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A global particle balance study has been investigated in the Large Helical Device (LHD) in which the first wall and the divertor tiles are made of stainless steel (SUS-316L) and carbon, respectivel...

The Large Helical Device (LHD) is a heliotron-type device employing large-scale superconducting magnets to enable advanced studies of net-current-free plasmas. The major goal of the LHD experiment ...

Spontaneous dynamics of magnetic islands depending on various plasma parameters in the Large Helical Device (LHD) are described. The structure of the magnetic island undergoes deformation during a ...

Transport characteristics of the stochastic magnetic boundary of the Large Helical Device (LHD) are investigated, based on three-dimensional Monte-Carlo Braginskii-type fluid model code, EMC3, coup...

Superdense plasma with a highly peaked electron density profile was obtained in the reduced recycling discharge in the Large Helical Device (LHD). During the discharge, a core region with a density as high as ∼5 × 1020 m−3 and a temperature of ∼0.8 keV is maintained by an Internal Diffusion Barrier (IDB) with a steep density gradient. In spite of such a high density at the core region, no serious impurity accumulation has been observed. According to the numerical calculation with the EMC3–EIRENE code, downstream parallel flow in the ergodic region plays an important role to screen the influx of impurities with its strong friction force. The divertor flux distribution and global recycling properties do not change so much as against the change in the dense core plasma.

"A high density regime with an internal diffusion barrier (IDB) has been extended to the helical divertor (HD) configuration in the Large Helical Device (LHD). Avoidance of the local enhancement of neutral pressure is necessary to enable IDB formation, which is consistent with earlier works by using the Local Island Divertor (LID) with efficient active pumping. The central pressure reached 1.3 times atmospheric pressure, where ne(0) = 6 × 1020 m?3 and Te(0) = 660 eV. The plasmas with an IDB are located in the plateau collisionality regime. The significant impurity effect has not been observed throughout the discharges in spite of the existence of a negative radial electric field. A central pressure limiting event is observed in the plasmas with an IDB using the HD. During this event which is referred to as the core density collapse (CDC), particles are flushed out from the core on the time scale of a few hundreds of microseconds. The suppression of the Shafranov shift by vertical elongation (κ) is effective to mitigate CDC. At κ = 1.2, the central β value is increased up to 6.6% at 1 T."

Material probes were installed at the inner walls along the toroidal and poloidal directions in LHD from the sixth to eighth experimental campaign in LHD. The boronization was three times conducted in each campaign. Gas species, H2, He, Ar and Ne, were employed in both main and glow discharges during the campaign. After each campaign, gas retention and boron deposition behavior was examined to clarify the change of wall surfaces and plasma wall interactions. The amounts of retained gases were observed to be large in the vicinity of anodes used for the glow discharges. The amount of retained helium was observed to be relatively large, only one or two orders of magnitude smaller than that of hydrogen. The amounts of retained argon and neon were about two orders of magnitude smaller than that of helium. The thickness of deposited boron was large in the vicinity of both anodes and gas inlets. The toroidal uniformity of the boron deposition increased to about 70% in the eighth campaign by adjusting the gas flow rate. Understanding for the plasma wall interactions in LHD was deepened by the present study.

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−3and 2 keV, respectively, at a 1.3 GJ discharge (#53776) and 0.4 × 1019 m−3and 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.

A new scenario to control transport of 'cold' ?-particles in a helical fusion device by changing poloidal field coil current during plasma discharge is proposed here. A way to enhance radial transport of ?-particles in the intermediate energy range is considered which relies on specific features of helical magnetic field. To remove helically trapped, cold ?-particles, a ?-induced change in the B/B0 modulation along the particle trajectory is employed. To remove passing particles resonant effects connected with the poloidal field coil currents are used.

In the Large Helical Device (LHD), the local island divertor (LID) experiment has been conducted to investigate how to improve the core confinement with edge plasma control. An advantage of the LID configuration over the intrinsic helical divertor is the high pumping efficiency with the closed divertor module (neutralizer and pumping duct) and high speed pump-system. The particle evacuation efficiency, evacuated particle number/fueled particle number, has been estimated in experiment. This efficiency depends on the relative position of the closed divertor module (divertor head) to m / n = 1/1 magnetic island’s outer separatrix corresponding to the divertor leg, and the efficiency is revealed to be 0.6–1. A new operational regime with high core density (>4 × 10 20 m −3 ) and a strongly peaked density profile has been discovered in the LID discharges. It is considered that a large effective pumping efficiency is necessary to enter this operational regime.

