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An overview is presented of recent experiments with radiating mantles on limiter and divertor machines, realizing simultaneously high confinement and high density at high-radiation levels. A variety of operational regimes has been observed and the characteristics of each are documented. High-performance plasmas (i.e. edge localized mode (ELM)-free H-mode confinement quality f(H93) greater than or equal to 1 and normalized beta values beta(n) greater than or equal to 2 simultaneously) with radiating mantles have been demonstrated under quasistationary conditions during the maximum flattop time of the machine (equal to tens of confinement times) on DIII-D and TEXTOR-94. Maximum values for beta(n) up to 4 and for the advanced tokamak confinement-stability product beta(n) x f(L89) up to 13, have been obtained in very high confinement mode (VH-mode) like discharges with radiating mantles in DIII-D. There is a striking similarity between improved ohmic confinement discharges (with or without Ne seeding) and radiating mantle discharges, indicating a possible common origin for the confinement improvement observed. Possible scenarios for the application of radiating mantles on larger machines such as JET and JT-60U are indicated.

A 140 GHz ECRH system is being installed at the ASDEX-Upgrade tokamak. The total power of 2 MW at a pulse length of 2 s — or 2.8 MW at 1 s, respectively — is generated in 4 gyrotrons and transferred to the plasma via 4 transmission lines, partly quasi-optically and partly through HE-11 wave guides. The power can be modulated up to 30 KHz. Rotatable mirrors inside the vacuum chamber facilitate the deposition of the rf power at any desired position within the cross section of the plasma. A first 0.5 MW — 0.5 s unit is operational since summer 1995. Gyrotron, transmission line and in-vessel components worked very reliably. Heating effects and the propagation of heat waves were analysed under various plasma conditions.

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.

Kinetic edge profiles are measured with high spatial and temporal resolution and their behaviour before and after the time of the L–H transition is analysed. It is shown that at low densities electron and ion channels are strongly decoupled and that the ion kinetic profiles determine the transition to H-mode.At medium densities, where electrons and ions are coupled, the H–L back-transition occurs at the same pressure gradients as the L–H transition. The threshold power for the H–L back-transition exhibits the same increase at low densities as the L–H transition.After the L–H transition the density development depends on the available neutral gas reservoir. Modelling of particle transport indicates that the temporal development of the density profiles can be reproduced assuming only a reduction in the edge diffusion coefficient.The influence of magnetic perturbation coils on the L–H transition power threshold strongly increases with density above a Greenwald factor of about 0.6.

Measurements of plasma core impurity ion temperature and toroidal rotation are obtained during the current ramp-up phases of plasma discharges in the ASDEX Upgrade tokamak, by means of charge exchange recombination spectroscopy utilizing short neutral beam injected blips. The phenomenology of these profiles as the current is varied is presented. The observations are reported for both ohmically heated and electron cyclotron heating assisted ramps, with auxiliary heating applied both on-axis and off-axis. The evolution of global parameters is also reported. A comparison between ion and electron kinetic profiles evolution is made. Power balance analysis is performed to establish the local confinement properties and their evolution during the current ramp. Transport modelling of several cases with a theory-based model is carried out and a recipe, which involves edge-relevant parameters, is proposed to account for the transport in the outer plasma region. Finally, linear gyrokinetic calculations of dominant microinstabilities allow us to correlate the observed transport with the background turbulence regime.

ASDEX Upgrade is equipped with an ICRH system consisting of 4 generators of 2 MW power each and 4 double loop antennas. The generators, tuneable in frequency from 30 to 120 MHz, cover several heating scenarios over a wide range of magnetic fields (1 T

Perturbative experiments are essential to understand the complex transport phenomena in fusion plasmas. The perturbative methods used for transport studies are summarized and the main properties discussed. Based on this approach, transport of particles, heat and momentum has been intensively investigated. The main results obtained for the different channels are described and illustrated with selected examples.

