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In fusion devices, the retention of the fusion fuel deuterium (D) and tritium (T) in plasma-facing components (PFCs) is a major concern. Measurement of their hydrogen isotope content gives insight into the retention physics. In FREDIS, two methods of thermal desorption are used for retention measurements: In Thermal Desorption Spectrometry (TDS) the samples are heated by 6 infrared lamps up to 1433 K with linear temperature ramps of up to 1.67 K/s. The desorbed gases are detected up to 100 amu/e with a double-QMS (Quadrupole Mass Spectrometer) that can distinguish between helium and D2 and uses an innovative differential pumping system. In a connected vacuum chamber, a ∅3 mm spot can be heated on the sample surface by a high energy Nd:YAG laser pulse (E0

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

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.

Electron- (ECWC) and ion- (ICWC) cyclotron wall conditioning are essential means for controlled fusion to modify the surface state of plasma-facing components in order to reduce impurity generation and fuel accumulation in the wall. Development of ECWC and ICWC requires characterization of neutral particle fluxes generated in discharges, because neutrals enhance the homogeneity of the conditioning, which may contribute to remote or shadowed areas, especially in the presence of a permanent magnetic field (e.g. W7-X, ITER). A time-of-flight neutral particle analyzer (ToF-NPA) with 4.07 m flight distance is employed to measure time- and energy-resolved low energetic (

In this work we present an analysis of fast-camera imaging of Alcator C-Mod's divertor region and of the plasma cross section in the W7-X stellarator. In C-Mod, the dynamics of intermittent edge plasma filaments are imaged at 397,100 frames per second in an oblique view. An automatic pipeline was developed to reconstruct the filament motion in the R, Z plane of the divertor using EFIT field line projections. Next, we added our fast-framing camera to W7-X's diagnostic suite for the 2017 run campaign, the first with a divertor. It has a tangential view of the machine cross section, including part of a divertor module. The view is coupled to the camera via a coherent fiber bundle. The spatial resolution is 2–4 cm and the typical plasma brightness limits the fastest usable frame rate to 30k frames/s. We observe filaments in the plasma boundary that rotate around the magnetic axis during radiative collapse; filaments were observed in steady state in the limiter configuration of W7-X by Kocsis et al. [1]. We also observe for the first time a 1–2 kHz quasi-coherent fluctuation in divertor target emissions in the W7-X “standard” magnetic configuration. Finally, a rough analysis of edge gas fueling efficiency using fast camera data and S/XB coefficients finds that fueling is less efficient at densities higher than 4 × 1019 m−3. Keywords: C-Mod, W7-X, Filaments, Blobs, Turbulence, Fluctuation

The toroidal magnetized system (TOMAS) plasma facility aims at complementary research on wall conditioning methods, plasma production and plasma–surface interaction studies. This paper explores for the first time the parameters in helium electron-cyclotron resonance (ECR) plasma and combined ECR + radio-frequency (RF) discharges in TOMAS. The ECR discharge in this work, at 2.45 GHz and 87.6 mT, is the main one for creating and maintaining the plasma, while the addition of RF power at 25 MHz allows to broaden the achievable electron temperature and density at a given gas flow, as evidenced by triple Langmuir probe measurements. This effect of the combined ECR + RF discharge provides flexibility to study particular aspects of wall conditioning techniques relevant to larger devices, or to approach plasma conditions relevant to fusion edge plasmas for particular surface interaction studies.

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.

Ion flow velocity measurement in the edge and scraper-off layer region is beneficial to understand the confinement related phenomenon in fusion devices such as impurity transport and plays an important role in impurity control. During the Wendelstein 7-X (W7-X) operation phase 1.2a, a multi-channel (MC) Mach probe mounted on the multi-purpose manipulator has been used to measure radial profiles of edge ion flow velocity. This MC-Mach probe consists of two polar and two radial arrays of directional Langmuir pins (28 pins in total) serving for different aims, of which the polar arrays could obtain a polar distribution of ion saturation current, while the radial arrays can be used to study the dynamic process of a radially propagated event. In this paper, we report the observation of the radially outward propagation of a low frequency mode with a speed of around 200 m/s. The first measurement of the radial ion flow velocity profile using the MC-Mach probe in the boundary plasma of the W7-X with an island divertor will also be presented.