The transport properties of the edge region in LHD have been investigated in order to clarify the divertor/SOL function of the heliotron type device. The momentum loss, mainly through the friction of counter-flows induced by the ergodic field lines, breaks the pressure conservation along flux tubes. This results in no high recycling regime even at high density operation, ¯ n » 7 £ 10 19 m i3 . The momentum loss is found to be higher than the case of W7-AS. This is because of the higher ratio of perpendicular and parallel transport scale length, » 10 i4 , and of the momentum loss due to the friction of counter-flows, which is more eective in the ergodic layer than the island divertor. In the heliotron configuration, a large temperature drop from LCFS to divertor, an order of magnitude, is easily realized due to the long connection length in the ergodic layer. This is certainly a favourable feature for future reactor in terms of reduction of heat power load at the divertor.

Boronization has been performed in many fusion plasma devices as one of several effective wall conditioning techniques. In LHD and JT-60U, boronization is operated as a typical wall conditioning method. Using material probes stayed in LHD during an experimental campaign, characteristics of a boronized wall are discussed in each device. Due to the complex three dimensional structure of the vacuum chamber, the thickness of boron film on the material probes depends on the distance from the anodes for glow discharge in LHD. On the other hand, such dependence is not clear in JT-60U due to thick carbon layers deposited on the material probes. To consider long-term operations, a deposited thick carbon layer with high boron-oxide concentration cannot be eroded completely by glow discharges, this layer is expected to be a saturated capacity of oxygen gettering. To keep the capacity of oxygen gettering, an additional boron layer is needed on the thick deposited layer. Since thin deposited layer on boron-coated area of LHD is eroded by glow discharge and flesh boron layer appears, a capacity of oxygen gettering is kept during long-term experiments.

Change of H α intensity profiles depending on magnetic configurations is observed in the divertor plasma. It can be explained by the magnetic field line structures in the ergodic layer and the divertor legs. The behavior of neutral particles in the plasma periphery is investigated by a three-dimensional neutral particle transport simulation code which assumes that the distribution of the plasma flow onto the divertor plates corresponds to that of the strike points calculated by magnetic field line traces. Vertical H α intensity profiles and polarization resolved H α spectra are calculated by the simulation code including the effect of Doppler broadening, fine structure splitting and polarization of the H α emission, which agree well with the measurements in various magnetic configurations. It shows spontaneous formation of high neutral density in inboard side of the torus, which is independent of the magnetic configurations in LHD.

The performance of net-current free heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fuelling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an internal diffusion barrier (IDB) by a combination of efficient pumping of the local island divertor function and core fuelling by pellet injection has realized a super dense core as high as 5 × 10 20 m -3 , which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5% and a discharge duration of 54 min with a total input energy of 1.6 GJ (490 kW on average) are also highlighted. The progress of LHD experiments in these two years is overviewed by highlighting IDB, high β and long pulse.

The large helical device is a heliotron device with L = 2 and M = 10 continuous helical coils and three pairs of poloidal coils, and all of coils are superconductive. Since the experiments started in 1998, the development of engineering technologies and the demonstration of large-superconducting-machine operations have greatly contributed to an understanding of physics in currentless plasmas and a verification of the capability of fully steady-state operation. In recent plasma experiments, the steady state and high-beta experiments, which are the most important subjects for the realization of attractive fusion reactors, have progressed remarkably and produced two world-record parameters, i.e. the highest average beta of 4.5% in helical devices and the highest total input energy of 1.6 GJ in all magnetic confinement devices. No degradation has been observed in the coil performance, and stable cryogenic operational schemes at 4.4 K have been established. The physics and engineering results from the LHD experiment directly contribute to the design study for a D-T fusion demo reactor FFHR with a LHD-type heliotron configuration.

