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Measurements and simulations with the UEDGE code of radiated power, and ion saturation currents and power loads to the target plates have been compared for density scans in ohmic and low confinement mode plasmas in DIII-D, ASDEX Upgrade, and JET. Simulations including cross-field drifts and assuming elevated chemical sputtering yields of 3%–4% move the numerical solutions closer to many of the measurements compared to omitting the drifts and using the published Davis–Haasz yields. Adopting these assumptions the simulations reproduce the measured currents and powers, and their functional dependence on upstream density to within a factor of 2, with the exception of the ion currents to the low field side target in ASDEX Upgrade and the high field side target in JET. The applicability of using enhanced sputtering yields is discussed by comparing measured and simulated emission from low charge state carbon in the divertor regions.

This contribution examines the impact on plasma radiation and on power load to the divertor (i) of large type I ELMs with energies 0.25–1.3 MJ in high current H-mode JET discharges and (ii) of disruptions. (i) The ELMs provoke strong radiation losses, mostly confined to the inner divertor region. Large type I ELMs with ΔWELM ⩾ 0.72 MJ show enhanced radiation losses which are associated with the ablation of carbon layers in the inner divertor. Such large ELMs are usually followed by a phase of type III ELMs with an increased radiation in the plasma core. (ii) The unmitigated disruptions exhibit small radiation fractions with strong poloidal radiation asymmetry during the thermal quench, which could cause Be melting by radiation in ITER. In dedicated experiments on massive gas injection, about 60% of the thermal energy and a significant part of the magnetic energy was converted in radiation and spread uniformly over the first walls.

Multi-fluid plasma edge code packages, usually coupled to Monte-Carlo transport codes for neutrals are in wide use in the fusion community to simulate the SOL and divertor. A status of the current code validation effort is presented using experimental data of Ohmic and L-mode discharges from ASDEX Upgrade for He in forward toroidal field, BT, only and for D fueled in forward and reversed BT in lower single null configurations. While satisfactory agreement between simulated and experimental data can be obtained for both divertors as long as one of the two divertors is not strongly recombining, the qualitative behavior as a function of line averaged density cannot be recovered in the simulations for higher recycling regimes. This holds independently of field direction or fueling species. Semi-empirical models and thus far not included physics models are discussed.

A major challenge in nuclear fusion research is the coherent combination of measurements from heterogeneous diagnostics. Different measurement techniques for measuring the same subset of physical p...

Detailed modelling of the plasma edge of ASDEX Upgrade in Ohmic and H-mode discharges was performed with the SOLPS (B2-EIRENE) code. A tendency of the code solutions to underestimate the divertor T e and overestimate the plasma density (thus, predicting cooler and denser divertor than in the experiment) has been reliably established. A possible cause of the observed discrepancies is the presence of a significant population of supra-thermal ions and electrons in the SOL and divertor, which cannot be modelled by the fluid plasma code B2.

In ASDEX Upgrade, a fast reciprocating probe positioned in the lower divertor, capable of accessing the low-field side (LFS) and high-field side (HFS) scrape-off layer (SOL) just below the x-point, as well as the private flux region, was equipped with a Mach head and used to investigate fast flow fluctuations and in–out divertor flow asymmetries during ELMs. We compare the flow behaviour during ELMs in the three separate regions. Flow enhancement is observed in the HFS SOL, with Mach number values reaching or exceeding M = 2, flow reversal in the LFS SOL, and complex fluctuating behaviour in the private flux region (which includes flow reversal). We discuss the possible mechanisms that could drive these observations.

