Conceptual design of the COMPASS upgrade tokamak

Pánek, R.; Markovič, T.; Cahyna, P.; Dejarnac, R.; Havlíček, J.; Horáček, J.; Hron, M.; Imríšek, M.; Junek, P.; Komm, M.; Šesták, D.; Urbán, J.; Varju, J.; Weinzettl, V.; Adámek, Jiřı́; Bílková, P.; Böhm, P.; Dimitrova, M.; Hacek, P.; Kovařík, K.; Krbec, J.; Mlynář, J.; Podolník, A.; Seidl, J.; Ştöckel, J.; Tomeš, M.; Zajac, F.; Mitošínková, K.; Peterka, M.; Vondráček, P.

doi:10.1016/j.fusengdes.2017.03.002

Cited by 52 publications

(24 citation statements)

References 12 publications

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“…It is limited by the surface water boiling which was, by the way, just recently overcome 12 . Testing survival under those extreme conditions (mostly using liquid metal heat shield 13 ) will be one of two major goals of tokamak COMPASS-Upgrade 14 , 15 , capable of reaching 100 MW/m 16 . This is, for better imagination, comparable to the Sun surface 63 MW/m radiation.…”

Section: Thermonuclear Fusion Reduction Due To Elm-induced Tungsten D...mentioning

confidence: 99%

Novel concept suppressing plasma heat pulses in a tokamak by fast divertor sweeping

Horáček

Lukes

Adámek

et al. 2022

Sci Rep

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One of the remaining challenges in magnetic thermonuclear fusion is survival of the heat shield protecting the tokamak reactor vessel against excessive plasma heat fluxes. Unmitigated high confinement edge localized mode (ELM) is a regular heat pulse damaging the heat shield. We suggest a novel concept of magnetic sweeping of the plasma contact strike point fast and far enough in order to spread this heat pulse. We demonstrate feasibility of a dedicated copper coil in a resonant circuit, including the induced currents and power electronics. We predict the DEMO ELM properties, simulate heat conduction, 3D particles motion and magnetic fields of the plasma and coil in COMSOL Multiphysics and Matlab. The dominant system parameter is voltage, feasible 18 kV yields 1 kHz sweeping frequency, suppressing the ELM-induced surface temperature rise by a factor of 3. Multiplied by other known mitigation concepts, ELMs might be mitigated enough to ensure safe operation of DEMO.

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Section: Thermonuclear Fusion Reduction Due To Elm-induced Tungsten D...mentioning

confidence: 99%

Novel concept suppressing plasma heat pulses in a tokamak by fast divertor sweeping

Horáček

Lukes

Adámek

et al. 2022

Sci Rep

Self Cite

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“…This simplified description is complicated in practice by plasma and/or impurity transport, plasma instabilities (ELMs, 9 ,p. 409] , sawtooth instabilities 9 ,p. 365] ), and plasma flows in the divertor region that these divertor legs occupy. Therefore, to support the DEMO divertor design, existing facilities are being upgraded 10 , and a new tokamak facility to study the plasma exhaust is under construction 11 , 12 .…”

Section: Introductionmentioning

confidence: 99%

Real-time feedback control of the impurity emission front in tokamak divertor plasmas

et al. 2021

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In magnetic confinement thermonuclear fusion the exhaust of heat and particles from the core remains a major challenge. Heat and particles leaving the core are transported via open magnetic field lines to a region of the reactor wall, called the divertor. Unabated, the heat and particle fluxes may become intolerable and damage the divertor. Controlled ‘plasma detachment’, a regime characterized by both a large reduction in plasma pressure and temperature at the divertor target, is required to reduce fluxes onto the divertor. Here we report a systematic approach towards achieving this critical need through feedback control of impurity emission front locations and its experimental demonstration. Our approach comprises a combination of real-time plasma diagnostic utilization, dynamic characterization of the plasma in proximity to the divertor, and efficient, reliable offline feedback controller design.

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“…( 29) with the proportionality constant given by the fitting procedure can be used to predict the SOL width for future tokamaks, such as ITER, COMPASS Upgrade, JT-60SA, and DTT. Considering the baseline scenario just before the L-H transition, one obtains λ q,th 3.5 mm for ITER (R 0 = 6.2 m, a = 2 m, q = 2, P SOL = 18 MW, κ = 1.4, ne = 4 • 10 19 m −3 , and B T = 5.3 T [78]), λ q,th 1.8 mm for COMPASS Upgrade (R 0 = 0.89 m, a = 0.27 m, q = 2.6, P SOL = 3.7 MW, κ = 1.8, ne = 2 • 10 20 m −3 , and B T = 5.0 T [79]), λ q,th 7.1 mm for JT-60SA (R 0 = 2.9 m, a = 1.2 m, q = 3, P SOL = 10 MW, κ = 1.9, ne = 6.3 • 10 19 m −3 , and B T = 2.3 T [80]), and λ q,th 3.0 mm for DTT (R 0 = 2.1 m, a = 0.6 m, q = 3, P SOL = 15 MW, κ = 1.7, ne = 1.8 • 10 20 m −3 , and B T = 6.0 T [81]). The theoretical value of λ q for ITER Lmode is within the range of values predicted by the experimental scaling laws derived in Ref.…”

Section: Near Solmentioning

confidence: 99%

Theory-based scaling laws of near and far scrape-off layer widths in single-null L-mode discharges

Giacomin,

Stagni,

Ricci

et al. 2021

Preprint

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Theory-based scaling laws of the near and far scrape-off layer (SOL) widths are analytically derived for L-mode diverted tokamak discharges by using a two-fluid model. The near SOL pressure and density decay lengths are obtained by leveraging a balance among the power source, perpendicular turbulent transport across the separatrix, and parallel losses at the vessel wall, while the far SOL pressure and density decay lengths are derived by using a model of intermittent transport mediated by filaments. The analytical estimates of the pressure decay length in the near SOL is then compared to the results of three-dimensional, flux-driven, global, two-fluid turbulence simulations of L-mode diverted tokamak plasmas, and validated against experimental measurements taken from an experimental multi-machine database of divertor heat flux profiles, showing in both cases a very good agreement. Analogously, the theoretical scaling law for the pressure decay length in the far SOL is compared to simulation results and to experimental measurements in TCV L-mode discharges, pointing out the need of a large multi-machine database for the far SOL decay lengths.

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Conceptual design of the COMPASS upgrade tokamak

Cited by 52 publications

References 12 publications

Novel concept suppressing plasma heat pulses in a tokamak by fast divertor sweeping

Novel concept suppressing plasma heat pulses in a tokamak by fast divertor sweeping

Real-time feedback control of the impurity emission front in tokamak divertor plasmas

Theory-based scaling laws of near and far scrape-off layer widths in single-null L-mode discharges

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