Divertor detachment is explored on the TCV tokamak in alternative magnetic geometries. Starting from typical TCV single-null shapes, the poloidal flux expansion at the outer strikepoint is varied by a factor of 10 to investigate the X-divertor characteristics, and the total flux expansion is varied by 70% to study the properties of the super-X divertor. The effect of an additional X-point near the target is investigated in X-point target divertors. Detachment of the outer target is studied in these plasmas during Ohmic density ramps and with the ion ∇B drift away from the primary X-point. The detachment threshold, depth of detachment, and the stability of the radiation location are investigated using target measurements from the wall-embedded Langmuir probes and two-dimensional CIII line emissivity profiles across the divertor region, obtained from inverted, toroidally-integrated camera data. It is found that increasing poloidal flux expansion results in a deeper detachment for a given line-averaged density and a reduction in the radiation location sensitivity to core density, while no large effect on the detachment threshold is observed. The total flux expansion, contrary to expectations, does not show a significant influence on any detachment characteristics in these experiments. In X-point target geometries, no evidence is found for a reduced detachment threshold despite a Nuclear Fusion Results from recent detachment experiments in alternative divertor configurations on TCVInternational Atomic Energy Agency a See the author list of 'Overview of progress in European Medium Sized Tokamaks towards an integrated plasma-edge/wall solution' by H. Meyer et al, to be published in the Nuclear Fusion

Transport analyses using first-principle turbulence codes and 112-D transport codes usually study radial transport properties between the tokamak plasma magnetic axis and a normalized minor radius around 0.8. In this region, heat transport shows significantly stiff properties resulting in temperature scalelength values (R∕LT) that are relatively independent of the level of the radial heat flux. We have studied experimentally in the tokamak à configuration variable [F. Hofmann et al., Plasma Phys. Controlled Fusion 36, B277 (1994)] the radial electron transport properties of the edge region, close to the last closed flux surface, namely, between ρV=V/Vedge=0.8 to 1. It is shown that electron transport is not stiff in this region and high R∕LTe values (∼20–40) can be attained even for L-mode confinement. We can define a “pedestal” location, already in L-mode regimes, where the transport characteristics change from constant logarithmic gradient, inside ρV = 0.8, to constant gradient between 0.8 and 1.0. In particular, we demonstrate, with well resolved Te and ne profiles, that the confinement improvement with plasma current Ip, with or without auxiliary heating, is due to this non-stiff edge region. This new result is used to explain the significant confinement improvement observed with negative triangularity, which could not be explained by theory to date. Preliminary local gyrokinetic simulations are now consistent with an edge, less stiff, region that is more sensitive to triangularity than further inside. We also show that increasing the electron cyclotron heating power increases the edge temperature inverse scalelength, in contrast to the value in the main plasma region. The dependence of confinement on density in ohmic plasmas is also studied and brings new insight in the understanding of the transition between linear and saturated confinement regimes, as well as of the density limit and appearance of a 2/1 tearing mode. The results presented in this paper provide an important new perspective with regards to radial transport in tokamak plasmas which goes beyond L-mode plasmas and explains some previous puzzling results. It is proposed that understanding the transport properties in this edge non-stiff region will also help in understanding the improved and high confinement edge properties.

Moderately peaked electron density profiles are observed in virtually all plasma conditions in TCV. The existence of an anomalous pinch is unambiguously demonstrated by the observation of peaked density profiles in stationary, fully relaxed, fully current driven electron cyclotron current drive (ECCD) discharges with V loop = 0. The behaviour of the density profiles from a database of 300 Ohmic L-and H-mode, as well as electron cyclotron heating and ECCD discharges, is compared to predictions of models based on the Ware pinch, the curvature pinch and anomalous thermodiffusion. Best overall agreement throughout the database is obtained with models combining an anomalous pinch mechanism, such as the curvature pinch, with the Ware pinch.

A large database of reciprocating probe data from the edge plasma of TCV (Tokamak à Configuration Variable) is used to test the radial velocity scalings of filaments from analytical theory [J. R. Myra, D. A. Russell, and D. A. D'Ippolito, Phys. Plasmas 13, 112502 (2006)]. The measured velocities are mainly scattered between zero and a maximum velocity which varies as a function of size and collisionality in agreement with the analytical scalings. The scatter is consistent with mechanisms that tend to slow the velocity of individual filaments. While the radial velocities were mainly clustered between 0.5 and 2 km/s, a minority reached outward velocities as high as 5km/s or inward velocities as high as-4km/s. Inward moving filaments are only observed in regions of high poloidal velocity shear in discharges with Bx∇B away from the X-point, a new finding. The filaments have diameters clustered between 3 and 11mm, and normalized sizes â clustered between 0.3 and 1.1, such that most filaments populate the resistive-ballooning regime, therefore, most filaments in TCV have radial velocities with little or no dependence on collisionality. Improvements in crosscorrelation techniques and conditional averaging techniques are discussed which reduce the sizes determined for the largest filaments, including those larger than the scrape-off layer (SOL).

The extreme shaping capabilities of the TCV tokamak have been used to investigate the effect of the plasma geometry on the confinement of non recycling trace impurities injected by means of the laser blow-off technique. The progression of the injected Silicon in the core of