In the Large Helical Device (LHD) operated with a metallic (stainless steel) first wall, it is found that the iron density, nFe, at the plasma core is fairly low (nFe ⩽ 108 cm−3) in general neutral beam (NB)-heated discharges, while the iron quickly increases with the appearance of impurity accumulation when a multi-hydrogen ice pellet is injected or the NB input power is largely reduced. Although the highest iron density (nFe ⩽ 1010 cm−3) at the plasma centre in the LHD is observed from such discharges, it suggests a still low iron concentration (nFe/ne < 10−3). Therefore, the edge iron transport in the ergodic layer, which determines the iron influx to the core plasma, is studied to clarify why the iron density in the core plasma is low. A line ratio of Fe XV located in the vicinity of the last closed flux surface to Fe VIII (or Fe IX) located in the ergodic layer decreases with density. The two-dimensional (2D) edge iron emission of Fe XVI and Fe IX is enhanced in the vicinity of the X-point with a larger number of magnetic field lines directly connected to divertor plates, which suggests that iron ions from the first wall move downstream. The density of edge Fe15+ ions giving the iron influx to the core plasma is analysed with the 2D distribution. The analysis also shows that the iron influx to the core plasma decreases with density. These results clearly indicate that the screening effect developed in the ergodic layer works well for iron ions coming from the first wall. A three-dimensional edge transport simulation with EMC3-EIRENE can also predict an effective impurity screening for heavy impurities compared to light impurities.

Remarkable progress in the physical parameters of net-current free plasmas has been made in the Large Helical Device (LHD) since the last Fusion Energy Conference in Chengdu, 2006 (O.Motojima et al., Nucl. Fusion 47 (2007. The beta value reached 5 % and a high beta state beyond 4.5% from the diamagnetic measurement has been maintained for longer than 100 times the energy confinement time. The density and temperature regimes also have been extended. The central density has exceeded 1.0×10 21 m -3 due to the formation of an Internal Diffusion Barrier (IDB). The ion temperature has reached 6.8 keV at the density of 2×10 19 m -3 , which is associated with the suppression of ion heat conduction loss. Although these parameters have been obtained in separated discharges, each fusion-reactor relevant parameter has elucidated the potential of net-current free heliotron plasmas. Diversified studies in recent LHD experiments are reviewed in this paper.

Recently, the properties of high temperature superconducting tapes have been in advance and high temperature superconducting magnets have been constructed and demonstrated. However, the high temperature superconducting tapes have different thermal characteristics compared with low temperature superconducting wires. Therefore, it is necessary to consider these characteristics of high temperature superconducting tapes at the magnet design stage. We proposed an optimal design method for superconducting coils wound with Bi2223/Ag tapes. In this paper, the configuration of 72 MJ SMES coils wound with Bi2223/Ag tapes are optimized.

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. High temperatures of approximately 1–2 keV, and steady-state plasma sustainment operations have been reported. Recently the steady-state operation regime was greatly extended in the Large Helical Device (LHD). A high-temperature plasma was created and maintained for 54 min with 1.6 GJ in the 2005FY experimental programme. The three-dimensional heat-deposition profile of the LHD helical divertor was modified, and during long-pulse discharges it effectively dispersed the heat load using a magnetic axis swing technique developed at the LHD. A sweep of only 3 cm in the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies and partially by electron cyclotron of fundamental resonance frequency. By accumulating the small flux of charge-exchanged neutral particles during the long-pulse operation, a high energy ion tail which extended up to 1.6 MeV was observed. This is the first experimental evidence of high energetic ion confinement of MeV range in helical devices. The long-pulse operations lasted until a sudden increase in radiation loss occurred, presumably because of metal wall flakes dropping into the plasma. The sustained line-averaged electron density and temperature were approximately 0.8 × 1019 m−3 and 2 keV, respectively, at a 1.3 GJ discharge (#53776) and 0.4 × 1019 m−3 and 1 keV at a 1.6 GJ discharge (#66053). The average input power was 680 kW and 490 kW, and the plasma duration was 32 min and 54 min, respectively. These successful long operations show that the heliotron configuration has a high potential as a steady-state fusion reactor.

High-ion-temperature ͑exceeding 5 keV͒ hydrogen plasmas have been successfully produced in the Large Helical Device ͓Iiyoshi et al., Nucl. Fusion 39, 1245 ͑1999͒; Motojima et al., Nucl. Fusion 47, S668 ͑2007͔͒ with the ion heat confinement improvement in the core region. The experimental ion heat diffusivity at the core region is found to be almost independent of the ion temperature, T i ͑even decreasing as T i increases͒. The neoclassical ͑NC͒ ripple transport is suppressed by the ambipolar radial electric field, E r ͑Ͻ0͒ predicted by NC transport fluxes. The temperature ratio, T i / T e , is one of the key parameters to reduce the NC ambipolar particle and heat fluxes. Thus, it is suggested that the selective ion heating ͑making T i / T e larger͒ is a plausible approach to further increase T i . Spontaneous rotation is evaluated in these high-T i plasmas, in which a co-directed component is recognized at the radial location with a large T i gradient, in addition to the tokamak-like counter-directed component expected for E r Ͻ 0.