The ITER hybrid scenario aims to exploit non-inductive current drive to enable burn times in excess of 1000 s. To achieve this, and optimize fusion performance, requires high βN (the plasma pressure normalized to a stability scaling) and energy confinement equal to or greater than that predicted for the baseline scenario. This paper discusses results from the JET candidate hybrid scenario, where βN,MHD ≤ 3.6 plasmas have been produced. Despite a different initial phase, confinement relevant plasma parameters evolve rapidly towards those of equivalent ELMy H-modes and are well described by IPB98(y, 2). In contrast to previous ELMy H-mode studies, a dedicated β scan experiment in the JET hybrid candidate scenario shows a strong negative dependence of global confinement on βN. Analysis indicates that the core transport remains consistent with weakly dependent electrostatic transport, whilst the edge confinement decreases strongly with increasing βN. By combining global confinement data from ASDEX Upgrade, DIII-D and JET hybrid scenario discharges, a multi-machine database is produced. In contrast to the JET case, confinement in ASDEX Upgrade and DIII-D is shown to be inconsistent with IPB98(y, 2) and alternative dependences are explored.

Confinement properties close to the H-mode power threshold are studied in the ASDEX Upgrade tokamak. The results show that good confinement can be obtained close to the threshold with Type-I ELMs. The existence of Type-I ELMs does not necessarily require the heating power to be higher than the H-Mode power threshold, but it requires collisionality to be low enough. At higher collisionality Type-III ELMs replace the Type-I ELMs and confinement time is reduced by about 20%.

Divertor impurity retention and exhaust capability were studied at ASDEX for two different divertor geometries. In one of these, bypass openings allow neutral gas exchange between divertor and main chamber (open divertor), in the other the bypasses are closed (closed divertor). Experiments were made with a closed upper divertor and an open lower one, as well as with both divertor closed, and discharges with double-null or single-null (up or down) magnetic configuration where measured. From the analysis of the observed up-down asymmetries for the different configurations, information is deduced on the performance of different divertor geometries. Results show that a closed divertor has a much better retention and exhaust capability and the plasma impurity content is drastically reduced.

Boronization is the plasmachemical deposition of thin boron/carbon or pure boron films onto the inner wall of a tokamak device. Since mid-April 1989 the ASDEX vessel has been boronized seven times in glow discharges, first with a gas mixture of 10% B 2 H 6 + 10% CH 4 +80% He (A), once with 10% B 2 +90% H 2 (B) and once with 10% B 2 H 6 +90% He (C). All boronization runs were done at room temperature vessel walls. Local surface analysis of the deposited layers yielded thicknesses of up to 2000A. Boronization provided considerably improved plasma conditions. In ohmic discharges, reductions of CO by a factor of more than five in the divertor chamber and of two for H 2 O were observed, in relation to a reduction of oxygen by a factor of more than five as measured by spectroscopy, but the reduction of hydrocarbons was only marginal. Iron disappeared from the spectra and even copper, the target plate material, was significantly reduced. As a consequence, Z eff is e > 3 × 10 19 m −3 . The density limit was improved by 25% (boronization A) and more (boronization B and C). The oxygen-diminishing effect of boron/carbon and pure boron layers was found to be longer lasting than the suppression of metal impurities. Deteriorated wall conditions occurred only after hundreds of discharges, including numerous additional heating shots with ICRH, NI, LH and combined scenarios, but pre-boronization conditions were never reached. The main emphasis of this paper is to describe the impact of boronization on aspects of plasma operation

New experimental results obtained in ASDEX Upgrade [O. Gruber, H.-S. Bosch, S. Gunter et al., Nucl. Fusion 39, 1321 (1999)] plasmas in low confinement mode with central electron cyclotron heating are presented in which transitions in both the particle and electron heat transport properties have been observed. A comprehensive albeit qualitative explanation for both the transport channels is provided in the framework of the theory of ion temperature gradient and trapped electron mode microinstabilities. The different transport behaviors are related to the dominant instability at play and to the collisionality regime. In particular, central electron heating induces a flattening of the density profile when the dominant instability is a trapped electron mode, and density peaking is observed to increase with decreasing collisionality.

Linear gyrokinetic calculations of low density electron heat dominated plasmas in ASDEX Upgrade [O. Gruber, H.-S. Bosch, S. Gunter et al., Nucl. Fusion 39, 1321 (1999)] are presented. It is found that the dominant mode is a trapped electron mode, which under the experimental conditions is sensitive to the density gradient, collisions, and magnetic shear. The quasilinear heat flux scales in good agreement with the experimental observed heat conductivity. Many experimentally observed phenomena can be explained on the basis of the presented calculations: the existence of a threshold, the fast propagation of the heat wave, and the moderate stiffness of the electron channel. The implications of the calculations for the empirical investigations are discussed.