A low density H-mode plasma has been selected for detailed inter-ELM modelling by the SOLPS code package, with the coupled treatment of its plasma (fluid code B2) and neutral (Monte-Carlo code Eirene) parts. Good quality measured midplane density and temperature profiles, covering the pedestal region and stretching far into the SOL, as well as several other parameters and profiles measured in the divertor, have enabled testing the consistency of code solutions with experiment. Once the upstream, midplane profiles have been fitted, and the global parameters (e.g. input power into the computational grid, radiated power) matched, the code reproduces experimental profiles and control parameters in the divertor with an accuracy within a factor of 2. Deviations of modelled parameters from the experiment were found around the strike point position where most of the power was deposited on the target. The deviations are consistent among themselves and all point to one common problem with the modelling: the predicted divertor electron temperature is too low and the density too high, compared with the experiment. The largest inconsistency between the code and experiment was in the magnitude of the peak Hα radiation in the outer divertor, which was larger by a factor of 2 in the code simulations. In addition, the code predicts a somewhat higher sub-divertor neutral flux but lower carbon impurity content in the edge plasma than in the experiment, as well as lower CIII emission. The discrepancy between Hα profiles can to a large degree be attributed to profile effects: the simulated Hα emission profiles are narrower than in the experiment, reflecting the tendency of the neutral–plasma mix to congregate excessively around the strike point in the modelling. At the same time, the integrated Hα emission matches very well with the experiment.Extensive sensitivity studies of the influence of variations in input parameters and assumptions of the code on the modelled divertor conditions have been conducted. They have not resulted in an identification of any SOLPS input/control parameters capable of removing the main disagreement between the code output and experiment. A possibility of parallel transport effects related to low collisionality to increase the effective plasma temperature near the strike point position or of increased perpendicular transport by neutrals (due to some missing reactions in Eirene) to widen the target profiles, will be explored in the future.

Heat deposition onto plasma facing components outside the divertor by ELMs and fast particles has been investigated in ASDEX Upgrade. Heat fluxes up to tens of megawatts per square meter are found during ELMs on limiters and a few megawatts per square meter in between. The total deposited power on non-divertor structures can be up to 10% of the heating power and has to be considered for high power next step experiments. The mechanisms behind the observed heat load patterns are qualitatively understood. The main contributions come from fast ions created by neutral beam injection (and ion cyclotron minority) heating, and from ELMy energy bursts from the plasma edge.

An integrated exhaust scenario has been developed in ASDEX Upgrade providing simultaneous control of particle exhaust, time averaged divertor power flux and ELM power via pace-making of the ELM frequency. Feedback quantities are the divertor neutral flux (actuator: deuterium gas puff) and the divertor temperature as measured by thermoelectric currents (actuator: argon gas puff). The ELM frequency is stabilised to a level high enough for moderate ELM power load with repetitive injection of small pellets from the high field side. The scenario is fully compatible with the present tungsten wall coating, which constitutes about 65% of the area of the plasma facing components. It is intended for application in a future full-tungsten ASDEX Upgrade with strongly reduced intrinsic carbon radiation in order to simulate a possible full-tungsten stage of ITER.

At certain values of the edge rotational transform ιa, the confinement quality of plasmas in the Wendelstein 7-AS (W7-AS) stellarator is found to react very sensitively to small modifications of the edge rotational transform ιa. As ιa can be reproducibly changed either by external fields or by a small plasma current, these transitions offer a precise way to systematically analyse differences in plasma turbulence between bad and good confinement cases. This paper presents results of the study of electron density fluctuations associated with confinement changes. Wavenumber and frequency spectra and radial profiles are compared. A slow and reproducible transition is induced by a small plasma current and the sequence of events leading to bad confinement is investigated. The laser scattering core plasma density fluctuation measurements are complemented by edge beam emission spectroscopy results and magnetic fluctuation measurements with Mirnov coils. Clear correspondence between plasma fluctuations and confinement degradation is observed: the weight of larger structures increases, fluctuations increase in the plasma core, the poloidal flow velocity decreases and regions of mode-like activity move radially in the plasma. These changes occur gradually and controllably if the magnetic configuration is steered by a small plasma current.

A new second harmonic heterodyne ECE system, which is suitable for the future W7-X stellarator, has been developed and tested at W7-AS. It completes the older W7-AS system. The design was aimed at using less components and improving the system noise compared to the older system. A technical description and first measurements are presented.

