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In the Wendelstein 7-X stellarator with its twisted magnetic geometry the investigation of plasma wall interaction processes in 3D plasma configurations is an important research subject. For the upcoming operation phase i.e. OP1.2, three different types of material probes have been installed within the plasma vessel for the erosion/deposition investigations in selected areas with largely different expected heat load levels, namely, ≤10 MW m−2 at the test divertor units (TDU), ≤500 kW m−2 at the baffles, heat shields and toroidal closures and ≤100 kW m−2 at the stainless steel wall panels. These include 18 exchangeable target elements at TDU, about 30 000 screw heads at graphite tiles and 44 wafer probes on wall panels, coated with marker layers. The layer thicknesses, surface morphologies and the impurity contents were pre-characterized by different techniques and subjected to various qualification tests. The positions of these probes were fixed based on the strike line locations on the divertor predicted by field line diffusion and EMC3/EIRENE modeling calculations for the OP1.2 plasma configurations and availability of locations on panels in direct view of the plasma. After the first half of the operation phase i.e. OP1.2a the probes will be removed to determine the erosion/deposition pattern by post-mortem analysis and replaced by a new set for the second half of the operation phase, OP1.2b.

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.

Significant progress has been made on ASDEX Upgrade during the last two years in the basic understanding of transport, in the extension of the improved H-mode in parameter space and towards an integrated operating scenario and in the development of control methods for major performance limiting instabilities. The important features were the understanding of particle transport and the control of impurity accumulation based on it, the satisfactory operation with predominantly tungsten-clad walls, the improved H-mode operation over density ranges and for temperature ratios covering (non-simultaneously) the ITER requirements on ν*, n/nGW and Te/Ti, the ELM frequency control by pellet injection and the optimization of NTM suppression by DC-ECCD through variation of the launching angle. From these experiments an integrated scenario has emerged which extrapolates to a 50% improvement in n T τ or a 30% reduction of the required current when compared with the ITER base-line assumptions, with moderately peaked electron and controllable high-Z density profiles.

Plasma sprayed B4C-layers considered as wall coatings for the W7X stellarator have been studied during and after exposure to TEXTOR and after arcing experiments in vacuum. Arcing through the B4C layer occurred favoured by high power fluxes and not restricted to less stable phases. But this arcing implies an especially noisy scrape-off layer (SOL). Instead of moving retrograde in the external magnetic field, the arc spot on the B4C-layer sticks to the same location for its whole lifetime. Consequently, the arc erodes the entire B4C layer, finally burning down to the Cu substrate. In the neighbourhood of craters the surface contains Cu originating from those craters. This material, hauled to the surface by the arc, is subject to subsequent erosion, transport, and redeposition by the SOL-plasma. The behaviour of arcs on B4C is most probably caused by the peculiar temperature dependences of the electrical and heat conductivity of B4C. 2003 Elsevier Science B.V. All rights reserved.

Analysis of Type I ELMs from present experiments shows that ELM energy losses decrease with increasing pedestal plasma collisionality ( ν ∗ ped ) and/or increasing τ Front ∥ , where ( τ ∥ Front =2π Rq 95 / c s ,ped ) is the typical ion transport time from the pedestal to the divertor target. ν ∗ ped and τ Front ∥ are not the only parameters that affect the ELMs, also the edge magnetic shear influences the plasma volume affected by the ELMs. ELM particle losses are influenced by this ELM affected volume and are weakly dependent on other pedestal plasma parameters. ‘Minimum’ Type I ELMs, with energy losses acceptable for ITER, where there is no change in the plasma temperature profile during the ELM, are observed for some conditions in JET and DIII-D. The duration of the divertor ELM power pulse is well correlated with τ Front ∥ and not with the duration of the ELM-associated MHD activity. Similarly, the time scale of ELM particle fluxes is also determined by τ Front ∥ . The extrapolation of present experimental results to ITER is summarised.

We found a complete extinction of the CD molecule emission in the inboard divertor of ASDEX Upgrade at high density with the transition to a recombining divertor plasma. The electron temperature in the recombining divertor plasma ranges from 1.2 down to 0.3 eV. Two effects may explain the extinction at such low temperatures, (1) a steep decay of the yield for chemical sputtering as theoretically predicted and (2) a suppression of the optical emission while the methane molecules dissociate unchanged via charge exchange.

In this paper we describe impurity line emission measurements in the divertor of ASDEX Upgrade during high power neutral beam heated discharges. We focus on detached conditions where the dominating part of the radiation comes from the X-point region. Spatially resolved line emission in the VUV and visible spectral region of the intrinsic carbon and additionally puffed impurities (neon and nitrogen) is presented. A simple interpretation of the line emission profiles is given and they are also compared to the results of bolometry.

