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We present an analysis of the spectral emission of Fe XIII near 200 A. High-resolution spectra were recorded at two densities (~2 × 1011 and ~1013 cm−3) in the laboratory and compared to collisional radiative model calculations based on the CHIANTI database, and to models using atomic data from distorted-wave and R-matrix calculations. The Fe XIII lines in this wavelength range are sensitive indicators of plasma density below ~ 1011 cm–3. The laboratory data thus test calculations in the astrophysical high-density limit. Significant differences between the measurements and models were found for several line ratios. Differences in the wavelengths employed in the different models also changed the agreement with the measurements. The best agreement was found for comparisons with CHIANTI.

Edge cooling experiments with a tracer-encapsulated solid pellet in the large helical device (LHD) show a significant rise in core electron temperature (the maximum rise is around 1 keV) as well as in many tokamaks. This experimental result indicates the possible presence of the nonlocality of electron heat transport in plasmas where turbulence as a cause of anomalous transport dominates. The nonlocal electron temperature rise in the LHD takes place in almost the same parametric domain (e.g. in a low density) as in the tokamaks. Meanwhile, the experimental results of LHD show some new aspects of nonlocal electron temperature rise, for example the delay in the nonlocal rise of core electron temperature relative to the pellet penetration time increases with the increase both in the collisionality in the core plasma and the electron temperature gradient scale length in the outer region of the plasma.

Experimental data on spatial distributions of a pellet cloud electron density are necessary for the development of many applications of pellet injection, namely, plasma fuelling, discharge control, and plasma diagnostics. An improved approach of electron density measurements inside the cloud of a polystyrene pellet ablating in hot plasma of the large helical device is described. Density values of (1-30) × 10(16) cm(-3) depending on the background plasma parameters and distance from the solid pellet were measured.

This paper is devoted to the analysis of the possible experiments with the D + D in Large Helical Device geometry and dynamics of fusion products. The transfer of energy from the fusion product to background plasma particles is discussed and the separation of the fusion products with the use of the drift resonances is considered.

The long-pulse plasma discharge experiment is an important experiment in the Large Helical Device, which has a superconducting coil system and the capability of steady-state operation. The experiment of long-pulse plasma discharge was carried out using mainly ion cyclotron range of frequencies heating. The maximum plasma duration is 31 min and 45 s, and the total injected heating energy reached 1.3 GJ. Swing of the magnetic axis is adopted as an effective method to scatter the local heat load on the dive rtor plate during the discharge. The plasma was terminated abruptly by the influx of metallic impurities accompanied by a spark in the vacuum vessel.

Cold pulse inversion in a helical plasma is observed in the Large Helical Device (LHD) and thus the strong non-local effects are evident in the helical device as well as in tokamaks. A hydrogen pellet or tracer encapsulated solid pellet is injected into the edge of the LHD plasmas. A significant rise of the electron temperature is observed in the central region in response to the edge cooling. Transient analysis indicates a heat flux jump despite the absence of a change in the local temperature gradient. The non-local temperature rise takes place in the low density and high temperature regime just as predicted by the TFTR scaling.

A long-pulse plasma discharge of more than 30 min duration was achieved on the Large Helical Device (LHD). A plasma of ne = 0.8 × 1019 m−3 and Ti0 = 2.0 keV was sustained with PICH = 0.52 MW, PECH = 0.1 MW and averaged PNBI = 0.067 MW. The total injected heating energy was 1.3 GJ. One of the keys to the success of the experiment was a dispersion of the local plasma heat load to divertors, accomplished by sweeping the magnetic axis inward and outward. Causes limiting the long pulse plasma discharge are discussed. An ion impurity penetration limited further long-pulse discharge in the 8th experimental campaign (2004).

A long-pulse plasma discharge for more than 30 min. was achieved on the Large Helical Device (LHD). A plasma of ne = 0.8× 1019 m−3 and Ti0 = 2.0 keV was sustained with PICH = 0.52 MW, PECH = 0.1 MW and averaged PNBI = 0.067 MW. Total injected heating energy was 1.3 GJ, which was a quarter of the prepared RF heating energy. One of the keys to the success of the experiment was a dispersion of the local plasma heat load to divertors, accomplished by shifting the magnetic axis inward and outward.