We study the evolution of turbulence at several radial plasma locations in the Axially Symmetric Divertor Experiment (ASDEX) Upgrade tokamak, using microwave reflectometry. We analyze two types of scenarios: H-mode plasmas with improved performance and plasmas with electron cyclotron resonance heating modulation. Profound modifications of the turbulence spectra, occurring at specific plasma regions, have been found in both cases, indicating the relation between turbulence changes and confinement. The results demonstrate the ability of the reflectometry system on ASDEX Upgrade to make localized measurements. Mapping the radial distribution of turbulence can help us to understand the links between plasma rotation, shear, E×B flows, and the plasma turbulence behavior.

The reflectometry system installed in the tokamak ASDEX Upgrade has 11 channels primarily aimed at profile measurements and two dedicated channels operating at fixed frequency to study density fluctuations. The channels used for profile measurements can also be operated at fixed frequency. In this article we describe the potential of the system to study important characteristics of the plasma turbulence, from combined fixed frequency and profile measurements (to localize the reflecting layers). The numerical tools used to treat fixed frequency data are briefly described. Using a set of similar discharges and the ability of the microwave reflectometry system to probe simultaneously different layers with several channels, we obtain the time evolution of the integrated power spectra of the different signals versus the minor radius of the plasma. In plasma with H-mode edge and internal transport barrier we could identify a zone at the plasma core and another at the edge with reduced fluctuation levels, coinci...

The paper summarizes the experiments performed with ion cyclotron resonance heating (ICRH) on ASDEX, from November 1984 until March 1986; the most interesting results are reported and discussed in detail. Heating and confinement studies using the hydrogen second harmonic scheme and the hydrogen minority scheme (PIC < 2.6 MW, tIC < 1.5 s) show a typical L-mode behaviour, i.e. a power dependent confinement degradation, which is rather similar to that found with neutral beam injection (NBI) heating. ICRH is accompanied by a slightly improved particle and energy confinement compared with that of NBI; this is also true for a combined ICRH + NBI scheme, up to Ptot ≈ 4.5 MW, absorbed in the plasma. Particular efforts have been devoted to investigations of the second harmonic regime in H/D plasmas with nH/ne ≈ 0.1 - 1, with a view to heating mixtures in reactor relevant plasmas. The achievement of H-mode transitions with ICRH alone in the hydrogen minority scheme at an absorbed RF power of about 1.1 MW supports the assumption of common confinement properties in auxiliary heated tokamaks, since they appear to be widely independent of the additional heating method. ICRH specific impurity problems, such as the strong release of iron from the vessel walls, have been overcome by applying extensive in situ wall carbonization. The mechanisms responsible for impurity generation have partly been identified and analysed; however, the problem still remains to be solved. Impurities preferentially released from the ICRH antenna do not pose problems.

The H mode in ion cyclotron-resonance-heated plasmas has been investigated with and without additional neutral beam injection. Ion cyclotron-resonance heating can cause the transition into a high-confinement regime (H mode) in combination with beam heating. The H mode, however, has also been realized: for the first time: with ion cyclotron-resonance heating alone in the D (H)-hydrogen minority scheme at an absorbed rf power of 1.1 MW.

The highest β-values derived from diamagnetic flux measurements in H-discharges are found to be close to the β-limit due to kink and ideal ballooning modes. Both energy and particle confinement are degraded in these discharges and do not recover during the injection period. The energy confinement time decreases by a typical factor of two, the electron thermal diffusivity being correspondingly enhanced. Electron heat conduction is again found to be the dominant energy loss channel. The degradation of confinement close to the ideal-MHD limit suggests that kink and ballooning instabilities occur and are responsible for the enhanced transport. This conclusion is supported by the correlation with the pressure profile and by the result that global and local transport do not change in a reference discharge which is stable against ideal-MHD modes.

