Conceptual design of a component test facility based on the spherical tokamak

Abstract

A small steady state spherical tokamak (ST) offers an attractive system for producing simultaneously the neutron, particle and heat fluxes necessary to effectively test and optimise blanket modules, first wall structures and other components under the required fusion power plant conditions. This component test facility (CTF) would complement and extend the qualification of materials by IFMIF and could operate in association with DEMO thus reducing the risk of delays, and extending the options, during this crucial stage of the development of commercial fusion power. The ST-CTF offers many advantages including low tritium consumption, ease of maintenance and a compact assembly and would operate in a strongly driven mode in which Q ∼ 1. The current drive would be provided by a mix of bootstrap current and neutral beam injection systems. The blanket modules under test are removed and replaced using a casking system and the entire centre column assembly can be relatively easily removed via a hydraulic lift system beneath the tokamak assembly. The single turn toroidal field coil has a water-cooled copper centre rod with multiple return limbs, which requires a low voltage, high current power supply. The poloidal field coils are also water-cooled but use a glass fibre reinforced cyanate ester resin insulation that offers higher radiation resistance and higher operating temperatures than the conventional epoxy resin systems. When operated in L-mode most of the exhaust power is directed to the outer legs of the double null divertor configuration where high power densities and high material erosion rates are developed. A novel divertor target based on the use of a cascading flow of pebbles is one option being developed for this application.