Transient transport experiments are performed in plasmas with and without internal transport barriers (ITB) on LHD and JT-60U. The dependence of χe on the electron temperature, Te, and on the electron temperature gradient, ∇Te, is analysed with an empirical non-linear heat transport model. In plasmas without an ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation: the χe depends on Te and ∇Te in JT-60U, while the ∇Te dependence is weak in LHD. Inside the ITB region, there is none or weak ∇Te dependence both in LHD and JT-60U. Growth of the cold pulse driven by the negative Te dependence of χe is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U).

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.

"It is very important to demonstrate the ability to sustain the plasma in a steady state on the Large Helical Device (LHD), which has external helical magnetic coils and is a superconducting device. The long pulse discharge experiment was carried out using the ion cyclotron range of frequencies (ICRF) heating mainly. The plasma discharge of 31 minutes and 45 seconds was achieved by a total injected heating energy of 1.3GJ. Swing of the magnetic axis to scatter the local heat load on the divertor plate was one of the key methods for the steady state operation. The repetitive hydrogen pellet injection was tried successfully to fuel the minority hydrogen ions for long pulse operation."

Transient transport experiments are performed in LHD plasma with electron internal transport barrier (e-ITB). Evidence for a reduction of electron heat diffusivity inside the ITB is observed from cold and heat pulse propagations. The observed enhancement of the cold pulse peak is explained by the temperature dependent electron heat diffusivity. The heat diffusivity inside the ITB decreases with an increase in the electron temperature in LHD.

The large helical device (LHD) is the world largest helical system having all superconducting coils. After completion of LHD in 1998, six experimental campaigns have been carried out successfully. The maximum stored energy, central electron temperature, and volume averaged beta value are 1.16 MJ, 10 keV, and 3.2%, respectively. The confinement time of the LHD plasma appears to be equivalent to that of tokamaks. One of the most important missions for LHD is to prove steady-state operation, which is also significant to international thermonuclear experimental reactor (ITER) and to future fusion reactors. LHD is quite appropriate for this purpose based upon the beneficial feature of a helical system, that is, no necessity of the plasma current. So far, the plasma discharge duration was achieved up to 150 s. The plasma density was kept constant by feedback control of gas puffing with real time information of the line density. The issue for demonstrating steady-state operation is whether divertor function to control particle and heat flux is effective enough. Relevant to this, LHD diagnostics should be consistent with the following: 1) continuous operation of main diagnostics during long-pulse operation for feedback control and physics understanding; 2) measurement of fraction of H, He, and impurities in the plasma; 3) heat removal and measure against possible damage or surface erosion of diagnostic components inside of the vacuum chamber; 4) data acquisition system for handling real time data display and a huge amount of data. Although there are already some achievements on the above subjects, there remain still several issues to be resolved. On the other hand, the long-pulse operation of the plasma gives benefits to the diagnostics. For example, the polarizing angle of ECE emission can be changed during the discharge, and the intensity dependence on the polarizing angle has been obtained. The spatial scanning of the neutral particle analyzer and the spectrometer can supply the spatial profiles of the fast neutral particle flux and the specific impurity lines. In this paper, the present status of these issues and future plans are described.

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.

By changing the distribution of the coil currents in the helical coils of the large helical device (LHD) the plasma could be scraped off on the stainless steel inboard wall at the last closed flux surface, eliminating the ergodic edge region and helical divertor (HD) plasmas. During such NBI-heated discharges slow oscillations (1–2.5 Hz) have been observed in the major global parameters. These oscillations exhibit many similarities to the ‘breathing’ relaxation phenomenon observed during long pulse experiments in LHD when the divertor material was stainless steel. These similarities include a large core radiation fraction, significant radiation from the metallic impurities and dependence of the oscillation frequency on density. In contrast, discharges using the graphite HD show hollow radiation profiles and reduced levels of radiation from heavy impurities. Modeling shows that differences in the density dependence of the oscillation frequency between the wall limiter and HD cases can be attributed to differences in the impurity diffusion between the core and ergodic edge regions.

