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In the previous divertor campaign, the Wendelstein 7-X (W7-X) device injected 3.6 MW of neutral beam heating power allowing for the achievement of densities approaching 2 × 1020 m−3, and providing the first initial assessment of fast ion confinement in a drift optimized stellarator. The neutral beam injection (NBI) system on W7-X is comprised of two beam boxes with space for four radio frequency sources each. The 3.6 MW of heating reported in this work was achieved with two sources in the NI21 beam box. The effect of combined electron-cyclotron resonance heating (ECRH) and NBI was explored through a series of discharges varying both NBI and ECRH power. Discharges without ECRH saw a linear increase in the line-integrated plasma density, and strong peaking of the core density, over the discharge duration. The presence of 1 MW of ECRH power was found to be sufficient to control a continuous density rise during NBI operation. Simulations of fast ion wall loads were found to be consistent with experimental infrared camera images during operation. In general, NBI discharges were free from the presence of fast ion induced Alfvénic activity, consistent with low beam betas. These experiments provide data for future scenario development and initial assessment of fast-ion confinement in W7-X, a key topic of the project.

The first neutral beam injector (NBI) experiments of the Wendelstein 7-X stellarator took place in summer 2018. The modelling of the fast ion production and slowing down processes predicts losses of the NBI fast ions to the first wall on the order of 15%. One location receiving a high load (possibly peaking at several M W/m2) is the immersion tube for optical and infrared monitoring of the divertor targets. The stainless steel face of the tube has three vacuum windows, which are sensitive to temperature gradients and overheating. To protect the windows from damage caused by the fast ions, different heat load mitigation techniques were investigated. Given the available time and resources until the first NBI experiments, a protective stainless steel collar mounted at the front of the immersion tubes was regarded the most realistic solution. This contribution describes the fast ion modelling of the loads, the new design, thermal modelling of the design, and finally experimental experience with the protective collar showing heat loads in excess of 1.5 M W/m2. The fast ion heat loads have been assessed computationally with the ASCOT code and experimentally with thermal imaging.

During its second operational campaign (OP 1.2a) W7-X operated with an edge island divertor. Much higher densities (1020m−3) than in to the previous, limiter campaign were achieved. In order to asses this improved performance and the nature of transport with regard to the expected higher densities and other parameters such as the electron and ion temperature, the electric field and plasma flow, measurements at the plasma edge were conducted with reciprocating probe, mounted on a Manipulator. The probe contains a set of Langmuir pins, Mach probes, compensation coils and a set of 3D pick up coils for the measurement of the local magnetic field and fluctuations, and an ion sensitive probe. It has been observed, in addition to differences due to the magnetic configurations, that the magnetic topology is sensitive to the plasma conditions, especially to changes due to a evolving toroidal plasma current. Measurements from divertor cameras showed that an increasing toroidal current in the standard configuration broadened the heat flux and shifted the strike line considerably. These down-stream observation can be complemented with the measurements at the mid-plane with the manipulator mounted probes. During the experiments the edge island remnant was identified with the measured edge plasma profiles and compared with the predictions from the field line tracing code. Those electron densities and temperatures have been used as an input for understanding edge transport with the EMC3-EIRENE modeling. The results from EMC3-EIRENE also help improve the caculated density profiles, since the modelling provides a profile of the effective charge Zeff. Keywords: W7-X, Langmuir probes, Magnetic islands, Plasma edge transport

One of the main aims of Wendelstein 7-X (W7-X), an advanced stellarator, is to investigate the quasi-steady state operation of magnetic confinement devices for nuclear fusion, for which power exhaust is an important issue. A dominant fraction of the energy leaving from the confined plasma region will be removed by 10 so-called island divertor units, which are designed to sustain a maximum heat flux of up to 10 MWm−2. An essential prerequisite for the safe operation of a steady-state device is automatic detection of hot spots and other abnormal events. Simple temperature limits in infrared (IR) thermographic images will not be enough on their own, because of plasma-generated surface coatings and other effects summarized in the following. To protect divertor elements from overheating, and to monitor power deposition onto the divertor elements, near real-time hotspot detection algorithms for the analysis of carbon plasma facing components (PFCs) were implemented and tested in the GLADIS facility.One of the difficulties in hotspot detection in a carbon-based machine is the deposition of plasma impurities as layers with a reduced thermal connection to the underlying bulk material. We have developed and successfully tested a method to classify surface layers and benchmarked the performance of the method with the Tore Supra IR data operating with actively cooled carbon PFCs. The surface layers can be detected in a steady plasma discharge during the initial rise and decay in temperature when a strike line touches parts of the divertor or wall. It can also be detected by modulating electron cyclotron resonance heating (ECRH) input power. This feature allows detection of overheated areas while reducing false positives. For the recent operational campaign, inertially cooled test divertor units (TDU) were installed to prepare for steady-state operation with water-cooled divertor units. Automatic, near real-time detection of hot spots and identification of surface layers in the W7-X divertor are presented. Results are compared with a best fit estimate of the heat transmission coefficient α which is used to calculate heat flux onto the divertor in the presence of surface layers. Keywords: W7-X, Divertor, Plasma facing components, Heat load, Infrared, Surface layers, Tore supra

The optimized superconducting stellarator device Wendelstein 7-X (with major radius , minor radius , and plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of with central electron temperatures were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of a temperature of with was transiently accomplished, which corresponds to with a peak diamagnetic energy of and volume-averaged normalized plasma pressure . The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above line integrated density and central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

The Multi-Purpose-Manipulator (MPM) situated at the outboard mid-plane is a versatile carrier system for diagnostic probes at Wendelstein 7-X (W7-X). It is able to probe into the last scrape-off (SOL) layer close to the last closed flux surface. The manipulator is also has the capability to puff gas for well-localized fueling/impurity seeding into the SOL. In the previous operational campaign of OP 1.2a and 1.2b both hydrogen and methane were injected from the manipulator mounted probes. The effect of hydrogen puffing on the main plasma was observed to be much stronger than that of the methane puffing. This difference is related to the difference in ionization lengths between the two gas species. To support this assumption a puffing of a similar amount of methane and hydrogen was modelled with ERO.

An important goal of Wendelstein 7-X, the most advanced operating fusion experiment of the stellarator line, is to demonstrate the ability of stellarators to perform steady-state discharges. In this respect, the monitoring and control of the heat loads on the plasma facing components, especially of the strike-lines in the ten island divertors, will be critical during next operation phase OP2. In this paper, it is shown that deep convolutional neural networks are able to learn the relationship between the heat-flux images, obtained by the analysis of thermographic data, and the applied control coil currents in standard magnetic configuration experiments. This study is carried out in view of understanding and modeling the relationship between the heat-flux distribution in the divertor strike-lines and the actuators influencing them.

