Divertor plasma characteristics in the Large Helical Device (LHD) have been investigated mainly by using Langmuir probes. The three-dimensional structure of the helical divertor, which is naturally produced in the heliotron-type magnetic configuration, is clearly seen in the measured particle and power deposition profiles on the divertor plates. These observations are consistent with the numerical results of field line tracing. The particle flux to the divertor plates increases almost linearly with the line averaged density. The high-recycling regime and divertor detachment, which are observed in tokamaks, have not been observed even during high density discharges with low input power. Both electron density and temperature decrease with increasing radius in the stochastic layer with open field lines, and at the divertor plate they become fairly low compared with those at the last closed flux surface. This means the reduction of pressure along the magnetic field lines occurs in the open field line region in LHD.

Large potential oscillations were detected in JIPPT-IIU tokamak plasmas in a wide range of plasma cross-sections in measurements using a multi-sample-volume heavy ion beam probe. These oscillations have large amplitudes reaching a few hundreds of volts and their frequencies are in the range of the geodesic acoustic mode (GAM). They are found over a wide range of plasma cross-sections and commonly have m = 0 structures. As they were Fourier analysed, it was found that the central frequency is higher in the core of the plasma and lower in the edge of the plasma. These observations agree with the properties of theoretically predicted GAM oscillations. It was also found that the frequency spectrum is peaked in the core and broad in the edge, which may have something to do with damping mechanisms of the GAM. The phase relation between the density and the electric field fluctuations was studied extensively in terms of the cross-correlation function. The level of the density fluctuation was low as it should be, and the expected 90° phase difference was found in a limited radial domain.

This article describes the design and performance of a multi-point ͑200͒ high repetition rate ͑4ϫ50 Hz͒ Thomson scattering diagnostic installed on the Large Helical Device. A unique feature of this system is its oblique back scattering configuration, which enables us to observe the entire plasma region along a major radius on the midplane under a severely restricted port constraint. High throughput collection optics using a mosaic mirror of 1.5 mϫ1.8 m area yield high quality data even with 0.5 J pulse energy delivered from 50 Hz repetition rate Nd: yttrium-aluminum-garnet lasers. High repetition and high spatial resolution ͑2-4 cm͒ of the system enable us to study island evolution in the plasma.

Extremely hollow profiles of impurities ͑denoted as "impurity hole"͒ are observed in the plasma with a steep gradient of the ion temperature after the formation of an internal transport barrier ͑ITB͒ in the ion temperature transport in the Large Helical Device ͓A. Iiyoshi et al., Nucl. Fusion 39, 1245 ͑1999͔͒. The radial profile of carbon becomes hollow during the ITB phase and the central carbon density keeps dropping and reaches 0.1%-0.3% of plasma density at the end of the ion ITB phase. The diffusion coefficient and the convective velocity of impurities are evaluated from the time evolution of carbon profiles assuming the diffusion and the convection velocity are constant in time after the formation of the ITB. The transport analysis gives a low diffusion of 0.1-0.2 m 2 / s and the outward convection velocity of ϳ1 m/ s at half of the minor radius, which is in contrast to the tendency in tokamak plasmas for the impurity density to increase due to an inward convection and low diffusion in the ITB region. The outward convection is considered to be driven by turbulence because the sign of the convection velocity contradicts the neoclassical theory where a negative electric field and an inward convection are predicted.