The shift and fine line shape of H α emissions in large helical device (LHD) plasmas have been analyzed by a high resolution spectroscopic system. The minimum detectable velocity for hydrogen atoms is 10 3 m/s. This is confirmed by a newly developed light source with a magnetic field of 1.13 T. In a long-pulse NBI plasma of LHD, hydrogen atoms move toward the plasma from the divertor with velocity of 10 4 m/s. Fine spectral profile of H α was measured with a linear polarizer and it shows an asymmetric structure. The spectrum consists of cold component shifted to the blue side and tail extended to 656.0 nm. The energy corresponding to the tail is from 8 to 100 eV.

The utility of ICRF heating in the LHD was demonstrated in the third campaign carried out in 1999. This paper summarizes the investigations made in 2000 with a focus on the optimization of ICRF heating. The flexibility of the LHD magnetic configuration was fully utilized as a key factor in the investigations. The experiments include (a) scan of magnetic field intensity, (b) scan of aspect ratio, (c) scan of magnetic axis shift, (d) scan of quadrupole magnetic field, and (e) cancellation of magnetic island. The performance of the ICRF heating was thus optimized with respect to magnetic parameters, while optimization was achieved mainly with respect to heating regime and wall conditioning in the third campaign. Some of these parameters are directly related to the orbit of trapped particles. Therefore, the relation between particle orbits and the performance of ICRF heating is also addressed in this paper based on analyses of the high-energy tail of ions.

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.

This paper reports on the behaviour of high-energy ions created by ion cyclotron range of frequency (ICRF) heating on the Large Helical Device (LHD). In the third experimental campaign conducted in 1999, it was found that minority heating has good heating performance, and high-energy particles were observed. In the fourth campaign in 2000, the temporal behaviour of high-energy ions was investigated in the minority heating regime using turnoff or modulation of ICRF power. The time evolution of the high-energy particle distribution was measured using a natural diamond detector (NDD) and a time-of-flight neutral particle analyser (TOF-NPA). It was found that the count number of higher-energy particles declines faster than that of lower-energy particles after ICRF turnoff. In the modulation experiments, the phase difference of the flux of high-energy particles with respect to the ICRF power modulation increased with energy. These results were explained qualitatively by the Fokker-Planck equation with a simple model. The pitch-angle dependence of the distribution function was also measured in the experiment by changing the line of sight of the TOF-NPA, and an anisotropy of the high-energy tail was found. This anisotropy was reproduced by solving the bounce-averaged Fokker-Planck equation. The second harmonic heating was conducted successfully for the first time in the LHD in high-β plasma, and high-energy particles were also detected in this heating regime.

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.

Neutral beam injection heated discharges at the density limit in the Large Helical Device [O. Motojima, H. Yamada, A. Komori et al., Phys. Plasmas 6, 1843 (1999)] are terminated with asymmetric radiative collapse (ARC) exhibiting several properties in common with the MARFE (multifaceted asymmetric radiation from the edge) phenomenon: (1) A highly poloidally asymmetric radiation profile which is stronger on the inboard side. (2) This asymmetry is well correlated with the signal from the multichord interferometer. (3) Moreover, evidence from several diagnostics at different toroidal locations supports the possibility that ARC may be toroidally symmetric. However in contrast to MARFE, ARC is only observed in the period just prior to the quench of the plasma.

A power modulation experiment was conducted in the third campaign of the LHD. In a conventional analysis of the modulation experiments, the energy confinement time and heating efficiency are taken as constant, disregarding their dependence on the plasma parameters. In this paper, their dependence on the plasma temperature and heating power is taken into consideration to improve the analysis of the power modulation experiments. Several models, with differing dependence on the plasma parameters, have been examined. There have been several reports suggesting that the transport coefficients obtained from the dynamical method are different from those obtained from power balance analyses. This paper finally concludes that the energy confinement time obtained from the power modulation experiments well agrees with that obtained from the power balance analysis, made including temperature dependence both in energy confinement time and heating efficiency.