The fundamental behavior of the W7-X island divertor under detached conditions, which has been theoretically predicted with the EMC3-Eirene code, is re-examined here under the experimental conditions achieved so far and compared with the first experimental results. Both simulations and experiments cover a range of divertor configurations and plasma parameters, and show the following common trends: (1) with rising impurity radiation, the target heat load decreases ‘uniformly’ over the entire target surface in the sense that both the peak and average heat loads can drop by an order of magnitude. Impurity radiation (mainly from intrinsic carbon) occurs primarily at the plasma edge and the resulting negative impact on the stored energy is less than 10%. (2) When the total radiation exceeds a critical level, the target particle flux (the recycling flux Γrecy) begins to fall and can drop by a factor of 3–5 at high radiation levels without an obvious indication of significant volume recombination. (3) While Γrecy decreases, the divertor neutral pressure continues to build up and reaches a maximum, at which point Γrecy has declined significantly. (4) During detachment, the electron temperature at the last closed flux surface falls in a way that is not quantitatively understandable from parallel classical heat conduction processes. This paper presents a physical explanation of the numerical/experimental results described above. Furthermore, using the EMC3-Eirene code as a diagnostic tool, we are able, apparently for the first time, to provide a full quantitative analysis of each transport channel in the island divertor, aiming to clarify how the island divertor plasma self-regulates to maintain particle, energy, and momentum balance under detached conditions.

The operational regimes in the first divertor campaign of W7-X were limited by unexpectedly high heat loads on certain baffle tiles. In the high-mirror configuration, a permanent hotspot on the baffle plate was detected by the thermographic system, which was further confirmed during the post-campaign inspections of the plasma-facing components. The maximum heat load on three baffle tiles reached about 4.5 MW m−2, which is almost an order of magnitude above its designed value of 0.5 MW m−2. The paper presents a detailed analysis to understand how the baffle plate, which is originally designed to screen the recycling neutrals and is therefore hidden from the hot plasma, can receive such a high heat load – a level that is expected for target plates. Three main causes have been identified: 1) the three baffle tiles are radially only about 5 mm away from a main heat channel towards a target, which is much shorter than the radial power decay length of several centimeters derived from the thermographic measurements in this region; 2) the three baffle tiles are intersected with field-lines of about 30 m length, which are long enough to collect considerable amount of heat via cross-field transport; 3) the baffle tiles have locally large grazing angles. In addition, the analysis method shown in this paper is of general use to estimate the heat flows in shadow areas of targets and thus to evaluate heat loads on the potentially critical components occurring in these areas, especially required for the design of complex 3D divertors like the one in W7-X.

The Wendelstein 7-X (W7-X) stellarator has recently successfully completed its experimental campaign with the passively cooled test divertor unit (TDU) made of graphite. TDU is the plasma facing component where the most intensive plasma-surface interactions take place in a controlled manner and it provides particle and heat exhaust. In addition to TDU the plasma vessel also consisted of other elements made of carbon and the stainless steel wall. Still impurity influx into the plasma due to the plasma-surface interactions is unavoidable. In addition the recessed areas can be also sources of impurities. Specifically, the low-Z oxygen and carbon were the main plasma impurities at W7-X. To tackle this issue boronization was applied – plasma-chemical in-situ deposition of amorphous boron-containing hydrogen films on all plasma-facing components in a helium glow discharge with 10$\%$ B$_2$H$_6$ [1]. The application of boronization has lead to one of the main achievements of the campaign: plasma operation at high core densities of more than 10$^{20}$ m$^{-3}$ due to the reduced radiation-induced density limit. This work mainly focuses on full characterization of the change of impurity influx at TDU, the corresponding effects of boronization on the performance of W7-X and the underlying mechanism. Photon influx spectroscopy of line emission of the atoms and ions of carbon, oxygen, boron and hydrogen allowed to deduce normalized (divided by hydrogen flux) influx of oxygen, carbon and boron. Those measurements were conducted with the overview spectrometer system during the complete experimental campaign. The photon emission spectroscopy data was supported by a number of other diagnostics provided at W7-X. In total three boronizations were applied during the second part of the TDU campaign. After the first boronization the oxygen to hydrogen flux ratio (normalized influx of oxygen) at the divertor substantially decreased by a factor of 10 and the carbon to hydrogen flux ratio (normalized influx of carbon) decreased by a factor of 4 as obtained from spectroscopy. In the same time, boron emission appeared in the spectra. Between the boronizations oxygen and carbon normalized influxes increased but never reached the pre-boronization values. With each subsequent boronization the oxygen level decreased even more, reaching the lowest values after the third boronization which were more than a factor of 100 lower than before the first boronization and the boron level was increasing simultaneously. Such a decrease in low-Z impurity concentration significantly extended the operation window of W7-X in terms of line-integrated electron density (from 4$\cdot$10$^{19}$ m$^{-2}$ to more than 1$\cdot$ 10$^{20}$ m$^{-2}$) and diamagnetic energy (from 330 kJ up to 510 kJ). Z$_{\text{eff}}$ decreased from 4.5 down to values close to 1.2 as obtained from bremsstrahlung measurements. The above mentioned changes in the line-integrated electron density, diamagnetic energy and Z$_{\text{eff}}$ are given for the two reference discharges before and after boronization. Plasma conditions did not significantly deteriorate between boronizations allowing device operation without glow-discharge cleaning. The decrease in impurity influx can be explained in terms of effective oxygen gettering by boron in the remote plasma-facing components of W7-X, while at the strike line area it is quickly eroded. Not only the freshly available boron plays an important role in this processes, but also redeposited layers provide newly available boron atoms. This thesis demonstrates boronization effects on the high performance of W7-X and answers question about the underlying mechanisms. These findings are discussed in the view of the future steady-state experimental campaign of W7-X, when steady-state discharges of up-to 30 min duration are planned. [1] J. Winter et al. “Boronization in Textor”. In:Journal of Nuclear Materials 162-164(1989), pp. 713–723.