In order to choose the best conditioning methods to remove hydrogen or tritium from the wall, the ion cyclotron conditioning (ICC) has been tested in fusion devices. The ICC is operated under strong magnetic fields of a few tesla and it is one of the important potential methods for future devices with superconducting coils, such as ITER. The successful initiations of ICC in TEXTOR, Tore Supra, HT-7, AUG, W7-AS and JET using standard ICRF antennas demonstrate the feasibility of this alternative technique. In the large helical device (LHD), which is a current-less steady-state helical device with all superconducting coils, a new conditioning method, the ICC has been started in 2005. An initial experiment of long-term ICC during a strong magnetic field was successfully carried out in LHD. The removal rate depends on the wide energy distributions of high-energy particles accelerated by ICRF power, and controlling this distribution is also important.

An overview of conceptual design activities on the LHD-type helical reactor FFHR is presented, mainly focusing on optimization studies on the reactor size and the proposal of a long-life blanket. A major radius of around 15 m is the present candidate under the constraints of the energy confinement achieved in LHD, a maximum magnetic field around 13 T with a current density around 30 A/mm2 and a neutron wall loading around 1.5 MW/m2. R&D on super-conducting magnet systems of large scale, high field and high current-density are new challenging targets based on the LHD. The development of new design tools has been started aiming at establishing a virtual power plant (VPP) and a virtual reality system for 3D design assisting. Next design issues are mainly on engineering optimization of the first wall thickness, the detailed 3D blanket system, and unscheduled replacements of breeder blankets.

We have investigated a relation between the edge magnetic structure of the Large Helical Device (LHD) and the fluctuation properties of the ion saturation currents I sat measured by a Langmuir probe array (16 channels) embedded in an outboard divertor plate. The edge magnetic structure is calculated by KMAG and Watanabe codes. The fluctuation properties are analyzed based on the probability distribution function(p.d.f.) and complex Morlet wavelet decomposition. Large positive skewness was obtained from I sat measured at a low field side of a divertor leg and negative skewness was observed at striking point of the divertor leg, where the connection length of magnetic line of force is long. The fluctuation characteristics has very weak dependence of plasma density and its gradient. Waiting time statistics was analyzed to indicate that the waiting time statistics is not a Poisson process and there is no clear signature of SOC paradigm.

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.

In order to understand the applicability of superpermeable membranes for tritium recovery from fusion exhaust gas containing impurities, the influence of oxygen and carbon on the performance of Nb and V membranes was examined by laboratory scale experiments. No significant degradation due to oxygen was observed if the membrane temperature was sufficiently high, i.e. above 700 K, and the bulk oxygen concentration was below 1.5 at.%. It was also possible to avoid the influence of carbon deposition by the operation at high temperatures through the conversion of carbon deposited layer into carbide layer.

Boron–titanium (B–Ti) was deposited on stainless steel substrate by evaporation of titanium and boron followed by annealing. The boron concentration at the surface region was comparable with the titanium concentration. Thermal desorption spectrum of deuterium was obtained after deuterium ion irradiation at room temperature, and the amount of retained deuterium and desorption temperatures of deuterium were measured. For comparison, titanium (Ti) and boron (B) films were deposited on the stainless steel substrate and the thermal desorption spectra were similarly obtained. The amount of retained deuterium in B–Ti was approximately 1/3 or 1/2 of that in Ti or B, respectively. The peak temperatures in B–Ti were 470 and 620 K, which were lower than those in B, 500 and 720 K. These results suggest that the use of B–Ti may reduce fuel hydrogen retention and recycling in fusion experimental devices, compared with the case of boron.

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."

One of the main objectives of the LHD is to extend the plasma confinement database for helical systems and to demonstrate such extended plasma confinement properties to be sustained in steady state. Among the various plasma parameter regimes, the study of confinement properties in the collisionless regime is of particular importance. Electron cyclotron resonance heating (ECRH) has been extensively used for these confinement studies of the LHD plasma from the initial operation. The system optimizations including the modification of the transmission and antenna system are performed with the special emphasis on the local heating properties. As the result, central electron temperature of more than 10 keV with the electron density of 0.6 x 10$^{19}$ m$^{-3}$ is achieved near the magnetic axis. The electron temperature profile is characterized by a steep gradient similar to those of an internal transport barrier observed in tokamaks and stellarators. 168 GHz ECRH system demonstrated efficient heating at over the density more than 1.0 x 10$^{20}$ m$^{-3}$. CW ECRH system is successfully operated to sustain 756 s discharge.