The energy confinement and thermal transport characteristics of net current free plasmas in regimes with much smaller gyroradii and collisionality than previously studied have been investigated in the Large Helical Device (LHD). The inward shifted configuration, which is superior from the point of view of neoclassical transport theory, has revealed a systematic confinement improvement over the standard configuration. Energy confinement times are improved over the International Stellarator Scaling 95 by a factor of 1.6 ±0.2 for an inward shifted configuration. This enhancement is primarily due to the broad temperature profile with a high edge value. A simple dimensional analysis involving LHD and other medium sized heliotrons yields a strongly gyro-Bohm dependence (τEΩ ∝ ρ*-3.8) of energy confinement times. It should be noted that this result is attributed to a comprehensive treatment of LHD for systematic confinement enhancement and that the medium sized heliotrons have narrow temperature profiles. The core stored energy still indicates a dependence of τEΩ ∝ ρ*-2.6 when data only from LHD are processed. The local heat transport analysis of discharges dimensionally similar except for ρ* suggests that the heat conduction coefficient lies between Bohm and gyro-Bohm in the core and changes towards strong gyro-Bohm in the peripheral region. Since the inward shifted configuration has a geometrical feature suppressing neoclassical transport, confinement improvement can be maintained in the collisionless regime where ripple transport is important. The stiffness of the pressure profile coincides with enhanced transport in the peaked density profile obtained by pellet injection.

A novel technology is suggested for making solid hydrogen shells around impurity pellets to be injected into plasmas of fusion devices with a view to looking into its transport properties. In proof-of-principle tests, a 3 mm long and 3 mm diameter cylindrical solid hydrogen shell was formed around a 0.2 mm diameter globular pellet at a temperature of 8–11 K within 5 min and accelerated in a pipe-gun barrel up to 1 km/s.

Pellet injection has been used as a primary fuelling scheme in the Large Helical Device. With pellet injection, the operational region of NBI plasmas has been extended to higher densities while maintaining a favourable dependence of energy confinement on density, and several important values, such as plasma stored energy of 0.88?MJ, energy confinement time of 0.3?s, ? of 2.4% at 1.3?T and density of 1.1 ? 1020?m -3, have been achieved. These parameters cannot be attained by gas puffing. Ablation and the subsequent behaviour of the plasma have been investigated. The measured pellet penetration depth estimated on the basis of the duration of the H? emission is shallower than the depth predicted from the simple neutral gas shielding (NGS) model. It can be explained by the NGS model with inclusion of the effect of fast ions on the ablation. Just after ablation, the redistribution of the ablated pellet mass was observed on a short timescale (~400?ms). The redistribution causes shallow deposition and low fuelling efficiency.

In the large helical device (LHD) radiation profiles measured using arrays of resistive metal, foil bolometers are used to investigate the change in radiation resulting from a replacement of stainless steel divertor plates with graphite tiles between the second and third experimental campaigns. In particular, for the magnetic configuration of R axis = 3.6 m and at line averaged densities below 3 x 10 19 m 3 , a reduction of the radiated power fraction from 35% in the second campaign to down to 12% in third campaign was observed. Comparing similar shots (in terms of discharge parameters) from the second and third cycles the core (0 < r/a < 0.79) radiation fraction was reduced from 44% to 30% of the total radiation. Spectroscopic measurements show a corresponding decrease in radiation from iron. Comparing long pulse discharges the radiated power fraction reduced from greater than 67% during the density limiting oscillation observed in the second campaign to 20% during the third campaign with a reduction in the respective core radiation fractions from 63% to 37%.

Semiconductor detectors based on natural diamonds have been installed on the Large Helical Device (LHD) heliotron to measure the energy distribution of charge exchange fast neutral particles from different viewing angles. Advantages of a natural diamond detector (NDD) are (1) very compact size, (2) relatively easy handling, and (3) high energy resolution. Although NDDs are sensitive to visible light, vacuum ultraviolet, and soft x rays, unfavorable pulses produced by such radiation were greatly reduced by choosing an appropriate stainless steel shield in this experiment. In LHD, the time-resolved energy distribution of counter-going beam ions and ion cyclotron range of frequency-produced energetic ions have been successfully obtained by means of an NDD. The performance of NDDs as a neutral particle analyzer and its good suitability to LHD plasmas were demonstrated throughout this work.

