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First, it is shown that during electromagnetic transients in COMPASS-U the poloidal field coils must drain sizeably the current from the vessel and, therefore, reduce disruption forces and their duration. Next, the role of poloidal eddy current (which is absent in some approaches) in the dynamics of vertical and radial forces is found to be essential. Finally, to verify the CarMa0NL modelling for COMPASS-U, the numerical results are cross-validated with general analytical predictions (Pustovitov 2015 Nucl. Fusion 55 113032): the computed vertical force on the tokamak wall is found to be almost zero during rapid (jump-like) transients, as it should be because of strong skin-effect. This test proves the credibility of the simulation model and computational realization. [ABSTRACT FROM AUTHOR]

The Thomson scattering (TS) diagnostic, one of the key diagnostics used on the tokamaks around the world, is planned for the COMPASS-U tokamak, which is recently under design and construction in the Institute of Plasma Physics in Prague, Czech Republic. This tokamak is supposed to be a world-unique, high magnetic field device with hot walls, allowing for the study of the plasma exhaust in advanced operational scenarios and testing cutting-edge technologies relevant to future fusion reactors, e.g., use of liquid metals. The core and edge TS systems are planned to be designed and operational, with a limited performance, already in the early stage of the tokamak operation. In this contribution, requirements and the most important constraints defining the TS system design are presented. The impact of both the possible collection lens location and spatial resolution on the plasma pedestal observation is simulated. Design considerations also take into account the high-resolution TS core and edge systems available from the COMPASS tokamak, which will be reused. The collection lenses will be newly built. Extension of the detection system will complete the plasma radius coverage in the future. The divertor TS is considered for later periods. [ABSTRACT FROM AUTHOR]

Substantial power dissipation in the edge plasma is required for the safe operation of ITER and next-step fusion reactors, otherwise unmitigated heat fluxes at the divertor plasma-facing components (PFCs) would easily exceed their material limits. Traditionally, such heat flux mitigation is linked to the regime of detachment, which is characterised by a significant pressure gradient between upstream and downstream scrape-off layer (SOL). However, the physics phenomena responsible for power dissipation and pressure loss are distinctly different, especially when the power dissipation is achieved by impurity seeding. In principle, it is possible to achieve substantial mitigation of the heat fluxes while maintaining conservation of the pressure along the open field lines in the SOL. This regime can be accessed by injection of medium- or high-Z impurities, which mostly radiate inside the last closed flux surface. The critical question related to such an approach is the effect on confinement and perspective fusion power generation in future thermonuclear reactors. In this work, we report on experiments at COMPASS tokamak, where neon and argon impurities were injected in ohmic or NBI-heated low confinement plasmas. With appropriate seeding waveform, stable scenarios were achieved, avoiding the radiative collapse of plasmas. Significant reduction of heat fluxes at the outer target was observed, with heat flux pattern similar to the one previously achieved by nitrogen seeding. The reduction of downstream pressure was, however, accompanied by an equal reduction of upstream pressure, indicating that the power dissipation occurred inside the separatrix. Indeed, the impurity cooling is causing a significant drop of edge temperature; however, the effect in the plasma centre is much less pronounced. [ABSTRACT FROM AUTHOR]

COMPASS upgrade (COMPASS-U) is a high-magnetic field, medium-sized tokamak with high-temperature (< 500 °C) operation. The scientific program is aimed to address the topics of plasma exhaust, liquid metals, enhanced confinement modes, and edge plasma physics. The plasma current is up to 2 MA and the toroidal magnetic field is up to 5 T. Therefore, plasma disruptions can produce large electromagnetic forces on the vacuum vessel (VV) and other conducting structures. This article presents the study of different VV design concepts, which were considered during the COMPASS-U conceptual design phase. This article describes the electromagnetic forces exerted during plasma disruption, high-temperature operation requirements, and other design constraints. INCONEL 625 is selected as a reference material for VVs. FE simulations of the 45° sector of the vacuum vessel were carried out for various load combinations. This article summarizes the wall thickness optimization that limits the vessel deformation and minimizes the stress in the shell. The results are broken down into different categories of stress according to the ASME code and compared with material limits. [ABSTRACT FROM AUTHOR]