Plasma discharges at W7-AS in the ‘electron root’ regime show highly peaked electron temperature- and pressure profiles. A switch-off of the ECRH leads to a fast temperature drop with time constants of less than 1 ms and a corresponding change of the Shafranov shift in the center of the plasma. For a temporally resolved comparison between calculated and measured temperature profiles, a fast method of gaining the mapping of the magnetic field to the radius along the ECE sightline is presented. Using these mappings to compare measured ECE signals with modeled temperatures (assuming plausible profiles of the radial electrical field), the time constant for the decay of the electric field after a ECRH power switch-off was found to be 0.6 ms.

Density-limit discharges in the W7-AS stellarator, with constant line-integrated density and a duration of up to 2 s, were studied at three values of the toroidal magnetic field (B = 0.8, 1.25 and 2.5 T). The central factor governing the physics of the density limit in stellarators was demonstrated to be the decreasing net power to the plasma when the centrally peaked radiated power density profile exceeds that of the deposited power density. The process was further accelerated by the peaking of electron density under these conditions. In discharges with B = 2.5 T, simulations of the centrally peaked radiation power density profiles could be shown to be due to peaked impurity density profiles. Laser blow off measurements clearly inferred an inward pinch of the injected aluminium. These discharges had the electron density profile form found in the improved confinement H-NBI mode on W7-AS. The aim of producing steady-state discharges at the highest possible density in stellarators is naturally of special interest for reactor operation. Such a scenario has been best achieved in H-mode discharges, in which ELMs restricted the impurity influx to the plasma and an equilibrium in the plasma parameters with suitably low radiation power levels was possible. A density scan in ECRH discharges highlights the need to control impurity sources and choose electron densities well below the density limit in order that steady-state operation can be attempted in discharges without ELMs. A simple model of bulk radiation predicted that the limiting density should depend on the square root of heating power and this was experimentally confirmed. The magnetic field scaling of the limiting density found experimentally in this simple model will partly depend on the term concerning the radial profile of the impurity density, which in turn is a function of the diffusion coefficient and inward pinch of the impurity ions. Theoretical studies have shown that an assumption about the B dependence of the thermal conductivity leads to density limit scaling laws with an explicit B dependence.

H-mode operation in the low-shear stellarator W7-AS is achieved for specific plasma edge topologies characterized by three `operational windows' of the edge rotational transform. An explanation for this strong influence of the magnetic configuration could be the increase of viscous damping if rational surfaces and thus island structures occur within the relevant plasma edge layer, thereby impeding the development of an edge transport barrier. Prior to the final transition to a quiescent state, the plasma edge passes a rich phenomenology of dynamic behaviour such as dithering and ELMs. Plasma edge parameters indicate that a quiescent H-mode occurs if a certain edge pressure is achieved.

The behaviour of carbon in the divertor of ASDEX-Upgrade is investigated by means of optical spectroscopy, bolometry and thermography. Apparent chemical erosion yields, derived from CH molecular band emission and H β spectroscopy are shown to exhibit a pronounced flux dependence and isotope effect, Y chem m H x Γ -0.7 , while no dependence on surface temperature (T surf < 100°C) and impact energy is seen. The flux and isotope dependences also show up in the CII, CIII and total carbon radiation. The weak correlation between the carbon yields in the divertor and the core plasma carbon concentration supports the assumption that main chamber sources are important for the core plasma carbon content under the low temperature, high recycling conditions of this study.

A new divertor was installed in ASDEX Upgrade and went into operation in spring of 1997. The divertor was designed to handle heat fluxes relevant to ITER-like scenarios. For this, the tiles expected to receive the maximum load (strike point modules) are hardened by the use of carbon fibre composites covering about 20% of the total divertor area. The maximum heat flux detected by thermography in a H-mode discharge is only 4 MW/m 2 or below for a total input power of 20 MW without radiating mantle, except some ELMs showing a moderately higher heat flux. A broad distribution of the power load found in the measured poloidal distribution of the heat flux as well as the energy distribution, shows that less than half of the power flowing into the divertor is received by the strike point modules. The remaining power is loaded to other tiles of the divertor, particularly the transition module and parts of the roof baffle near to the strike point modules. The reconstructed radiation pattern reveals that the fraction of power radiated outside the divertor is comparable for both geometries. But the radiated fraction inside the divertor is increased by 10-15% in the Lyra shaped divertor.