Recent experimental results are summarized. One of them is the bifurcation phenomenon of the electric potential which manifests itself in various aspects of CHS plasmas. Here, electrons heated with ECH play an essential role. The CHS will be operated, after being moved to the Toki site, so as to perform complementary experiments to LHD.

The initial experiments on the Large Helical Device (LHD) have extended confinement studies on currentless plasmas to a large scale (R = 3.9 m, a = 0.6 m). Heating by NBI of 3 MW produced plasmas with a fusion triple product of 8 × 1018m-3keVs at a magnetic field strength of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.1 keV were achieved simultaneously at a line averaged electron density of 1.5 × 1019 m-3. The maximum stored energy reached 0.22 MJ with neither unexpected confinement deterioration nor visible MHD instabilities, which corresponds to β = 0.7%. Energy confinement times reached a maximum of 0.17 s. A favourable dependence of energy confinement time on density remains in the present power density (~40 kW/m3) and electron density (3 × 1019 m-3) regimes, unlike the L mode in tokamaks. Although power degradation and significant density dependence are similar to the conditions on existing medium sized helical devices, the absolute value is enhanced by up to about 50% from the International Stellarator Scaling 95. Temperatures of both electrons and ions as high as 200 eV were observed at the outermost flux surface, which indicates a qualitative jump in performance compared with that of helical devices to date. Spontaneously generated toroidal currents indicate agreement with the physical picture of neoclassical bootstrap currents. Change of magnetic configuration due to the finite β effect was well described by 3-D MHD equilibrium analysis. A density pump-out phenomenon was observed in hydrogen discharges, which was mitigated in helium discharges with high recycling.

The effects of the plasma profile on the global energy confinement have been studied in Heliotron E with special regard to differences between heating methods (ECH, NBI and NBI + ECH). With high power NBI, peaked Ti and peaked ne profiles (Ti(0)/Ti 2.7, ne(0)/ne 4.5) were simultaneously achieved under low recycling conditions. A peaked ne profile (ne(0)/ne 2.5 ) could lead to the high Ti mode where the ion heat transport in the central region is substantially reduced. By changing the ECH launching condition (on-axis, off-axis and toroidally oblique injection), the peakedness of the Te profile could be controlled in the range 1.3 Te(0)/Te 4.5. A peaked Te profile and a flat ne profile (3.5 Te(0)/Te, ne(0)/ne 1.8) were brought about by the well focused on-axis ECH. The ECH plasma with a peaked Te profile has higher stored energy than that with a moderately peaked Te profile for the same injected ECH power and the same density region. The global energy confinement time normalized by the LHD scaling, τE G/τE LHD, showed ne(0)/ne dependence for the low Ti mode NBI plasmas. For the high Ti mode, the ne(0)/ne dependence of τE G/τE LHD was weak. These findings suggest that the LHD scaling should be modified to scale the global energy confinement of the helical plasmas in a wide range of ne(0)/ne.

Variation of the plasma position relative to the centre of the helical coil winding is a very effective means of controlling the MHD stability and the trapped particle confinement in heliotron/torsatron systems, but improving one of these two characteristics with this parameter simultaneously has a detrimental effect on the other. The inward shifted configuration is favourable for drift orbit optimization but is predicted to be unstable according to the Mercier criterion. Various physics problems, such as electric field structure, plasma rotation and MHD phenomena, have been studied in the Compact Helical System (CHS) with a compromise intermediate position. With this standard configuration, CHS has yielded experimental results that contribute to the understanding of general toroidal confinement physics and low aspect ratio helical systems. In the recent experiments, it was found that a wide range of inward shifted configurations give stable plasma discharges without any restriction to the special pressure profile. Such an enhanced range of operation made it possible to study experimentally the drift orbit optimized configuration in heliotron/torsatron systems. The effect of configuration improvement was studied with plasmas in a low collisionality regime.