In this paper we present first measurements of the plasma potential close to the separatrix in ASDEX Upgrade using a self emitting tip. The probe was inserted into the edge plasma of AUG using the midplane manipulator. Assuming Maxwellian plasmas, the observations agree with the predicted voltage drop in the plasma sheath, V pl ps − V fl = 2.5 T e , where V pl ps is the plasma potential at the presheath boundary and V fl the floating potential. Applying this technique a rapid change of the plasma potential was observed close to the separatrix during Ohmic discharges. From the gradient we derive a radial electric field E r of about −5 kV / m close to separatrix. Further out the field strength changes sign and we find up to +7 kV / m in the SOL.

The erosion and redeposition of silicon and tungsten evaporated onto the divertor plates of ASDEX Upgrade have been investigated. For comparison with the experimental results the computer program, ERO, has been used. The erosion of Si markers is smaller than expected from plasma ion sputtering, but still exceeds the W erosion by more than an order of magnitude. The observed erosion can only be explained by assuming a protective effect of carbon ions coming from the plasma which are deposited on the marker surfaces, thus reducing the marker erosion. A redeposition of the sputtered tungsten atoms has been found close to the original marker spots mainly because of prompt redeposition during the first gyration.

The stellarator Wendelstein 7-X will allow for quasicontinuous operation with the duration only being limited to two 30 min discharges per day, at a continuous heating power of 10 MW electron cyclotron resonance heating (ECRH) at 140 GHz, by the capacity of the cooling water reservoir. This will result in high thermal loads on all plasma facing components of 50-100 kW/m(2) from radiation alone and of up to about 500 kW/m(2) on components additionally exposed to convective loads. In high density scenarios toroidally varying ECRH stray radiation levels of 50-200 kW/m(2) need to be coped with, requiring careful material selection and different shielding and hardening techniques. Furthermore, a gradual buildup of coatings on plasma facing optical components, which without any measures being taken, would lead to high transmission losses already within a few days of long pulse operation (equivalent to about 1 year of operation in pulsed devices like JET or ASDEX-upgrade) and therefore needs to be prevented as much as possible. In addition in situ cleaning as well as absolute calibration techniques need to be developed for all plasma facing optical systems. Here we report about some of our efforts to find, for various types of diagnostics, ways to cope with these adverse effects. Moreover, we give a few examples for individual diagnostic specific issues with respect to quasicontinuous operation, such as the development of a special integrator for the magnetic diagnostics as well as special interferometer types which can cope with unavoidable vibrations and slow path length changes due to, e.g., thermal expansion of the plasma vessel.

ASDEX Upgrade has recently finished its transition towards an all-W divertor tokamak, by the exchange of the last remaining graphite tiles to W-coated ones. The plasma start-up was performed without prior boronization. It was found that the large He content in the plasma, resulting from DC glow discharges for conditioning, leads to a confinement reduction. After the change to D glow for inter-shot conditioning, the He content quickly dropped and, in parallel, the usual H-Mode confinement with H factors close to one was achieved. After the initial conditioning phase, oxygen concentrations similar to that in previous campaigns with boronizations could be achieved. Despite the removal of all macroscopic carbon sources, no strong change in C influxes and C content could be observed so far. The W concentrations are similar to the ones measured previously in discharges with old boronization and only partial coverage of the surfaces with W. Concomitantly it is found that although the W erosion flux in the divertor is larger than the W sources in the main chamber in most of the scenarios, it plays only a minor role for the W content in the main plasma. For large antenna distances and strong gas puffing, ICRH power coupling could be optimized to reduce the W influxes. This allowed a similar increase of stored energy as yielded with comparable beam power. However, a strong increase of radiated power and a loss of H-Mode was observed for conditions with high temperature edge plasma close to the antennas. The use of ECRH allowed keeping the central peaking of the W concentration low and even phases of improved H-modes have already been achieved.

Dedicated experiments have been carried out in vacuum to study breakdown and arc burning on graphite cathodes covered by a thick (lμm) tungsten layer. The motivation originates from the usage of similar layers as first wall material in fusion devices where a large number of arcs appeared during plasma operation. Besides the standard measurements of current and voltage optical observations of the developing cathode spot and the retrograde motion in a transverse magnetic field on a short time scale have been carried out to determine the arc velocity reliably. Special attention was payed to non-reversible modifications of the layer by arcing. Electron microscopy (SEM) has been applied to specify the geometric structure of the tracks together with the distribution of elements. Main result is the observation that a lμm tungsten layer is permanently perforated by passing of an arc exposing carbon from the underlying graphite to all subsequent erosion processes. The effect depends nonlinearly on the arc current and is used to estimate qualitatively arc currents from remnant tracks found on plasma lacing components. The erosion rate is about one order of magnitude higher than for arcing on bulk W.

Feedback-controlled puffing of neon and deuterium has been applied to control the edge-localized-mode behavior and the target plate power deposition during high-power H -mode discharges in ASDEX Upgrade. A regime has been found in which more than 90% of the heating power is lost through radiation and divertor detachment occurs, without deterioration of the energy confinement. The plasma remains in the H mode, exhibiting small-amplitude, high-frequency ELM's, which do not penetrate to the target plates in the strike zone region.

