Community Research and Development Information Service - CORDIS


The most challenging issue in the design of future fusion experiments is a divertor concept which simultaneously provides tolerable power loading at the target plates and sufficiently large pumping rates. With respect to the target power load a reduction by at least one order of magnitude seems to be required. If this is to be achieved by upstream losses, one possibility is the proposed cold gas divertor concept, in which the energy is transferred mainly to the divertor side walls by CX-neutrals before reaching the target plate. This report discusses exploratory 2D simulations of divertors in present and future tokamaks (such as ITER). The main focus is on neutral gas effects rather than impurity radiation.

Additional information

Authors: SCHNEIDER R ET AL., Max-Planck-Institut für Plasmaphysik, Garching bei München (DE);REITER D, Forschungszentrum Jülich GmbH (KFA) (DE);BAELMANS M, Katholieke Universiteit Leuven, Heverlee (BE);BRAAMS B, New York University, Courant Institute (US)
Bibliographic Reference: Article: Proceedings of the 20th European Conference on Controlled Fusion and Plasma Physics, Vol. II (1993) pp. 775-778
Record Number: 199311156 / Last updated on: 1994-11-29
Original language: en
Available languages: en