Radiation losses in divertor I of ASDEX Upgrade have been studied at high density when the divertor is detached. Spectroscopic measurements in the visible and VUV spectral range allow a characterization of both the impurity source and radiation. Despite very low target temperatures of 1 eV, carbon is released from the target with high rates by chemical sputtering. Power losses are estimated from the resonance lines of deuterium and the carbon ions. A comparison with bolometry shows that 80% of the emission can be ascribed to carbon and deuterium. The composition of the carbon emission is compared with results obtained on JT-60 and DIII-D. Using the Braams-Eirene model of the detached divertor the influence of transport on the power losses is investigated. By means of a realistic radiative loss function we find that the enhancement of maximum power input into the divertor by non-coronal effects is not more than a factor of two.

A simple analytic model of divertor radiation is used as a framework for identifying different routes to divertor detachment. Three means of achieving detachment are demonstrated experimentally including, increasing plasma density, purposely increasing the impurity level and increasing the scrape-off layer connection length. The resulting low electron temperatures in the divertor causes friction processes to dominate ionization in the recycling region, giving rise to plasma pressure loss along field lines. The pressure loss is in approximate agreement with simple analytic modelling.

Power exhaust is one of the critical questions to be solved for ITER, especially with the side condition of sufficient particle exhaust. Since the H-mode threshold might require a high power flux over the separatrix, only small power losses inside the bulk plasma might be permitted, and thus eventually a large fraction of the power must be removed in the scrape-off layer and divertor plasma. Neutral energy losses can only remove a small fraction of the power, and therefore most of the power flux has to be radiated, either by hydrogen isotopes, by the intrinsic impurities (mainly carbon) or by additionally seeded impurities. However, all these scenarios with reduced power flow to the divertor imply also a reduced particle flow into the divertor. This might be critical for particle exhaust, especially concerning the helium ash. Reduction of the power flow to the divertor by enhanced radiation losses has been investigated in the large divertox experiments like JET, DIII-D, ASDEX Upgrade, and JT-60U, and an experimental overview of the present status will be given in this paper, together with a discussion of impurity exhaust (mainly He and Ne) in these scenarios.

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.

Phase-space time-resolved measurements of fast-ion losses induced by edge localized modes (ELMs) and ELM mitigation coils have been obtained in the ASDEX Upgrade tokamak by means of multiple fast-ion loss detectors (FILDs). Filament-like bursts of fast-ion losses are measured during ELMs by several FILDs at different toroidal and poloidal positions. Externally applied magnetic perturbations (MPs) have little effect on plasma profiles, including fast-ions, in high collisionality plasmas with mitigated ELMs. A strong impact on plasma density, rotation and fast-ions is observed, however, in low density/collisionality and q(95) plasmas with externally applied MPs. During the mitigation/suppression of type-I ELMs by externally applied MPs, the large fast-ion bursts observed during ELMs are replaced by a steady loss of fast-ions with a broad-band frequency and an amplitude of up to an order of magnitude higher than the neutral beam injection (NBI) prompt loss signal without MPs. Multiple FILD measurements at different positions, indicate that the fast-ion losses due to static 3D fields are localized on certain parts of the first wall rather than being toroidally/poloidally homogeneously distributed. Measured fast-ion losses show a broad energy and pitch-angle range and are typically on banana orbits that explore the entire pedestal/scrape-off-layer (SOL). Infra-red measurements are used to estimate the heat load associated with the MP-induced fast-ion losses. The heat load on the FILD detector head and surrounding wall can be up to six times higher with MPs than without 3D fields. When 3D fields are applied and density pump-out is observed, an enhancement of the fast-ion content in the plasma is typically measured by fast-ion D-alpha (FIDA) spectroscopy. The lower density during the MP phase also leads to a deeper beam deposition with an inward radial displacement of approximate to 2 cm in the maximum of the beam emission. Orbit simulations are used to test different models for 3D field equilibrium reconstruction including vacuum representation, the free boundary NEMEC code and the two-fluid M3D-C1 code which account for the plasma response. Guiding center simulations predict the maximum level of losses, approximate to 2.6%, with NEMEC 3D equilibrium. Full orbit simulations overestimate the level of losses in 3D vacuum fields with approximate to 15% of lost NBI ions.

