The design study of PROTO-SPHERA, a novel compact torus configuration, has been completed. It is composed of a spherical torus (ST) (with closed flux surfaces) and a force-free screw pinch (SP) (with open flux surfaces and fed by electrodes). PROTO-SPHERA is formed at spherical-tokamak-like densities (∼10 19 m −3) with low voltage (∼200 V) between the electrodes. The idea of replacing the metal centrepost current (I tf) of the spherical tokamaks with the SP plasma electrode current (I e) is aimed mainly at getting rid of the rod at the centre of the plasma configuration, which is the most critical component of spherical tokamak design. As a consequence it should be possible to decrease the aspect ratio A = R/a (R = ST major radius, a = ST minor radius) in the course of experiment and to increase the ratio between the toroidal plasma current (I ST) and the plasma electrode current, I ST /I e 1. Matching two plasma configurations, i.e. an open flux-surface SP and a closed flux-surface ST, brings to life several radically new issues. The purpose of this paper is to analyse the equilibrium, the ideal MHD stability and the formations and modelling issues of such a combined magnetic confinement system. The MULTI-PINCH experimental setup, which is being assembled inside the START vacuum vessel (now in Frascati), will represent the first phase of PROTO-SPHERA: its goal is to prove the feasibility of a stable disc-shaped SP around the electrodes.
Helium-cooled divertor concepts are considered suitable for use in fusion power plants for safety reasons, as they enable the use of a coolant compatible with any blanket concept, since water would not be acceptable, e.g. in connection with ceramic breeder blankets using large amounts of beryllium. Moreover, they allow for a high coolant exit temperature for increasing the efficiency of the power conversion system. Within the framework of the European power plant conceptual study, different helium-cooled divertor concepts based on different heat transfer mechanisms are being investigated at ENEA Frascati, Italy, and Forschungszentrum Karlsruhe, Germany. They are based on a modular design which helps reduce thermal stresses. The design goal is to withstand a high heat flux of about 10–15 MW m−2, a value which is considered relevant to future fusion power plants to be built after ITER. The development and optimization of the divertor concepts require an iterative design approach with analyses, studies of materials and fabrication technologies and the execution of experiments. These issues and the state of the art of divertor development shall be the subject of this report.
This paperpresents an overview of results of the 1994/95 experimental campaign on JET with the new pumped divertor and draws implications for ITERin the areas of detached and radiative divertorplasmas, theuseofberyllium as a divertor target tile malerial, the confmement properties of discharges with the same dimensionless parameters (exceptforthedimensionless Larmorradius) as lT!3R and the effect of varying the toroidal magnetic field ripple in the FTER relevant range.Discbarges withhigh fusionpe~ormance athighcurrentjn steadystate with ELMS and in the ELM-free hot -ion H-mode, are also reported. Limits to operations are discussed and projections to D-T performance are made.
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