The first plasma experiments on the W7AS advanced stellarator were conducted in October 1988, after magnetic surface mapping. The characteristics of the device are described. During the first phase of operation, 70 GHz ECF was used to generate and heat a 'currentless' plasma which was maintained in quasi-steady state for typically 0.5 s. Effects of the magnetic configuration on the confinement and measures to deal with the observed plasma current (bootstrap current and ECF-driven current) were investigated. Preliminary results of transport analysis are presented and compared with predictions of transport models.

The authors give an overview of the different confinement regimes observed on ASDEX and compare the changes during the transition phases with qualitative tendencies suggested by theoretical models. The transitions discussed are those between purely Ohmic heating and additional heating in the L-regime between the L- and the H-regime and between discharges with flat and peaked electron density profiles.

Interaction of lower hybrid waves at 1.3 GHz with ions and electrons was studied in the density range 0.2-5*1013 cm-3 in the ASDEX tokamak. At high densities, ne>or approximately=4*1013 cm-3, fast ions with mainly perpendicular velocities are produced by the RF power at the plasma periphery. They are not well confined and do not lead to any bulk plasma heating. At lower densities, 2*1013

Discharges fuelled by stationary pellet injection (PI), gas puffing (GP) or a combination of the two methods are compared with respect to recycling in the divertor and particle confinement. Fuelling by PI yields much better global particle confinement than by GP. This has been found for both low and high recycling. In the low-recycling case this improvement is due to the deeper particle deposition for PI than for GP since the transport in the inner plasma is not reduced. For high recycling the improvement results from both the deeper deposition and a reduction in the transport. The best global particle confinement was found for phases with low or no GP. This, however, can be reached for short times only. Since with PI alone it is impossible to keep the recycling on a high level, GP is unavoidable for sustaining the favourable high-recycling condition.

First ICRF experiments on ASDEX have been performed at 67 MHz, corresponding to 2 Omega CH-heating of a hydrogen plasma at B0=2.2 T. Despite divertor operation ICRH is accompanied by a significant increase of impurity production which can drastically be reduced by means of wall carbonisation. RF power up to 2.3 MW is routinely coupled to the plasma for pulse lengths of up to 1 sec. The RF heating is found to depend strongly on plasma preheating. In combination with neutral beam injection the ICRF heating efficiency is even higher than the one of NI. Confinement degrades with ICRH to values in between NI-L-type and OH confinement.

Neutral particle fluxes onto the walls of ASDEX have been investigated using a time-of-flight (TOF) method. The energy distributions of the neutrals could be determined in the range of 10–1000 eV/amu. Ohmic divertor and limiter discharges with equal plasma currents and densities have been compared for He and D2. The He0 outflux at ∼2000 eV from He discharges is 110 of the corresponding D0 flux in D2 discharges. At lower energies this difference is much smaller. In all cases many more He neutrals were observed than was anticipated from the CX rate-coefficients for He2+. The impurity fluxes due to sputtering by the CX-neutrals show no significant difference for He and D2 discharges. For divertor discharges CX-sputtering can fully account for the Fe impurity content determined spectroscopically.

A review is given on investigations in the ASDEX Tokamak on impurities in ohmically, NI and ICRH heated plasmas. For ohmic discharges in H2 and D2 it is found that iron release from the wall can be explained by sputtering due to neutral charge exchange (CX) atoms. In the case of He, however, significant contributions caused by ion sputtering are inferred. Comparing discharges with C limiters in He and D2 suggests that in the case of hydrogen chemical processes are involved in C sputtering. By means of wall carbonization the concentrations of metal ions in the plasma could be substantially reduced. This achievement is of particular importance for NI counter-injection and ICRH, where under non-carbonized conditions severe impurity problems occur. We studied impurity confinement in the case of various heating scenarios by means of the laser injection technique. The poorest confinement is found for the L-phase of NI. Metal injection into the high confinement H-phase generally causes temporary suppression of the edge localized modes (ELM's). With respect to ICRH we conclude that enhanced wall erosion — probably due to the production of high energy ions in the boundary — together with a slightly increased impurity confinement is the dominant reason for the increase of the metallic concentrations. Impurity sputtering as an alternative strong erosion process was experimentally ruled out.