Li pellets of large size were injected into electron cyclotron resonance (ECR) heated plasmas and neutral beam injection (NBI) heated plasmas of Heliotron E. The discharge behaviour, pellet ablation and wall conditioning were studied. The electron pressure is doubled after injection into the NBI plasma and remains unchanged in the case of ECR heating. This may be due to the energy exchange between the electrons and thermal ions with the fast ions from the neutral beam. The observed discrepancy between the experimental and modelled ablation rates may be caused by both the plasma cooling due to pellet ablatant and the ablation stimulated by the fast ions in the NBI-heated regime and by the fast electrons in the ECR-heated regime. In preliminary experiments on wall conditioning by Li pellet injection, no improvement of plasma performance after Li pellet injection was observed in the divertor or limiter configuration, with the limiter radii .

A device for producing a tracer-encapsulated cryogenic pellet is constructed for an accurate transport diagnostic system to measure particle transport both parallel and perpendicular to the magnetic-field lines in magnetic confinement devices. As for the typical configuration of the tracer-encapsulated pellet, it is proposed that a 50–250 μm diam tracer made of a light atom such as lithium, carbon, etc., is encapsulated in the center of a 1–3 mm diam cylindrical pellet of hydrogen. For demonstration of the device operation, a 240 μm diam carbon sphere is encapsulated in the center of a 3 mm diam cylindrical pellet of hydrogen, and accelerated by a light gas gun to velocities of 400–800 m/s in a test chamber. The pellet has been photographed simultaneously from two directions and the two two-dimensional images are reconstructed to a three-dimensional image. Thus, the proof of principle of the device operation has been demonstrated.

The Heliotron E device is characterized by a large aspect ratio, , large rotational transform, and , high shear, , an and M = 19 helical coil, two vertical field coils and 19 toroidal coils (19 toroidal coils were removed to increase neutral beam injection (NBI) heating power up to 4 MW in 1985), where is the rotational transform, is the average minor radius, R is the major radius, is the pole number and M is the helical period number of the magnetic field. Theoretical characteristics of Heliotron E are summarized in section 2. Initial experiments in 1980 - 3 were mainly for studying ohmically heated plasmas. Topics relating to current-carrying plasmas are reviewed in section 3. In Heliotron E the first currentless plasma was produced by ECRH using a gyrotron with a frequency of 28 GHz in 1982. Then the gyrotron frequency was increased to 53 GHz to produce higher density currentless plasmas. After NBI heating of the target plasma produced by ECRH was successful, currentless plasmas were studied intensively in Heliotron E. In section 4, the production and heating of currentless plasmas by ECRH, NBI and ion cyclotron range of frequency (ICRF) are summarized. MHD properties of currentless Heliotron E plasmas are summarized in section 5. Confinement and transport are discussed for ECRH, NBI and ICRF plasmas in section 6. Characteristics of edge plasma and plasma - wall interaction in Heliotron E are summarized in section 7. Configuration improvement and future directions based on Heliotron E results are discussed in section 8.

Ignition characteristics in the D-T high field helical reactor of R=20 m, a=2 m are analyzed by the `operation path method` on the P{sub ht}{tau}{sub E}{sup 2}-T plane and `POPCON`. Based on the LHD scaling law, the strong magnetic field B{sub o}=12 to nearly 15 T makes a reactor operable with a low confinement enhancement, leading to the low helium ash accumulation. POPCON analyses with the synchrotron radiation losses, the density limit and confinement time saturation effect show that marginal confinement factors to have the fusion power of 3 GW are 0.9 to 1.3 for B{sub o}=15 T, and 1.1 to 1.6 for 12T, respectively. 11 refs., 5 figs.

Recent results of Heliotron E experiment are reviewed. Detailed studies of the plasma transport have been achieved in Heliotron E, in which the plasma wall interaction has been changed using the ECH boronization technique. The boronization realized low-density, high-T{sub i} plasmas whose ion collisionality deeply estended to the 1/{nu}{sub i}** regime. The local electron and ion thermal diffusivities in the NBI plasma are analyzed and discussed by comparison with theory-based transport models. The particle transport in ECH plasmas is investigated with edge fluctuation measurements. Both electrostatic and magnetic fluctuations are well correlated with the global particle confinement. Characteristic structures of the electric field or potential distribution are observed during the particle confinement degradation phase. It is considered that dc convective flow across the last closed flux surface is caused by the potential structure. Edge plasma flows have been discussed in connection with these studies and with asymmetric divertor heat/particle load. 10 refs., 9 figs.

