J. Kisslinger

Physics and engineering design for Wendelstein VII-X

AbstractThe future experiment Wendelstein VII-X (W VII-X) is being developed at the Max-Planck-Institut fur Plasmaphysik. A Helical Advanced Stellarator (Helias) configuration has been chosen because of its confinement and stability properties. The goals of W VII-X are to continue the development of the modular stellarator, to demonstrate the reactor capability of this stellarator line, and to achieve quasi-steady-state operation in a temperature regime >5 keV. This temperature regime can be reached in W VII-X if neoclassical transport plus the anomalous transport found in W VII-A prevail. A heating power of 20 MW will be applied to reach the reactor-relevant parameter regime.The magnetic field in W VII-X has five field periods. Other basic data are as follows: major radius R0 = 6.5 m, magnetic induction B0 = 3 T, stored magnetic energy W ≈ 0.88 GJ, and average plasma radius a = 0.65 m. Superconducting coils are favored because of their steady-state field, but pulsed water-cooled copper coils are also bei...

First island divertor experiments on the W7-AS stellarator

1. Abstract In the past, under limiter conditions, it has been impossible to produce high-power, highdensity, quasi-stationary neutral beam injection (NBI) discharges in W7-AS. Such discharges tended to evince impurity accumulation, lack of density control and subsequent radiation collapse (Normal Confinement). Presently, W7-AS is operating with a modular, open island divertor similar to that foreseen for W7-X. The divertor enables access to a new NBI heated, high density (ne up to 4·10 20 m -3 ) operating regime (High Density H-mode). It is extant above a threshold density, and is characterized by flat density profiles, high energyand low impurity confinement times and edge-localized radiation. The HDH-mode shows strong similarity to ELM-free H-mode scenarios previously observed in W7-AS, but in contrast to these avoids impurity accumulation. These new features enable full density control and quasi steady-state operation over many confinement times (at present only technically limited by the availability of NBI) also under conditions of partial detachment from the divertor targets. In HDH-mode, even in attached discharges, the divertor target load is considerable reduced. This is mainly due to favourable upstream conditions (higher nes), edge localized radiation and increased power deposition width. The benefits of the HDH-mode do not restrict only to hydrogen plasmas. They also occur ‐ albeit in a modified manner ‐ in deuterium plasmas. Undoubtedly, there are clear isotope effects between hydrogen and deuterium discharges. The results obtained in W7-AS render good prospects for W7-X and support the island divertor concept as a serious candidate for devices with magnetic islands at the edge. 2. Results Fig. 1 summarizes the behaviour of the energy confinement time E =W/Pabs, the normalized radiated power Prad/Pabs, and separatrix density nes obtained from quasi-stationary discharges with Pabs=1.4 MW as a function of the line-averaged density ne. E-values in NC follow the scaling E ISS95 =0.26· a 0.4 ·Bt 0.83 ·a 2.21 ·R 0.65 ·ne 0.51 ·Pabs -0.59 , [2], whereas for the HDH-mode one finds E ~ 2· E ISS95 . P rad /P abs grows smoothly with ne until partial plasma detachment, where a jump in the normalized radiated power occurs. The separatrix density n es increases sharply at the NC HDH-mode transition point, then continues to climb with ne and saturates

3D Fluid Modelling of the Edge Plasma by Means of a Monte Carlo Technique

An extended version of the 3D Monte Carlo edge plasma transport code EMC3 including more complex physics is presented. The balance equations for mass, momentum and energies are formulated in a general conservation form suited for direct application of the Monte Carlo solving algorithm. The new extended version is applied to the planned W7-AS island divertor. First results of the investigations, which are focused on neutral pump-out efficiency, power load distribution on the plates and high recycling performance, are presented.

Transport in island divertors: physics, 3D modelling and comparison to first experiments on W7-AS

Basic plasma transport properties in island divertors are compared to those of standard tokamak divertors. A realistic plasma transport modelling of high-density discharges in island divertors has become possible by implementing a self-consistent treatment of impurity transport in the EMC3-EIRENE code. In contrast to standard tokamak divertors, the code predicts no high recycling prior to detachment, with the downstream density never exceeding the upstream density. This is mainly due to momentum losses arising from the cross-field transport associated with the specific island divertor geometry. This momentum loss is effective already at low densities, high temperatures and is responsible for the high upstream densities needed to achieve detachment. Numerical scans of carbon concentration for high-density plasma typically show first a smooth, then a sharp increase of the carbon radiation, the latter being accompanied by a sharp drop of the downstream temperature and density indicating detachment transition. The jumps of the radiation and temperature are due to a thermal instability associated with the form of the impurity cooling rate function and can be reproduced by a simple 1D radial energy model based on cross-field transport and impurity losses. This model is used as a guideline to illustrate and discuss the detachment physics in details, including detachment condition and thermal instability. Major EMC3-EIRENE code predictions have been verified by the first W7-AS divertor experiments. A comparison of calculations and measurements is presented for high-density, high-power W7-AS divertor discharges and the physics related to rollover and detachment is discussed in detail. The code has been recently extended to general SOL configurations with open islands and arbitrary ergodicity by using a new highly accurate field-line mapping technique. The method correctly reproduces flux surfaces and islands over a high number of toroidal field periods, thus ensuring a clear distinction between parallel and radial transport. The technique has been tested successfully on W7-AS, W7-X, LHD and TEXTOR DED, and first applied to solve the coupled heat conduction equations for a typical ergodic W7-AS configuration.