New design approaches are proposed for the LHD-type heliotron D–T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R&D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated.

A three-dimensional neutral particle transport simulation code is applied to identify a source of the hydrogen outgas in a LHD long pulse discharge by ICRF heating. Radiation collapse induced by uncontrollable plasma density rise terminated the long pulse discharge. Toroidally non-uniform increases in the H α emission and divertor temperature were observed. A CCD camera detected local heating of a vertically installed divertor plate, which is consistent with the strike points of the trajectories of protons accelerated by ICRF waves. The neutral particle transport simulation with the proton trajectory analyses shows that outgas from the divertor plates heated by the protons is necessary for explaining the observed toroidal non-uniform distribution of the H α emission in the long pulse discharge.

The capability of a niobium panel to absorb atomic hydrogen was investigated for particle control application in divertors. The absorption rate (1.3 · 10 20 H/m 2 s) stayed almost constant up to a dose of 4 · 10 23 H/m 2 with no hydride formation occurring. The absorption panel can be easily regenerated. The absorbed hydrogen can be retained in the panel even after 48 h at room temperature. The absorption rate was found to decrease with panel temperature for T < 200 � C, remarkably dropping in the vicinity of room temperature. This is connected most probably with the formation of a multilayer impurity coverage. The multilayer carbon coverage deposited during the panel exposure in the divertor region of LHD decreased the absorption rate, but the panel properties could be completely restored by heating at 850 � C in spite of panel carbidization. � 2004 Elsevier B.V. All rights reserved.

In order to evaluate the retention of hydrogen and helium and the ion impact desorption in stainless steel, the hydrogen and helium glow discharge were alternately conducted in the glow discharge apparatus with a stainless steel liner. The retained amount of hydrogen or helium in the stainless steel wall was measured based upon a residual gas analysis. The wall temperature was taken in the range from room temperature to 473 K. The retained amount of helium was observed to be large, comparable with that of hydrogen. The helium ion impact desorption for the retained hydrogen was observed to be effective, and increased with the wall temperature. The fraction of desorbed hydrogen became approximately a half of the retained amount of hydrogen. The hydrogen ion impact desorption for the retained helium, however, was observed to be negligible small. The impurity desorption owing to the hydrogen or helium glow discharge increased with the wall temperature. For reduction of the surface impurities, the hydrogen glow discharge was more effective than the helium glow discharge. The present results are useful to understand the retention and desorption of helium and hydrogen in stainless steel wall.

In the low-density plasmas of the Large Helical Device, the shape of the electron temperature profile changes depending on the direction of the tangential neutral beam injection (NBI) when the magnetic axis position is inward-shifted at R = 3.50 m. Core flattening was observed in plasmas heated by counter-NBI. The electron thermal diffusivities in co-NBI and counter-NBI-heated plasmas are compared. The diffusivity becomes large at the central region in the case of counter-NBI. This result shows that the flattening in the electron temperature profile is not caused simply by a change in the power deposition only. Some magnetic fluctuations are seen during counter-NBI. On the other hand, it is a promising feature that the electron thermal diffusivity at the peripheral region does not increase with the heating power in co-NBI plasmas.

The electron internal transport barrier (ITB) is formed with centrally focused electron cyclotron resonance heating superposed on plasmas heated by neutral beam injection in the Large Helical Devic...

