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.

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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.

Divertor power load reduction is one crucial issue for magnetically confined fusion reactors. Increased edge radiation can dissipate power before reaching divertor plates. Stable sustainment of such enhanced radiation, i.e. radiative divertor (RD) or detached divertor, is, however, still an open issue. In this paper, we present experimental evidence of edge radiation stabilization effect and potential of divertor power load control with resonant magnetic perturbation (RMP). Figure 1 shows time evolution of peak divertor power load, radiation intensity (P rad ) and line averaged density ( e n ) during density ramp up experiments with and without RMP. RMP has m/n=1/1 mode, which has resonance layer in the edge stochastic region, and creates remnant island. The perturbation strength is kept constant at ≈ 0 / B b r 0.1% throughout the discharge in the case with RMP. The plasma is heated by neutral beam injection (NBI) with ~ 8MW of deposited power in both cases. The divertor power load is estimated with Langmuir probe. The radiation is obtained with photo diode array viewing almost entire plasma at specific toroidal location. Without RMP (gray lines), the radiation intensity gradually increases with increasing density. The rapid increase of radiation intensity at t~3.8 sec with concomitant density rise indicates onset of thermal instability. The instability grows so rapidly that it is difficult to stabilize the density rise, leading to discharge termination. With RMP (black lines), on the other hand, transition to enhanced radiation state occurs at t~3 sec, and it leads to divertor power load reduction by a factor of 3 ~ 10. The RD operation is successfully sustained by gas puff feedback control up to the end of NBI. The results show stabilization effect of RMP on the radiating edge plasma. The enhanced radiation with RMP is due to increased volume of low T e (~10 eV) region caused by temperature flattening at the Opoint. 3D edge transport simulation result, which is consistent with the radiation profile measurement, show that the radiation increases further around X-point of the island (1) , where the code predicts n e > 10 20 m -3 and T e ~ a few eV.The well structured edge radiation with RMP such as the selective cooling around X-point is considered to provide stabilization effect by holding the intense radiation there and thus avoids it penetrating inward. Fig.2 shows dependence of controllability of RMP assisted RD on radial location of the island X-point,

Spectral profiles of the Hα line emitted from the large helical device plasma [O. Motojima et al., Phys. Plasmas 6, 1843 (1999)] have been measured with polarization-separation optics and a high-resolution spectrometer. Besides the underlying high-temperature component, which probably arises from charge-exchange recombination, the profiles are interpreted as superpositions of Zeeman profiles for two different magnetic field strengths. The emission locations are thus identified on the magnetic field map; the emissions are localized in the inner and outer regions just outside the ergodic layer, and each field-strength contribution to the overall Zeeman profile represents two radiator temperatures, and inward atom flow velocities in the range of (1–7)×103m∕s.

It is found that the remnant island structure created by n/m=1/1 resonant magnetic perturbation field in the stochastic magnetic boundary of the Large Helical Device (LHD) [A. Komori et al., Nucl. Fusion 49, 104015 (2009).] has a stabilizing effect on formation of radiating plasma, realizing stably sustained divertor detachment operation with the core plasma being unaffected. The data from the several diagnostics, (profiles of electron temperature & density, radiation and temporal evolution of divertor particle flux) indicate selective cooling around X-point of the island and thus peaked radiation there, which is stabilized outside of the last closed flux surface throughout the detachment phase. The VUV spectroscopy measurements of high Z impurity (iron) emission shows significant decrease during the detachment, indicating core plasma decontamination. The results from the 3D edge transport code EMC3 (Edge Monte-Carlo 3D) [Y. Feng et al., Contributions to Plasma Physics, 44, 57 (2004).]-EIRENE [D. Reiter et al., Fusion Sci. Technol., 47 172 (2005).] show similar tendency in the radiation pattern. The island size and its radial location are varied to investigate the magnetic topology effects on the detachment control. The divertor particle flux and neutral pressure exhibit intermittent oscillation as well as modification of recycling pattern during the detachment, which are found to reflect the island structure.