For the study of electrostatic and magnetic properties of filaments characterizing the edge region of the stellarator experiment W7-X, a specifically designed insertable probe head was constructed within the framework of EUROfusion WP.S1 work package in collaboration between Consorzio RFX, IPP Greifswald and Forschungszentrum Mich. The probe head, named High Resolution Probe (HRP), was conceived to be installed on the mid-plane multi-purpose fast reciprocating manipulator on W7-X. Electromagnetic filamentary turbulent structures are found to characterize the edge region of different magnetic configurations including Reversed Field Pinch, stellarator and tokamak, where strong currents are associated also to ELM filamentary structures. The study of those phenomena in W7-X stellarator is of particular interest as the electromagnetic features of filaments are expected to become more relevant with the increase of the local plasma beta. In particular the aim is to provide information about the presence and the features of parallel current density associated to filamentary turbulent or ELM-like structures. Furthermore, the possibility to measure the time evolution of the flow radial profiles using the Mach probe array was considered as a further interesting part of the study, given the strong interplay expected between the turbulent fluctuations and the average flows. Further important information provided is the radial propagation of turbulent flux. The contribution will present the design development, the R&D studies and the applied solutions for the sensors embedded in the probe head. In particular, the presence of 140 GHz ECRH plasma environment represents one of the main challenges for reliable magnetic fluctuation measurements. First measurements were performed during the W7-X experimental campaign OP1.2b, where electromagnetic features of turbulence were measured with the probe, at different magnetic configurations.

For the in situ application of LID (Laser-Induced Desorption) as a space-resolved tritium retention diagnostic in ITER, the desorption behaviour of co-deposited deuterium (D) from beryllium (Be) layers is studied. In particular, the desorption efficiency dependence on laser pulse parameters is investigated for pulse durations of 1–20 ms and absorbed energy densities up to 5 MJ m−2. For these parameter scans homogenous Be/D layers were produced by High Power Impulse Magnetron Sputtering, with 10 μm thickness and 1.6 at% D. Almost 99% of the initial D can be desorbed with a single LID pulse. As the layers show a high D desorption temperature (ca. 800 K) in slow Thermal Desorption Spectrometry, an LID efficiency of only 50% is reached before Be melting. Microscopy reveals that in molten regions holes are formed, which could serve as desorption channels to facilitate gas release above the melting point. Hill formation and cracking are further modifications, but no layer destruction was observed in general.

Plasma-wall interaction (PWI) research is an active field of study in long-pulse operation in current magnetic confinement fusion devices, such as the Experimental Advanced Superconducting Tokamak (EAST). It is an urgent requirement to be able to investigate several key PWI issues, such as fuel retention, by in situ diagnostic methods. In this work, an in situ diagnostic method of laser-induced breakdown spectroscopy (LIBS) combined with optical emission spectroscopy (OES) is developed. The whole system is applied to study PWI and fill the research gap concerning the correlation between fuel retention and edge plasma conditions during long-pulse plasma operation conditions in EAST. The fuel retention intensity from LIBS on the first wall and the edge plasma condition from OES are monitored simultaneously during the long-pulse plasma operation in EAST. The results indicate that the deuterium (D) retention amount increases as the local edge D particle fluence increases. The results effectively demonstrate the potential of the LIBS method for in situ investigation of the fuel retention for PWI study in upcoming long-pulse fusion devices such as the International Thermonuclear Experimental Reactor (ITER).

