Plasma potential in toroidal devices: T-10, TJ-II, CHS and LHD

Abstract

Direct measurements of the electric potential ?, using the Heavy Ion Beam Probing, have been undertaken in the T-10 tokamak, and in the helical devices with various magnetic topology: TJ-II, CHS and LHD. L-mode plasmas were considered. Despite the large differences in machine sizes, heating methods and the topology of the magnetic field, the observed ? shows the striking similarities: (i) Similar magnitudes of Er; (ii) For low densities, ne < 0.5 ×1019 m-3, ? is positive, and an increase in ne is associated with the decrease of positive ? and formation of a negative Er; (iii) For higher densities, ne > (0.5-1)×1019 m-3, both ? and Er tends to be negative despite the use of different heating methods: Ohmic and ECR heating in T-10, ECRH and/or NBI in TJ-II, CHS and LHD; (iv) Application of ECRH, causing a rise in Te, results in more positive values for ? and Er. The analysis show that the main features of the ? dependences on the ne and Te agree with neoclassical predictions on the four devices within experimental and simulation precisions. new winding law for the continuous helical coils is proposed for Large Helical Device (LHD) type fusion reactors to satisfy the requirements for a wide blanket space and large plasma volume. Helical coils wound along the geodesic line of a torus with an elongated cross section can produce a magnetic configuration having a D-shaped magnetic surface with a magnetic well in the core region and high magnetic shear in the peripheral regions. The DT alpha particle confinement performance is greatly improved by increasing the elongation factor ? of the cross section of the winding frame for the helical coils. The results suggest that a smaller LHD-type fusion reactor can be realized. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.