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The local structures of 122-type paradium arsenides, namely BaPd2As2 and SrPd2As2, are examined by As K-edge extended x-ray absorption fine structure measurements to find a possible correlation between the variation of their superconducting transition temperature and the local structure. The local atomic distances are found to be consistent with average distances measured by diffraction techniques. The temperature dependence of mean square relative displacements reveal that, while BaPd2As2 is characterized by a local As-Pd soft mode, albeit with larger atomic disorder, SrPd2As2 shows anomalous As-Pd correlations with a kink at similar to 160 K due to hardening by raising temperature. We have discussed implications of these results and possible mechanisms of differing superconducting transition temperature in relation with the structural instability.

A remote steering antenna was recently developed for electron cyclotron heating and current drive (ECH/ECCD) experiments on the TRIAM-1M tokamak. This is the first application of the remote steering antenna concept for ECH/ECCD experiments, which have conditions relevant to the International Thermonuclear Experimental Reactor (ITER). Fundamental ECH and ECCD experiments were conducted in the ITER frequency from the low field using this antenna system. In addition to the angles near 0°, the launcher was a symmetric direction antenna with an extended steering-angle capability of ±(8°–19°). The output beam from the antenna was a well-defined Gaussian with a proper steering angle. The Gaussian content and the steering-angle accuracy were 0.85 and −0.5°, respectively. The high power tests measured the antenna transmission efficiency at 0.90–0.94. The efficiencies obtained in the low and high power tests were consistent with the calculations using higher-order modes. In order to excite the pure O/X-modes in the oblique injection, two polarizers were used to control the elliptical polarization of the incident beam for the ECCD experiments. The fundamental O/X-mode ECH/ECCD was applied to lower hyrid current drive plasmas at the optimized incident polarization. In the X-mode experiment, at medium density (~1 × 1019 m−3), clear differences in the plasma current and the hard x-ray intensity were observed between the co- and counter-steering injections due to the ECCD effect on the coupling of forward fast electrons.

Two types of improved core confinement were observed during centrally focused electron cyclotron heating (ECH) into plasmas sustained by counter (CNTR) and Co neutral beam injections (NBI) in the Large Helical Device. The CNTR NBI plasma displayed transition phenomena to the high-electron-temperature state and had a clear electron internal transport barrier, while the Co NBI plasma did not show a clear transition or an ECH power threshold but showed broad high temperature profiles with moderate temperature gradient. This indicated that the Co NBI plasma with additional ECH also had an improved core confinement. The electron heat transport characteristics of these plasmas were directly investigated using heat pulse propagation excited by modulated ECH. These effects appear to be related to the m/n = 2/1 rational surface or the island induced by NBI beam-driven current.

This paper reports on the results of an experimental investigation into electron cyclotron resonance heating (ECRH) using the fundamental O-mode in the large helical device (LHD). The aim of the experiment is to understand the mechanism of the selective excitation of the O-mode EC wave and, thereby, optimize the performance of the ECRH. The polarization angle and the ellipticity of the electric field of the incident high power millimetre waves were varied in the experiment using a pair of grating polarizers. The responses of the plasma stored energy and electron temperature were measured to evaluate the effects. The optimum values of the polarization angle and the ellipticity for the ECRH were identified experimentally. A simple model of power branching between the O- and X-modes is introduced in order to interpret the observed experimental results. Specific to the ECRH in the LHD configuration, the effect of magnetic shear in the plasma peripheral region is considered.

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.

Internal transport barriers with respect to electron thermal transport (eITB) were observed in the large helical device, when the electron cyclotron resonance heating (ECH) power was highly localized on the centre of a plasma sustained by neutral beam injection. The eITB is characterized by a high central electron temperature of 6–8 keV with an extremely steep gradient, as high as 55 keV m−1 and a low electron thermal diffusivity within a normalized average radius ρ≈0.3 as well as by the existence of clear thresholds for the ECH power and plasma collisionality.

The behaviour of high-energy ions accelerated by an ion cyclotron range of frequency (ICRF) electric field in the large helical device (LHD) is discussed. A better confinement performance of high-energy ions in the inward-shifted magnetic axis configuration was experimentally verified by measuring their energy spectrum and comparing it with the effective temperature determined by an electron slowing down process. In the standard magnetic axis configuration a saturation of the measured tail temperature was observed as the effective temperature was increased. The ratio between these two quantities is a measure of the quality of transfer efficiency from high-energy ions to a bulk plasma; when this efficiency was compared with Monte Carlo simulations the results agreed fairly well. The ratio of the stored energy of the high-energy ions to that of the bulk plasma was measured using an ICRF heating power modulation method; it was deduced from phase differences between total and bulk plasma stored energies and the modulated ICRF heating power. The measured high energy fraction agreed with that calculated using the injected ICRF heating power, the transfer efficiency determined in the experiment and the confinement scaling of the LHD plasma.

