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International audience; Infrared (IR) diagnostics are used to measure plasma-facing components (PFC) surface temperature in fusion devices. However, the interpretation of such images is complex in all-reflective environments because of unknown emissivity and multiple reflections issues. In order to assess these challenges an iterative inversion method based on a fast photonic model, the radiosity method, has been developed. This method is applied to two different direct models based on different geometries, Sec-Tore and RADIOS, in order to estimate temperatures from experimental-like data simulated with a Monte Carlo ray-tracing code with diffuse reflective surfaces or specularly reflective surfaces. RADIOS allows retrieving temperature on colder targets (lower than 200°C) with errors of 33% and the peak temperatures with errors of 6%.

During a regular vessel baking after a manned access a leakage of a heating/cooling pipe released about 100 l of hot water into the vacuum vessel of ASDEX Upgrade (AUG). Erosion of a ten years old Cu gasket by water during baking causes the leak. At plasma facing components the water steam forms white remnants as it reacted with the boron-hydride layers used for wall conditioning and cause locally oxidation of stainless steel. After cleaning no significant damages remained. As some remote parts of AUG are not heated water condensates at these locations and stays there for weeks. At these ports serious damages was found at certain kinds of electrical feed through and Al gaskets. Completely removal of water within 4 weeks is needed to avoid this kind of problems. To identify and fix the most dominate leak a step-by-step approach was used. After repair AUG was successful operated for 9 months till the next regular maintenance.

The electron cyclotron resonance heating system at ASDEX Upgrade (AUG) is currently being extended to eight similar Gyrotrons in total. Each Gyrotron operates at 105 and 140 GHz and is designed for up to 1 MW millimetre wave output power. A substantial part of the AUG program will focus on experimental conditions, where the plasma density may be above the X-2 cut-off density at 140 GHz. In order to cope with the high density, the heating system will operate in the O-2 mode scheme with potentially incomplete absorption in the first pass. Reflecting gratings installed into the heat shield on AUG’s inner column allow for a controlled second pass of the beam’s unabsorbed fraction. Thermocouple measurements serve to control the beam position on the grating. The beam geometry is being finalized for the launchers #1-4. Beam propagation is simulated with the TORBEAM code and previous high density experiments are used as a database. The geometry is optimized using three criteria: central deposition, high absorption and robustness of the beam dump after the second pass. The experimental conditions, and the plasma electron density in particular, may vary such that the Gaussian beam parameters of the incoming beam on the grating deviate from the design values. It is proposed to model the effect of the grating with an equivalent ellipsoidal mirror. Laboratory measurements are shown, which support this model.

We compare various diagnostics characterizing an ASDEX Upgrade upper single-null discharge to EMC3-EIRENE simulations now including volumetric recombination and main chamber plasma-wall interaction but not yet drifts. The discharge is in a small-ELM regime and is approaching detachment due to a density ramp and nitrogen seeding. Time/ELM averaged upstream density, electron and ion temperature measurements as well as downstream ion saturation current, electron temperature and power flux measurements match the simulation within the error bars before and after reaching detachment, qualifying the code for this regime. Although the Ion Cyclotron Resonance Heating (ICRH) antenna limiters are 35 mm away from the separatrix, i.e. ten times the typical near-SOL power fall-off length λnearq = 3.8 mm, they receive power fluxes of several hundreds of kW exceeding the ones to the outer target in the detached phase and strongly contribute to the particle fueling. This is explained by a substantially larger far-SOL power fall-off length λfarq = 25 mm possibly caused by the small ELMs in the detached phase and the occurrence of a density shoulder.

Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m2 and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first results from the operation in the ASDEX Upgrade experiment are presented.

The ASDEX Upgrade electron cyclotron resonance heating operates at 105 GHz and 140 GHz with flexible launching geometry and polarization. In 2016 four Gyrotrons with 10 sec pulse length and output power close to 1 MW per unit were available. The system is presently being extended to eight similar units in total. High heating power and high plasma density operation will be a part of the future ASDEX Upgrade experiment program. For the electron cyclotron resonance heating, an O-2 mode scheme is proposed, which is compatible with the expected high plasma densities. It may, however, suffer from incomplete single-pass absorption. The situation can be improved significantly by installing holographic mirrors on the inner column, which allow for a second pass of the unabsorbed fraction of the millimetre wave beam. Since the beam path in the plasma is subject to refraction, the beam position on the holographic mirror has to be controlled. Thermocouples built into the mirror surface are used for this purpose. As a protective measure, the tiles of the heat shield on the inner column were modified in order to increase the shielding against unabsorbed millimetre wave power.

