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.

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.

OVERVIEW OF THE LARGE HELICAL DEVICE PROJECT. The Large Helical Device (LHD) has successfully started running plasma confinement experiments after a long construction period of eight years. During the construction and machine commissioning phases, a variety of milestones were attained in fusion engineering which successfully led to the first operation, and the first plasma was ignited on 31 March 1998. Two experimental campaigns are planned in 1998. In the first campaign, the magnetic flux mapping clearly demonstrated a nested structure of magnetic surfaces. The first plasma experiments were conducted with second harmonic 84 and 82.6 GHz ECH at a heating power input of 0.35 MW. The magnetic field was set at 1.5 T in these campaigns so as to accumulate operational experience with the superconducting coils. In the second campaign, auxiliary heating with NBI at 3 MW has been carried out. Averaged electron densities of up to 6 × 10 19 m-3 , central temperatures ranging from 1.4 IAEA-F1-CN-69/OV1/4 2 to 1.5 keV and stored energies of up to 0.22 MJ have been attained despite the fact that the impurity level has not yet been minimized. The obtained scarling of energy confinement time has been found to be consistent with the ISS95 scaling law with some enhancement.

Abstract. Edge impurity transport has been investigated in the stochastic layer of Large Helical Device (LHD) and the scrape-off layer (SOL) of Huan Liuqi-2A (HL-2A) tokamak, as a comparative analysis based on the three-dimensional (3D) edge transport code EMC3-EIRENE and on the carbon emission profile measurement. The 3D simulation predicts impurity screening effect in the both devices, but also predicts different impurity behavior against collisionality and impurity source location between the two devices. The difference is caused by geometrical structures of the magnetic field lines in the stochastic layer and X-point poloidal divertor SOL, i.e., number of poloidal turns of flux tubes affecting poloidal distribution of plasma parameters and impact of perpendicular transport on parallel pressure conservation and energy transport. These processes have an influence on the impurity screening efficiency at upstream and downstream positions of field lines. The carbon emission measured in the stochastic layer of LHD clearly indicates the screening effect in high density region. The result can be qualitatively interpreted by the present modeling, although the modeling shows a slight difference in the quantitative behavior of carbon ions in the stochastic layer of LHD. On the other hand, comparison of the carbon emission profile from HL-2A with the modeling is not straightforward. It is found that the impurity distribution in the HL-2A SOL is very sensitive to the impurity source location. In order to interpret the experimental observation a further study is necessary, in particular, on the impurity source distribution in the divertor plate and the first wall.