The Third IAEA Technical Meeting on Fusion Data Processing, Validation and Analysis took place at the IAEA Headquarters in Vienna, Austria, from 28-31 May and brought together more than 60 scientists and engineers from 19 Member States, the European Commission and ITER Organisation working on data analysis and machine learning (ML) methods for the processing of fusion data, collected either from experimental diagnostics or from plasma simulations. 'Accurate data processing leads to a better understanding of the physics related to nuclear fusion research. It is essential for the careful estimate of the error bars of the raw measurements and processed data,' said D. Mazon (Chair, CEA, France) in his introductory talk and progress has been shown in this direction during the meeting. In particular the meeting discussed new developments in fusion R&D applications in the following areas: inversion techniques, such as tomography; magnetic topology reconstruction, such as equilibrium reconstructions; system identification; scaling laws determination and their accuracy for extrapolation from current machines to fusion reactors; model-based algorithms for control applications; identification of spurious and undesired events, such as disruptive phenomena or hot spots in infra-red images using sophisticated mathematical techniques like neural network and support vector machines. Discussions also focused on the potential use of these techniques for ITER, in particular, on their relevance to the first ITER plasma. The use of Integrated Modelling & Analysis Suite (IMAS) infrastructure - a convenient platform that allows implementation of different simulation codes in the same format - and synthetic diagnostics that could be coupled with this structure, in order to test the different mathematical approaches developed by the meeting participants, were explored in detail. Recent results and progress in the development of new tools will be discussed during the next meeting on the topic, scheduled in 2021.

The recently modified EMC3-EIRENE code package has been widely applied as an edge-plasmaanalysis tool and resulted in successful validation against various measured trends seen instellarator and tokamak plasma boundaries. It has been shown that the code package applied forWendelstein 7-X (W7-X) discharges in the interpretive mode can assess the impact of impurityeffects on the electron density, measured by a set of Langmuir probes. In particular the spatialquantification of impurities and effects from the effective charge state Zeff and effective massmeff, which are non-trivial to record by diagnostics, were examined. The results showed thatearlier assumptions of the effective charge-state distribution and effective mass for reportedLangmuir probe measurements must be revised. Subsequently, reprocessing these measurementswith code-interpreted spatial profiles of the effective charge state and effective mass led to anoverall improved physical consistency.

The Wendelstein 7-X (W7-X) fusion experiment is aimed at proving that the stellarator concept is suitable for a future fusion reactor. Therefore, it is designed for steady-state plasmas of up to 30 min, which means that the thermal control of the plasma-facing components (PFCs) is of vital importance to prevent damage to the device.In this paper an overview of the design of the Near Real-Time Image Diagnostic System (hereinafter called “the System”) for PFCs protection in W7-X is presented. The goal of the System is to monitor the PFCs with high risk of permanent damage due to local overheating during plasma operations and to send alarms to the interlock system. The monitoring of the PFCs is based on thermographic and video cameras, and their video streams are analyzed by means of graphics processing unit–based computer vision techniques to detect the strike line, hot spots, and other thermal events. The video streams and the detected thermal events are displayed online in the control room in the ...

Observations on confinement in the first experimental campaign on the optimized Stellarator Wendelstein 7-X are summarized. In this phase W7-X was equipped with five inboard limiters only and thus the discharge length restricted to avoid local overheating. Stationary plasmas are limited to low densities

For the first time, the optimized stellarator Wendelstein 7-X has operated with an island divertor. An operation regime in hydrogen was found in which the total plasma radiation approached the absorbed heating power without noticeable loss of stored energy. The divertor thermography recorded simultaneously a strong reduction of the heat load on all divertor targets, indicating almost complete power detachment. This operation regime was stably sustained over several energy confinement times until the preprogrammed end of the discharge. The plasma radiation is mainly due to oxygen and is located at the plasma edge. This plasma scenario is reproducible and robust at various heating powers, plasma densities, and gas fueling locations. These experimental results show that the island divertor concept actually works and displays good power dissipation potential, producing a promising exhaust concept for the stellarator reactor line.

The paper presents procedures which have been developed for a quantitative analysis of the divertor power deposition at Wendelstein 7-X. The development of these tools is motivated by the need to compare and verify scientific and engineering predictions with experimental measurements. The measurements have been performed by means of the thermographic diagnostic system, capable of exploring the divertor heat loads, with the aim to study the heat load symmetry, compare footprint patterns with theoretical expectations, but also investigate leading edges and divertor misalignment. In order to compare measurements and numerical calculations, an accurate mapping between the camera data, the divertor geometry and the 3D CAD models has been constructed. This mapping allows to find a correspondence between the data in different representations, simplifying data interpolation and visualization. This also provides a high resolution model of the target surface to compare numerical heat deposition calculations with experimental results from different cameras.

Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened. Peer Reviewed Article escrit per 365 autors/autores: Labit, B.; Eich, T.; Harrer, G. F.; Wolfrum, E.; Bernert, M.; Dunne, M. G.; Frassinetti, L.; Hennequin, P.; Maurizio, R.; Merle, A.; Meyer, H.; Saarelma, S.; Sheikh, U.; Adamek, J.; Agostini, M.; Aguiam, D.; Akers, R.; Albanese, R.; Albert, C.; Alessi, E.; Ambrosino, R.; Andr be, Y.; Angioni, C.; Apruzzese, G.; Aradi, M.; Arnichand, H.; Auriemma, F.; Avdeeva, G.; Ayllon-Guerola, J. M.; Bagnato, F.; Bandaru, V. K.; Barnes, M.; Barrera-Orte, L.; Bettini, P.; Bilato, R.; Biletskyi, O.; Bilkova, P.; Bin, W.; Blanchard, P.; Blanken, T.; Bobkov, V.; Bock, A.; Boeyaert, D.; Bogar, K.; Bogar, O.; Bohm, P.; Bolzonella, T.; Bombarda, F.; Boncagni, L.; Bouquey, F.; Bowman, C.; Brezinsek, S.; Brida, D.; Brunetti, D.; Bucalossi, J.; Buchanan, J.; Buermans, J.; Bufferand, H.; Buller, S.; Buratti, P.; Burckhart, A.; Calabr, G.; Calacci, L.; Camenen, Y.; Cannas, B.; Cano Megías, P.; Carnevale, D.; Carpanese, F.; Carr, M.; Carralero, D.; Carraro, L.; Casolari, A.; Cathey, A.; Causa, F.; Cavedon, M.; Cecconello, M.; Ceccuzzi, S.; Cerovsky, J.; Chapman, S.; Chmielewski, P.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Coda, S.; Coelho, R.; Colas, L.; Colette, D.; Cordaro, L.; Cordella, F.; Costea, S.; Coster, D.; Cruz Zabala, D. J.; Cseh, G.; Czarnecka, A.; Cziegler, I.; D’Arcangelo, O.; Dal Molin, A.; David, P.; De Carolis, G.; De Oliveira, H.; Decker, J.; Dejarnac, R.; Delogu, R.; Den Harder, N.; Dimitrova, M.; Dolizy, F.; Domínguez-Palacios Durán, J. J.; Douai, D.; Drenik, A.; Dreval, M.; Dudson, B.; Dunai, D.; Duval, B. P.; Dux, R.; Elmore, S.; Embréus, O.; Erds, B.; Fable, E.; Faitsch, M.; Fanni, A.; Farnik, M.; Faust, I.; Faustin, J.; Fedorczak, N.; Felici, F.; Feng, S.; Feng, X.; Ferreira, J.; Ferr, G.; Février, O.; Ficker, O.; Figini, L.; Figueiredo, A.; Fil, A.; Fontana, M.; Francesco, M.; Fuchs, C.; Futatani, S.; Gabellieri, L.; Gadariya, D.; Gahle, D.; Galassi, D.; Gałązka, K.; Galdon-Quiroga, J.; Galeani, S.; Gallart, D.; Gallo, A.; Galperti, C.; Garavaglia, S.; Garcia, J.; Garcia-Lopez, J.; Garcia-Mu oz, M.; Garzotti, L.; Gath, J.; Geiger, B.; Giacomelli, L.; Giannone, L.; Gibson, S.; Gil, L.; Giovannozzi, E.; Giruzzi, G.; Gobbin, M.; Gonzalez-Martin, J.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Grekov, D. 1; Grenfell, G.; Griener, M.; Groth, M.; Grover, O.; Gruca, M.; Gude, A.; Guimarais, L.; Gyergyek, T.; Hacek, P.; Hakola, A.; Ham, C.; Happel, T.; Harrison, J.; Havranek, A.; Hawke, J.; Henderson, S.; Hesslow, L.; Hitzler, F.; Hnat, B.; Hobirk, J.; Hoelzl, M.; Hogeweij, D.; Hopf, C.; Hoppe, M.; Horacek, J.; Hron, M.; Huang, Z.; Iantchenko, A.; Iglesias, D.; Igochine, V.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Ivanova-Stanik, I.; Jacobsen, A.; Jakubowski, M.; Janky, F.; Jardin, A.; Jaulmes, F.; Jensen, T.; Jonsson, T.; Kallenbach, A.; Kappatou, A.; Karpushov, A.; Kasilov, S.; Kazakov, Y.; Kazantzidis, P. V.; Keeling, D.; Kelemen, M.; Kendl, A.; Kernbichler, W.; Kirk, A.; Kocsis, G.; Komm, M.; Kong, M.; Korovin, V.; Koubiti, M.; Kovacic, J.; Krawczyk, N.; Krieger, K.; Kripner, L.; Křivská, A.; Kudlacek, O.; Kulyk, Y.; Kurki-Suonio, T.; Kwiatkowski, R.; Laggner, F.; Laguardia, L.; Lahtinen, A.; Lang, P.; Likonen, J.; Lipschultz, B.; Liu, F.; Lombroni, R.; Lorenzini, R.; Loschiavo, V. P.; Lunt, T.; MacUsova, E.; Madsen, J.; Maggiora, R.; Maljaars, B.; Manas, P.; Mantica, P.; Mantsinen, M. J.; Manz, P.; Maraschek, M.; Marchenko, V.; Marchetto, C.; Mariani, A.; Marini, C.; Markovic, T.; Marrelli, L.; Martin, P.; Martín Solís, J. R.; Martitsch, A.; Mastrostefano, S.; Matos, F.; Matthews, G.; Mayoral, M.-L.; Mazon, D.; Mazzotta, C.; Mc Carthy, P.; McClements, K.; McDermott, R.; McMillan, B.; Meineri, C.; Menkovski, V.; Meshcheriakov, D.; Messmer, M.; Micheletti, D.; Milanesio, D.; Militello, F.; Miron, I. G.; Mlynar, J.; Moiseenko, V.; Molina Cabrera, P. A.; Morales, J.; Moret, J.-M.; Moro, A.; Moulton, D.; Nabais, F.; Naulin, V.; Naydenkova, D.; Nem, R. D.; Nespoli, F.; Newton, S.; Nielsen, A. H.; Nielsen, S. K.; Nikolaeva, V.; Nocente, M.; Nowak, S.; Oberkofler, M.; Ochoukov, R.; Ollus, P.; Olsen, J.; Omotani, J.; Ongena, J.; Orain, F.; Orsitto, F. P.; Paccagnella, R.; Palha, A.; Panaccione, L.; Panek, R.; Panjan, M.; Papp, G.; Paradela Perez, I.; Parra, F.; Passeri, M.; Pau, A.; Pautasso, G.; Pavlichenko, R.; Perek, A.; Pericoli Radolfini, V.; Pesamosca, F.; Peterka, M.; Petrzilka, V.; Piergotti, V.; Pigatto, L.; Piovesan, P.; Piron, C.; Piron, L.; Plyusnin, V.; Pokol, G.; Poli, E.; Pölöskei, P.; Popov, T.; Popovic, Z.; Pór, G.; Porte, L.; Pucella, G.; Puiatti, M. E.; Pütterich, T.; Rabinski, M.; Juul Rasmussen, J.; Rasmussen, J.; Rattá, G. A.; Ratynskaia, S.; Ravensbergen, T.; Réfy, D.; Reich, M.; Reimerdes, H.; Reimold, F.; Reiser, D.; Reux, C.; Reznik, S.; Ricci, D.; Rispoli, N.; Rivero-Rodriguez, J. F.; Rocchi, G.; Rodriguez-Ramos, M.; Romano, A.; Rosato, J.; Rubinacci, G.; Rubino, G.; Ryan, D. A.; Salewski, M.; Salmi, A.; Samaddar, D.; Sanchis-Sanchez, L.; Santos, J.; Särkimäki, K.; Sassano, M.; Sauter, O.; Scannell, R.; Scheffer, M.; Schneider, B. S.; Schneider, P.; Schrittwieser, R.; Schubert, M.; Seidl, J.; Seliunin, E.; Sharapov, S.; Sheeba, R. R.; Sias, G.; Sieglin, B.; Silva, C.; Sipilä, S.; Smith, S.; Snicker, A.; Solano, E. R.; Hansen, S. K.; Soria-Hoyo, C.; Sorokovoy, E.; Sozzi, C.; Sperduti, A.; Spizzo, G.; Spolaore, M.; Stejner, M.; Stipani, L.; Stober, J.; Strand, P.; Sun, H.; Suttrop, W.; Sytnykov, D.; Szepesi, T.; Tál, B.; Tala, T.; Tardini, G.; Tardocchi, M.; Teplukhina, A.; Terranova, D.; Testa, D.; Theiler, C.; Thorén, E.; Thornton, A.; Tilia, B.; Tolias, P.; Tomes, M.; Toscano-Jimenez, M.; Tsironis, C.; Tsui, C.; Tudisco, O.; Urban, J.; Valisa, M.; Vallar, M.; Vallejos Olivares, P.; Valovic, M.; Van Vugt, D.; Vanovac, B.; Varje, J.; Varju, J.; Varoutis, S. 1; Vartanian, S.; Vasilovici, O.; Vega, J.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vicente, J.; Viezzer, E.; Villone, F.; Voitsekhovitch, I.; Voltolina, D.; Vondracek, P.; Vu, N. M. T.; Walkden, N.; Wauters, T.; Weiland, M.; Weinzettl, V.; Wensing, M.; Wiesen, S.; Wiesenberger, M.; Wilkie, G.; Willensdorfer, M.; Wischmeier, M.; Wu, K.; Xiang, L.; Zagorski, R.; Zaloga, D.; Zanca, P.; Zaplotnik, R.; Zebrowski, J.; Zhang, W.; Zisis, A.; Zoletnik, S.; Zuin, M.

