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Electron cyclotron resonance heating (ECRH) at the W7-AS stellarator is performed two systems of comparable heating power at 70 GHz (0.8 MW, 3 s) and 140 GHz (0.9 MW, 0.4 s). Experiments with 140 GHz open a new parameter window with high density operation up to 1.1 × 1020 m−3. H mode transitions were observed in high density discharges at 2.5 T and at lower density at 1.25 T. The operational window for the stellarator H mode with emphasis on the density and power threshold and the influence of gas puffing on the H transitions are discussed. Perturbation experiments with modulated ECRH power were performed to determine the heat transport and the power deposition profile. Density control is achieved in combined heating with neutral beam injection (NBI) despite the beam particle fuelling, whereas with NBI alone a steady density rise is observed. The impact of combined heating on the impurity confinement was investigated. Electron cyclotron current drive (ECCD) was studied in different magnetic field configurations and the influence of trapped particles on the ECCD efficiency was examined experimentally and compared with theory.

In W 7-AS the [ital H] mode has been observed for the first time in a currentless stellarator plasma. [ital H] modes are achieved with 0.4 MW electron cyclotron resonance heating at 140 GHz at high density. The [ital H] phases display all characteristics known from tokamak [ital H] modes including edge localized modes (ELMs). The achievement of the [ital H] mode in a shear-free stellarator without toroidal current has consequences on [ital H]-mode transition and ELM theories.

There exist two classes of discharges on ASDEX which overcome the severe confinement deterioration at high densities or heating power. The first class is characterized by unusually peaked density profiles and the second one by exceptionally broad ones. These profiles and the concomitant confinement enhancement imply a change in both the particle and heat transport. Whereas the changes in particle transport are not fully understood, transport analyses point out that the improved heat transport in the first profile class can be explained by reduced ion contribution losses coming close to the neoclassical ones. The different results for the ion transport with flat and peaked density profiles are qualitatively consistent with that expected from eta i-modes. This conclusion is supported by momentum transport studies. The analyses, however, cannot yet explain the electron heat transport which is anomalously high in all regimes.

Following the report of Fleischmann and Pons,(1) we (The Bavarian Bubble Bottle Team) have attempted to reproduce their claims of cold nuclear fusion, and failed. We note that our measurements would not be able to detect neutrons at the level of Joneset al.(2) Three electrolytic cell experiments were conducted using palladium cathodes and Platinum anodes, in a 0.1-Molar solution of LiD in heavy water, without any signs of neutrons, tritium, or gammas above backgrounds, and within ±0.3 watt accuracy calorimetry, no excess heating. Excess heating at the levels of F&P would have been easily detected, if present. Intrinsic tritium, differing from each D2O bottle tested, was however observed. The longest duration experiment ran for 21 days, and was an attempt to duplicate the large “melting incident” of F&P. This was terminated on April 28, 1989, by throwing the vacuum-cast 22 gram, deuterium-loaded palladium cathode directly into liquid nitrogen, immediately next to a bare BF3 counter (backed by 25 cm of moderator), in order to attempt one of the Italian ENEA neutron production variants. No neutrons above backgrounds were seen, while counting for 1 hour, and also none while the piece warmed to room temperature over the next hour. Post-mortem analysis of the darkened, hardened Pd piece showed large crystal grains (up to 2 mm × 2 mm), and continuing evolution of gas bubbles at the grain boundaries even days after the experiment was ended. eight weeks after loading, the catalytically-active palladium piece continued to create heavy water (with exposure to oxygen in the air).

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.

Pulsator was operated by IPP Garching from 1973 to 1979. It made two important contributions to international tokamak research. 1. In Pulsator, for the first time a tokamak plasma was influenced locally by an external, resonant, helical field. The accompanying formation of islands on magnetic surfaces, which affects transport and stability, provided the starting point for experimental investigation of the disruptive instability and its theoretical description. 2. By gas puffing during the discharge it was possible to attain a high-density regime which dramatically raised the then prevailing nτE values by almost two orders of magnitude. In this high-density regime an inward directed particle drift velocity and, as a consequence, under certain conditions an accumulation of impurities in the plasma core was found.

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.

