The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the re-availability of all flywheel-generators allowed high-power operation with up to 20 MW heating power at I p up to 1.2 MA. Implementation of alternative ECRH schemes (140 GHz O2- and X3-mode) facilitated central heating above n e = 1.2 × 1020 m−3 and low q 95 operation at B t = 1.8 T. Central O2-mode heating was successfully used in high P/R discharges with 20 MW total heating power and divertor load control with nitrogen seeding. Improved energy confinement is obtained with nitrogen seeding both for type-I and type-III ELMy conditions. The main contributor is increased plasma temperature, no significant changes in the density profile have been observed. This behaviour may be explained by higher pedestal temperatures caused by ion dilution in combination with a pressure limited pedestal and hollow nitrogen profiles. Core particle transport simulations with gyrokinetic calculations have been benchmarked by dedicated discharges using variations of the ECRH deposition location. The reaction of normalized electron density gradients to variations of temperature gradients and the T e/T i ratio could be well reproduced. Doppler reflectometry studies at the L–H transition allowed the disentanglement of the interplay between the oscillatory geodesic acoustic modes, turbulent fluctuations and the mean equilibrium E × B flow in the edge negative E r well region just inside the separatrix. Improved pedestal diagnostics revealed also a refined picture of the pedestal transport in the fully developed H-mode type-I ELM cycle. Impurity ion transport turned out to be neoclassical in between ELMs. Electron and energy transport remain anomalous, but exhibit different recovery time scales after an ELM. After recovery of the pre-ELM profiles, strong fluctuations develop in the gradients of n e and T e. The occurrence of the next ELM cannot be explained by the local current diffusion time scale, since this turns out to be too short. Fast ion losses induced by shear Alfvén eigenmodes have been investigated by time-resolved energy and pitch angle measurements. This allowed the separation of the convective and diffusive loss mechanisms.

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.

Recent experiments at ASDEX Upgrade have achieved advanced scenarios with high βN (>3) and confinement enhancement over ITER98(y, 2) scaling, HH98y2 = 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, βNHITER89-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, q95 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).

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 K-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 Bur 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-beta ablatant. Higher densities are achievable in H-mode compared with low-held 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 II-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.

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 600 plasma discharges have been performed to date, around 300 of which were auxiliary heated with heating powers up to 10 MW. The production of tungsten in the divertor was monitored by a W I line at 400.8 nm. In the plasma centre an array of spectral lines at 5 nm emitted by ionization states around W XXX was measured. From the intensity of these lines the W content was derived. Under normal discharge conditions W-concentrations around or even lower were found. The influence on the main plasma parameters was found to be negligible. The maximum concentrations observed decrease with increasing heating power. In several low power discharges accumulation of tungsten occurred and the temperature profile was flattened. The concentrations of the intrinsic impurities carbon and oxygen were comparable to the discharges with the graphite divertor. Furthermore, the density and the limits remained unchanged and no negative influence on the energy confinement or on the H-mode threshold was found. Discharges with neon radiative cooling showed the same behaviour as in the graphite divertor case.

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 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.

The results of divertor studies on ASDEX Upgrade, at currents of up to 1.2 MA and heating powers up to 10 MW are described, with emphasis on the ELMy H-mode. The spatial and temporal characteristics of their heat load, and the simulation of ELMs by a time-dependent scrape-off layer code are described. High gas puff rates were found to lead to a large increase in divertor neutral pressure, at modest changes in ne, and to a strong reduction in time-averaged power flow and complete detachment from both target plates in between ELMs. Using pre-programmed puffs of neon and argon, the radiative power losses could be raised to 75% of the heating power, in H-regime discharges, and the regime of enhanced divertor neutral pressure was found also to lead to an improved pumping of recycling impurities.

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.

The objective of ASDEX Upgrade is to investigage open poloidal divertor configurations as produced by a reactor-compatible, poloidal field coil system. The accessible poloidal and toroidal fields and hence the plasma current provide a sufficiently large plasma parameter regime to ensure the functioning of the open divertor.