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In low density discharges of a Large Helical Device (LHD), anisotropic pressure is expected because the LHD has powerful tangential neutral beam injection systems. We study the behaviour of the ratio of the observed saddle loop flux to the diamagnetic flux, and the results are compared with the predicted beam pressure anisotropy by a Monte Carlo technique and the steady state Fokker–Planck solution. We show the possibility of the degree of pressure anisotropy being estimated by magnetic measurements in the LHD.

In the Large Helical Device (LHD), the highest operational averaged beta value has been expanded from 3.2% to 4% in the last 2 years by increasing the heating capability and exploring a new magnetic configuration with a high aspect ratio. Although the magneto-hydrodynamic (MHD) stability properties are considered to be unfavourable in the new high aspect configuration, the heating efficiency due to neutral beams and the transport properties are expected to be favourable in a high-beta range. In order to clarify the effect of the global ideal MHD unstable mode on the operational regimes in helical systems, especially the beta gradients in the peripheral region and the beta value, the MHD analysis and the transport analysis are performed in a high-beta range of up to 4% in LHD. In a high-beta range of more than 3%, the maxima of the observed thermal pressure gradients at a low order rational magnetic surface in the peripheral region are marginally unstable to the low-mode-number ideal MHD instability. Though a gradual degradation of the local transport in the region has been observed as beta increases, a disruptive degradation of the local transport does not appear in the beta range up to 4%.

The Large Helical Device is a heliotron device with L = 2 and M = 10 continuous helical coils with a major radius of 3.5–4.1 m, a minor radius of 0.6 m and a toroidal field of 0.5–3 T, which is a candidate among toroidal magnetic confinement systems for a steady state thermonuclear fusion reactor. There has been significant progress in extending the plasma operational regime in various plasma parameters by neutral beam injection with a power of 13 MW and electron cyclotron heating (ECH) with a power of 2 MW. The electron and ion temperatures have reached up to 10 keV in the collisionless regime, and the maximum electron density, the volume averaged beta value and stored energy are 2.4 × 1020 m−3, 4.1% and 1.3 MJ, respectively. In the last two years, intensive studies of the magnetohydrodynamics stability providing access to the high beta regime and of healing of the magnetic island in comparison with the neoclassical tearing mode in tokamaks have been conducted. Local island divertor experiments have also been performed to control the edge plasma aimed at confinement improvement. As for transport study, transient transport analysis was executed for a plasma with an internal transport barrier and a magnetic island. The high ion temperature plasma was obtained by adding impurities to the plasma to keep the power deposition to the ions reasonably high even at a very low density. By injecting 72 kW of ECH power, the plasma was sustained for 756 s without serious problems of impurities or recycling.

An overview of the progress with experiments in the HT-7 during 2003–4 is presented. The H-mode, negative reversed shear and high li operational scenarios were investigated for quasi-steady-state high performance plasma discharges. Ion Bernstein wave (IBW) heating at 30 MHz produced a typical edge H-mode plasma. Transport in both electron and ion channels were reduced in the outer half portion of the plasma. H-mode was produced in off-axis lower hybrid current drive (LHCD) plasmas by launching a lower hybrid wave (LHW) with a greater parallel refractive index, , of 3.1. The improved confinement status in such plasma discharges could be sustained for more than 400τE. A high inductance, li, plasma was created using a fast plasma current ramp down and sustained by central LHCD and IBW heating for a duration of >1 s with a strongly peaked electron temperature profile. The highest central electron temperature obtained was 4.5 keV. An increase in the energy confinement time with li was observed. It was found that the IBW heating could improve the plasma confinement at the same li if part of the LHW power was replaced by IBW. Stationary internal transport barriers with normalized performance H89βN > 1–3 were obtained with combined injection of LHW and IBW for a duration of several hundred energy confinement times in the weak negative magnetic shear (RS). Increasing the total injection power to 1 MW did not degrade the plasma confinement significantly in the RS operational scenario. Long pulse discharges were performed, using three feedback controls for the plasma current and position, the central line-averaged electron density and the magnetic swing flux of the transformer. The longest plasma discharge, with a duration of 240 s, Te(0) ∼ 1 keV and the central electron density > 0.8 1019 m−3 was achieved in 2004. The wall saturation and refreshment (wall pumping) were first observed in such long pulse discharges. A fully LHW current driven plasma without using ohmic current in the central solenoid coils was sustained for 80 s. The main limitation for the pulse length was due to the recycling, which caused an uncontrollable rise in the electron density.

Energy confinement and heat transport of net-current-free NBI heated plasmas in the large helical device (LHD) are discussed with emphasis on density and power deposition profile dependences. Although the apparent density dependence of the energy confinement time has been demonstrated in a wide parameter range in LHD, the loss of this dependence has been observed in the high density regime under specific conditions. Broad heat deposition due to off-axis alignment and shallow penetration of neutral beams degrades the global energy confinement while the local heat transport maintains a clear temperature dependence, lying between Bohm and gyro-Bohm characteristics. The central heat deposition tends towards an intrinsic density dependence like τE∝(n¯e/P)0.6 from the state where density dependence is lost. The broadening of the temperature profile due to the broad heat deposition profile contrasts with the invariant property that has been observed widely as profile resilience or stiffness in tokamak experiments. The confinement improvement as a result of the inward shift of the magnetic axis is obvious in the core region, which emphasizes the improvement of transport because of the geometry being unfavourable for the central heating of NBI in this configuration. The edge pressure, clearly, does not depend on the magnetic axis position. Unlike a tokamak H-mode, the edge pressure is determined by transport and can be increased by increasing the heating power.

Significant progress has been made with ion cyclotron range of frequencies (ICRF) heating in the Large Helical Device. This is mainly due to better confinement of the helically trapped particles and less accumulation of impurities in the region of the plasma core. During the past two years, ICRF heating power has been increased from 1.35 to 2.7 MW. Various wave-mode tests were carried out using minority-ion heating, second-harmonic heating, slow-wave heating and high-density fast-wave heating at the fundamental cyclotron frequency. This fundamental heating mode extended the plasma density range of effective ICRF heating to a value of 1×1020 m−3. This use of the heating mode was its first successful application in large fusion devices. Using the minority-ion mode gave the best performance, and the stored energy reached 240 kJ using ICRF alone. This was obtained for the inward-shifted magnetic axis configuration. The improvement associated with the axis-shift was common for both bulk plasma and highly accelerated particles. For the minority-ion mode, high-energy ions up to 500 keV were observed by concentrating the heating power near the plasma axis. The confinement properties of high-energy particles were studied for different magnetic axis configurations, using the power-modulation technique. It confirmed that with the inward-shifted configuration the confinement of high-energy particles was better than with the normal configuration. By increasing the distance of the plasma to the vessel wall to about 2 cm, the impurity influx was sufficiently reduced to allow sustainment of the plasma with ICRF heating alone for more than 2 min.