The completely detached high confinement (CDH) regime established recently in high power ASDEX-Upgrade single null divertor discharges is described. The standard CDH mode scenario with feedback controlled external deuterium and neon puffing is analysed in detail with respect to its divertor detachment and confinement properties. Impurity transport and pumping effects and the density operation windows of the CDH mode are outlined. First results with other impurity radiators (N,Ar) are given. Finally the CDH mode operation space is discussed in terms of main chamber and scrape-off layer plus divertor radiation and possible consequences for future fusion experiments are outlined.

Detailed experimental divertor data are presented on the profiles of density and temperature in the inner and outer divertor fans, the radiated power distribution, the gas pressure and the spectroscopically derived particle fluxes, all as a function of the discharge density. At low and medium density, the inner divertor is cold and dense compared to the outer divertor. At high density, strong X-point MARFE and separatrix radiation partially detaches the inner divertor. Probe measurements which penetrate into the X-point MARFE at the outer divertor are presented.

The H-mode obtained in ASDEX Upgrade by three heating methods (Ohmic, NBI, ICRF) is analysed. The power threshold is shown to be relatively low compared with other tokamaks and to be independent of the heating method. The operational window for the different types of H-phases, (dithering, ELMing with type-III and I ELMs, ELM-free) is given and discussed. The features of the H-mode discharges, in particular quasi-stationary ELMing phases with type I ELMs, are described and discussed. Finally, it is shown that the power flux through the plasma edge is a key parameter for H-mode operation and control.

The dynamic behaviour of the L-H transition in ASDEX Upgrade is described. We focus on two dynamic phenomena: Dithering Cycles and ELMs. The Dithering Cycles are characterized experimentally; then, a theoretical model is presented which explains the cycle as a limit cycle oscillation at the transition. With regard to ELMs, the modulation of heat and particle fluxes into the divertor due to ELMs is studied, quantitative results from thermography of the target plates are given.

ASDEX Upgrade is a poloidal divertor experiment very similar to ITER with respect to magnetic field properties and especially to the plasma boundary geometry. A large part of the programme on ASDEX Upgrade is therefore dedicated to investigating and optimizing reactor-relevant plasma boundary issues. This contribution is concerned with H-mode studies and edge physics in general. The L/H-power threshold, H-mode confinement, details of ELM dynamic and 'dithering' L/H-transitions will be discussed. The edge physics investigations are concerned with model calculations and validations, the characteristic 'approach' to the density limit (DL), target plate sputtering and the influence of hydrocarbons, asymmetries and drifts depending on the direction of the toroidal magnetic field, and finally first results of high-Z material experiments.

Deuterium fueling profiles across the separatrix have been calculated with the edge fluid codes UEDGE, SOLPS and EDGE2D/EIRENE for lower single null, ohmic and low-confinement plasmas in DIII-D, ASDEX Upgrade and JET. The fueling profiles generally peak near the divertor x-point, and broader profiles are predicted for the open divertor geometry and horizontal targets in DIII-D than for the more closed geometries and vertical targets in AUG and JET. Significant fueling from the low-field side midplane may also occur when assuming strong radial ion transport in the far scrape-off layer. The dependence of the fueling profiles on upstream density is investigated for all three devices, and between the different codes for a single device. The validity of the predictions is assessed for the DIII-D configuration by comparing the measured ion current to the main chamber walls at the low-field side and divertor targets, and deuterium emission profiles across the divertor legs, and the high-field and low-field side midplane regions to those calculated by UEDGE and SOLPS.

The extension of the computational region of the coupled fluid plasma, Monte-Carlo neutrals code, B2-Eirene, to the plasma center is discussed. The simulation of completely detached H-mode plasma is presented, as is the modelling of He and Ne compression.