Wendelstein 7-X is a highly optimized stellarator that went into operation in 2015. With a 30 cubic meter volume, a superconducting coil system operating at 2.5 T, and steady-state heating capability of eventually up to 10 MW, it was built to demonstrate the benefits of optimized stellarators at parameters approaching those of a fusion power plant. We report here on the first results with the test divertor installed, during the second operation phase, which was performed in the second half of 2017. Operation with a divertor, and the addition of several new fueling systems, allowed higher density operation in hydrogen as well as helium. The effects that higher density operation had on both divertor operation and global confinement will be described. In particular, at high densities detachment was observed, and the highest fusion triple product for a stellarator was achieved.

The control of rotational transform in Wendelstein 7-X (W7-X) is key to both the island divertor operation and safety of plasma facing components. The island divenor concept in W7-X relies on an edge flux surface with rotational transform of t = 1 resonating with an intrinsic n/m = 5/5 resonance to form a five lobed island chain. This island chain intersects with divertor plates to give rise to the island divertor. Changes in the relative position of the rational surface and the divertor plates can result in changes in divenor performance, thus the control of the rotational transform is essential to operation of the W7-X device. During the first divertor campaign electromagnetic loads resulted in elastic deformations of the shaped modular stellarator coils. Such deformations made these coils more planar, reducing the vacuum rotational transform, subsequently shifting the t = 1 resonance outward. Unintended plasma wall interactions provided the first clear evidence of this effect during plasma operation. Flux surface measurements were utilized to estimate the correct level of current in the planar coils for correction of t, and found to be around similar to-100 A. Scans the planar coil current for iota correction were performed during plasma operation. These measurements suggest planar coil currents between -250 and similar to 0 A would place the strike lines at the designed distance from the pumping gaps. Divertor Langmuir and upstream probe measurements confirm these estimates along with measurements of divertor neutral gas pressure.

The paper presents experimental observations and simulations for the effects of toroidal plasma current on divertor power depositions on W7-X. With increasing toroidal current accompanying changes in the island geometry result in a sweep of the strike line and a redistribution of the heat flux footprints. Good agreement between experiments, which partly used electron cyclotron current drive to generate an additional toroidal current contribution, and modelling using field line tracing in vacuum magnetic fields including an ad-hoc toroidal current on the magnetic axis is found for both standard and low-iota magnetic configurations.

International audience; The Wendelstein 7-X (W7-X) optimized stellarator fusion experiment,which went into operation in 2015, has been operating since 2017 with anun-cooled modular graphite divertor. This allowed first divertor physicsstudies to be performed at pulse energies up to 80 MJ, as opposed to 4MJ in the first operation phase, where five inboard limiters wereinstalled instead of a divertor. This, and a number of other upgrades tothe device capabilities, allowed extension into regimes of higher plasmadensity, heating power, and performance overall, e.g. setting a newstellarator world record triple product. The paper focuses on the firstphysics studies of how the island divertor works. The plasma heat loadsarrive to a very high degree on the divertor plates, with only minorheat loads seen on other components, in particular baffle structuresbuilt in to aid neutral compression. The strike line shapes andlocations change significantly from one magnetic configuration toanother, in very much the same way that codes had predicted they would.Strike-line widths are as large as 10 cm, and the wetted areas alsolarge, up to about 1.5 m2, which bodes well for futureoperation phases. Peak local heat loads onto the divertor were ingeneral benign and project below the 10 MW m?2 limit of thefuture water-cooled divertor when operated with 10 MW of heating power,with the exception of low-density attached operation in the high-iotaconfiguration. The most notable result was the complete (in all 10divertor units) heat-flux detachment obtained at high-density operationin hydrogen.

The 3D effects on divertor heat loads have been investigated for performance-optimized island divertor configurations at Wendelstein 7-X with 3D modeling and IR camera measurements. A new high mirror configuration optimized for more stable island divertor operation due to reduced bootstrap currents and a more even heat load distribution between the main divertor targets has been investigated for the first time numerically and experimentally.Transport calculations with EMC3-EIRENE show a strong dependence of the heat flux distributions on the configurations and the details of the 3D island geometry. IR camera measurements confirm the predictions concerning the global heat load distributions for the standard configuration but show more even distributions between the main wetted divertor plates in the high mirror configurations. The local heat load profiles show offsets in their position of up to 5 cm to each other and averaged peak heat fluxes varying between 0.5 MWm−2 and 2.2 MWm−2 depending on the divertor module considered. These heat flux asymmetries complicate the local matching of profiles between experiment and 3D modeling.The 3D equilibrium of a high mirror high-performance scenario predicted by the HINT code has been investigated with EMC3-EIRENE for the first time to anticipate plasma response in higher performance scenarios. The island divertor is preserved for β=3% despite enhanced stochasticity. However, the islands are increased in size while the path lengths for parallel transport are reduced causing a substantial change in the divertor heat flux patterns. Keywords: Stellarator, Island divertor, Scrape-off layer, Heat fluxes, Equilibrium

a versatile gas injection system and an absolute calibrated spectroscopic observation system with more than 200 lines-of-sight have been installed in two divertor locations in the Wendelstein 7-X stellarator. This powerful setup enables among other measurements to infer Te and ne in the divertor region by means of line-ratio spectroscopy on thermal helium. This is a robust and routinely used diagnostic at W7-X and this work describes its main characteristics and shows some examples of Te and ne profiles. In particular it is shown that the inclusion of high Rydberg states in the helium collisional-radiative model is necessary to obtain a realistic estimate of Te. The applicability of the diagnostic is currently being extended towards the low Te and high ne conditions occurring during detachment by means of line-ratio spectroscopy on neon. Initial Te, ne profiles obtained with a neon-beam are here presented.

