J. Hofmann

The Isotope Effect in ASDEX

The paper describes the effect of the isotopic mass on plasma parameters as observed in the ASDEX tokamak. The paper comprises Ohmic as well as L mode, H mode and H* mode scenarios. The measurements reveal that the ion mass is a substantial and robust parameter, which affects all the confinement times (energy, particle and momentum) in the whole operational window. Both core properties such as the sawtooth repetition time and edge properties such as the separatrix density change with the isotopic mass. Specific emphasis is given to the edge parameters and changes of the edge plasma due to different types of wall conditioning, such as carbonization and boronization. The pronounced isotope dependences of the edge and divertor parameters are explained by the secondary effect of different power fluxes into the scrape-off layer plasma and onto the divertor plates. Finally, the observations serve to test different transport theories. With respect to the ion temperature gradient driven turbulence, the isotope effect is also studied in pellet refuelled discharges with peaked density profiles. The results from ASDEX are compared with the results from other experiments

Measurement of the radial electric field in the ASDEX tokamak

Estimates of the radial electric field Er at the plasma periphery are obtained by measuring the drift velocities of low-Z impurity ions (He II, B IV, C III). The drift velocities are determined from the differential Doppler shift of visible line emission observed along opposite viewing directions. The principle of the measurement, including contributions from the diamagnetic drift, as well as radial gradients in the excitation rate and the effect of integrating along the line of sight are discussed in detail. The measured line of sight averaged drift velocities can be strongly influenced by the location and shape of the Er profile, especially if, as measured on other tokamaks, it is localized to a narrow region just within the separatrix. Values of Er estimated by assuming a constant radial profile underestimate maximum local values. During the H*-phase, however, high line of sight averaged perpendicular drift velocities of the B IV ions of a least 15 km/s in the electron diamagnetic drift direction are observed. From this, the presence of a strong negative radial electric field of at least 25 kV/m in the plasma edge region is inferred. Values of the B IV ion poloidal drift velocity calculated from an appropriate neoclassical theory are in the same direction as those measured. However, the calculated line of sight averaged values are much smaller than the measured ones. This reinforces the conclusion that a strong negative radial electric field is present just within the separatrix during the H-mode

H-mode of W7-AS stellarator

In W7-AS the H-mode has been observed for the first time in a currentless stellarator plasma. H-modes are achieved with 0.4 MW Electron Cyclotron Resonance Heating with 140 GHz at 2.5 T and high density, with 70 GHz at 1.25 T and lower density and with neutral beam injection. The H-phases display all characteristics known from tokamak H-modes including the development of an edge transport barrier, an increase of the poloidal impurity flow velocity at the edge, the reduction of edge turbulence and ELMs. The power threshold for the H-mode seems to be lower than that in tokamaks and is in agreement with an neBT scaling. Major differences to the divertor H-mode is the small increase in energy content of maximally 30%, the lack of a strong isotope effect both in threshold and in H-mode characteristics and a peculiarly narrow operational range in iota.

Density Limit Investigations on ASDEX

Density limit investigations on ASDEX have been performed under a variety of conditions: ohmically heated and neutral injection heated plasmas in H2, D2 and He have been studied in different divertor configurations, after various wall coating procedures, with gas puff and pellet fuelling, and in different confinement regimes with their characteristically different density profiles. A detailed description of the parametric dependence of the density limit, which in all cases is a disruptive limit, is given. This limit is shown to be a limit to the density at the plasma edge. Therefore, the highest densities corresponding to neRqa/Bt>30*1019 m-2.T-1 are obtained with centrally peaked ne profiles. Radiation from the main plasma at the density limit is always significantly below the total input power. The plasma disruption is due to an m=2 instability which for medium and high qa is preceded by one or more minor disruptions. In this range of qa, the disruptive instability is initiated by the occurrence of a Marfe on the high field side as a consequence of strong plasma cooling in this region. The duration of the Marfe increases with increasing distance between the plasma edge and the q=2 surface. After penetrating onto closed flux surfaces the Marfe leads to a current contraction and a subsequent destabilization of the m = 2 mode. In helium plasmas a strongly radiating, poloidally symmetric shell is observed before the density limit instead of a Marfe. An instantaneous destabilization of this mode is observed at low qa. Detailed measurements of plasma edge and divertor parameters close to the density limit indicate the development of a cold, dense divertor plasma before the disruption. Models describing the scrape-off layer and the divertor region predict an upper limit to the edge density at low divertor temperatures according to power balance considerations. Their relations to the experimental findings, especially the low field side cooling, ar

Diagnostic strategy of the W7-X stellarator

The Wendelstein 7-X experiment is a concept test for properties of reactor relevant plasmas in advanced stellarators. Prominent features include a modular superconducting coil assembly, a fivefold toroidal symmetry, and a helical magnetic axis. Due to the optimization process, W7-X is characterized by a vacuum magnetic field configuration with smooth magnetic surfaces, improved equilibrium properties with a weak dependence of rotational transform and shear on the plasma pressure β, good magneto-hydrodynamic stability properties due to magnetic well stabilization, reduced neoclassical transport losses and negligible bootstrap current in the long mean-free-path regime, good collisionless α-particle confinement in an equivalent reactor, and, as a technical aspect, good feasibility of the superconducting modular coils. W7-X will be heated by continuous electron cyclotron resonance heating and pulsed neutral beam injection and ion cyclotron resonance heating. The envisaged parameters are Te⩽10 keV, Ti⩽6 keV ce...

Heating and Confinement in the Ion Cyclotron Range of Frequencies on the Divertor Tokamak ASDEX

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.

Confinement regime transitions in ASDEX

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.

Boronization of ASDEX

Abstract 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

Impurity Accumulation in Plasma Regimes with High Energy Confinement

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.

Long-Pulse Heating of ASDEX Plasmas

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.