The characteristics of neutral-beam-heated ASDEX discharges exhibiting either low (L)- or high (H)-confinement are described. H-mode discharges, which are by now observed over a wide range of operating conditions, show a spontaneous improvement in particle and energy confinement after a short L-phase at the beginning of neutral injection. H-discharges yield high βp values comparable to the aspect ratio A (βp≤2.65~0.65 A). The most important parameter for transition to the H-mode seems to be a high edge electron temperature: Te values of -600 eV only a few centimeters inside the separatrix with radial gradients of ~300 eV/cm are measured. This requirement of high edge temperatures explains the lack of success in obtaining the H-regime in limiter discharges. Numerical simulation of the broad ne and Te profiles typical of H-mode plasmas indicates a reduction in electron thermal diffusivity by a factor of typically 2 over the entire plasma. H-mode energy confinement times are found to scale linearly with current, but to have little dependence on plasma density and absorbed beam power (PNI≤3.4MW). The confinement is degraded by a fast growing mode localized at the plasma edge that may be identified as a kink or tearing mode driven unstable by the high current densities at the edge.

Systematic investigations of the scrape-off layer (SOL) in the midplane of ASDEX have been carried out in He, D2 and H2 for diverted ohmic discharges over a wide range of plasma conditions: ne ∼ 0.5−4.7 × 1013 cm−3, Ip = 200–450 kA, BT = 16–23 kG, qa∼ 2.4–4.4 and POH = 200–480 kW. For the first two cm outside the separatix, ne is found to decay exponentially with an e-folding length λn given by λn = kqα (He, k = 1.32 cm, α = 0.52; D2, k =1.29 cm, α = 0.35; H2, k = 1.18 cm, α = 0.4) when from which follows for qa = 3: λn(D2) ∼ λn(H2) ∼ 0.8 λn(He). The qα scaling is roughly predicted by the simple formula λn = D⊥ Lυ∥ under the assumption D⊥ ∝ mi−0.5 (as has been observed on ASDEX for H2 and D2). There appears to be no explicit λn dependence on heating power. λn varies strongly with ne in the range ne ≤ 1 × 1013 cm−3, decreasing for example (D2,H2; qa = 3.0), from λn ≥ 3 cm at ne ∼ 0.5 × 1013 cm−3 to λn ∼ 1.9 cm for ne ≥ 1.5 × 1013 cm−3, ne at the separatrix is primarily a function of ne.

The sawtooth oscillations in tokamak discharges with Ohmic and neutral-beam heating could be suppressed when a large part of the plasma current was driven by lower-hybrid waves (I/sub HF//I/sub p/approx. =0.5). The stabilization is due to a flattening of the current profile j(r) and an increase of q(0) above 1. Higher central electron temperatures are obtained with neutral-beam heating if the sawteeth are stabilized. The increase in total energy content in this case was 30% higher than in the presence of sawteeth.

Neutral beam (NI), ion cyclotron resonance (ICRH) and lower hybrid resonance (LHRH) heating on ASDEX are discussed with regard to their effect on plasma confinement. Comparison of NI and ICRH shows that the L and H-regimes are universal confinement modes of auxiliary-heated tokamak plasmas (i.e. independent of the heating method), and that the edge electron temperature (or a related parameter) dictates which mode prevails. In this connection it is noted that carbonization of the vessel walls impedes transition to the H-mode in the case of NI heating. Studies of energy confinement in the intermediate regime from ohmic to NI heating reveal a gradual transition from ohmic ( approximately ne) to neutral injection L-mode ( approximately I) scaling. At the same time a remarkable invariance of electron temperature profile shape with increasing heating power is observed. Changing the NI power deposition profiles from central to off-axis leaves gross energy confinement times unchanged while central confinement is substantially improved.

The thermal wave of a minor disruption can initiate the H phase of a neutral-beam--heated divertor tokamak discharge. Its propagation is used to probe the plasma edge conditions at the H transition. The results show the existence of a transport barrier which forms at the plasma edge and impedes the flow of particles and energy across the plasma surface, giving rise to improved confinement properties. Location and extension of the barrier coincide with the edge zone of increased shear specific to the divertor configuration.

