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ITER-like geometry with a single-null-divertor (Fig. 1)
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Clear high confinement mode (H-mode) in ohmic regime
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Unique set of saddle coils for the resonant magnetic perturbation.
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Flexible neutral beam injection heating system (NBI) for ion heating
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Comprehensive set of diagnostics focused on the edge plasma
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High flexibility in tokamak operation and planning
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Low-cost fusion experiment
Model of the COMPASS vacuum vessel, toroidal and poloidal coils
Fig. 1: Comparison of the plasma cross-section of tokamaks with ITER-relevant plasma shapes (left) and model of the COMPASS vacuum vessel, toroidal and poloidal coils (right)