Long-pulse operation and high-energy particle confinement properties were studied using ion cyclotron range of frequency (ICRF) heating for the Large Helical Device. For the minority-ion mode, ions with energies up to 500 keV were observed by concentrating the ICRF heating power near the plasma axis. The confinement of high-energy particles was studied using the power-modulation technique. This confirmed that the confinement of high-energy particles was better with the inward-shifted configuration than with the normal configuration. This behavior was the same for bulk plasma confinement. Long-pulse operation for more than 2 min was achieved during the experimental program in 2002. This was mainly due to better confinement of the helically trapped particles and accumulation of fewer impurities in the region of the plasma core, in conjunction with substantial hardware improvements. Currently, the plasma operation time is limited by an unexpected density rise due to outgassing from the chamber materials. The temperature of the local carbon plates of the divertor exceeded 400 deg, C, and a charge-coupled device camera observed the hot spots. The hot spot pattern was well explained by a calculation of the accelerated-particle orbits, and those accelerated particles came from outside the plasma near the ICRF antenna.

A local island divertor (LID) experiment has begun in the Large Helical Device (LHD) to demonstrate improved plasma confinement, and fundamental LID functions were demonstrated in the sixth experimental campaign in 2002-2003. It was clearly shown that when an m/n = 1/1 island is generated by adding a resonant perturbation field to the LHD magnetic configuration, the particle flow is guided along the island separatrix to the backside of the island, where carbon plates are located on a divertor head. The particles recycled there are pumped out efficiently so that the line-averaged core plasma density is reduced by a factor of {approx}2 at the same gas puff rate, compared with non-LID discharges. Obvious improvement of the global plasma confinement was, however, not observed yet, because the discharge could not be optimized, due to a large amount of outgas from the divertor head to the core plasma. The size of the divertor head was found to be larger than the optimum one; hence, the core plasma impacted slightly on the core plasma-facing portion of the divertor head with which the core plasma was not expected to collide.

Characteristics of transport in electron internal transport barriers (ITB) and in the vicinity of a rational surface with a magnetic island are studied with transient transport analysis as well as with steady state transport analysis. Associated with the transition of the radial electric field from a small negative value (ion-root) to a large positive value (electron-root), an electron ITB appears in the Large Helical Device [M. Fujiwara et al., Nucl. Fusion 41, 1355 (2001)], when the heating power of the electron cyclotron heating exceeds a power threshold. Transport analysis shows that both the standard electron thermal diffusivity, chie, and the incremental electron thermal diffusivity, chieinc (the derivative of normalized heat flux to temperature gradient, equivalent to heat pulse chie), are reduced significantly (a factor 5 10) in the ITB. The chieinc is much lower than the chie by a factor of 3 just after the transition, while chieinc is comparable to or even higher than chie before the transition, which results in the improvement of electron transport with increasing power in the ITB, in contrast to its degradation outside the ITB. In other experiments without an ITB, a significant reduction (by one order of magnitude) of chieinc is observed at the O-point of the magnetic island produced near the plasma edge using error field coils. This observation gives significant insight into the mechanism of transport improvement near the rational surface and implies that the magnetic island serves as a poloidally asymmetric transport barrier. Therefore the radial heat flux near the rational surface is focused at the X-point region, and that may be the mechanism to induce an ITB near a rational surface.

In order to examine the behavior of hydrogen and helium during a glow discharge in Large Helical Device (LHD), simulation experiments using a glow discharge apparatus were conducted. The hydrogen and helium glow discharges were repeated alternately to the wall of a liner installed in the discharge apparatus, and then the retained and reduced gas amounts were measured. The amount of retained helium was approximately 1/4 of retained hydrogen. It was obvious that the retention of hydrogen or helium significantly depended on the discharge sequence. The reduction of retained helium due to hydrogen discharge was very small, whereas hydrogen retention was largely reduced by the helium discharge. The amounts of retained hydrogen and helium decreased with an increase in the liner temperature, whereas the amount of impurity desorption increased with the liner temperature.

he Large Helical Device (LHD) has been extending an operational regime of net-current free plasmas towardsthe fusion relevant condition with taking advantage of a net current-free heliotron concept and employing a superconducting coil system. Heating capability has exceeded 10 MW and the central ion and electron temperatureshave reached 7 and 10 keV, respectively. The maximum value of β and pulse length have been extended to 3.2% and 150 s, respectively. Many encouraging physical findings have been obtained. Topics from recent experiments, which should be emphasized from the aspect of theoretical approaches, are reviewed. Those are (1) Prominent features in the inward shifted configuration, i.e., mitigation of an ideal interchange mode in the configuration with magnetic hill, and confinement improvement due to suppression of both anomalous and neoclassical transport, (2) Demonstration ofbifurcation of radial electric field and associated formation of an internal transport barrier, and (3) Dynamics of magnetic islands and clarification of the role of separatrix.

