is scheduled for the end of September 2022, and scientific plasma operation will start at the end of November. The Fast Interlock System (FIS) is an important component of the Safety Instrumented System (SIS) of W7-X. The main task of the FIS is to protect the inner plasma vessel components from thermal overload from both plasma heating systems (e.g. neutral beam injection) and the plasma itself. There are strict limits for reaction times after detection of interlock relevant events, which have to be observed in order to prevent overload situations of the first wall. The requirements for the FIS were substantially modified and extended, in order to address the challenges due to the actively cooled divertor. This led to extensive revision of the technical implementation and the software of the FIS. In this paper, after a short introduction of FIS design and operation in OP1, the modification of FIS for OP2.1 operation is presented. Finally, the first results during the OP2.1 commissioning and operation of the FIS are discussed.

Analysis and design of the actively cooled JT-60 SA divertor

In the frame of the broader approach activities, Fusion for Energy (F4E) is in charge of the design and manufacturing the JT-60 SA Actively Cooled Divertor (ACD). The ACD will be an upgrade of the already existing Inertially Cooled Divertor (ICD) designed by the Japanese National Institute for Science and Technology (QST).The ACD consists of 36 units in total, each unit being approximately 1.8 m wide and 1.3 m tall. Each unit is composed of a Stainless Steel Cassette Frame that supports the High Heat Flux (HHF) and the Normal Heat Flux (NHF) elements. The HHF components are made of Graphite tiles and a TZM heat sink. They are actively cooled with water at an inlet pressure of 2 MPa. The NHF components are made of Graphite tiles bolted to explosion bonded CuCrZr/SS plates cooled with water at 1.4 MPa. The main difference between the ACD and the ICD is the presence of HHF Elements in the vertical targets, which are then suitable for steady state operation. Due to the additional presence of these elements, the supports have been re-designed in order to sustain the mentioned electromagnetic and thermal loads. The design was aimed also at easing and improving the manufacturing processes, in particular harmonizing all cassette and support variants (e.g. for diagnostics installation) and by making most of the components (cassette frames, NHF supports, etc.) compatible with all ports locations.In this paper, the main design features of the new divertor are described in detail, together with the electromagnetic, thermal and mechanical analyses that have been performed to justify the feasibility of the proposed design. The design justification plays a crucial role in the tendering and manufacturing phases, which has started in June 2022 with the HHF elements and will continue until late 2025 with the manufacturing of the other components and the integration of the divertor units.

First feedback-controlled divertor detachment in W7-X: Experience from TDU operation and prospects for operation with actively cooled divertor

In the last experimental campaign (OP1.2b) of the stellarator Wendelstein 7-X (W7-X), boronisation as a mean for first wall conditioning was applied for the first time which led to strongly reduced impurity fluxes from plasma-facing components. Thermal detachment at the uncooled target plates of the test divertor unit (TDU) was reached at higher plasma densities and was accompanied by high recycling of neutrals at the target plate (Feng et al., 2021), (Jakubowski et al., 2021) [1] . A feedback control system was established in W7-X to actively control the gas injection (actuator) for plasma fuelling and impurity seeding (Effenberg et al., 2019) [2] through the divertors. It allowed very successful stabilisation of the detached plasma condition state as well as mitigation of thermal overloads to some baffle tiles. Different routinely available diagnostic signals were used as input parameters (sensors). We describe the setup of the feedback control system, its performance and provide some example results with the main focus on the development of the control scheme which led to the detachment stabilisation over the entire longest (30 s) high-power discharge at W7-X so far. In view of the achieved very successful detachment stabilisation and the necessity to include simultaneous optimisation of the core performance in the future, the feedback control system is being upgraded for the upcoming campaign (OP2.1) in which the water cooled and hereby inherently steady-state capable divertor has been currently installed. The prospects and some experiment ideas for active detachment control are discussed.

Erosion and redeposition pattern on the W7-X graphite test divertor unit tile

A set of plasma-facing components was retrieved from the W7-X vessel after OP1.2b campaign with an inertially cooled Test Divertor Unit (TDU) and analyzed postmortem to provide data on the plasma-surface interactions processes in the complex device geometry. In the present study, the lower TDU horizontal target (HT) graphite tile HM19TM400hTE2 exposed to the plasma during both OP 1.2 campaigns was examined. The surface modification of the material caused by the plasma-wall interactions was determined. The study employed several microscopy methods including SEM, FIB, STEM, TEM, EDX, and optical profilometry which showed that (i) distinct erosion zone was located at the inner HT strike line position, (ii) the area covered by a pronounced deposit was located adjacent the erosion zone, (iii) the target finger substructures positioned close to the outboard side was unaltered, and (iv) significant changes to surface roughness values along the length of the plate occurred.