An instrumented scoop limiter probe is being operated on TEXT to test the concept of limiter cooling and improved particle removal efficiencies using an externally-applied resonant magnetic field perturbation (the resonant helical divertor concept). Cooling of the limiter face has been demonstrated for limiter positions ranging from r L = 29.0 cm inward to r L = 25.5 cm (the Text primary poloidal hoop limiter radius r a = 27.0 cm). Pressure rises in the limiter throat of approximatel 40% are observed under optimized conditions. Interchangable limiter heads with thicknesses of 1.0 cm and 0.3 cm have been used to examine particle ducting into the scoop aperture. Experimental results are discussed along with observations of the limiter floating potential, H α recycling emissions, pressure measurements, and edge density and temperature measurements.

Investigations of impurity accumulation phenomena in ASDEX are reviewed. There are four different operating regimes where pronounced accumulation is observed and these regimes are also characterized by improved energy confinement. In particular, medium-Z metallic ions are involved in accumulation processes whereas low-Z ions appear almost unaffected. The rapid accumulation observed in the case of metallic ions may be explained by neoclassical inward drifts if we assume that the anomalous diffusion is sufficiently suppressed, some indication of this being found from laser blow-off studies. The present results, however, can only be partly explained by neoclassical theory, according to which accumulation of low-Z impurities should also occur. The temporal behaviour of accumulation and the retarding effect of proton dilution for collision dominated transport are also discussed. Finally, we conclude that the full benefits of improved energy confinement can be achieved only if the impurity influxes are kept to a sufficiently low level. Expressed in terms of concentrations under low confinement conditions we have to postulate, for ASDEX, concentrations ≲ 10−4 for metals and ≲ 2% for all light impurities.

The H-mode transition can lead to a rapid increase in tokamak plasma confinement. A semiempirical transport model was derived from global OH and L-mode confinement scalings and then applied to simulation of H-mode discharges. The radial diffusivities in the model depend on local density and pressure gradients and satisfy an appropriate dimensional constraint. Examples are shown of the application of this model and of similar models to the detailed simulation of two discharges which exhibit an H-mode transition. The models reproduce essential features of plasma confinement in the Ohmic heating and the low- and highconfinement phases of these discharges. In particular, the evolution of plasma energy content through the H-mode transition can be reproduced without any sudden or ad hoc modification of the plasma transport formulation.

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.

A toroidal large area limiter was installed in the divertor tokamak ASDEX. The limiter basically consists of a plane annular stainless steel sheet which is tangential to the plasma and which is placed in the bottom of the main plasma chamber. In a second series of experiments this limiter was covered with graphite plates. With the steel limiter clean discharges could not be obtained. Energy deposition on the limiter was only 12% of the input. With the graphite limiter Z eff was below 3, energy deposition on the limiter was 25% (up to 40% in low density operation). Energy deposition was toroidally symmetrical. Results, including some experiments with a single null divertor are presented.

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.

The Divertor Tokamak ASDEX, its neutral injection system and its ICRH system have been modified to permit additional heating with a power of 6 MW for pulse lengths up to 10 s. The paper summarizes the arguments for long-pulse heating, describes the technical modifications of the divertor performed, their effect on the operational behaviour of the tokamak and presents a few typical results of recent experiments exploiting the long-pulse heating facilities.

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.

The fueling efficiency for gas puffing, i.e. the fraction of the external gas flux that is ionized inside the separatrix, is reduced in divertor discharges since part of it is ionized in the scrape-off layer and pumped off by the divertor. The fueling efficiency is determined by switching-off the gas feed during the stationary phase of a discharge and dividing the time derivative of the total number of particles inside the separatrix by the external gas flux. The determination of this time derivative must take into account profile changes. In ASDEX the fueling efficiency ranges from close to 1.0 for discharges with a stainless steel poloidal limiter and decreases to about 0.2 at high densities ( 6 × 10 13 cm −3 line average) for diverted discharges. These results are compared with estimates of the fueling efficiency which include molecular disintegration, plasma albedo for neutral atoms and imperfect wall reflection.

Use of the ASDEX divertor permits the production of stable low-density discharges (n/sub e/> or approx. =10/sup 12/ cm/sup -3/) with extremely low resistivity lasting for more than 10 s. While the distribution functions of electrons and ions show suprathermal tails, runaway electrons in the megaelectronvolt range are found to disappear with decreasing density. There are indications that in these discharges the energy confinement is improved compared with ALCATOR scaling.

