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1 Author(s)
Morris, A.W. ; Culham Sci. Centre, EURATOM/CCFE Fusion Assoc., Abingdon, UK

MAST, alongside other spherical Tokamaks (STs), provides new perspectives on Tokamak physics for ITER and beyond and a platform to explore the potential of the ST as the core of a Component Test Facility (CTF) to test and develop components and technology for fusion power plants. MAST is one of the two largest STs in the world, the other being NSTX (PPPL, U.S.). Recent physics results include new measurements of the edge pedestal, the use of magnetic perturbations to influence edge-localized modes, pellet fueling, H-mode access, core transport, confinement scaling, fast-particle physics, and MHD instabilities. The program makes use of a set of unusually high resolution and wide-viewing diagnostics on MAST, since, in most experiments, the focus is on understanding the mechanisms, to support predictive modeling and comparison with other types of Tokamak. The first stage of a major upgrade is under way which will raise the heating power and toroidal field by 50% and almost double the inductive flux swing to allow long pulses and sustained noninductively driven plasmas. In particular, 17 poloidal field coils and an in-vessel cryopump will be added to allow an expanded divertor based on the “super-X” concept, as well as a conventional configuration. These upgrades will transform the operating space, pulselength, and flexibility of MAST for goal-oriented research and as a user facility. The heating and pulselength upgrades are aimed at testing off-axis current drive and fast-particle effects (important for sustained Tokamak fusion plasmas) and developing steady-state scenarios for a CTF. The divertor development is part of a growing international effort to find solutions for the power exhaust which ease the technology and material advances needed for the divertor plasma-facing components in a fusion power plant.

Published in:

Plasma Science, IEEE Transactions on  (Volume:40 ,  Issue: 3 )

Date of Publication:

March 2012

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