A new electron cyclotron heating (ECH) system has been installed on the Heliotron-E helical device. The gyrotron output 106 GHz TE 12,2 mode is converted into a gaussian beam by the quasi-optical Vlasov convertor. The gaussian beam is coupled to the HE 11 waveguide mode, then transmitted by the corrugated waveguides. The transmitted beam is well focused and launched to the Heliotron-E device. Plasma has been produced and heated by this 106 GHz ECH. The second harmonic breakdown was investigated and compared with the fundamental harmonic breakdown. The improvement of the particle confinement was observed during 106 GHz second harmonic ECH. Compared with the 53 GHz fundamental ECH phase, the averaged electron density was increased by 20% without gas puffing. The outflux to the divertor plate was reduced.

We have measured extreme ultraviolet (EUV) spectra in the 20 nm region from tungsten ions in plasmas produced in the Large Helical Device (LHD) at the National Institute for Fusion Science. The spectra are recorded by a grazing incidence spectrometer following an injection of a tungsten pellet. A quasicontinuum feature has been observed in the 20 nm range under low temperature conditions below 1.0 keV, where large contributions of charged states lower than W27+ are expected. In order to give an insight into this spectral feature, we have carried out atomic structure calculations for W7+-W27+ using Cowan's code. The calculated distributions of wavelengths and gA values suggest that the emission largely arises from n = 5-5 transitions in stages lower than W27+.

Extreme ultra‐violet (EUV) emission spectra from highly charged tin, xenon and tungsten ions have been measured in optically thin high‐temperature plasmas produced in the Large Helical Device (LHD) at the National Institute for Fusion Science by using a grazing incidence spectrometer and a tracer‐encapsulated solid pellet (TESPEL) injector. Quasi‐continuous spectral features arising from unresolved transition array (UTA) of open 4d subshell ions were observed for tin, xenon and tungsten around 13.5, 11, and 5 nm, respectively, when edge plasma was cooled sufficiently. The spectral appearance obviously depends on edge electron temperature and atomic number. In the case of intermediate edge temperature, sharp discrete lines from highly charged open 4s or 4p subshell ions are clearly observed for tin and xenon in the longer wavelength side of the UTAs but not the case for tungsten. Assignments of the strong discrete lines have been performed by comparison with other experimental data and the theoretical calculations generated using the Cowan code. In order to interpret the tungsten spectra it was necessary to include the contribution of open 4f subshell ions.

The authors present a neutral gas shielding model including the distribution function of particles carrying the heat for ablation. In an ohmically heated tokamak, high energy electrons in the tail part of the Maxwellian with E approximately=6-8 Te contribute significantly to the ablation process at the pellet surface. A simple scaling for the pellet ablation rate is given which takes into account the Maxwellian electron energy distribution. The pellet penetration depth obtained with the model presented agrees well with the multigroup approximation for an energy distribution during ablation by Houlberg et al. (1988). Pellet ablation dominated by hot ions was obtained in a special NBI heated Heliotron E plasma with an electron temperature much lower than the ion temperature

A comparative study on edge plasma properties between the original “natural” divertor configuration and the material limiter configuration was performed for Heliotron E currentless plasma. By insertion of a carbon rail limiter, the particle and heat fluxes to the divertor footprint were reduced to 20–50% of that in the divertor configuration. Such a reduction of influx resulted in the decrease of metal impurities sputtered from the wall surface. There were no clear differences in the edge density and heat flux profiles for NBI plasma between the limiter and divertor scrape-off. In the limiter configuration, a highly heat loaded area was observed locally on only one side of the limiter head. A perturbation field affected this local heat load.

