Mappings of magnetic field lines in poloidal divertor tokamaks

Abstract

Another important concept to control the plasma edge in tokamaks is the socalled poloidal divertor tokamaks (see Wesson (2004)). The magnetic configuration of these tokamaks contains a magnetic surface (a magnetic separatrix) sharply separating closed field lines on nested magnetic surfaces from open field lines hitting the walls of a fusion device. It has one (or two) singular points, X-points, on the poloidal section where the poloidal components of the magnetic field are zeros. These configurations are schematically shown in Fig. 12.1): a) a so-called single-null poloidal divertor; b) a double-null poloidal divertor. Such configurations of the magnetic field are created by one or two external current coils parallel to the plasma current, respectively. Magnetic fusion devices with a poloidal divertor provide an improved energy confinement of the plasma and diverts particles and heat efficiently into divertor plates in a special volume, from where they are pumped away. The future International Thermonuclear Experimental Reactor, ITER, is designed as a poloidal divertor tokamak. © 2006 Springer.