Topology of toroidal helical fields in non-circular cross-sectional tokamaks

The ordinary differential magnetic field line equations are solved numerically; the tokamak magnetic structure is studied on Hefei Tokamak-7 Upgrade (HT-7U) when the equilibrium field with a monotonic q-profile is perturbed by a helical magnetic field. We find that a single mode (m, n) helical perturbation can cause the formation of islands on rational surfaces with q = m/n and q = (m +/- 1, +/- 2, +/- 3,...)/n due to the toroidicity and plasma shape (i.e. elongation and triangularity), while there are many undestroyed magnetic surfaces called Kolmogorov-Arnold-Moser (KAM) barriers on irrational surfaces. The islands on the same rational surface do not have the same size. When the ratio between the perturbing magnetic field B-r(r) and the toroidal magnetic field amplitude B(phi)0 is large enough, the magnetic island chains on different rational surfaces will overlap and chaotic orbits appear in the overlapping area, and the magnetic field becomes stochastic. It is remarkable that the stochastic layer appears first in the plasma edge region.

Heating, current drive and confinement regimes with the JET ICRH and lhcd systems

During its 1990 operation, 2 large RF systems were available on JET. The Ion Cyclotron Resonance Heating (ICRH) system was equipped with new beryllium screens and with feedback matching systems. Specific impurities generated by ICRH were reduced to negligible levels even in the most stringent H-mode conditions. A maximum power of 22 MW was coupled to L-mode plasmas. High quality H-modes (tau-E greater-than-or-equal-to 2.5 tau-EG) were achieved using dipole phasing. A new high confinement mode was discovered. It combines the properties of the H-mode regime to the low central diffusivities obtained by pellet injection. A value of n(d) tau-E T(i) = 7.8 x 10(20) m-3 s keV was obtained in this mode with T(e) approximately T(i) approximately 11 keV. In the L-mode regime, a regime, a record (140 kW) D-He-3 fusion power was generated with 10 - 14 MW of ICRH at the He-3 cyclotron frequency. Experiments were performed with the prototype launcher of the Lower Hybrid Current Drive (LHCD) systems with coupled power up to 1.6 MW with current drive efficiencies up to < n(e) > R I(CD)/P = 0.4 x 10(20) m-2 A/W. Fast electrons are driven by LHCD to tail temperatures of 100 keV with a hollow radial profile. Paradoxically, LHCD induces central heating particularly in combination with ICRH. Finally we present the first observations of the synergistic acceleration of fast electrons by Transit Time Magnetic Pumping (TTMP) (from ICRH) and Electron Landau Damping (ELD) (from LHCD). The synergism generates TTMP current drive even without phasing the ICRH antennae.

Optimization of positions and currents of tokamak poloidal field coils using genetic algorithms

Plasma equilibrium geometry has a great influence on the confinement and magnetohydrodynamic stability in tokamaks. The poloidal field (PF) system of a tokamak should be optimized to support the prescribed plasma equilibrium geometry. In this paper, a genetic algorithm-based method is applied to solve the optimization of the positions and currents of tokamak PF coils. To achieve this goal, we first describe the free-boundary code EQT Based on the EQT code, a genetic algorithm-based method is introduced to the optimization. We apply this new method to the PF system design of the fusion-driven subcritical system and plasma equilibrium geometry optimization of the Experimental Advanced Superconducting Tokamak (EAST). The results indicate that the optimization of the plasma equilibrium geometry can be improved by using this method.