Ion temperature at the plasma centre has been measured from Doppler broadening of Ti XXI (2.61 A) and Ar XVII (3.95 A) x-ray lines using a newly installed crystal spectrometer with CCD detector in ECH, NBI and ICRF plasmas of Large Helical Device (LHD). The ion temperature obtained in a range of 0.6 and 3.5 keV was analysed with electron density and compared with electron temperature. A new parameter range of Ti>Te was found in low-density (ne

In LHD discharges, the NBI heated plasmas are terminated in two ways: (a) thermal decay (TD) after the termination of NBI and (b) radiative collapse (RC) during the NBI heating. The basic characteristics of the TD and RC discharges are compared. It is found that the decay and collapse of the plasma are mainly governed by the heating power and the plasma density. The critical density c for the collapse of RC plasmas is similar to the scaling laws obtained in other helical devices, i.e. c∝(PB/V)0.5, where P, B and V denote heating power, magnetic field and plasma volume, respectively. Moreover, measurements using multichannel bolometric diagnostics indicate that the total radiation profiles in TD and RC plasmas are usually inboard-outboard symmetric and asymmetric, respectively, at the end of the discharge. In RC discharges, the total radiation profile develops in several phases. Before the onset of the thermal instability (TI), the radiation profile is rather symmetric, while after that, the radiation profile evolves from being symmetric in the initial period towards being asymmetric eventually with high radiation on the inboard side. Corresponding variations are shown in the time evolutions of the density and temperature profiles, and a substantial contraction of the plasma column is observed immediately after TI onset. The spatial and temporal coincidence of the asymmetries in the radiation, density and temperature is similar to that observed with multifaceted asymmetric radiation from the edge (MARFE) in tokamaks. But, unlike MARFEs, the asymmetric radiation (AR) in LHD is rather transient since it appears just before the end of RC discharges. The underlying cause for the development of radiation asymmetry was investigated and compared with existing instability models. The result suggests that the high inboard radiation is a manifestation of an enhanced local thermal instability, and the AR results from asymmetric developments of TI on the inboard-outboard sides during the final stage of RC discharges.

The experimentally optimized magnetic field configuration of the Large Helical Device [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)], where the magnetic axis is shifted inward by 15 cm from the early theoretical prediction, reveals 50% better global energy confinement than the prediction of the scaling law. This configuration has been investigated further from the viewpoints of high energy particle confinement and magnetohydrodynamic (MHD) stability. The confinement of high energy ions is improved as expected. The minority heating of ion cyclotron range of frequency was successful and the heating efficiency was improved by the inward shift. The confinement of passing particles by neutral beam injection was also improved under low magnetic field strength, and there could be obtained an almost steady high beta discharge up to 3% in volume average. This was a surprising result because the observed pressure gradient exceeded the Mercier unstable limit. The observed MHD activities became as high as beta but they did not grow enough to deteriorate the confinement of high energy ions or the performance of the bulk plasma, which was still 50% better than the scaling. According to these favorable results, better performance would be expected by increasing the heating power because the neoclassical transport can also be improved there.

During the first two years of the LHD experiment the following results have been achieved: (i) higher Te (Te(0) = 4.4 keV at ne = 5.3 × 1018 m-3 and Pabs = 1.8 MW); (ii) higher confinement (τE = 0.3 s, Te(0) = 1.1 keV at ne = 6.5 × 1019 m-3 and Pabs = 2.0 MW); (iii) higher stored energy, Wpdia = 880 kJ at B = 2.75 T. High performance plasmas have been realized in the inward shifted magnetic axis configuration (R = 3.6 m) where helical symmetry is recovered and the particle orbit properties are improved by a trade-off of MHD stability properties due to the appearance of a magnetic hill. Energy confinement was systematically higher than that predicted by the International Stellarator Scaling 95 by up to a factor of 1.6 and was comparable with the ELMy H mode confinement capability in tokamaks. This confinement improvement is attributed to configuration control (inward shift of the magnetic axis) and to the formation of a high edge temperature. The average beta value achieved reached 2.4% at B = 1.3 T, the highest beta value ever obtained in a helical device, and so far no degradation of confinement by MHD phenomena has been observed. The inward shifted configuration has also led to successful ICRF minority ion heating. ICRF powers up to 1.3 MW were reliably injected into the plasma without significant impurity contamination, and a plasma with a stored energy of 200 kJ was sustained for 5 s by ICRF alone. As another important result, long pulse discharges of more than 1 min were successfully achieved separately with an NBI heating of 0.5 MW and with an ICRF heating of 0.85 MW.

The bootstrap current and electron thermal conductivity have been investigated in ECRH plasmas on the CHS heliotron/torsatron. The neoclassical transport coefficients have been externally controlled over a wide range by changing the ellipticity of the plasma cross-section. While local thermal conductivity, as well as global confinement, have shown inconsistency with the neoclassical predictions, the observed toroidal net current has been successfully explained by considering the three dimensional geometry with the available neoclassical theory. The present experimental results indicate that the transport parallel to the magnetic field lines (bootstrap current) is likely to be neoclassical but that the perpendicular transport (thermal conductivity) is anomalous, which is common in current less plasmas as well as tokamaks

In LHD discharges a significant enhancement of the global energy confinement has been achieved for the first time in a helical device with an edge thermal barrier, which exhibits a sharp gradient at the edge of the temperature profile. Key features associated with the barrier are quite different from those seen in tokamaks: (i) almost no change in particle (including impurity) transport, (ii) a gradual formation of the barrier, (iii) a very high ratio of the edge temperature to the average temperature, and (iv) no edge relaxation phenomenon. These features are very attractive in applying the thermal barrier to future reactor grade devices.