We present a detailed overview and first results of the new laser blow-off system on the stellarator Wendelstein 7-X. The system allows impurity transport studies by the repetitive and controlled injection of different tracer ions into the plasma edge. A Nd:YAG laser is used to ablate a thin metal film, coated on a glass plate, with a repetition rate of up to 20 Hz. A remote-controlled adjustable optical system allows the variation of the laser spot diameter and enables the spot positioning to non-ablated areas on the target between laser pulses. During first experiments, clear spectral lines from higher ionization stages of the tracer ions have been observed in the X-ray to the extreme ultraviolet spectral range. The temporal behavior of the measured emission allows the estimate of transport properties, e.g., impurity transport times in the order of 100 ms. Although the strong injection of impurities is well detectable, the global plasma parameters are barely changed.

The paper reports on the optimization process of the soft X-ray pulse height analyzer installed on the Wendelstein 7-X (W7-X) stellarator. It is a 3-channel system that records X-ray spectra in the range from 0.6 to 19.6 keV. X-ray spectra, with a temporal and spatial resolution of 100 ms and 2.5 cm (depending on selected slit sizes), respectively, are line integrated along a line-of-sight that crosses near to the plasma center. In the second W7-X operation phase with a carbon test divertor unit, light impurities, e.g., carbon and oxygen, were observed as well as mid- to high-Z elements, e.g., sulfur, chlorine, chromium, manganese, iron, and nickel. In addition, X-ray lines from several tracer elements have been observed after the laser blow-off injection of different impurities, e.g., silicon, titanium, and iron, and during discharges with prefill or a gas puff of neon or argon. These measurements were achieved by optimizing light absorber-foil selection, which defines the detected energy range, and remotely controlled pinhole size, which defines photon flux. The identification of X-ray lines was confirmed by other spectroscopic diagnostics, e.g., by the High-Efficiency XUV Overview Spectrometer, HEXOS, and high-resolution X-ray imaging spectrometer, HR-XIS.

Plasma-wall interaction (PWI) is a crucial issue for high-performance operation and first wall lifetime of future fusion reactors. An in situ measurement and the control of fuel retention in the vacuum-vessel will be an important aspect for the safe operation of nuclear fusion devices. Laser-Induced Ablation Spectroscopy (LIAS) is an attractive option with promising capabilities for in situ characterization of fuel retention and co-deposition during plasma discharges. In this work, an in situ LIAS wall diagnostic system is developed for real-time during discharge diagnosis of the first wall on the EAST superconducting tokamak. A dedicated timing system, which offers the function of multiple temporal settings with high time resolution for LIAS measurement during long-pulse discharges, is successfully commissioned. The experimental setup of the LIAS system for the characterization of the first wall on EAST is presented in detail. First LIAS results obtained in EAST during high power discharges are presented. The results indicate that in situ measurements of deposition, co-deposition and dynamic retention are possible.

In the first experimental campaign of Wendelstein 7-X (W7-X), a combined probe head mounted on the multi-purpose manipulator has been used to measure the scrape-off layer (SOL) turbulence characteristics. The preliminary experimental results are summarized to illustrate the SOL turbulence properties in the limiter configuration on W7-X. In a standard limiter configuration, significant electrostatic fluctuations can be found in the near SOL, and the dominant frequency of fluctuation power is below 100 kHz. The auto-correlation spectrum power law decay factor is α ≈ −1 below 40 kHz and α ≈ –2 between 50 and 200 kHz. A broadband spectrum appears between 240 and 380 kHz with a low spectral power density, but a high cross-correlation coefficient. The statistical characteristics of turbulence are calculated by the two-point cross-correlation technique. A clear poloidal dispersion relation is found in the spectrum S(kθ, f), propagating along the ion diamagnetic drift direction with a group velocity (below 100 kHz) about 0.56 km/s in the near SOL in the laboratory frame. The poloidal correlation length is around 5–10 mm in SOL. The turbulence phase velocity is about 0.5–1 km/s when close to the last closed flux surface, which is comparable with the poloidal E × B drift speed.

