Abstract
The characterisation of plasma current quench and understanding of its underlying physical processes play a crucial role when designing large fusion devices such as ITER. For the first time, an extensive analysis of the COMPASS tokamak disruption database is presented.
A unique set of magnetic diagnostics allows the investigation of local toroidal and poloidal vessel currents, including currents flowing along the open magnetic field lines from the plasma to the vacuum vessel (VV) (i.e., halo currents). Area-normalised current quench times are in agreement with the ITPA 1.67 ms m(-2) lower limit.
Extremely fast I-p quench rates of up to 0.6 MA ms(-1) are observed during runaway electron campaigns at COMPASS, which are under the ITPA lower limit. Vertical movement of the plasma column is accelerated by the I-p quench during major disruptions.
Toroidal vessel currents of around 2% - 4% of the predisruptive plasma current I-disr(p) I-p quench. Net poloidal eddy currents are obtained by Mirnov coils and diamagnetic loop, reaching 3% of I-disr(p) I-p.
Geometric features of the VV structure and in-vessel component positions on the poloidal vessel current measurements are discussed.