The reduction of metallic impurity was found in a carbon limiter configuration in Heliotron E. In order to investigate the impurity penetration into the plasma, the distance between the outermost magnetic surface and the vacuum vessel wall was varied by inserting a carbon rail limiter. The analysis of vacuum ultraviolet impurity spectra and charge exchange neutral fluxes shows that the decrease of the impurity production rates by charge exchange neutral particles is the most probable reason to explain the experimental results. Moreover the electron temperature and density profiles in the limiter shadow provide an ability to screen metallic impurity influxes from the wall.

The optimum confinement of plasmas produced by electron cyclotron heating (ECH) in Heliotron E and some other significant transport characteristics of ECH plasmas have been studied experimentally by changing the magnetic field configuration. It is found that the plasma confinement can be improved by minor changes of the magnetic field configuration. The main results are as follows: (1) Plasmas with a minor radius reduced (by means of a limiter) to two thirds of the value in the original (non-limited) configuration have optimum confinement properties when the magnetic axis is shifted inward by 20 mm. (2) The electron pressure in an optimized plasma configuration (confirmed by neutral beam injection) is higher than that in a standard ECH plasma. (3) The insertion of a carbon limiter causes the plasma line density to increase, with no significant degradation of the core temperature or density. (4) Magnetic field perturbations hardly affect the temperature and density profiles, unless the magnetic axis is shifted outward by a large (50 mm) distance. (5) Transport analysis indicates that significant improvement in the confinement properties can be achieved by slightly manipulating the magnetic field configuration.

A two‐photon‐excited laser‐induced fluorescence (LIF) was applied to measurement of atomic hydrogen densities in the edge region of high‐temperature plasmas of Heliotron E for the first time. In this application, fluorescence signals were observed coaxially to the laser beam. This was possible because, in two‐photon‐excitation LIF, the visible fluorescence is localized at the laser focus and stray light can be completely eliminated by optical spatial filters, allowing much easier optical access than in previous LIF experiments. The densities at the initiation of the discharge and at the quasisteady state of the discharge were determined to be 3×1016 and 1×1016 m−3, respectively within an error of a factor of 2. Optimization to maximize the signal‐to‐noise ratio of the observation and resulting detection limit were discussed and confirmed the above results.

Parametric scaling studies of radiofrequency and neutral beam heated currentless Heliotron E plasmas have been performed. The parametric local electron transport analyses show that the electron energy transport in electron cyclotron heating (ECH) plasmas is nearly of the same magnitude as the neoclassically predicted transport inside the 2/3 radius, while neutral beam injection (NBI) plasmas and plasmas in the ion cyclotron range of frequencies (ICRF) are dominated by anomalous electron transport in the entire region. Scaling studies on the global energy confinement time reveal that ECH, NBI and ICRF plasmas obey approximately identical scalings that are characterized by continuous power degradation and favourable positive density dependence. The global energy confinement time is thought to be affected by the anomalous transport in the peripheral plasma regions in the same way for ECH, NBI and ICRF plasmas although the core plasma properties – such as local electron transport – seem to be different for the individual plasmas.

The magnetic configuration of the Heliotron-E device was controlled by adding auxiliary toroidal and/or vertical fields. The changes in the plasma edge and the “divertor trace” were experimentally studied by using a double probe, calorimeters, a thermal Li-beam probe and a laser Thomson scattering system. The variation of the vacuum configuration was also confirmed by the “stellarator diode method” with a small hot cathode. It was found that the observed change in the edge region basically agreed with that expected from the line-tracing calculation.

Experiments involving plasma improvement, confinement scaling, bootstrap currents, and edge fluctuations have been carried out in the Advanced Toroidal Facility (ATF) torsatron [Fusion Technol. 10, 179 (1986)]. Average densities ne≤9×1019 m−3 have been obtained, with global energy confinement times τ*E≤20 msec. Confinement times generally follow the stellarator/torsatron empirical scaling law, τSL =0.17×P−0.58n0.69eB0.84a2R0.75 (with τSL in seconds, power P in megawatts, density ne in 1020 m−3, and plasma radius a and major radius R in meters). Gas injection during neutral beam injection (NBI) causes increases in ne, so that τ*E does not decrease during NBI. Edge plasma fluctuations are found to exhibit a mode change near the peak of the energy confinement time. Plasma currents observed during electron cyclotron heating have been identified as bootstrap currents.