This report summarizes the contributions presented at the 3rd IAEA technical meeting on divertor concepts, held in Vienna, Austria, 4–7 November 2019. The meeting brought together more than 70 experts from nuclear fusion research sites worldwide to discuss the different aspects that the divertor design and fusion machine operation involve, from ITER divertor developments to innovative technologies for future DEMO divertor. The main topics of the meeting were: divertor and confinement; radiative power exhaust; scrape-off layer (SOL) and divertor physics; steady state operation and transient heat loads; plasma facing components materials and heat exhaust for steady state operation; and divertors for DEMO and future power reactors.

openaire: EC/H2020/633053/EU//EUROfusion The neutral beam deposition model in the BEAMS3D code is validated against neutral beam attenuation data from Wendelstein 7-X (W7-X). A set of experimental discharges where the neutral beam injection system of W7-X was utilized were reconstructed. These discharges scanned the magnetic configurations and plasma densities of W7-X. The equilibrium reconstructions were performed using STELLOPT which calculates three-dimensional self-consistent ideal magnetohydrodynamic equilibria and kinetic profiles. These reconstructions leveraged new capabilities to incorporate electron cyclotron emission and X-ray imaging diagnostics in the STELLOPT code. The reconstructed equilibria and profiles served as inputs for BEAMS3D calculations of neutral beam deposition in W7-X. It is found that if reconstructed kinetic profiles are utilized, good agreement between measured and simulated beam attenuation is found. As deposition models provide initial conditions for fast-ion slowing down calculations, this work provides a first steptowards validating our ability to predict fast ion confinement in stellarators.

Two graphite divertor elements called scrapers have been installed on the Wendelstein 7-X stellarator in the throat of the magnetic island divertor. To diagnose one, we have designed, built, calibrated, and installed a new infrared/visible imaging endoscope system to enable detailed observations of the plasma interactions and heat loads at one of the scrapers and the nearby divertor surfaces. The new system uses a shuttered pinhole-protected pair of 90° off-axis 228 mm focal length aluminum parabolic mirrors, and two flat turning metal mirrors, to send light to a sapphire vacuum window 1.6 meters away, beyond which we have co-located telephoto lens-based infrared and visible cameras. The back-to-back off-axis parabolas serve to cancel out most aberrations, enabling the use of off-the-shelf commercial optics outside of the vessel. For the infrared, we use a 3-5 μm 1-megapixel FLIR SC8303HD camera and for the visible, a 5-megapixel CMOS PCO 5.5 edge camera. A short 1-m quartz pickoff fiber is used to send 200-1100 nm light to a compact spectrometer, also located in the same iron shield box as the cameras. The camera field of view covers the 700 mm length of the scraper, and includes locations monitored by thermocouples and Langmuir probes embedded in some of the scraper tiles. Predicted and actual optical test performances of the overall system are compared.

Wendelstein 7-X aims at quasi-steady state operation with up to 10 MW of heating power for 30 min. Power exhaust will be handled predominantly via 10 actively water cooled CFC (carbon-fiber-reinforced carbon) based divertor units designed to withstand power loads of 10 MW/m2 locally in steady state. If local loads exceed this value, a risk of local delamination of the CFC and failure of entire divertor modules arises. Infrared endoscopes to monitor all main plasma facing components are being prepared, and near real time software tools are under development to identify areas of excessive temperature rise, to distinguish them from non-critical events, and to trigger alarms. Tests with different cameras were made in the recent campaign. Long pulse operation enforces additional diagnostic design constraints: for example, the optics need to be thermally decoupled from the endoscope housing. In the upcoming experimental campaign, a graphite scraper element, in front of the island divertor throat, will be tested as a possible means to protect the divertor pumping gap edges during the transient discharge evolution.

Wendelstein 7-X (W7‐X) is the world’s largest superconducting nuclear fusion experiment of the optimized stellarator type. In the first Operation Phase (OP1.1) helium and hydrogen plasmas were studied in limiter configuration. The heating energy was limited to 4 MJ and the main purpose of that campaign was the integral commissioning of the machine and diagnostics, which was achieved very successfully. Already from the beginning a comprehensive set of diagnostics was available to study the plasma. On the path towards high-power, high-performance plasmas, W7‐X will be stepwise upgraded from an inertially cooled (OP1.2, limited to 80 MJ) to an actively cooled island divertor (OP2, 10 MW steady-state plasma operation). The machine is prepared for OP1.2 with 10 inertially cooled divertor units, and the experimental campaign has started recently. The paper describes a subset of diagnostics which will be available for OP1.2 to study the plasma edge, divertor and scrape-off layer physics including those already available for OP1.1, plus modifications, upgrades and new systems. The focus of this summary will be on technical and engineering aspects, like feasibility and assembly but also on reliability, thermal loads and shielding against magnetic fields.

Transport properties of impurities in stellarators are sensitive to potential variations on flux-surfaces, and thermodynamic forces from the bulk plasma which might be affected by changes in magnetic field properties. In this contribution, the effect of planar coil current changes on the core impurity behavior in the first operating phase of Wendelstein 7-X (W7-X) is described. Hydrogen plasmas, at different heating power levels were investigated showing Core-Electron-Root-Confinement (CERC) conditions, characterized by peaked Te profiles and positive radial electric fields in the plasma core. Impurity spectra in the x-ray emission range between 1 and 10 keV were measured by the Pulse Height Analysis (PHA) diagnostic. The line-integrated measurements of transitions in sulphur, chlorine and argon are in the focus of this analysis. First observations indicate that the scans of the magnetic configuration did not affect strongly the impurity concentration in the core. Clear differences in the impurity content, however, have been observed when the ECRH power was changed, which affects the bulk plasma properties.