Very High spatial Resolution IR thermography diagnostic positioning system upgrade in WEST Tokamak

The Very High spatial Resolution (VHR) infrared thermography diagnostic developed by CEA-IRFM is installed in the WEST tokamak to measure surface temperature of the actively cooled W-monoblocks (MB) of the Plasma Facing Units (PFUs), which are similar to the International Thermonuclear Experimental Reactor (ITER) ones. It is a major diagnostic to investigate the behavior of PFU during experimentation thanks to its very high spatial resolution (0.1 mm / pixel), movable field of view (FoV) and the wide-range temperature measurement. In particular, it can study the evolution of monoblocks mounted in the tokamak, the influence of the shaping of these components submitted to high heat load and the behavior of the leading edges regarding the assembling tolerances between adjacent monoblocks. Finally, such diagnostic directly contributes to the specification assessment of the ITER divertor units. This diagnostic was operational during the 2019 WEST experimentation and its enhanced design for the 2020 WEST experimentation campaign. In order to monitor 30 × 12 mm monoblocks temperature, the VHR IR diagnostic is able to move its narrow FoV (64 × 51 mm) to specific targets on the observable divertor area (360 × 420 mm). Position and focus steering are achieved through motor-controlled mirrors and lenses operated by a dedicated Labview application on a remote unit. Mirror positioning occurs between pulses under normal WEST operation conditions, and can withstand thermal radiation and disruptions during pulses. Repeatability, hysteresis, mechanical components reliability are critical parameters to provide accurate positioning (<1 mm) of the FoV. The tuning and calibration of these actuators in accordance with WEST environment is crucial to guarantee the best position of the FoV and optical performances. This paper gives the FoV positioning implementation details and performance of the VHR in WEST.

Carbon erosion/deposition on the divertor of W7-X during the operational period OP 1.2b

Carbon net erosion and deposition at the test divertor unit (TDU) of Wendelstein 7-X (W7-X) were measured for the discharge period OP 1.2b in the year 2018 using 18 specially prepared target elements in all 10 TDUs. These had lengths between 30 and 60 cm and were coated with marker layers for erosion/deposition investigations of 5–10 μm carbon on top of about 300 nm molybdenum. The marker layer thicknesses were measured by elastic backscattering spectrometry (EBS) before and after plasma exposure using 2.5 MeV protons; the surface morphology was investigated using scanning electron microscopy (SEM) and focused ion beam cross-sectioning (FIB), the surface roughness was determined using a two-dimensional optical profiler. Plasma-exposed surfaces were considerably smoother than unexposed surfaces with decreased mean roughness and a shift of the inclination angle distribution towards lower values. The erosion on the 10 TDUs was unequal within a factor of about two. During the discharge period in total 20.4 ± 5.7 g carbon was eroded from the 10 TDUs. Adjacent to the strike line some deposition of carbon was observed. Compared to the discharge period OP 1.2a in the year 2017, the net carbon erosion rate dropped by a factor of 5–6 due to regular boronizations, which reduced the oxygen (and subsequently also the carbon) content in the plasma by 1–2 orders of magnitude. The significance of erosion/deposition processes for long-pulse discharges is discussed.

Laser ablation molecular isotopic spectroscopy: A novel tool to characterise the distribution of 13C and 12C on graphite after 13CH4 tracer injection in Wendelstein 7-X

Wendelstein 7-X (W7-X) operation with inertially cooled graphite Test Divertor Unit (TDU) was finalised in a series of identical hydrogen plasmas with injection of 13C isotopically marked methane. The break-up of 13CH4, injected through gas inlets at one toroidal position of W7-X, and the subsequent transport and deposition of 13C in the device was studied ex-situ. The TDU was therefore extracted from the vessel and some of the removed TDU elements were examined regarding their 13C distribution in deposited layers on the surface using Laser Ablation Molecular Isotopic Spectroscopy (LAMIS). This study shows the 13C/12C distinction capabilities of LAMIS used to analyse the carbon deposition pattern on Plasma-Facing Components (PFCs) on two of the removed TDU elements. The resulted pattern was compared with complementary results from Nuclear Reaction Analysis (NRA) and simulations of 13C deposition by ERO2.0, a material transport and plasma-surface interaction code. Ultra- short laser pulses of a Nd:YAG laser were used in collinear double-pulse configuration for LAMIS. The first pulse (l1 = 355 nm, t1 = 35 ps, F = 2.3 J/cm 2 ) produced a laser- induced plasma on the surface and 50 ns later, the second laser pulse (l2 = 1064 nm, t2 = 35 ps) was directed into this plasma to enhance the signal. Each pulse pair ablates in total 200 nm of the surface material. The analysis is conducted in 1.5 mbar N2 environment for improved signal-to-noise ratio relative to vacuum conditions. A high throughput custom-made spectrometer in Littrow-arrangement (f = 750 mm, A = 6000 at l = 473 nm, lspan = 14 nm) was used to analyse the C2 Swan bands (d3 ⊓ g - a3 ⊓ u, v = 1 to v’ = 0) of laser-induced plasma emission. • Laser Ablation Molecular Isotopic Spectroscopy is a method for isotopical analysis of materials. • Similar to Laser-Induced Breakdown Spectroscopy with analysis of molecular spectra. • This method has in-situ potential for material deposition studies in fusion devices. • This study shows good agreement with results from Nuclear Reaction Analysis. • General agreement with ERO2.0 simulations were found far from the injection location.