Optical and laser-based diagnostics in ITER will use mirrors to transmit plasma radiation and laser light to the corresponding detectors and cameras. Mirrors will be sputtered by the fast plasma particles and contaminated by impurities leading to the degradation of the reflectivity and hampering the performance of corresponding diagnostics. Dedicated measures were proposed to minimize the impurity deposition comprising the use of shutters and fins inside diagnostic ducts to trap impurities on their way toward the mirror located in the end of these ducts. Modeling results predict at least 7-fold suppression of the deposition for the duct having four fins located at the distance of a half of a diameter from each other. The Mirror Station (MS) was designed to validate modeling predictions and to study the suppression of deposition inside of diagnostic ducts. The MS contained cylindrical and cone-shaped tubes of different lengths with smooth and shaped geometry of ducts. The MS was exposed in the midplane port of TEXTOR for about 3960 s of plasma operation. After exposure, no drastic suppression of deposition was observed in the cylindrical ducts with fins. In the conical tubes no deposition was detected outlining the advantages of a cone form.

Disruptions in tokamaks seem to be unavoidable. Consequences of disruptions are (i) high heat loads on plasma-facing components, (ii) large forces on the vacuum vessel, and (iii) the generation of runaway electron beams. In ITER, the thermal energy of the plasma needs to be evenly distributed on the first wall in order to prevent melting, forces from vertical displacement events have to be minimized, and the generation of runaway electrons suppressed. Massive gas injection using fast valves is a concept for disruption mitigation which is presently being explored in many tokamaks. Fast disruption mitigation valves based on an electromagnetic eddy current drive have been developed in Julich since the 1990s and models of various sizes have been built and are in operation in the TEXTOR, MAST, and JET tokamaks. A disruption mitigation valve for ITER is of necessity larger with an estimated injected gas volume of ∼20 kPa m3 [7] for runaway electron suppression and all materials used have to be resistant to much higher levels of neutron and gamma radiation than in existing tokamaks. During the last 5 years, the concept for an ITER prototype disruption mitigation valve has been developed up to the stage that a fully functional valve could be built and tested. Special emphasis was given to the development and functional testing of some critical items: (i) the injection chamber seal, (ii) the piston seal, (iii) the eddy current drive, and (iv) a braking mechanism to avoid too fast closure of the valve, which could damage the injection chamber seal. The concept of the valve will be introduced and detailed solutions and testing results for the critical items presented.

Wendelstein 7-X (W7-X), started operation in December 2015 with a limiter configuration. In conjunction with the multi-purpose manipulator, a carrier for fast reciprocating probe systems, the combined probe has been installed. This combined probe is able to measure the local electron temperatures and densities, magnetic field, the electric field and the plasma flow. These parameters are very useful in ascertaining the edge plasma perfomance. In addition, the field line tracing feature of the W7-X webservices was used to calculate the connection length along the path of the probe, for each configuration.

A laser based method combined with spectroscopy, such as laser-induced breakdown spectroscopy (LIBS) and laser-induced ablation spectroscopy (LIAS), is a promising technology for plasma-wall interaction studies. In this work, we report the development of in situ laser-based diagnostics (LIBS and LIAS) for the assessment of static and dynamic fuel retention on the first wall without removing the tiles between and during plasma discharges in the Experimental Advanced Superconducting Tokamak (EAST). The fuel retention on the first wall was measured after different wall conditioning methods and daily plasma discharges by in situ LIBS. The result indicates that the LIBS can be a useful tool to predict the wall condition in EAST. With the successful commissioning of a refined timing system for LIAS, an in situ approach to investigate fuel retention is proposed.