A new solid tungsten divertor for the fusion experiment ASDEX Upgrade is under construction at present. A new divertor tile design has been developed to improve the thermal performance of the current divertor made of tungsten coated fine grain graphite. Compared to thin tungsten coatings, divertor tiles made of massive tungsten allow to extend the operational range and to study the plasma material interaction of tungsten in more detail. The improved design for the solid tungsten divertor was tested on different full scale prototypes with a hydrogen ion beam. The influence of a possible material degradation due to thermal cracking or recrystallization can be studied. Furthermore, intensive Finite Element Method (FEM) numerical analysis with the respective test parameters has been performed. The elastic–plastic calculation was applied to analyze thermal stress and the observed elastic and plastic deformation during the heat loading. Additionally, the knowledge gained by the tests and especially by the numerical analysis has been used to optimize the shape of the divertor tiles and the accompanying divertor support structure. This paper discusses the main results of the high heat flux tests and their numerical simulations. In addition, results from some special structural mechanic analysis by means of FEM tools are presented. Finally, first results from the numerical lifecycle analysis of the current tungsten tiles will be reported.

Recent experiments in ASDEX Upgrade aimed at improving the physics base for ITER and DEMO to prepare operation and aid the design. In order to increase its exhaust capabilities and operational flexibility, a new bulk W divertor as well as an adjustable cryopump had been installed prior to the 2014 campaign. In experiments with high-field-side pellet injection, central electron densities twice as high as the Greenwald density limit could be achieved without strongly increasing the pedestal density and deleterious effect on confinement. Due to its large installed heating power, a large normalized heat flux $P_{\mathrm{ sep}}/R = 10$ MWm $^{-1}$ has been reached, representing two-thirds of the ITER value, under partially detached conditions with a peak target heat flux well below 10 MWm $^{-2}$ . The divertor load could be further reduced by increasing the core radiation, still keeping the confinement in the range of $H_{98y2}\approx 1$ . Suppression of edge-localized modes (ELMs) at low collisionality has been observed in a narrow spectral window in contrast to earlier results at high densities. The ITER $Q = 10$ baseline scenario has been investigated, matching as close as possible the triangularity, the plasma beta, $q_{95}$ , and the distance to the L-H threshold. It turned out that the ELM frequency is low and consequently the energy ejected by a single ELM is very high and ELM mitigation appears to be difficult. As a possible alternative, a scenario has been developed achieving a similar performance at a lower plasma current (and consequently higher $q_{95})$ . Experiments using electron cyclotron current drive (ECCD) with feedback-controlled deposition have allowed successfully testing several control strategies for ITER, including automated control of (3, 2) and (2, 1) neoclassical tearing modes during a single discharge. Concerning advanced scenarios, experiments with central ctr-ECCD have been performed in order to modify the q-profile. A strong reversal of the q-profile could be stationarily achieved and an internal transport barrier could be triggered. In disruption mitigation studies with massive gas injection (MGI), a runaway electron beam could be provoked and mitigated by a second MGI. Ongoing enhancements aim at strengthening the power supplies in order to allow full use of the installed heating power, the exchange of two ion cyclotron resonance heating (ICRH) antennas to reduce the W influx during ICRH, and the upgrading of the electron cyclotron resonance heating (ECRH) system to 7–8 MW for 10 s.

A table-top pellet injector (TATOP) has been designed to fulfill the following scientific aims: to study the ELM triggering potential of impurity pellets, and to make pellet injection experiments comparable over several fusion machines. The TATOP is based on a centrifugal accelerator therefore the complete system is run in vacuum, ensuring the compatibility with fusion devices. The injector is able to launch any solid material (stable at room temperature) in form of balls with a diameter in the 0.5–1.5 mm range. The device hosts three individual pellet tanks that can contain e.g. pellets of different materials, and the user can select from those without opening the vacuum chamber. A key element of the accelerator is a two-stage stop cylinder that reduces the spatial scatter of pellets exiting the acceleration arm below 6°, enabling the efficient collection of all fired pellets. The injector has a maximum launch speed of 450 m/s. The launching of pellets can be done individually by providing TTL triggers for the injector, giving a high level of freedom for the experimenter when designing pellet trains. However, the (temporary) firing rate cannot be larger than 25 Hz. TATOP characterization was done in a test bed; however, the project is still in progress and before application at a fusion oriented experiment.

