New approaches for magnetic diagnostic of future fusion devices

New approaches for magnetic diagnostic of future fusion devicesI. Ďuran1, S. Entler1, K. Kovařík1, A. Torres1,11, P. Turjanica2, J. Reboun2, L. Viererbl3, D. Najman1,4, M. Kočan5, G. Vayakis5, K. Výborný6, Z. Šobáň6, M. Kohout6, V. Mortet6, A. Taylor6, P. Moreau7, F.P. Pellissier7, A. Le-Luyer7, P. Spuig7, W. Biel8, T. Franke9,101Institute of Plasma Physics of the CAS, Praha, Czech Republic2Faculty of Electrical Engineering, University of West Bohemia, Plzeň, Czech Republic3Research Centre Rez, 250 68 Husinec-Řež, Czech Republic4Czech Technical University in Prague, Praha, Czech Republic5ITER Organization, St Paul Lez Durance Cedex, France6Institute of Physics of the CAS, Praha, Czech Republic7IRFM, CEA, F-13108 Saint Paul lez Durance, France8Institut für Energie und Klimaforschung, Forschungszentrum Jülich GmbH, Germany9EUROfusion Power Plant Physics and Technology (PPPT) department, Garching, Germany10Max-Planck-Institut für Plasmaphysik, Garching, Germany11Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, Lisbon, PortugalThis contribution will review main challenges associated with the deployment of magnetic diagnostic for the EU DEMO fusion reactor, starting from return of experience from ITER, particularly in terms of operational temperature, neutron loads, required accuracy, and limited maintainability. New approaches to local magnetic field sensors design and construction will be introduced, namely metal-ceramic Hall sensors and inductive sensors manufactured using Thick Printed Copper (TPC) technology. Key lessons learned from the process of development, manufacturing, and calibration of ITER outer vessel steady state magnetic diagnostic based on bismuth Hall sensors will be outlined. The development of steady state magnetic sensors for a DEMO reactor is an even more challenging task compared to the ITER sensor system primarily due to approximately two orders of magnitude higher life time neutron fluence at envisaged sensor locations and also due to the higher operational temperature of the sensors. An outlook on some of the perspective design solutions for steady state magnetic sensors tackling the more demanding DEMO requirements will be given. Special attention will be paid to the design of the Hall sensors control electronics, and advanced methods of signal detection which are essential to ensure accurate detection of inherently low output voltages of Hall sensors in the noisy environment of a fusion reactor. Regarding inductive sensors, TPC technology will be introduced as an alternative to Low Temperature Co-fired Ceramics (LTCC) technology for the manufacturing of inductive local magnetic field sensors for DEMO and for their proof-of-principle application in COMPASS-U tokamak. Finally, potential synergy between steady state and traditional inductive approaches of local magnetic field measurements will be highlighted

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

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