The Multi-Purpose Manipulator (MPM) for Wendelstein 7-X, at the Max-Planck-Institute for Plasma Physics in Greifswald (MPIPP), is used to transport electrical probes and targets to the edge of the inner vessel. It is a lock system, which is attached to the outer port in the equatorial plane of the cryostat vessel. At the parking position, the tip of the probe coincides with the inner vessel wall; a fully controlled movement into the edge plasma for all magnetic field configurations is feasible. The whole functionality of the MPM system is maintained by a distributed control system (DCS) based on Siemens PCS7, which is the recommended standard for machine and diagnostic control at W7-X. Beside standard scenarios like target exchange and generation of ultra-high vacuum conditions a high variety of parameter selections for stroke depth, velocity or acceleration is selectable in the sequence control system within safety limits. A sophisticated error handling facilitates a reliable remotely controlled operation without manual access over longer periods.

A group of edge diagnostics and modelling has been developed for investigation of synergy between 3D edge physics and plasma-wall interactions on Wendelstein 7-X (W7-X). Two endoscopes have been designed for visible and ultraviolet spectroscopy and tomography of the plasma edge, along with infrared thermography of the divertor tiles. 2D profiles of impurities (e.g. helium, carbon) will be measured by two endoscopes viewing the island divertor region in the plasma edge with a spatial resolution of

The investigation of edge plasmas at the stellarator W7-X requires a flexible tool for integration of a variety of different diagnostics as e.g. electrical probes, probing magnetic coils, material collection, or material exposition probes, and gas injection. A multi-purpose manipulator (MPM) system has been developed and attached to the W7-X vessel before the operational phase 1.1. The system was designed as user facility for many diagnostics, which can be mounted on a unique interface without breaking the W7-X vacuum. The manipulator system, located in the equatorial plane, transports the inserted diagnostic probe to the edge of the inner vacuum vessel. From there, the probe can be moved over a maximum distance of 350 mm to different positions inside the plasma with a maximum acceleration and deceleration of 30 m/s2. Acceleration, speed, and stroke depth are individually adjustable and programmable by a PLC system within predetermined limits. The MPM system can be equipped with multifunctional probes and is prepared for cooling/heating of the probe head, gas injection, and a flexible setting of the electrical diagnostic. In the paper, the MPM components, their functionalities, and the interface to the probe head are described. The scope of physical parameters is presented for the development of diagnostic probes applicable in the MPM user facility. This also includes the operational parameters for the movements and the electrical environment for measurements. The operation of the MPM during the OP1.1 is demonstrated on the application of a multipurpose probe head, including electric, magnetic, and thermal probes.

With the objective of understanding the energetic-particle loss mechanism in three-dimensional plasmas, a Faraday-cup-based fast-ion loss detector (FILD) was developed and installed in OP1.2b in Wendelstein 7-X (W7-X) as a collaboration between the National Institute for Fusion Science and Max-Planck-Institute for Plasma Physics. The FILD, which consists of double apertures and thin aluminum foils, was based on a magnetic spectrometer using the magnetic field of the fusion device. The double aperture limits the thermal ion, but allows the energetic ion to enter the FILD box. The thin aluminum foils serve as the ion collector. Orbit-following calculations were performed in order to find a suitable position for the FILD. The results indicated that barely co- and counter-going transit ions reached the FILD position mounted on the multi-purpose manipulator installed on W7-X. Moreover, because the injection angle of neutral beams injector installed on W7-X was relatively perpendicular, the target range of pitch angle was set to from 91 degrees to 150 degrees. An energy-and-pitch-angle map was created using a grid calculation code in order to decide on the position/size of aperture and aluminum foil. The grid calculation code indicates the position where the energetic ion would strike on the aluminum foil. Here, the map was used as the basis for designing an aluminum foil pattern with two energies and four pitch angle ranges. Also, note that the lower energy range was designed so that it accommodates a neutral beam injection energy of 55 kV. Measurements of beam ion losses were performed in neutral beam (NB) blip experiments, where concurrent increases and decreases of barely co-going transit beam ion losses due to NB injections were observed. Furthermore, this study validated that the FILD installed on a midplane manipulator probe is capable of probing a range of radii spanning 1.0–1.5 cm, over which the beam ion loss current would vary significantly.