Physics of island divertors as highlighted by the example of W7-AS

Based on theoretical analysis, numerical simulations and experimental results, the paper outlines a self-consistent physics picture of the island divertor transport in W7-AS, as it emerges from the present understanding, documented over the past several years of theoretical and experimental research on the subject. Key function elements of a divertor, such as particle flux enhancement, neutral screening, impurity retention, thermal power removal via impurity line radiation and detachment, are examined for the island divertor and assessed with respect to tokamak divertors. The paper focuses on describing the global scrape-off layer (SOL) transport behaviour associated with the specific island topology and aims at illustrating the elementary differences and similarities in divertor physics between a tokamak and a typical helical device. Shown and analysed are also the correlation between the SOL and core plasma and the role of the island divertor for improving the global plasma performance. Discussion is mainly based on simple models and estimations, while three-dimensional modelling calculations serve only for control of self-consistency and for determining basic functional dependences not accessible otherwise. The island divertor physics is presented within a theoretical frame with most key issues, however, being related to experimental results.

Influence of magnetic field configurations on divertor plasma parameters in the W7-AS stellarator

Abstract The new island divertor in W7-AS enables quasi steady-state operation with NBI at very high density including scenarios with stable detachment from the targets. Experiments with reversed B-field indicate that the interaction zones on the targets are affected in first order by E×B drifts. Stable detachment is restricted to magnetic field configurations with sufficiently large separation between x-points and targets and not too small field line pitch inside the islands. It is always partial in the sense that it does not extend over the full target area. This inhomogeneity is ascribed to an in/out asymmetry of the electron temperature at the upstream separatrix position.

Physics of the geometry-related detachment stability in W7-AS

This paper presents a detailed analysis of the transport behaviour of the detached plasmas in W7-AS based on an extended numerical study using the EMC3-EIRENE code, aimed at understanding the underlying physics responsible for the geometry-dependent detachment stability observed in W7-AS island divertor experiments. Here, a stable detachment can only be established when the control coils are switched on to generate sufficiently large islands with relatively short connection lengths. Special attention will be paid to a discussion of the carbon radiation, location and dynamics of the radiation layer, the neutral screening efficiency specific to the island divertor geometry and its impact on the detachment stability. Based on the three-dimensional simulation results, a linear stability model is presented in order to obtain some insight into the mechanisms driving the instability. The radiation behaviour and the location and evolution of the radiation zone in the island divertor will be discussed with respect to those of tokamak-MARFEs.

Island divertor studies on W7-AS

Abstract Basic topological features of the island divertor concept for low shear stellarators are discussed with emphasis on the differences to tokamak divertors. Extensive measurements of the edge structures by two-dimensional plasma spectroscopy and by target calorimetry are in excellent agreement with predicted vacuum and equilibrium configurations, which are available up to central β values of ∼ 1%. For this β value the calculated field-line pitch inside the islands is twice that of the corresponding vacuum case. Video observations of the strike points indicate stability of the island structures for central β values up to ∼ 3.7%. The interpretation of the complex island divertor physics of W7-AS has become possible by the development of the three-dimensional plasma transport code EMC3 (Edge Monte Carlo 3D), which has been coupled self-consistently to the EIRENE neutral gas code. Analysis of high density NBI discharges gives strong indications of stable high recycling conditions for n e ≥ 10 20 m −3 . The observations are reproduced by the EMC3/EIRENE code and supported by calculations with the B2/EIRENE code adapted to W7-AS. Improvement of recycling, pumping and target load distribution is expected from the new optimized target plates and baffles to be installed in W7-AS.

A simple highly accurate field-line mapping technique for three-dimensional Monte Carlo modeling of plasma edge transport

The paper presents a new simple and accurate numerical field-line mapping technique providing a high-quality representation of field lines as required by a Monte Carlo modeling of plasma edge transport in the complex magnetic boundaries of three-dimensional (3D) toroidal fusion devices. Using a toroidal sequence of precomputed 3D finite flux-tube meshes, the method advances field lines through a simple bilinear, forward/backward symmetric interpolation at the interfaces between two adjacent flux tubes. It is a reversible field-line mapping (RFLM) algorithm ensuring a continuous and unique reconstruction of field lines at any point of the 3D boundary. The reversibility property has a strong impact on the efficiency of modeling the highly anisotropic plasma edge transport in general closed or open configurations of arbitrary ergodicity as it avoids artificial cross-field diffusion of the fast parallel transport. For stellarator-symmetric magnetic configurations, which are the standard case for stellarators,...

The Trim Coils for the Wendelstein 7-X Magnet System

The magnet system of the fusion experiment Wendelstein 7-X (W7-X) consists of superconducting as well as normal conducting coils. 50 non planar superconducting coils are forming the main field, 20 planar superconducting coils allow varying the shape of the plasma. Inside of the plasma vessel 10 normal conducting control coils will be placed to modify the strike points of the plasma at the divertor. In addition a set of five normal conducting trim coils has been designed to allow the correction of error fields and to increase the experimental flexibility. The coils will be placed at the outer surface of the outer vessel of W7-X. Four out of five coils have identical size and shape. They have dimensions of 3.5 × 3.3 meters with 48 turns and will be operated with currents of up to 1.8 kA. The other coil type has a smaller size of 2.8 × 2.2 meters, but a higher number of turns and a higher operating current of 1.95 kA. Both types of trim coils will be made of square copper hollow profile with an integrated cooling channel. Five independent power supplies will be used to energize the coils. The present concept is based on four-quadrant power supplies. The control system will allow the local control as well as the remote control of the five power supplies from an external control room.