During the first operational phase (referred to as OP1.1) of the new Wendelstein 7-X (W7-X) stellarator, five poloidal graphite limiters were mounted on the inboard side of the vacuum vessel, one in each of the five toroidal modules which form the W7-X vacuum vessel. Each limiter consisted of nine specially shaped graphite tiles, designed to conform to the last closed field line geometry in the bean-shaped section of the standard OP1.1 magnetic field configuration (Sunn Pedersen et al 2015 Nucl. Fusion 55 126001). We observed the limiters with multiple infrared and visible camera systems, as well as filtered photomultipliers. Power loads are calculated from infrared (IR) temperature measurements using THEODOR, and heating patterns (dual stripes) compare well with field line mapping and EMC3-EIRENE predictions. While the poloidal symmetry of the heat loads was excellent, the toroidal heating pattern showed up to a factor of 2× variation, with peak heat loads on Limiter 1. The total power intercepted by the limiters was up to ~60% of the input ECRH heating power. Calorimetry using bulk tile heating (measured via post-shot IR thermography) on Limiter 3 showed a difference between short high power discharges, and longer lower power ones, with regards to the fraction of energy deposited on the limiters. Finally, fast heating transients, with frequency >1 kHz were detected, and their visibility was enhanced by the presence of surface coatings which developed on the limiters by the end of the campaign.

International audience; This conference report summarizes the contributions to, and discussions at, the 21st Joint EU-US Transport Task Force workshop, held in Leysin, Switzerland, during 5-8 September 2016. The workshop was organized under 8 topics: progress towards full-F kinetic turbulence simulation; high and low Z impurity transport, control and effects on plasma confinement; 3D effects on core and edge transport (including MHD, external fields and stellarators); predictive experimental design; electron heat transport and multi-scale integration; understanding power decay length in the SOL; role of the SOL in the L-H transition; validation of fundamental turbulence properties against turbulence measurements. This report follows the same structure.

Classical particle drifts are known to have substantial impacts on fluxes of particles and heat through the edge plasmas in both tokamaks and stellarators. Here we present results from the first dedicated investigation of drift effects in the W7-X stellarator. By comparing similar plasma discharges conducted with a forward- and reverse-directed magnetic field, the impacts of drifts could be isolated through the observation of up-down asymmetries in flux profiles on the divertor targets. In low-density plasmas, the radial locations of the strike lines (i.e. peaks in the target heat flux profiles) exhibited discrepancies of up to 3 cm that reversed upon magnetic field reversal. In addition, asymmetric heat loads were observed in regions of the target that are shadowed by other targets from parallel flux from the core plasma. A comparison of these asymmetric features with the footprints of key topological regions of the edge magnetic field on the divertor suggests that the main driver of the asymmetries at low density is poloidal E × B drift due to radial electric fields in the scrape-off layer and private flux region. In higher-density plasmas, upper and lower targets collected non-ambipolar currents with opposite signs that also inverted upon field reversal. Overall, in these experiments, almost all up-down asymmetry could be attributed to the field reversal and, therefore, field-dependent drifts.

Radiative power exhaust by impurity seeding was demonstrated for the first time in island divertor configurations at the stellarator Wendelstein 7-X. Feasibility of stable plasma operation was shown during seeding with both neon (Ne) and nitrogen (N2). High radiative power losses (>=80%) were found to reduce the divertor heat loads globally by 2/3 with both seeding gases injected at a single toroidal location into one of five magnetic islands. Heat flux detachment was achieved for the price of a loss of (>=-15%) in the stored energy. Ne seeding allows for sustained enhancement of edge radiation with a very slow decay of line emission of several tens of seconds after the end of the injection indicating a high recycling of this noble gas at the carbon main plasma facing components. In N2 seeded discharges it is shown that a response of line emission and plasma parameters is in correlation to the puff duration which indicates a higher level of absorption of this seeding gas in the wall. Continuous N2 seeding results in global cooling of the scrape-off layer (SOL) and decay of radiation over several seconds after the injection. Damping of counter-streaming SOL flows, and divertor particle fluxes induced by Ne and N2 seeding have been measured and provide evidence for a reduction of the convective part of the divertor heat fluxes. Losses in density in response to seeding can be compensated by feedback controlled divertor fueling. The controlled reduction of heat fluxes within this complex 3D edge island geometry is a very promising finding concerning detachment control in a future all-metal divertor.

The W7-X scrape-off layer (SOL) with its characteristic magnetic island chain has been investigated using electric probes mounted on a reciprocating manipulator close to the outboard mid-plane. A survey of the W7-X configuration space shows that the presence and particular topology of magnetic islands significantly affects the SOL profiles of electron temperature, density, electric field and plasma flows. Particularly relevant for divertor operation, very wide SOL heat flux profiles have been observed in some magnetic configurations, which we link to the presence of magnetic islands. In these situations, the islands can feature a local minimum of the plasma potential accompanied by a direction reversal of E x B driven dynamics measured by probe arrays.

Impurity seeding experiments during the start-up limiter campaign of Wendelstein 7-X provide first evidence for a localization effect of the 3D magnetic edge structure on the seeded impurities and their radiation distribution and cooling effect. Moreover, species dependencies have been seen. Nitrogen was observed to cool the entire edge plasma, with a stronger electron temperature reduction measured at the downstream position at the limiter. The radiation was limited at the periphery of the confined region. Both the temperature reduction and the radiation enhancement were directly correlated to the injection of the coolant gas. Mitigation of the limiter heat loads was also measured. Neon was observed to affect also the confined plasma with a long-lasting radiation and a cooling of the entire plasma.

Wendelstein 7-X, a superconducting optimized stellarator built in Greifswald/Germany, started its first plasmas with the last closed flux surface (LCFS) defined by 5 uncooled graphite limiters in December 2015. At the end of the 10 weeks long experimental campaign (OP1.1) more than 20 independent diagnostic systems were in operation, allowing detailed studies of many interesting plasma phenomena. For example, fast neutral gas manometers supported by video cameras (including one fast-frame camera with frame rates of tens of kHz) as well as visible cameras with different interference filters, with field of views covering all ten half-modules of the stellarator, discovered a MARFE-like radiation zone on the inboard side of machine module 4. This structure is presumably triggered by an inadvertent plasma-wall interaction in module 4 resulting in a high impurity influx that terminates some discharges by radiation cooling. The main plasma parameters achieved in OP1.1 exceeded predicted values in discharges of a length reaching 6 s. Although OP1.1 is characterized by short pulses, many of the diagnostics are already designed for quasi-steady state operation of 30 min discharges heated at 10 MW of ECRH. An overview of diagnostic performance for OP1.1 is given, including some highlights from the physics campaigns.