Distributions of deposits and hydrogen on the upper and lower TDUs3 target elements of Wendelstein 7-X

Distributions of deposits and hydrogen (H) on the graphite divertor target elements TM4h4 and TM3v5 in the test divertor units 3 (TDUs3) of Wendelstein 7-X (W7-X) are studied. The TM4h4 and TM3v5 are located at the magnetically symmetric positions in the upper and lower divertor. The microstructure of the deposition layer is characterized by a transmission electron microscope (TEM) combined with a focused ion beam (FIB). Metallic deposits such as iron (Fe), molybdenum (Mo), chromium (Cr) are detected in the deposition layer by energy-dispersive x-ray spectroscopy (EDS). The depth-resolved distribution patterns of boron (B) and metallic deposits on upper and lower horizontal (h) divertor target elements TDUs3-TM4h4 as well as upper and lower vertical (v) divertor target elements TDUs3-TM3v5 are clarified by glow discharge optical emission spectrometry (GDOES). Results for both TDUs3-TM4h4 and TDUs3-TM3v5 show that the B deposition regions exhibit higher H retention due to the co-deposition with deposits. On the other hand, up-down asymmetries in B deposition caused by particle drift exist on both TDUs3-TM4h4 and TDUs3-TM3v5. The B deposition amount on upper TDUs3-TM4h4 is 40% smaller than that on lower TDUs3-TM4h4. While for the vertical target elements, the B deposition amount on upper TDUs3-TM3v5 is 35% larger than that on lower TDUs3-TM3v5. Meanwhile, a shift of around 3 cm in B deposition peaks is observed on upper and lower TDUs3-TM4h4 and TDUs3-TM3v5. Results of numerical simulation of carbon deposition/erosion profiles on the target elements using ERO2.0 code and power flux measured by infrared cameras are shown and compared with the above mentioned B profiles.

Investigation of boron distribution and material migration on the W7-X divertor by picosecond LIBS

One set of horizontal target elements of the Test Divertor Units (TDU), retrieved from the Wendelstein 7-X (W7-X) vessel after the end of second divertor Operation Phase (OP1.2B) in Hydrogen (H), were investigated by picosecond Laser-Induced Breakdown Spectroscopy (ps-LIBS). The Boron (B) distribution, H pattern and the material erosion/deposition pattern on these target elements were analyzed with high depth resolution and mapped in the poloidal direction of W7-X. From the spectroscopic analysis, B, H, Carbon (C) and Molybdenum (Mo) were clearly identified. A non-uniformly distributed B pattern on these divertor target elements was determined by the combination of B layer deposition during the three boronizations and W7-X plasma operation with multiple erosion and deposition steps of B. Like the TDU, the analyzed target elements are made of fine grain graphite, but have two marker layers which allow us to determine the material migration via the ps-LIBS technique. Two net erosion zones including one main erosion zone with a peak erosion depth of 6.5 μm and one weak erosion with a peak erosion of 1.3 μm were determined. Between two net erosion zones, a net deposition zone with width of 135 mm and a thickness up to 3.5 μm at the peak deposition location was determined by the ps-LIBS technique. The B distributions are correlated with the erosion/deposition pattern and the operational time in standard magnetic configuration of W7-X in the phases after the boronizations. The thickness of the containing B layer on these target elements also correlates with the erosion/deposition depth, in which the thickness of the containing B layer varies spatially in poloidal direction between 0.1 μm and 6 μm. Complementary, Focused Ion Beam combined with Scanning Electron Microscopy (FIB-SEM) was employed also to verify and investigate the deposition layer thicknesses at typical net erosion and net deposition zones as well as to identify the three boronizations in depth.

Real-Time Detection of Overloads on the Plasma-Facing Components of Wendelstein 7-X

Wendelstein 7-X (W7-X) is the leading experiment on the path of demonstrating that stellarators are a feasible concept for a future power plant. One of its major goals is to prove quasi-steady-state operation in a reactor-relevant parameter regime. The surveillance and protection of the water-cooled plasma-facing components (PFCs) against overheating is fundamental to guarantee a safe steady-state high-heat-flux operation. The system has to detect thermal events in real-time and timely interrupt operation if it detects a critical event. The fast reaction times required to prevent damage to the device make it imperative to automate fully the image analysis algorithms. During the past operational phases, W7-X was equipped with inertially cooled test divertor units and the system still required manual supervision. With the experience gained, we have designed a new real-time PFC protection system based on image processing techniques. It uses a precise registration of the entire field of view against the CAD model to determine the temperature limits and thermal properties of the different PFCs. Instead of reacting when the temperature limits are breached in certain regions of interest, the system predicts when an overload will occur based on a heat flux estimation, triggering the interlock system in advance to compensate for the system delay. To conclude, we present our research roadmap towards a feedback control system of thermal loads to prevent unnecessary plasma interruptions in long high-performance plasmas.