Ion heating experiments have been carried out in the large helical device using ECH (82.5, 84.0, 168 GHz, ≤1 MW), ICRF (38.5 MHz, ≤2.7 MW) and NBI (H° beam: 160 keV, ≤9 MW). The central ion temperature has been observed from the Doppler broadening of Ti XXI (2.61 A) and Ar XVII (3.95 A) x-ray lines, which are measured using a newly installed crystal spectrometer with a charge-coupled device. Recently, in ECH discharges, on-axis heating became possible. As a result, a high Te(0) of 6–10 keV and a high ion temperature of 2.2 keV were obtained at ne = 0.6×1013 cm−3. A clear increment of Ti was also observed with the enhancement of the electron–ion energy flow when the ECH pulse was added to the NBI discharge. These results demonstrate the feasibility towards ECH ignition. A clear Ti increment was observed also in ICRF discharges at low density ranges of (0.4–0.6)×1013 cm−3 with appearance of a new operational range of Ti(0) = 2.8 keV > Te(0) = 1.9 keV. In low power ICRF heating (1 MW), the fraction of bulk ion heating is estimated to be 60% of the total ICRF input power, which means Pi>Pe. Higher Ti(0), up to 3.5 keV, was obtained for a combined heating of NBI ( Te(0), whereas the Ti(0) remained at relatively low values of 2 keV in H2 and He NBI discharges due to less Pi. The main reasons for the high Ti achievement in the Ne discharges are: (1) 30% increment of deposition power, (2) increase in Pi/ni (five times, Pi/niPe/ne, Pi

An internal transport barrier (ITB) was observed in the electron temperature profile in the Large Helical Device [O. Motojima et al., Phys. Plasmas 6, 1843 (1999)] with a centrally focused intense electron cyclotron resonance microwave heating. Inside the ITB the core electron transport was improved, and a high electron temperature, exceeding 10 keV in a low density, was achieved in a collisionless regime. The formation of the electron-ITB is correlated with the neoclassical electron root with a strong radial electric field determined by the neoclassical ambipolar flux. The direction of the tangentially injected beam-driven current has an influence on the electron-ITB formation. For the counter-injected target plasma, a steeper temperature gradient, than that for the co-injected one, was observed. As for the ion temperature, high-power NBI (neutral beam injection) heating of 9 MW has realized a central ion temperature of 5 keV with neon injection. By introducing neon gas, the NBI absorption power was increased in low-density plasmas and the direct ion heating power was much enhanced with a reduced number of ions, compared with hydrogen plasmas.

The superconducting machine LHD has conducted long pulse experiments for four years to achieve long-duration plasmas with high performance. The operational regime was largely extended in discharge duration and plasma density. In this paper, the plasma characteristics, in particular, plasma performance and impurity behaviour in long pulse discharges are described. Confinement studies show that global energy confinement times are comparable to those in short pulse discharges. Long sustainment of high performance plasma, which is equivalent to the previous achievement in other devices, was demonstrated. Long pulse discharges enabled us to investigate impurity behaviour in a long timescale. Intrinsic metallic impurity accumulation was observed in a narrow density window (2–3×1019 m−3) only for hydrogen discharges. Impurity transport study by using active impurity pellet injection shows a long impurity confinement time and an inward convection in the impurity accumulation window, which is consistent with the intrinsic impurity behaviour. The pulsed neon gas injection experiment shows that the neon penetration into the plasma core is caused by the inward convection due to radial electric field. Finally, impurity accumulation control with an externally induced magnetic island at the plasma edge was demonstrated.

Ergodicity in various vacuum magnetic configurations in large helical device was quantitatively estimated, using the Kolmogorov length as a measure. It is found that the edge electron temperature profile changes its gradient at the position where the ergodicity begins to increase several cm outside the last closed flux surface (LCFS). In the profiles no distinguished change of plasma parameters was seen at the LCFS, which suggests the existence of a region just outside the LCFS where the confinement performance is relatively as good as that in the closed region. The radial electron heat conductivity was also estimated using a simple energy balance equation and it was confirmed that the conductivity is strongly affected by the ergodicity. Using the relationship between ergodicity and electron heat conductivity, a new practical definition of the LCFS based on the experiments is proposed.