Abstract: The COMPASS tokamak, a divertor device with clear H-mode and ITER-relevant geometry, has been re-installed in IPP Prague after its transport from CCFE in UK. The first plasma was achieved in December 2008. Many new diagnostic tools with both high temporal and spatial resolutions have been developed to address the scientific programme of COMPASS focused on H-mode physics and pedestal investigations. In the paper, an overview of existing and in near future planned diagnostics (magnetic, spectroscopy, microwave, probe and beam/particle) on COMPASS is given including their basic technical specifications and achievements. [Copyright &y& Elsevier]

COMPASS Upgrade is a new medium size, high magnetic field tokamak (R = 0.9 m, B t = 5 T, I p = 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. [ABSTRACT FROM AUTHOR]

• Main constraints influencing the conceptual design of diagnostics for COMPASS-U were identified. • Diagnostic solutions for the hot vacuum vessel, high plasma densities and high heat flux densities are proposed. • Spatial constraints at COMPASS-U are reviewed and possible solutions indicated. COMPASS-U, a high magnetic field tokamak with hot walls, will be designed and built at IPP Prague. Unique features of this new device bring noticeable constraints and requirements on plasma diagnostics, which make their development highly demanding. In this paper, the main expected constraints influencing the conceptual design of diagnostic tools for COMPASS-U (high temperature of the vacuum vessel, high plasma density, high heat flux density, strong auxiliary plasma heating, spatial constraints, liquid metals in the divertor) are reviewed and possible solutions are indicated. [ABSTRACT FROM AUTHOR]

This paper contains the description of the generalized Sequential Probability Ratio Test (gSPRT) method used for automatic ELM detection in different diagnostic signals collected on the COMPASS tokamak. After determination of H-mode region based on D α signal in a given shot, ELMs are automatically detected in different diagnostic time traces (e.g. magnetic signals and also in D α signals). The onset time, the maximum location and the peak value, and the duration of each detected ELM is determined. Analyzed diagnostic signals came from different radial positions of the studied plasma volume, thus from arrival times of given ELM into different detectors, the average radial propagation velocity of the ELM event can be also estimated. Comparison of results from gSPRT method with two different, commonly used automatic ELM detection method (threshold technique and correlation based technique) is also reported. [ABSTRACT FROM AUTHOR]

This article presents a basis for the material selection of toroidal field coils of COMPASS-U, a medium size tokamak currently under construction. The main focus is given to a 0D model which produces temporal evolution of toroidal field coil parameters (e.g. voltage, temperature, energy) during a plasma discharge based on the specification of the coils, materials and power supply system. Linkage of these parameters, together with the design requirements of the tokamak, allows the reduction of design parameters space. The model was used for the selection of copper alloy candidate materials for the coils. Mechanical and electrical material tests and results from the model used to select material candidates are presented. [Display omitted] • Toroidal field coils provide 5 T at a major radius of 0.894 m for at least 1 s. • 0D numerical model was used in material selection process. • Electrical and mechanical tests of materials at room and cryogenic temperatures. [ABSTRACT FROM AUTHOR]

O-mode frequency-modulated continuous wave (FMCW) reflectometry provides an alternative to magnetic measurements in the determination of the plasma separatrix position for plasma position control. This type of measurement proves to be particularly attractive for the control of future fusion reactors where the harsh radiation environment may damage magnetic probes or induce non-compensable measurement drifts. Plasma position reflectometry (PPR), first demonstrated in ASDEX-Upgrade, is a control technique that is increasingly important to validate in diversified experimental devices and relevant plasma regimes. The COMPASS tokamak provides suitable conditions for such advanced demonstrations and regular PPR operation and development, thanks to its O-mode reflectometer and Multi-Threaded Application Real-Time executor (MARTe) based real-time control system. Herein we present the integration of a PPR system on COMPASS, both at hardware and software levels. Reflectometry swept measurements require signals to be acquired in bursts of data and streamed to the corresponding MARTe-PPR node through PCIe ® fibre-optic links. The data transferred in real-time is used to reconstruct the radial density profiles from which the outer separatrix position is estimated. This estimate is then delivered to the central MARTe controller node via a dedicated Xilinx ® Aurora ® link at a rate matching COMPASS's 500 μ s slow control cycle. The implemented system systematically met the required latency specifications, being able to deliver an estimation of the plasma radial position capable of successfully replacing the corresponding magnetic measurements in the plasma position feedback control loops. [ABSTRACT FROM AUTHOR]