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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.

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.

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.

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 first plasma in ASDEX Upgrade (AUG) was produced on 21stMarch '91 after more than 8 years of design, manufacture, and assembly. During the first operational period until August '92, with 1949 shots, the control and safety systems were commissioned in parallel with plasma discharges. Most discharges were run with He, but also H and D discharges were executed up to a plasma current of 800 kA at Bo = 2 T. With additional heating by ICRH, of up to 2.2 MW power and up to 2 s duration, the H-mode was obtained at a threshold of 1.2 MW after boronization. Discharges with line-densities of up to ne = 4*1019m−3and central electron temperatures of up to Teo =1.5 keV could adequately be diagnosed (ne-, Te-profiles, radiated power, ELM and MHD-activities, impurity influx …). With a CCD-camera the plasma boundary and critical in vessel components were observed. The magnetic diagnostics provided input for elaborate feedback control and reconstruction of the flux surfaces. The main effort of the experimental program was put on establishing the SN-divertor configuration, investigating the vertical n = 0 instability (VDE) and ICR heating scenarios. This paper describes the control systems, discharge essentials, vessel conditioning and the specific VDE behaviour.

ASDEX Upgrade, a medium-size tokamak, is characterized by the following features: a poloidal field system with all coils located outside the TF magnet; elongated plasmas up to s = 1.6 in double-null (DN), single-null (SN) or limiter-controlled (L) configurations; a plasma current of up to I p ≈ 2 MA, which ensures reactor-like plasma boundary parameters; auxiliary heating (probably ICRH, 12 MW, 8 antennae) producing a reactor-equal energy flux density of 0.3 MW/m2 in the plasma boundary. The vacuum vessel installation is primarily adapted to SN divertor operation using two toroidal divertor plate arrays ≈ 3 m2 in area each. It allows DN divertor operation with reduced parameters. Proper adjustment of target plates, antennae and heat shields also allows full-parameter DN and L configurations to be studied. The aims of ASDEX Upgrade are to study the reactor-compatible open divertor configuration in the high-recycling, low-boundary-temperature regime. Boundary layer models promise adequate solutions for taking the plasma energy to the material structure in such a way that impurity production is kept sufficiently low and that proper steady-state plasma conditions can prevail. The condition for helium pumping can also be studied. The detailed design of the tokamak system is about to be completed; the first components have been ordered and construction work is scheduled for completion by the end of 1987.

The magnetic field of the poloidal coils creates plasma equilibrium conditions, plasma breakdown and heating and may be formed by two separate coil systems (OH; V). All coils without the central OH coil (OH 1) are normally stressed; less than 100 MPa. The extremely stressed OH central coil (350 MPa)must be specially designed for keeping stress and temperature rise low. The start temperature of OH 1 has to be less than 80 K for keeping the temperature rise i=9.1 \frac{KA}{cm^{2}} ; t pulse = 4.5 s. All coils are designed for direct cooling. Cool medium can be Li N2 or supercritical He at high pressure.

ASDEX Upgrade is a tokamak experiment with external poloidal field coils that is now under construction at IPP Garching. It can produce elongated single-null (SN), double-null (DN), and limiter (L) configurations. The SN is the reference configuration with asymmetric load distributions in the poloidal field (PF) system and the toroidal field (TF) magnet. Plasma control and stabilization require a rigid passive conductor close to the plasma. The design principles of the coils and support structure are described.

ASDEX Upgrade (UG), a divertor tokamak with a minor plasma radius of a=0.5 m and a plasma current of 1.2 to 1.5 MA, is intended to succeed ASDEX. A major target of this experiment is to investigate a reactor-compatible plasma boundary. This requires according to /1/ a toroidal field (B O ) at the plasma centre normalized to the aspect ratio (A) of B o /A ≈ 1.2. The optimum with due allowance for physical requirements and technical constraints of the complete tokamak system was obtained in this case for B o =3.9 T (A=3.25) and a plasma radius of 1.63m. The toroidal field (TF) magnet designed to meet these requirements is presented here. Aspects of its turnover structure such as force transfer, coil housing and access to the plasma vessel are described. The coil concept developed in collaboration with industry is also presented, in particular the conductor concept and the current terminals and water manifold designs. Finally, the results of the cooling and stress analysis are summarized.

The ASDEX UG tokamak poloidal field coil system will comprise three ohmic heating coils and three vertical field coils. The ohmic heating coils consist of a highly excited coil in the central bore of the toroidal field magnet and coils outside the toroidal magnet arranged in pairs symmetric to the torus mid-plane. This coil arrangement is expected to generate a magnetic flux swing of about 9.5 Vs for a total discharge time of the plasma of 10s. The vertical field coils can be arranged a) all outside the toroidal magnet system, and b) partially between the toroidal magnet system and the vacuum vessel, likewise in pairs symmetric to the torus midplane.

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.