This special issue of Fusion Science and Technology is a comprehensive review of the main physical and technological results achieved with the A xial Symmetric Divertor EXperiment (ASDEX) Upgrade located at the Max-Planck-Institut fur Plasmaphysik in Garching, Germany. It includes special chapters presenting details of machine operation and control as well as results from the main topics of scientific investigations. This introductory chapter gives a survey of the aims of ASDEX Upgrade and the evolution of the scientific goals, which were accompanied by technical modifications during the more than 10 yr of operation. It presents a short overview on hardware modifications and on the scientific program. This chapter is also a summary of major scientific results, which are then described in more detail in the following chapters.

Studies in ASDEX Upgrade of the phenomenology and scaling of the H-mode transition, of edge-localized modes (ELMs), and characterization of the H-mode edge transport barrier carried out in various experimental campaigns between 1996 and 2001 are described. The H-mode transition is recognized by formation of a radial electrical field at the plasma boundary, which in ASDEX Upgrade is detected by an associated increase of the neutral particle charge exchange flux from ripple trapped particles. A scaling for the critical local edge temperature for the H-mode transition threshold is found. Similarity experiments with ASDEX Upgrade and Joint European Torus plasmas for the H-mode transition indicate that the H-mode transition can be obtained at the same values of dimensionless parameters {rho}{sup *}, {nu}{sup *}, and {beta} at the plasma edge, indicating that the threshold scaling is normally not dominated by atomic physics processes. Energy losses due to ELMs are examined. Different types of ELMs can be obtained, depending on plasma edge temperature and magnetics configuration. An interesting regime is the type II ELMy H-mode for configurations near double null, where the peak heat flux to the target is much reduced compared to large type I ELMs. High-resolution Thomson scattering measurements show thatmore » the edge transport barrier width in ASDEX Upgrade shows only weak variations, while the pedestal top electron pressure and pressure gradient strongly depend on the plasma current, or value of B{sub t}/q{sub 95}.« less

An overview of edge and divertor physics research on ASDEX Upgrade of relevance for next-step fusion devices like ITER is presented. The results described were primarily obtained in lower single-nu...

Analysis of Type I ELMs from ongoing experiments shows that ELM energy losses are correlated with the density and temperature of the pedestal plasma before the ELM crash. The Type I ELM plasma energy loss normalized to the pedestal energy is found to correlate across experiments with the collisionality of the pedestal plasma (ν*ped), decreasing with increasing ν*ped. Other parameters affect the ELM size, such as the edge magnetic shear, etc, which influence 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. In JET and DIII-D, under some conditions, ELMs can be observed (`minimum' Type I ELMs with energy losses acceptable for ITER), that do not affect the plasma temperature. The duration of the divertor ELM power pulse is correlated with the typical ion transport time from the pedestal to the divertor target (τ||Front = 2πRq95/cs,ped) and not with the duration of the ELM-associated MHD activity. Similarly, the timescale of ELM particle fluxes is also determined by τ||Front. The extrapolation of the present experimental results to ITER is summarized.

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.

Radiation distribution and heat load on the target plates of ASDEX Upgrade are investigated over a wide experimental parameter range with discharges up to 20 MW heating power and n e =(0.4–0.8)n GW . Integrating the radiation density over the whole plasma yields a total radiated power of about 80% of the input power, and radiation from the divertor and X-point region is about 40% of the input power. Whereas the radiated power in the main plasma is comparable for the former open divertor I (Div I) and the new closed LYRA divertor (Div II), the radiation fraction in the divertor has nearly doubled for the LYRA divertor. Simultaneous to the increased divertor radiation in the LYRA divertor, the maximum heat flux density to the target plates, measured by thermographic cameras, is reduced by a factor of about two compared to that in Div I for comparable attached conditions. Furthermore, especially in the inner divertor fan, the radiation contributes to a large part of the measured heat flux.

On ASDEX Upgrade, the inner leg of the divertor is diagnosed by Langmuir probes, spectroscopy, and thermography. All of these observe a narrow peak near the strike point and a second, broad peak located 10–20 cm higher. The profile form in and between ELMs is very similar, so it is not the temporal superposition of two single peaks. Numerical modeling shows each of the peaks, depending on plasma conditions, but never both peaks at the same time, so an explanation based on axisymmetric physics appears unlikely. The observation in a discharge where the strike point is swept that the upper peak remains fixed in space while the lower peak follows the strike point favors a geometric origin, even though the observations by different diagnostics are similar and the probes were designed to be outside the shadows of the neighboring divertor tiles. Alternatively, toroidally asymmetric physics could produce such a profile.