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 relation between edge and helical divertor plasmas in the Large Helical Device are presented, in particular electron density ne and temperature Te behaviors, by line-averaged density n e scan ( n e ∼1–8×10 19 m−3) in the two typical magnetic configurations with relatively ‘thin’ and ‘thick’ naturally stochastic boundaries, respectively. The relations appear to be different in these configurations. With increasing n e , detachment-like behaviors of ne and Te are observed in the thick stochastic boundary configuration. On the other hand, both ne and Te at the divertor plate are almost proportional to them at near the last closed flux surface independent of n e . The different characteristics of particle screening and radiation power profile are suggested as causes of the different edge–divertor relations in these stochastic boundaries.

In general, helium gas does not remain on vacuum chamber walls because of its small activation energy. However, outgassing of helium gas from the walls has been observed in the LHD plasma vacuum chamber with a very long time constant after helium glow discharge cleaning (GDC), and absorption of helium atoms has been observed in plasma discharge experiments using helium gas. The helium partial pressure before the daily experiments was determined only by the gas species of the last GDC. No dependence has been found between the helium partial pressure and the species of the fueling gas of the last plasma experiment fueling gas. Considering that the outgassing rate of the helium gas is almost the same each morning after He GDC, the retention of helium atoms in the wall after the GDC is almost at the same level. The concentration of helium atoms in the wall before the daily experiments is estimated. The outgassing rate after the GDC is 2×10−4 Pa m3/s and the concentration is 4.7×1016 atoms/cm2. These results are of the same order as in another experiments. During helium gas plasma experiments, about a half of the amount of the inlet gas disappears with the missing particles remaining in the wall. The stainless steel wall, which is saturated with He GDC, may still have the capacity to trap high energy helium atoms. However, the energy dependence of trapping helium atoms presently is not clear.

The time-of-flight-type neutral particle analyzer has an ability of horizontal scanning from 40 to 100° of the pitch angle. The information from the spatially resolved energy spectrum gives not only the ion temperature but also the information of the particle confinement and the electric field in plasmas. We have been studying the energy distributions at various magnetic configurations in the neutral beam injection (NBI) plasma. The spatially resolved energy spectra can be observed during long discharges of the NBI plasma by continuous scanning of the neutral particle analyzer. The shape of spectra is almost similar from 44° to 53°. However, the spectra from 55° are strongly varied. They reflect the injection pitch angle of the beam. The pitch angle scanning experiment during the long discharge of NBI plasma has also been made under the reversal of the magnetic field direction. NBI2 becomes counter injected with the reversal. We can easily observe the difference between co- and counter injections of NBI. During the electron cyclotron heating in the low-density plasma for the formation of the internal thermal barrier, large neutral particle increase or decease can be observed. The degree of the increase/decrease depends on the energy and the density. The reason for the variation of the particle flux is that the orbit of the trapped particle changes due to the electric field formed by the strong electron cyclotron heating.

Radial electric field (Er) properties in LHD have been investigated based on the neoclassical transport theory and have also been applied to LHD experimental results. The effects of the helicity of the magnetic configuration on the condition required to realize the electron root are examined. The larger helicity makes the threshold temperature lower for the same electron density. A higher threshold temperature is anticipated to be required in the plasma core region based on this fact and also due to the larger density there. This high electron temperature (Te) has been successfully obtained with a centerfocused ECH. There is a threshold for the ECH power to achieve a steep gradient of Te, and it seems to be qualitatively consistent with the transition of Er, at least in the sense that the abrupt increase of Te occurs after entering the anticipated electron root regime. These experimental results, consistent with those of analysis of the neoclassical ambipolar Er, indicate that the transition phenomena of Er in LHD are predominantly governed by neoclassical features.