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.

The high heat flux ion beam test facility MARION (Material Research ION Beam test Stand) has been used for ion source conditioning and material testing within the IEK-4 for several years. Mainly small samples have been tested in the past by perpendicular irradiation. In these experiments only 25% of the available power density of MARION (>80 MW/m 2 ) was used so far. The activation of these resources for material testing was realized in the first test phase of the new tungsten divertor element for the JET tokamak. The enhancement of MARION by a new protection scraper is shown. During the last experiments the need of a more flexible handling of the sample itself and of the facility became clear. Based on those experiences an upgrade concept was developed to prepare MARION for future demands allowing the handling of much bigger prototypes. Additionally main emphasis is given to the access to improved (surface) diagnostics for all orientations of the target.

In the frame of the ITER-like Wall (ILW) for the JET tokamak, a divertor row made of bulk tungsten material has been developed for the position where the outer strike point is located in most of the foreseen plasma configurations. In the absence of active cooling, this represents a formidable challenge when one considers the temperature reached by tungsten (TW,surf > 2000 °C) and the vertical gradient ∂T/∂z = 5 × 104 K/m. As the development is drawing to an end and most components are in production, actual 1:1 prototypes are exposed to an ion beam with a power density around 7 MW/m2 on the plasma-facing surface. Advantage is taken of the flexibility of the Marion facility to bombard the tungsten stack under shallow angles of incidence (∼6°) with a powerful beam of ions and neutrals (>70 MW/m2 on axis). The shallow angles are important, with respect to the toroidal wetted surface, for properly simulating the expected performance under actual tokamak conditions. The M arion tests have been used to validate for a few typical cases the thermal calculations that were steadily developed along with the tungsten tile and, at the same time, to gather information on the actual temperatures of individual components. The latter is an important factor to a finer estimation of the power handling capabilities.

A solid tungsten divertor row has been designed for JET in the frame of the ITER-like Wall project (ILW). The plasma-facing tiles are segmented in four stacks of tungsten lamellae oriented in the toroidal direction. Earlier estimations of the expected tile performance were carried out mostly for engineering purposes, to compare the permissible heat load with the power density of 7 MW/m2 originally specified for the ILW as a uniform load for 10 s. The global thermal model developed for the W modules delivers results for more realistic plasma footprints: the poloidal extension of the outer strike point was reduced from the full lamella width of 62 mm to ⩾15 mm. Model validation is given by the experimental exposure of a 1:1 prototype stack in the ion beam facility MARION (incidence ∼6°, load E ⩽ 66 MJ/m2 on the wetted surface). Spreading the deposited energy by appropriate sweeping over one or several stacks in the torus is beneficial for the tungsten lamellae and for the support structure.

Controlling the heat and particle fluxes in the plasma edge and on the plasma facing components is important for the safe and effective operation of every magnetically confined fusion device. This was attempted on Wendelstein 7-X in the first operational campaign, with the modification of the magnetic configuration by use of the trim coils and tuning the field coil currents, commonly named iota scan. Ideally, the heat loads on the five limiters are equal. However, they differ between each limiter and are non-uniform, due to the (relatively small) error fields caused by the misalignment of components. It is therefore necessary to study the influence of the configuration changes on the transport of heat and particles in the plasma edge caused by the application of error fields and the change of the magnetic configuration. In this paper the up-stream measurements conducted with the combined probe are compared to the downstream measurements with the DIAS infrared camera on the limiter.

Experimental evidence supporting the theory of hot tail runaway electron (RE) generation has been identified in TEXTOR disruptions. With higher temperature, more REs are generated during the thermal quench. Increasing the RE generation by increasing the temperature, an obvious RE plateau is observed even with low toroidal magnetic field (1.7 T). These results explain the previously found electron density threshold for RE generation.

