Validation of equilibrium tools on the COMPASS tokamak

SOFT 2014 conference, submitted to Fusion Engineering and DesignInternational audienceVarious MHD (magnetohydrodynamic) equilibrium tools, some of which being recently developed or considerably updated, are used on the COMPASS tokamak at IPP Prague. MHD equilibrium is a fundamental property of the tokamak plasma, whose knowledge is required for many diagnostics and modelling tools. Proper benchmarking and validation of equilibrium tools is thus key for interpreting and planning tokamak experiments. We present here benchmarks and comparisons to experimental data of the EFIT++ reconstruction code [L.C. Appel et al., EPS 2006, P2.184], the free-boundary equilibrium code FREEBIE [J.-F. Artaud, S.H. Kim, EPS 2012, P4.023], and a rapid plasma boundary reconstruction code VacTH [B. Faugeras et al., PPCF 56, 114010 (2014)]. We demonstrate that FREEBIE can calculate the equilibrium and corresponding poloidal field (PF) coils currents consistently with EFIT++ reconstructions from experimental data. Both EFIT++ and VacTH can reconstruct equilibria generated by FREEBIE from synthetic, optionally noisy diagnostic data. Hence, VacTH is suitable for real-time control. Optimum reconstruction parameters are estimated

Experimental and theoretical study of utilization of probe methods for plasma diagnostics

The ball-pen probe is a unique probe recently developed at the Institite of Plasma Physics in Prague. It has been designed for direct measurement of plasma potential at the CASTOR tokamak. Nowadays, it is used routinely at several tokamaks in Europe, and the first tests in low-temperature plasma have also already been performed. The aims of the thesis are primarily experimental. A ball-pen probe has been constructed from available materials, which is suitable for systematic measurements of radial profiles in the low-temperature plasma of a cylindrical magnetron. By means of comparison to other diagnostics, it was proved that ball- pen probe is able to directly measure plasma potential in a certain range of plasma parameters even though its current-voltage characteristic is not symmetric, which is in contradiction with the simplified theory for ball-pen probe. Powered by TCPDF (www.tcpdf.org

Měření potenciálu plazmatu pomocí ball-pen a Langmuirovy sondy

The ball-pen probe is a unique probe recently developed at the Institite of Plasma Physics in Prague. It has been designed for direct measurement of plasma potential at the CASTOR tokamak. It has also been succesfully tested on several other high-temperature plasma devices in Europe. The aims of the bachelor work are primarily experimental. A ball-pen probe has been constructed from available materials, which is suitable for measurement in the low-temperature plasma of a cylindrical magnetron. Although its parameters are much different from those in high-temperature plasma devices, the main principle of measurement with ball-pen probe has been proven to apply also in this brand-new conditions

Měření potenciálu plazmatu pomocí ball-pen a Langmuirovy sondy

The ball-pen probe is a unique probe recently developed at the Institite of Plasma Physics in Prague. It has been designed for direct measurement of plasma potential at the CASTOR tokamak. It has also been succesfully tested on several other high-temperature plasma devices in Europe. The aims of the bachelor work are primarily experimental. A ball-pen probe has been constructed from available materials, which is suitable for measurement in the low-temperature plasma of a cylindrical magnetron. Although its parameters are much different from those in high-temperature plasma devices, the main principle of measurement with ball-pen probe has been proven to apply also in this brand-new conditions

Edge plasma study using a fast VIS camera in the COMPASS tokamak

International audienc

Overview of the COMPASS results

COMPASS addressed several physical processes that may explain the behaviour of important phenomena. This paper presents results related to the main fields of COMPASS research obtained in the recent two years, including studies of turbulence, L–H transition, plasma material interaction, runaway electron, and disruption physics: • Tomographic reconstruction of the edge/SOL turbulence observed by a fast visible camera allowed to visualize turbulent structures without perturbing the plasma. • Dependence of the power threshold on the X-point height was studied and related role of radial electric field in the edge/SOL plasma was identified. • The effect of high-field-side error fields on the L–H transition was investigated in order to assess the influence of the central solenoid misalignment and the possibility to compensate these error fields by low-field-side coils. • Results of fast measurements of electron temperature during ELMs show the ELM peak values at the divertor are around 80% of the initial temperature at the pedestal. • Liquid metals were used for the first time as plasma facing material in ELMy H-mode in the tokamak divertor. Good power handling capability was observed for heat fluxes up to 12 MW m−2 and no direct droplet ejection was observed. • Partial detachment regime was achieved by impurity seeding in the divertor. The evolution of the heat flux footprint at the outer target was studied. • Runaway electrons were studied using new unique systems—impact calorimetry, carbon pellet injection technique, wide variety of magnetic perturbations. Radial feedback control was imposed on the beam. • Forces during plasma disruptions were monitored by a number of new diagnostics for vacuum vessel (VV) motion in order to contribute to the scaling laws of sideways disruption forces for ITER. • Current flows towards the divertor tiles, incl. possible short-circuiting through PFCs, were investigated during the VDE experiments. The results support ATEC model and improve understanding of disruption loads

Preliminary design of the COMPASS upgrade tokamak

COMPASS Upgrade is a new medium size, high magnetic field tokamak (R = 0.9 m, Bt = 5 T, Ip = 2 MA) currently under design in the Czech Republic. It will provide unique capabilities for addressing some of the key challenges in plasma exhaust physics, advanced confinement modes and advanced plasma configurations as well as testing new plasma facing materials and liquid metal divertor concepts. This paper contains an overview of the preliminary engineering design of the main systems of the COMPASS Upgrade tokamak (vacuum vessel, central solenoid and poloidal field coils, toroidal field coils, support structure, cryostat, cryogenic system, power supply system and machine monitoring and protection system). The description of foreseen auxiliary plasma heating systems and plasma diagnostics is also provided as well as a summary of expected plasma performance and available plasma configurations