A high-resolution imaging system, consisting of megapixel mid-IR and visible cameras along the same line of sight, has been prepared for the new W7-X stellarator and was operated during Operational Period 1.1 to view one of the five inboard graphite limiters. The radial line of sight, through a large diameter (184 mm clear aperture) uncoated sapphire window, couples a direct viewing 1344 × 784 pixel FLIR SC8303HD camera. A germanium beam-splitter sends visible light to a 1024 × 1024 pixel Allied Vision Technologies Prosilica GX1050 color camera. Both achieve sub-millimeter resolution on the 161 mm wide, inertially cooled, segmented graphite tiles. The IR and visible cameras are controlled via optical fibers over full Camera Link and dual GigE Ethernet (2 Gbit/s data rates) interfaces, respectively. While they are mounted outside the cryostat at a distance of 3.2 m from the limiter, they are close to a large magnetic trim coil and require soft iron shielding. We have taken IR data at 125 Hz to 1.25 kHz frame rates and seen that surface temperature increases in excess of 350 °C, especially on leading edges or defect hot spots. The IR camera sees heat-load stripe patterns on the limiter and has been used to infer limiter power fluxes (∼1-4.5 MW/m

As shown on DIII-D edge localized modes (ELMs) can be either completely eliminated or mitigated with resonant magnetic perturbation (RMP) fields. Two infrared cameras, separated 105° toroidally, were used to make simultaneous measurements of ELM heat loads with high frame rates. Without the RMP fields ELMs display a variety of different heat load dynamics and a range of toroidal variability that is characteristic of their 3D structure. Comparing radial averages there is no asymmetry between two toroidal locations. With RMP-mitigated ELMs, the variability in the radially averaged power loads is significantly reduced and toroidal asymmetries in power loads are introduced. In addition to RMP ELM suppression scenarios an RMP scenario with only very small ELMs and very good confinement has been achieved.

Measurements of electron density n e ( r , t ) and temperature T e ( r , t ) fields in the far edge ( r / a > 0.98) of the tokamak TEXTOR during application of a fast rotating resonant magnetic perturbations (RMP) show an in-phase modulation of both n e and T e coherent to the rotation frequency of the external RMP field, which is correlated to the local rotating magnetic topology. The comparison provides experimental evidence that a three-dimensional scrape-off layer is preserved for high RMP field rotation frequencies of ν RMP = 974 Hz. Additionally, we obtain the geometrical dimensions of the 3D helical flux tubes in terms of the helical and radial width and compare them to the magnetic topology modeled in vacuum approximation in a quasi-static approach. The geometrical dimensions of the experimentally detected flux tubes increase linearly with increasing perturbation current as prescribed by the magnetic topology. For high RMP amplitudes the pressure drop inside the helical flux tube is limited by an effectively enhanced radial transport.

The three-dimensional (3D) features of plasma edge profiles and wall interaction patterns induced by edge resonant magnetic perturbation fields (RMP) are discussed comparing TEXTOR and DIII-D. We show that the scrape-off layer (SOL) profiles and decay lengths depend the edge safety factor, the RMP base mode as well as on the plasma rotation during RMP application indicating modification of SOL transport by the 3D perturbation fields. This is compatible with channelling of particle and heat efflux along open perturbed field lines in the very edge of the plasma boundary into a completely re-arranged, helically striated 3D divertor footprint. The distribution of the measured divertor heat and particle fluxes at DIII-D match the vacuum modelled magnetic footprint topology in L-mode while in H-mode the striation width exceeds the modelled footprint width by 15–30%. This 3D structure of the measured heat and particle fluxes results in a new situation for the material erosion properties and initial quantification of the net-erosion within the 3D footprint shows in L-mode a 50% decrease of the chemical erosion yield and evidence for a comparably small 15-20% increase in physical sputtering. Extrapolation of these findings to ITER by vacuum modelling of the magnetic footprint for the actual ELM control coils shows a similar vacuum magnetic footprint topology as found at DIII-D during RMP ELM suppression. However, the open field lines escape the CFC covered ITER divertor area potentially transferring net-erosion characteristics from the CFC domain onto the Tungsten including so far unconsidered heat and particle loads on this sensitive material.

Stochastic boundaries in fusion devices have been investigated in tokamaks, stellarators and reversed field pinch experiments for many years. However, since edge localized modes (ELMs) have been successfully eliminated in H-mode plasmas at the DIII-D tokamak [1,2] with small, edge resonant magnetic perturbations, they have become a widely investigated topic in tokamaks. In DIII-D stochastic boundaries are produced by coils external to the plasma. The magnetic field there consists of field lines with very different connection lengths, which produces a three dimensional, heterogenous structure of stochastic volume. The most obvious manifestation of the perturbed plasma edge is the strike line splitting observable in heat and particle fluxes, which changes with collisionality. The interaction of the magnetic perturbation and the magnetic equilibrium is of a resonant nature and the structure of the stochastic volume is a strong function of q(95). This is observed as a modulation of e.g. electron temperature as measured by ECE or Thomson scattering. In this work we summarize recent experimental findings on properties of the stochastic boundary in DIII-D. (C) 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

Two very contrary particle confinement stages were obtained at TEXTOR-DED by application of resonant magnetic perturbations. On the one hand a spontaneous build up of the total number of particles N tot with correlated increase in the particle confinement time τ p was observed and on the other hand a controlled decrease of N tot and τ p – the so called stochastic particle pump out is seen. Numerical analysis of the perturbed magnetic field topology shows that both domains can be distinguished by the ratio of short connection length field lines touching a specific resonant flux surface (here the q = 5 / 2 surface) to the complete perturbed layer width. During improved particle confinement, the hyperbolic fixed points (X-points) of the pitch resonant islands are directly connected to the DED target followed by an ≲ 40 % increase in τ p . The subsequent increase in the E × B shear rate Ω E × B at the q = 5 / 2 surface and a steepening of ∇ n e ( r ) suggests a reduction of the radial particle transport. On the opposite, complete stochastisation of this island chain, i.e. a predominant diffusive field line characteristics, causes a ≲ 30 % decrease of τ p with a reduction in Ω E × B at the q = 5 / 2 surface and ∇ n e ( r ) indicating enhanced effective outward particle transport.

For given conditions, significant change in main particle and carbon transport is observed in TEXTOR under the action of the Dynamic Ergodic Divertor, DED, operating both in 6/2 and 3/1 basic modes. In particular, the stochastic layer created at the plasma edge by the applied perturbing field, is responsible for the so called pump out, PO, effect (observed in DIII-D and on JET during experiments of ELM mitigation) characterized by a decrease in the plasma density and the reduction in the intrinsic carbon concentration in the plasma core. For a sufficiently high value of the applied field strength, in DED 3/1 basic operational mode a 3/1 island connects to the wall, with related further enhancement of particle transport. With respect to carbon, this phenomenon simply leads to a further carbon de-contamination from the central plasma. (C) 2009 Elsevier B.V. All rights reserved.