The ability to exhaust helium as the fusion born plasma impurity is a critical requirement for burning plasmas. We demonstrate in this paper that resonant magnetic perturbation (RMP) fields can be used to actively manipulate helium exhaust characteristics. We present results from puff/pump studies at TEXTOR as example for a tokamak with a pumped limiter and from the Large Helical Device (LHD) with the closed helical divertor as example for a heliotron/stellarator device. For LHD, the effective helium confinement time τ p , H e ∗ is a factor of 7–8 higher in the low and high density regimes explored when compared to TEXTOR discharges. This is attributed to ion root impurity transport which is one particular impurity transport regime assessed experimentally at LHD and which facilitates helium penetration to the plasma core. However, when an edge magnetic island is induced by externally applied RMP fields, τ p , H e ∗ is decreased by up to 30 % and hence τ p , H e ∗ values closer to those of TEXTOR can be established. The combination of TEXTOR and LHD results suggest that a magnetic island induced by the RMP field in the plasma source region is an important ingredient for improving helium exhaust. The reduction in τ p , H e ∗ seen is caused by a combination of improved helium exhaust due to an enhanced coupling to the pumping systems, increased outward transport and a reduced fueling efficiency for the helium injected and recycling from the wall elements.

Within the development of new fusion devices like ITER and DEMO, material testing becomes more and more relevant. The material ion beam test facility MARION which was primarily designed for beam source development and commissioning is also used for high heat flux testing up to 120 MW/m2 power density. In order to improve this second field of activity the concept of MARION is adapted to material testing. To increase the flexibility and to decrease the costs for material testing operation the vacuum system has been changed by excluding the cryogenic pumps combined with an enhancement of the conventional vacuum pumping part. The reduction in pumping capacity has made a new optimization of the operation parameters necessary. The experimental results of this optimization can be used as calibration base for future heat load tests. They are shown in this paper, combined with the changes in operation range for this new, additional mode.

For the development of the tritium monitoring system in ITER the hydrogen isotope release by Laser-Induced Desorption (LID) from Be layers is studied to determine the laser parameters for a high desorption efficiency while minimising dust production and surface modifications is also pursued. Be layers of 1 µm thickness with 25–30 at% D and 3 × 1022 D/m2 comparable to JET-ILW areal concentrations [1] have been produced by High Power Impulse Magnetron Sputtering (HiPIMS) on ITER grade W. Laser pulses of 1, 5 and 10 ms duration heat the layer in vacuum in the Fuel REtention DIagnostic Setup (FREDIS) and release the retained D thermally. By mass spectrometry in FREDIS and subsequent Nuclear Reaction Analysis (NRA) inside the laser spot the desorbed and remaining D is quantified. While a pulse duration of 1 ms cannot fully desorb the deuterium, it is found that a single 5 or 10 ms laser pulse with an absorbed energy density of ca. 1.5 MJ/m2 corresponding to a heat flux factor around 20 MW√s/m2 leads to nearly complete desorption of the retained D. This encourages the development of a useful tritium monitoring system, although the present layers produce some dust due to local delamination of the layer on at least 11% of the heated surface (at 1.4 MJ/m2 absorbed energy within 5 ms) and lead to unavoidable crack formation. Keywords: Fuel retention, Beryllium, Tritium monitoring, Laser, Desorption, FREDIS

The Multi-Purpose-Manipulator (MPM) has been operated, as a versatile carrier system for probes, since the first campaign OP. 1.1 in 2015 at Wendelstein 7X. The MPM is mounted at the outboard midplane and is able to perform fast plunges through the entire scrape off layer up to the last closed flux surface. In addition, a system for gas injection was commissioned that allowed the use of a dedicated probe head and a port on the existing probes for fueling and impurity seeding experiments. A combined probe, a combination of Langmuir, Mach and magnetic probes, was used. For the second campaign OP. 1.2a in 2017, with an island divertor, an upgraded combined probe, a fluctuation probe, a retarding field analyzer, a Mach probe array, a laser blow off target and a material probe holder were added to the range of plasma edge diagnostics. The new diagnostic probe heads were designed to address specific tasks, such as measuring the edge fluctuations, the ion temperature and the radial Mach number profiles. However, all fast plunging probes were designed such that they provide the most important quantities for SOL characterization (electron temperature, density, radial electric field).