In the aperture limiter shadow of the T-10 tokamak, the temporal evolution of plasma density and electron temperature during single discharges is studied by using a Langmuir double probe. It is observed that the plasma of the scrape-off layer (SOL) passes through a stationary phase when the Ohmic heat input Uloop Ip also does, suggesting that the stationarity of the SOL is correlated with the stationarity of the discharge power balance. During this stationary phase the plasma density at the aperture limiter edge was found to be approximately proportional to the square of the line averaged core plasma density. – It is shown in the paper that this quadratic density scaling can be easily understood within the framework of a stationary particle balance model of the recycling process, i.e. the interaction of plasma and neutral particles of the working gas in the edge region. Quadratic density scaling is expected to hold generally if the plasma density is high enough to prevent the neutrals from reaching the centre. For high densities, on the other hand, there is a density limit defined by the requirement that the discharge should supply enough power to sustain recycling.

Wall erosion by unipolar arcs has been found i n the T-10 tokamak of the Kurchatov Institute, Moscow, by the use of a movable probe. Arc current and arc duration are deduced from photographs of the arc tracks, yielding values about 100 A, 50 ?s, respectively. Some correlations between arcing and the main discharge are discussed.

After more than five years of operation of the T-10 tokamak a visual inspection of the internal liner surface was made and many arc tracks were found and photographed. Both the distribution and the direction of the tracks suggest the existence of an m = 1, n = 1 plasma mode lasting for at least 1 ms. A simple model is proposed to match the observed structures and to yield information about the plasma column.

Impurity confinement behaviour has been studied by using a deposition probe in conjunction with pellet injection. Generally, an exponential decay of the impurity efflux and nearly symmetric ion/electron side toroidal flows have been observed. During phases of strong plasma disturbances, asymmetric flow is seen, indicative of edge transport and prompt recycling from local sources. The application of ECRH may cause such disturbances.

A vacuum transfer system has been developed that allows samples such as deposition probes, electrical probes, etc., to be inserted from a lock-section into the near wall plasma of a tokamak. AES and SIMS diagnostics are installed in the analysis chamber 13 m away from the tokamak behind the biological screen for in situ investigations of the changes in surface composition of deposition probes to determine impurity fluxes in the tokamak edge plasma. The system is constructed of uhv components (stainless steel, copper sealed CF-flanges). Viton sealed valves isolate the various sections of the 18 m long, 150 mm ID transfer line. The sample size is limited by the smallest constriction within the components (60 mm in diameter for the deposition probe). The remote controlled sample transfer takes about 1 min.

Densities achievable in ASDEX Upgrade discharges are restricted by a disruptive limit in the L-mode caused by an edge-power imbalance which is linking divertor detachment, Marfe formation and the separatrix density. The attainable average densities depend then on the internal particle sources and the core transport and can exceed the empirical Greenwald density. In H-mode an upper density limit is found which represents a non-disruptive H - L back transition, which is preceded by the occurrence of type-III ELMs. Close to the Greenwald limit this H - L transition cannot be avoided at any power flux across the separatrix and - at high external neutral gas fluxes - confinement compared with ITER H-92P scaling degrades even before the back transition. The H-mode operational window is determined by local edge-barrier parameters and their gradients, respectively. The boundaries are represented by the L - H transition-temperature threshold, the ideal ballooning edge-pressure gradient limit, the upper temperature limit for type-III ELMs and an upper H-mode barrier density limitation. The cause for the last limitation is not yet identified; it may be due to resistive ballooning modes or the separatrix density limit. Despite the limited edge densities the Greenwald density could be surpassed by a factor of three with pellet refuelling from the low magnetic-field side. Pellet injection from the high-field side gains from the strong increase of fuelling efficiency due to the assisting toroidal outward drift of the formed high- ablatant. Higher densities are achievable in H-mode compared with low-field side injection and diminished convective losses avoid confinement degradation up to the Greenwald density. In gas-puffed type-I ELMy H-modes the plasma thermal energy and the edge-pressure gradients, which are limited by ballooning stability, are linked via a robust temperature-profile stiffness and the flat density profiles resulting from dominant edge refuelling at high densities. Their confinement does not improve with increasing density (and neutral gas fluxes) and may even slightly degrade. Therefore, the superior confinement of type-I ELMy H-modes compared with type-III ELMy ones at medium densities is actually offset at densities close to the Greenwald density. In contrast to the temperature-profile resilience density profiles can be changed both by deep refuelling (with pellets) and intrinsic transport improvements connected with density peaking (observed in CDH-modes), which offers the combination of high confinement and high density operation. The possible alliance with radiation cooling, divertor detachment and divertor compatible type-III ELMs could solve the power exhaust problem.