Plasma–surface interaction in the stellarator W7-X: conclusions drawn from operation with graphite plasma-facing components

W7-X completed its plasma operation in hydrogen with island divertor and inertially cooled test divertor unit (TDU) made of graphite. A substantial set of plasma-facing components (PFCs), including in particular marker target elements, were extracted from the W7-X vessel and analysed post-mortem. The analysis provided key information about underlying plasma–surface interactions (PSI) processes, namely erosion, transport, and deposition as well as fuel retention in the graphite components. The net carbon (C) erosion and deposition distribution on the horizontal target (HT) and vertical target (VT) plates were quantified and related to the plasma time in standard divertor configuration with edge transform ι = 5/5, the dominant magnetic configuration of the two operational phases (OP) with TDU. The operation resulted in integrated high net C erosion rate of 2.8 mg s−1 in OP1.2B over 4809 plasma seconds. Boronisations reduced the net erosion on the HT by about a factor 5.4 with respect to OP1.2A owing to the suppression of oxygen (O). In the case of the VT, high peak net C erosion of 11 μm at the strike line was measured during OP1.2B which converts to 2.5 nm s−1 or 1.4 mg s−1 when related to the exposed area of the target plate and the operational time in standard divertor configuration. PSI modelling with ERO2.0 and WallDYN-3D is applied in an interpretative manner and reproduces the net C erosion and deposition pattern at the target plates determined by different post-mortem analysis techniques. This includes also the 13C tracer deposition from the last experiment of OP1.2B with local 13CH4 injection through a magnetic island in one half module. The experimental findings are used to predict the C erosion, transport, and deposition in the next campaigns aiming in long-pulse operation up to 1800 s and utilising the actively cooled carbon-fibre composite (CFC) divertor currently being installed. The CFC divertor has the same geometrical design as the TDU and extrapolation depends mainly on the applied plasma boundary. Extrapolation from campaign averaged information obtained in OP1.2B reveals a net erosion of 7.6 g per 1800 s for a typical W7-X attached divertor plasma in hydrogen.

Erosion of tungsten marker layers in W7-X

In order to get first insight into net tungsten erosion in W7-X, tungsten (W) marker layers were exposed during the operational phase OP 1.2b at one position of the Test Divertor Unit (TDU), at 21 different positions of the inner heat shield, and at two scraper elements. The maximum tungsten erosion rate at the TDU strike line was 0.13 nm s−1 averaged over the whole campaign. The erosion was inhomogeneous on a microscopic scale, with higher erosion on ridges of the rough surface inclined towards the plasma and deposition of hydrocarbon layers in the recessed areas of the rough surface. The W erosion at the inner heat shield was below the detection limit of 3–6 × 1012 W-atoms/cm2s, and all inner heat shield tiles were covered with a thin B/C/O layer with thickness in the range 2 × 1017–1018 B + C atoms/cm2 (about 20–100 nm B + C). W-erosion of the marker layers on the scraper elements was also below the detection limit.

Plasma-wall interaction studies in W7-X: main results from the recent divertor operations

Wendelstein 7-X (W7-X) is an optimized stellarator with a 3-dimensional five-fold modular geometry. The plasma-wall-interaction (PWI) investigations in the complex 3D geometry of W7-X were carried out by in situ spectroscopic observations, exhaust gas analysis and post-mortem measurements on a large number of plasma-facing components extracted after campaigns. The investigations showed that the divertor strike line areas on the divertor targets appeared to be the major source of carbon impurities. After multistep erosion and deposition events, carbon was found to be deposited largely at the first wall components, with thick deposits of >1 μm on some baffle tiles, moderate deposits on toroidal closure tiles and thin deposits at the heat shield tiles and the outer wall panels. Some amount of the eroded carbon was pumped out via the vacuum pumps as volatile hydrocarbons and carbon oxides (CO, CO2) formed due to the chemical processes. Boron was introduced by three boronizations and one boron powder injection experiment. Thin boron-dominated layers were found on the inner heat shield and the outer wall panels, some boron was also found at the test divertor unit and in redeposited layers together with carbon. Local erosion/deposition and global migration processes were studied using field-line transport simulations, analytical estimations, 3D-WallDYN and ERO2.0 modeling in standard magnetic field configuration.