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.

Local transport in the L (low βp) and H (high βp) types of double-null divertor discharges in ASDEX is investigated by computer modelling. For the L type, two subregimes with different local transport behaviour can be distinguished: Homogeneous electron heat diffusivities χe = e or flat χe profiles and diffusion coefficients D = 0.2 χe are found for e < 5.2 x 104 cm2s–1 (L-I), while above this limit, steep χe and flat D profiles are determined (L-II). The observed decrease in energy confinement time with increasing neutral injection power is found to be due to enhancement of the electron heat diffusivity. In the L-II regime, χe strongly increases with radius, while the diffusion coefficient remains flat. The diffusivities show a rather weak dependence on time as in the L-I regime. In both L regimes they rise with increasing injection power and decreasing plasma current, but are independent of density. In the H regime, transport is characterized by flat χe profiles and D = 0.2 χe. The diffusivities are reduced by a factor of about two in the whole plasma compared with the L type. They show the variation with plasma current found in L discharges and are also independent of density. One more common feature of L and H regimes is that the ion heat diffusivity is in the range of 1 to 3 times the neoclassical values.

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.

A series of initial resonant helical divertor (RHD) experiments have been carried out in ohmically and auxiliary heated JIPP T-IIU plasmas. Disruptive and MHD instabilities make the interpretation of the RHD results difficult but an apparent increase in the energy confinement time is observed when the helical magnetic perturbation is applied. This may be due to the suppression of MHD activity or to a reduction in the edge convective heat losses. Magnetic island effects have been observed on the floating potential of a Langmuir probe array and energy scrape-off layer widths have been measured with and without helical perturbations during ICRF operation. Basic pump limiter data is presented including ion temperatures and C4+ impurity profiles. Energy confinement times are reported in ohmically and NBI heated discharges.

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.

In the first experimental phase the divertor tokamak ASDEX was run with a closed SS-limiter. The cleaning procedure for the vessel consisted of baking to 120°C, carbon removal by glow discharge in hydrogen, and oxygen removal with a continuous low power 50 Hz AC discharge. Wall contact of the plasma was reduced by carefully positioning the plasma with a feedback system. Discharges with plasma currents up to 280 kA and a disruption-free duration of up to 1 s were reliably produced with a filling pressure of 5 × 10 −5 mbar and a programmed current rise. No change in start-up conditions and discharge behaviour was observed in material limiter discharges with superimposed divertor field.

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.

Both in ohmically and beam-heated L-mode discharges of ASDEX, the electron-temperature (T/sub e/) profile shape can be varied over a wide range by the choice of the safety factor q/sub a/. The power-deposition profile, on the contrary, has no effect on the T/sub e/-profile shape. In current-free WVII-A stellarator plasmas, no such invariance property is found. An independent constraint seems to fix the current distribution j(r) of the tokamak, which defines the conditions of electron heat transport.

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 ...

Strongly peaked electron density profiles have been obtained in ASDEX by different refuelling methods: pellet fuelling (ohmic and co-injection heating), NBI counter-injection and recently by reduced gas puff fuelling scenarios. These discharges show in common increased density limits, a canonical electron temperature profile independent of the density profile and an improvement of the particle and energy confinement. Whereas the changes in particle transport are not fully understood, transport analyses point out that the improved energy transport can be explained by reduced ion conduction losses coming close to the neoclassical ones. The different results for the ion transport with flat and peaked density profiles are quantitatively consistent with that expected from eta i-driven modes. The analyses cannot yet explain the anomalous electron energy transport, apart from identified continuous trends such as inverse scaling with the isotope mass and enhancement with heating power.

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.

For the particle losses of a weakly ionized plasma which result from the torus drift in a curved magnetic field, an expression is derived which is valid for certain parameters of the positive column of a gas discharge. To check this theory the “AMBIPOL” device was built. With this device it was possible to determine simultaneously the losses both in the toroidal and in the linear magnetic field by measuring the longitudinal electric field strength. As theory predicts, with growing magnetic field strength a weaker decrease of the longitudinal electric field was observed in the toroidal part of the discharge as compared to the linear part. The measured values of the relative electric field strength, however, exceed the theoretical limit, although the measurements of the electric field in the straight part and the measurements of the particle density and of the electron temperature in the curved part are consistent with theory. Moreover, contrary to the expectations, the onset of the KADOMTSEV instability occurs at lower critical magnetic fields in the toroidal part than in the straight part. Several possible explanations are discussed. In a later paper it will be attempted to make a choice among the three most probable ones.