The ITER divertor power deposition database is described and analysed in this paper. The database contains experimental measurements from the major divertor experiments in L-mode and H-mode regimes. These measurements are used to derive multi-machine scaling laws for the peak divertor heat flux and width, particularly of their machine size dependence. The physical basis for these scalings is discussed and the laws obtained are used to extrapolate from existing experiments to the parameters expected in the ITER-EDA device.

Detailed experiments have been carried out in ASDEX-Upgrade to characterize detached plasma conditions in divertor II. This paper presents measurements of the hydrogen emission in the visible and VUV spectral range that distinguishes between the ionization front and the recombining divertor plasma. Plasma parameters in the recombination dominated region were obtained from the Balmer continuum and line emission. By means of the ADAS programme package the rate of volume recombination is evaluated including the effect of opacity. The optical thickness was experimentally determined from the line ratio of L β and H α . In the recombining plasma we have observed an inverse response of H α light during the spatial broadening of ELMs. Inverse ELMs show signatures of convective transport.

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.

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.

Tungsten coated tiles, manufactured by plasma spray on graphite, were mounted in the divertor of the ASDEX Upgrade tokamak and cover almost 90% of the surface facing the plasma in the strike zone. Over 500 plasma discharges, among which around 300 were heated with heating powers up to 10 MW, were performed up to now. The tungsten flux in the divertor was monitored by a WI line at 400.8 nm. In the plasma centre an array of spectral lines at 5 nm emitted by ionisation states around W XXX was monitored. Under normal discharge conditions W-concentrations of around 10 −5 or even lower were found. The influence on the main plasma parameters was negligible. In a few low power discharges accumulation of tungsten occurred and the temperature profile was flattened. The concentrations of the intrinsic impurities carbon and oxygen are comparable to the discharges with graphite divertor. Furthermore, the density-limits and the β-limits remained unchanged and no negative influence on the energy confinement as well as on the H-mode threshold was found.

The authors characterize the divertor target plate heat and particle fluxes that occur due to Edge-Localized-Modes (ELMs) during H-mode in DIII-D and ASDEX-Upgrade. During steady-state ELMing H-mode the fraction of main plasma stored energy lost with each ELM varies from 6% to 2% as input power increases above the H-mode power threshold. The ELM energy is deposited near the strikepoints on the divertor target plates in a fast time scale of {le} 1 ms. The spatial profile of the ELM heat pulse is flatter and broader, up to about a factor of 2, than that of the heat flux between ELMs. On ASDEX-Upgrade the inboard strike-point receives the greatest fraction, {ge} 75%, of ELM divertor heat flux, while on DIII-D the in/out split is nearly equal. The toroidal asymmetry of the heat pulse has produced a peaking factor on DIII-D of no more than 1.5. The particle flux, as measured by Langmuir probes, has also been found to be localized near the divertor strike-points. The increased particle flux during ELMs is a significant fraction of the total time-integrated divertor plate particle flux.

During the 1994/1995 discharge period two graphite tiles at the outer and the inner lower divertor plates of the tokamak ASDEX-Upgrade have been replaced by tiles of RG-Ti-91 which is a graphite, doped with about 2 at% of Ti and subsequently treated at high temperatures and pressures to achieve a high thermal conductivity perpendicular to one direction. The RG-Ti-91 tiles have been installed with the directions of high thermal conductivity normal to the plasma exposed surface. The plates have been exposed to about 670 tokamak discharges each lasting about 2 to 3 s. The performance of the discharges has not been influenced by the presence of the RG-Ti-91 tiles, i.e., no additional impurities, such as Ti, have been observed in the plasma. The temperature increase of the RG-Ti-91 tiles during high power loads was about a factor 2 lower than the temperature increase of fine grain graphite plates. The surfaces of the tiles showed preferential erosion of the carbon, and TiC crystallites staying out on the surfaces. This is similar to the surface structure which develops at D-ion bombarded surfaces of this RG-Ti-91 material in the laboratory.