The main topics of ICRF investigations in ASDEX are the influence of the divertor on impurity production and transport in ICRH heated discharges, and the heating efficiency and plasma confinement in various scenarios (minority and harmonics regimes). The first experiments were conducted in November 1984 at 67 MHz, corresponding to second harmonic heating of a hydrogen plasma at B0=2.2 T. A transmitted power of 2.5 MW has been reached so far, the total capability being 3 MW. A linear increase of the central electron and ion temperature with the rf power is observed in Ohmically preheated plasmas (ΔTe∼280 eV, ΔTi∼500 eV, ΔWp∼17 kJ at a power of 1.2 MW coupled to the plasma and ne=3.5×1013 cm−3). The total radiation increases linearly with the power, too, and the ratio Prad/Ptot∼0.35 stays approximately constant. However, first investigations indicate that with a divertor, ICRF operation is also accompanied by a significant increase in impurity production. The presence of neutral beam injection in addition ...

This letter reports on two experiments undertaken to evaluate the retention of gaseous and target produced impurities in the ASDEX divertor. The retention for gaseous impurities was determined by puffing Ar into the main chamber and simulating the time behaviour of the Ar XVI line intensity with a time dependent impurity transport code including a simple divertor model. During Ohmic heating a factor of 3 and 4.5 enhancement of impurity retention if found relative to the vacuum time constant (90 ms) of the divertor chamber, for ne = 2 × 1013 cm−3 and ne = 3.5 × 1013 cm−3, respectively, while a drastic breakdown of the retention occurs during high power NI heating. – To deduce the retention of impurities generated at the divertor plates, a segment (3.5%) of the plates was covered with copper, a metal previously not used in ASDEX. By measuring the Cu influx at the target plates and the line intensity of the Cu XX line (11.38 A) in the core plasma and by using the transport code, it is found that during NI heating (ne ≤ 2 × 1013 cm−3) Cu atoms originating from the target plates have a ≤ 3.5 times higher probability to penetrate into the core plasma than if they had when originating from the main chamber walls.

Recent measurements of plasma boundary and divertor scrape-off parameters for ohmically and neutral injection heated plasmas are presented. For these data the power flow onto the divertor plates and the sputtering rates at the plates are calculated and compared with separate measurements. The impurity behaviour in front of the plates is also discussed.

A far infrared laser interferometer with a band-width of 10MHz is designed. FIR lasers and involved problems are briefly described. Present results concerning the lasers and beats between two FIR lasers are given. The minimum fluctuation of electronic density that can be detected is estimated.

Recharging of the Ohmic-heating transformer of a tokamak by means of lower-hybrid waves is demonstrated experimentally in ASDEX. The results are analyzed on the basis of a simple transformer circuit. A recharging efficiency is defined and found to depend on rf power, plasma density, and plasma resistivity modified by the applied rf power. Up to now, we achieved in our recharging experiments in ASDEX a flux swing of FI${\ifmmode \dot{}\else \.{}\fi{}}_{\mathrm{OH}\mathrm{Mdt}=0.24}$ V sec, at an rf power of ${P}_{\mathrm{RF}}$=690 kW, with a pulse duration of 1 sec, while maintaining a plasma with n${\ifmmode\bar\else\textasciimacron\fi{}}_{e}$=4\ifmmode\times\else\texttimes\fi{}${10}^{12}$ ${\mathrm{cm}}^{\mathrm{\ensuremath{-}}3}$ and ${I}_{p}$=290 kA.

A fast FIR laser interferometry technique, using CO 2 -pumped CH 3 I lasers at 447 μm, is employed in the Petula B tokamak to diagnose low-frequency (ω ci ) electron-density fluctuations. Digitally implemented FFT and coherence analyses are made to determine various characteristics of the fluctuations.

Lower hybrid current drive can be applied in a tokamak in a number of different schemes, namely for start-up of the discharge, for plasma current ramp-up, for driving a stationary plasma current or for recharging the Ohmic heating transformer in a quasistationary tokamak operation. Recent experimental work related to these topics can be found in Refs. /1–10/. In this paper we report experiments performed in ASDEX.

Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER and DEMO operation. This is of particular importance for the bulk ion heating capabilities of ICRF waves. Efficient bulk ion heating with the standard ITER ICRF scheme, i.e. the second harmonic heating of tritium with or without 3He minority, was demonstrated in experiments carried out in deuterium-tritium plasmas on JET and TFTR and is confirmed by ICRF modelling. This paper focuses on recent experiments with 3He minority heating for bulk ion heating on the ASDEX Upgrade (AUG) tokamak with ITER-relevant all-tungsten PFCs. An increase of 80% in the central ion temperature Ti from 3 to 5.5 keV was achieved when 3 MW of ICRF power tuned to the central 3He ion cyclotron resonance was added to 4.5 MW of deuterium NBI. The radial gradient of the Ti profile reached locally values up to about 50 keV/m and the normalized logarithmic ion temperature gradients R/LTi of about 20, which are unusually large for AUG plasmas. The large changes in the Ti profiles were accompanied by significant changes in measured plasma toroidal rotation, plasma impurity profiles and MHD activity, which indicate concomitant changes in plasma properties with the application of ICRF waves. When the 3He concentration was increased above the optimum range for bulk ion heating, a weaker peaking of the ion temperature profile was observed, in line with theoretical expectations. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

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.

This paper describes the updates to and analysis of the International Tokamak Physics Activity (ITPA) Global H-Mode Confinement Database version 3 (DB3) over the period 1994–2004. Global data, for the energy confinement time and its controlling parameters, have now been collected from 18 machines of different sizes and shapes: ASDEX, ASDEX Upgrade, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M, JT-60U, MAST, NSTX, PBX-M, PDX, START, T-10, TCV, TdeV, TFTR and TUMAN-3M. The database now contains 10382 data entries from 3762 plasma discharges, including data from deuterium–tritium experiments, low-aspect ratio plasmas, dimensionless parameter experiments and plasmas. DB3 also contains an increased amount of data from a range of diverted machines and further data at high triangularity, high density and high current. A wide range of physics studies has been performed on DB3 with particular progress made in the separation of core and edge behaviour, dimensionless parameter analyses and the comparison of the database with one-dimensional transport codes. The errors in the physics variables of the database have also been studied and this has led to the use of errors in variables fits. A key aim of the database has always been to provide a basis for estimating the energy confinement properties of next step machines such as ITER, and so the impact of the database and its analysis on such machines is also discussed.

The first two-frequency gyrotron has been installed in the new multi-frequency ECRH system at the ASDEX Upgrade Tokamak experiment. In its final stage the system consist of 4 gyrotrons, tunable in the range 105-140 GHz with a total power of 4 MW and a pulse length of 10 s. It employs a fast steerable launcher for feedback controlled deposition that allows for poloidal steering of 10/spl deg/ within 100 ms. Transmission line elements, such as corrugated waveguides, polarizer mirrors and vacuum windows, are designed to cope for this frequency band.

Summary form only given. The new ECRH system which is currently under construction at ASDEX Upgrade is aimed for a total power of 4 MW, generated by 4 gyrotrons, and a pulse length of 10 sec. The system should allow for central heating at different neoclassical tearing modes (NTM) through the possibility of current very localized feedback controlled power deposition with a narrow beamwidth by using a fast steerable launcher. The transmission line components are especially designed to cope for the large bandwidth of gyrotron to be installed can work at 105 GHz and at 140 GHz. The operating modes are TE/sub 17.6/ and TE/sub 22.8/. Here we make use of the 3/spl lambda//2 and 4/spl lambda//2 resonances of the single disk synthetic diamond vacuum window at these frequencies. A second gyrotron is step-tunable, i.e. it can work at several frequencies within the same frequency range. For each frequency the gyrotron oscillates in a different cavity mode. Short pulse test gyrotrons at FZK and IAP have demonstrated efficient generation of power in excess of 1 MW at all these frequencies. The frequency can be changed between two ASDEX Upgrade pulses and requires a resetting of the cryomagnetic field at the cavity, the gun magnetic field and the collector magnetic field. There is one complicating feature with such a gyrotron: the output beam leaves the gyrotron window at slightly different angles and positions so that mirror adjustment in the matching optics unit is necessary. In order to limit the number of required phase correcting mirror sets we choose 4 frequencies as our main operating modes. The torus window required for the step-tunable gyrotron should be in principle broadband, e.g. a Brewster window. To avoid constraints in the polarization and the quasi-optical design, a tunable-disk window with a remote controlled adjustment of the distance between the disks is used. For the two-frequency gyrotrons a single-disk window is sufficient.