The Toroidal Magnetized System device has been significantly upgraded to enable development of various wall conditioning techniques, including methods based on ion and electron cyclotron (IC/EC) range of frequency plasmas, and to complement plasma–wall interaction research in tokamaks and stellarators. The toroidal magnetic field generated by 16 coils can reach its maximum of 125 mT on the toroidal axis. The EC system is operated at 2.45 GHz with up to 6 kW forward power. The IC system can couple up to 6 kW in the frequency range of 10 MHz–50 MHz. The direct current glow discharge system is based on a graphite anode with a maximum voltage of 1.5 kV and a current of 6 A. A load-lock system with a vertical manipulator allows exposure of material samples. A number of diagnostics have been installed: single- and triple-pin Langmuir probes for radial plasma profiles, a time-of-flight neutral particle analyzer capable of detecting neutrals in the energy range of 10 eV–1000 eV, and a quadrupole mass spectrometer and video systems for plasma imaging. The majority of systems and diagnostics are controlled by the Siemens SIMATIC S7 system, which also provides safety interlocks.

A bi-directional multi-channel retarding field analyzer (RFA) probe has been successfully developed for the first time on the Wendelstein 7-X (W7-X) stellarator boundary plasma. Modifications to the RFA prototype hardware and its upgrade for the two W7-X island divertor campaigns are presented, including the electronics. In this paper the experiences and challenges operation and customizing an RFA at W7-X are discussed, as well as the data analysis using a maximum coefficient of determination method to obtain the ion temperature based on the measured modulated ion current. Edge ion temperature profiles have been measured in the standard and high iota configurations. Previous article in issue

This paper reports on the design and the performance of the recently upgraded X-ray imaging spectrometer systems, X-ray imaging crystal spectrometer and high resolution X-ray imaging spectrometer, installed at the optimized stellarator Wendelstein 7-X. High resolution spectra of highly ionized, He-like Si, Ar, Ti, and Fe as well as H-like Ar have been observed. A cross comparison of ion and electron temperature profiles derived from a spectral fit and tomographic inversion of Ar and Fe spectra shows a reasonable match with both the spectrometers. The also measured impurity density profiles of Ar and Fe have peaked densities at radial positions that are in qualitative agreement with the expectations from the He-like impurity fractional abundances, given the measured temperature profiles. Repeated measurements of impurity decay times have been demonstrated with an accuracy of 1 ms via injection of non-recycling Ti, Fe, and Mo impurities using a laser blow-off system.

Knowledge of the edge ion temperature (T i) is important to understand the edge physics in fusion plasmas. The T i edge profile has been successfully measured at W7-X in an island divertor configuration using a retarding field analyzer probe on a reciprocating manipulator. The experimental observations show that in the standard configuration, an ion temperature shoulder at the plasma boundary has been observed near the location where a sudden change of the magnetic field connection length appears in the scrape-off layer (SOL) region. The value of the T i shoulder was observed to be dependent on the edge island structure modified by the control coils, SOL kinetic power, and electron density. A significant reduction and flattening in the edge ion temperature profile has been also observed during the nitrogen seeding from the upper horizontal divertor port. In addition, the edge ion-to-electron temperature ratio has been calculated, which gradually increases with the major radius and decreases with the normalized ion collisionality. An extended two-point model including upstream ion temperature is introduced to understand the experimental measurement.