A new solid tungsten divertor for the fusion experiment ASDEX (Axial Symmetric Divertor EXperiment) Upgrade is under construction at present. For special purposes of the plasma diagnostic in the divertor region a special formed solid tungsten divertor tile is required. A so-called Langmuir probe is used to determine the ion temperature, ion density and ion potential of the plasma. With the aim to place the probe on the right position some of the divertor tiles (9 at the device circumference) have been adequately adapted. This paper discusses the main results of the numerical analysis of the thermo-mechanical behavior under heat load. Primary, the elastic-plastic calculation was applied to analyze thermal stress and the observed elastic and plastic deformation during the heat loading. The influence of a possible material degradation due to thermal cracking was studied. Additionally, the knowledge gained by the numerical analysis was used for the shape optimization of the divertor tile. The first results from the numerical life cycle analysis of the tungsten tiles are reported. Finally, based on the knowledge gained by the numerical analysis, in the light of problem complexity, it is recommended to perform some additional thermal tests. These tests should be performed with the aim to increase the reliability of the special shaped tungsten tile during operation.

Injection of impurity pellets has been shown to be a successful method of reducing thermal and mechanical loads during disruptions. The evolution of the quenching plasma after pellet injection in ASDEX Upgrade is described and the requirements of such a method for mitigating disruptions in future devices are discussed

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.

Puffing of impurities (neon, argon) and deuterium gas in the main chamber is used to feedback control the total radiated power fraction and the divertor neutral particle density simultaneously in the ASDEX Upgrade tokamak. The variation of Psep=Pheat-Prad(core) by impurity radiation during H mode shows a similar effect on the ELM behaviour as that obtained by a change of the heating power. For radiated power fractions above 90%, the ELM amplitude becomes very small and detachment from the divertor plates occurs, whilst no degradation of the global energy confinement is observed (completely detached high confinement mode). Additional deuterium gas puffing is found to increase the radiated power per impurity ion in the plasma core owing to the combined effect of a higher particle recycling rate and a lower core penetration probability. The outer divertor chamber, which is closed for deuterium neutrals, builds up a high neutral pressure, the magnitude of which is determined by the balance of particle sources and pumping. For this particular situation, the effective pumping time of neon and argon is considerably reduced, to less than 0.3 s, mainly owing to an improved divertor retention capability. The radiation characteristics of discharges with a neon driven radiative mantle are modelled using a 1-D radial impurity transport code that has been coupled to a simple divertor model describing particle recycling and pumping. The results of simulations are in good agreement with experiment

Due to the open field lines, the scrape-off layer and divertor region of tokamak plasmas is a complex, two-dimensional system, involving transport parallel and perpendicular to the magnetic field, as well as interaction of the plasma with surfaces and with the neutral gas. Therefore sophisticated two-dimensional codes are required to model the divertor and edge physics. In this paper, the B2-EIRENE code package is used to simulate the ASDEX-Upgrade scrape-off layer plasma and the neutral gas dynamics in a fully self-consistent way. Specific ASDEX-Upgrade discharges are modelled using the actual magnetic configuration and in-vessel components. Single fluid as well as multifluid calculations including self-consistent target and wall erosion of carbon are described. At given input power and bulk plasma radiation, typical divertor plasma profiles from Langmuir probes are fitted by varying the separatrix density and the transport coefficients. On the basis of such multifluid fits, spectroscopic divertor diagnostics are numerically modelled and compared with measured profiles, and reasonable agreement is found.

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.

In the ASDEX Upgrade tokamak, complex power deposition structures on the divertor target plates during type-I edge-localized modes (ELMs) have been discovered by fast (few microseconds), two-dimensional ($40\ifmmode\times\else\texttimes\fi{}40\text{ }{\mathrm{c}\mathrm{m}}^{2}$) infrared thermography. In addition to the usual axisymmetric power deposition line near the separatrix, there appear, statistically distributed, several laterally displaced and inclined stripes, mostly well separated from each other and from the main strike zone. These structures are interpreted as footprints of approximately field aligned, helical perturbations at the low field side of the main plasma edge related to the nonlinear ELM evolution. Based on this picture, the ELM related mode structure can be derived from the target load pattern, yielding on average toroidal mode numbers in a range of 8\char21{}24.

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.

The surface of movable rail limiters (graphite and stainless steel) and limiter-like graphite probes were analysed after operation in T-10. The redeposition of limiter material was indicated by a peculiar morphology at the edge of the limiters and the elemental composition in the surface layer. Arcing was found to be strongly influenced by the actual composition of the involved regions. The arc ignition probability and the efficiency of droplet production by arcing seemed to be enhanced by a high metal concentration.

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.