In a previous paper31 discrepancies between theory and experiment were found on investigating the positive column in a curved magnetic field. The approximation derived in 31 for the torus drift in a weakly ionized magnetoplasma is therefore checked here (Part I) with a refined theory which also yields the transverse electric field strength. Experimentally, both the transverse electric fields and the density profiles in the DC discharge were determined in addition to the longitudinal electric field strength. The discrepancies occurring in 31 are ascribed to the fact that the plasma concentrates at the cathode end of the magnetic field coils, this effect having a considerable influence on the form of the transverse density profile and on the stability behaviour. Part II later will show how the influence of this concentration can be eliminated and what effect in the current-carrying toroidal plasma causes a marked reduction of the charge carrier losses.

A concept for a helical divertor is proposed, which utilizes a resonance effect, and reduces the required divertor currents to a small fraction (typically 10 -2 ) of the plasma current. This divertor should have a larger screening efficiency than the poloidal field divertor, and could operate with divertor coils outside the vacuum vessel or the blanket.

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.

The high- and Low-Z impurity make-up of the plasma both at Low and high density Tokamak discharges has been investigated. At 4 ms into the discharge the oxygen impurity content is estimated from the OVI (1032 Aring;) Line intensity to be about 3 % of the electron density. Later in the discharge the quantitative estimate of the oxygen content is more difficult to make because only the oxygen flux is measured. At late times the total oxygen flux is found to be 2 × 1014 cm−3s−1 which is 1.3 % of the hydrogen flux. The high-Z impurities come into the plasma either from the stainless steel wall (Fe, Ni, Cr) or from the limiter (Mo). The L-lines of these impurities are observed. Their intensity is found to rise during the first 40 to 50 ms of the discharge, which indicates a continuous accumulation of high-Z impurities in the bulk of the plasma. Utilizing global rate coefficients for the L-lines, one can obtain the temporal evolution and the radial distribution of the high-Z impurities in the plasma. No accumulation of the high-Z impurities in the center of the discharge is observed. On the contrary, one finds for both the low and the high density discharges an inverted high-Z impurity profile with a weak minimum in the center of the discharge. The low density discharge is usually (for clean wall conditions) metal dominated (Zeff≈ 6 to 8), but at high densities the oxygen begins to play a major role (Zff≈ 1.5 to 2). The high density discharges are obtained by allowing the hydrogen gas to flow into the discharge at some chosen time (usually at t = 40 ms). During the fast increase in electron density - typically from 2.5 × 1013 to 1.8 × 1014 cm−3 - one observes that the absolute density of the high-Z impurity decreases by almost a factor of 4. Comparison of these results with the QWAASS measurements show a rough quantitative agreement.

A new operational regime has been observed in neutral-injection-heated ASDEX divertor discharges. This regime is characterized by high ${\ensuremath{\beta}}_{p}$ values comparable to the aspect ratio $A({\ensuremath{\beta}}_{p}l~0.65A)$ and by confinement times close to those of Ohmic discharges. The high-${\ensuremath{\beta}}_{p}$ regime develops at an injection power \ensuremath{\ge}1.9 MW, a mean density ${\overline{n}}_{e}g~3\ifmmode\times\else\texttimes\fi{}{10}^{13}$ ${\mathrm{cm}}^{\ensuremath{-}3}$, and a $q(a)$ value \ensuremath{\ge}2.6. Beyond these limits or in discharges with material limiter, low ${\ensuremath{\beta}}_{p}$ values and reduced particle and energy confinement times are obtained compared to the Ohmic heating phase.

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.

The xenon spectrum, excited in ECR- and NBI-heated plasmas with central electron densities of around 1020 m-3, and central electron temperatures from 0.7 to 2.5 keV, has been studied photoelectrically with a multi-channel grazing-incidence spectrometer. Besides numerous well-known lines of Zn- and Cu-like Xenon, more than 50 additional lines which have not yet been published in the literature have been found and partly identified,.