Recent experiments at ASDEX Upgrade have achieved advanced scenarios with high β N (>3) and confinement enhancement over ITER98(y, 2) scaling, H H98y2 = 1.1-1.5, in steady state. These discharges have been obtained in a modified divertor configuration for ASDEX Upgrade, allowing operation at higher triangularity, and with a changed neutral beam injection (NBI) system, for a more tangential, off-axis beam deposition. The figure of merit, β N H ITER89-P , reaches up to 7.5 for several seconds in plasmas approaching stationary conditions. These advanced tokamak discharges have low magnetic shear in the centre, with q on-axis near 1, and edge safety factor, q 95 in the range 3.3-4.5. This q-profile is sustained by the bootstrap current, NBI-driven current and fishbone activity in the core. The off-axis heating leads to a strong peaking of the density profile and impurity accumulation in the core. This can be avoided by adding some central heating from ion cyclotron resonance heating or electron cyclotron resonance heating, since the temperature profiles are stiff in this advanced scenario (no internal transport barrier). Using a combination of NBI and gas fuelling line, average densities up to 80-90% of the Greenwald density are achieved, maintaining good confinement. The best integrated results in terms of confinement, stability and ability to operate at high density are obtained in highly shaped configurations, near double null, with δ = 0.43. At the highest densities, a strong reduction of the edge localized mode activity similar to type II activity is observed, providing a steady power load on the divertor, in the range of 6 MW m -2 , despite the high input power used (> 10 MW).

Though helium plasmas are one option for the low activation phase of ITER, little effort has thus far been devoted to studying them in a large, diverted tokamak. A recent campaign on JET has therefore sought to address some of the important questions related to helium operation (He concentrations near 90%) in single null configuruations, particularly with regard to edge and divertor physics. This contribution compiles a selection of results from these experiments, in which, in each case, discharges have been chosen to match as closely as possible previous, well characterised D plasmas in both L and ELMing H-modes. These matched pulses are used to draw conclusions regarding the principle source and location of carbon production in D plasmas, to compare and contrast the mechanisms of the density limit and the detachment process in D and He, to investigate the nature of cross-field power transport in the SOL and to gain insight into the process by which ELM energy is transported to the divertor targets. (C) 2003 Elsevier Science B.V. All rights reserved.

Recent experiments at ASDEX Upgrade have concentrated on advanced scenarios, specifically ELMy H-mode discharges without an internal transport barrier which obtain a high βN = 3.5 in steady state. A highly shaped plasma configuration is used at 800 kA, with a triangularity δ = 0.42 and edge safety factor, q95, in the range 3.6-4.1. The product βNH89-P reaches up to 8.0 in plasmas approaching a stationary conditions, and βNH89-P{} = 11.5 is achieved transiently. These discharges are at high density, 80-90% of the Greenwald density limit. Discharges at the highest density show a strong reduction of the ELM activity, similar to type II activity, eliminating transient high heat loads on the divertor target. Finally, the non-inductively driven current approaches 70% of the total current making this regime a serious candidate for an advanced scenario in a reactor, and a benchmark for other advanced scenarios.

The beam intensity of electron cyclotron resonance heating at ASDEX Upgrade has the potential to seriously damage in-vessel components, whenever not fully absorbed by the plasma. Operation is, therefore, interlocked with both plasma current and density above a given threshold. Microwave protection detectors installed in several ports on the low field side switch the heating system off, in case the stray radiation exceeds a given threshold. During regular inspections, however, damages were reported in the vicinity of the launchers and in particular around the tiles of the heat shield. On one hand, it was found that insulating material, which may not face the plasma, degraded due to millimetre wave absorption. The waves entered the free space behind the heat shield through gaps. On the other hand, local damage even of metallic components was observed on surfaces, which were directly exposed to the microwave beam. Polarisation errors, which led to a local shine through of significant beam power, were responsible. We note that this happened mainly on the high field side in a certain distance to the microwave protection detectors, which were not triggered by the events. In order to increase the level of protection, we identify three necessary measures: Firstly, polarisation control is to be automated such, that mode content and shine through can be monitored. Secondly, by installing additional detectors, the spatial coverage of stray radiation monitoring is enlarged. Thirdly, the heat shield tiles will be redesigned in order to increase the shielding against millimetre waves.

International audience; Infrared (IR) diagnostics are used to measure plasma-facing components (PFC) surface temperature in fusion devices. However, the interpretation of such images is complex in all-reflective environments because of unknown emissivity and multiple reflections issues. In order to assess these challenges an iterative inversion method based on a fast photonic model, the radiosity method, has been developed. The radiosity method is based on strong hypotheses including all diffuse surfaces. The inversion method allows retrieving the true surface temperature of PFC in two steps: a step of the target emissivity estimation in a baking scene and the use of the emissivity map to retrieve the temperature of metallic components with errors up to 3% during a plasma scenario.