Electron cyclotron emission (ECE) diagnostics have been developed in the Large Helical Device (LHD). The ECE is transmitted from the antenna to the spectrometer for 80 m by means of a corrugated waveguide system, which has a small transmission loss (˜30%) in LHD. Although the field angle changes from −40 to +30 degrees on the sight-line of the ECE antenna, it has been observed that the polarization of ECE, which is determined at the plasma edge in LHD, is identical for all frequencies. Thus, the electron temperature profile can be measured using the second harmonic X-mode of ECE. The ECE measured by the Michelson is calibrated with a hot source in order to reduce noise. The ECE temperature measurement agrees well with the Thomson scattering measurement over a wide range of electron densities. The radiometer is cross-calibrated to the Michelson, and is employed for the study of fast phenomena such as internal transport barrier (ITB) formation. Tangentially emitted ECE, in addition to the normal ECE, is investigated in LHD in order to explore the potential of ECE diagnostics.

Two significant problems that need to be solved for any future fusion device are heat removal and particle control. A very promising method to attack these problems in tokamaks and helical devices is the use of a divertor, providing a controlled interaction zone between plasma and wall. By carefully designing a divertor, conditions can be created in front of the divertor targets, which lead to a sufficient reduction of the power load on the targets by strong radiation redistribution. Any solution of course needs to allow for an energy confinement which is at least sufficient for the realization of a fusion reactor. Since energy confinement has been found to be strongly related to edge anomalous transport and edge plasma profiles, the ultimate aim is to find an integral solution which is optimum with respect to exhaust, heat load and energy confinement. Two different types of divertors are presently being investigated in helical devices: the `helical divertor' and the `island divertor'. So far divertor concepts have been investigated only in a few helical devices. Theoretical and experimental efforts have mainly concentrated on the suitability of divertor magnetic field structures, while detailed studies of the divertor plasma properties for the two types of divertor configurations have only recently begun. In the course of this exploration, a promising new high-density H-mode (HDH) plasma operational regime has been discovered on the Wendelstein stellarator W7-AS. It benefits from high-energy (up to twice the value of the International Stellarator Scaling ISS95) and low impurity confinement times, complemented by edge radiated power fractions of up to 90% in detached regimes. This allowed quasi-steady-state operation for up to 50 energy confinement times and so far was only constrained by machine operability.

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.

Experimental results are summarized for major helical devices in Japan: Large Helical Device (LHD), Compact Helical System (CHS), and Heliotron J. The LHD and CHS have planar magnetic axes, while Heliotron J has a nonplanar magnetic axis.The LHD, the largest superconducting device in the world, has the following machine parameters: major radius R of 3.9 m, average minor radius a of 0.6 m, magnetic field on axis B of 3 T, multipolarity l of 2, toroidal period number m of 10, and auxiliary heating power P of {approx}14 MW. The LHD achieved the maximum stored energy W{sub pdia} of > 1 MJ, the maximum value of the volume averaged beta 3.0% at B of 0.5 T, high confinement time {tau}{sub E} of 0.3 s at T{sub e}(0) of 1.1 keV and of 6.5 x 10{sup 19} m{sup -3}, and long pulse operations up to 120 s at high temperature.The CHS has the following machine parameters: R = 1 m, a = 0.2 m, l = 2, and m = 8. The parameters of neutral beam heated plasmas are in the range with of {approx}4 to 5 x 10{sup 19} m{sup -3}, T{sub e}(0)more » of 500 to 700 eV, and energy confinement time of several milliseconds. Progress has been made in studies of bifurcation phenomena of electric potential, neoclassical internal transport barriers, and plasma flows in the toroidal and poloidal directions.The most important achievement from recent CHS and LHD experiments is to realize the internal transport barriers on the basis of potential bifurcation.Heliotron J, which was converted from the Heliotron E (H-E) device, employs a helical magnetic axis by the use of l = 1 continuous helical coil and auxiliary coils such as poloidal and toroidal coils. The machine parameters are as follows: R = 1.2 m, a = 0.1 to 0.2 m, and B = 1 to 1.5 T. Initial results show the maximum stored energy W{sub p} {approx} 0.7 kJ and {approx} 0.2%.The range of plasma parameters has been greatly expanded by the LHD, CHS, H-E, and Heliotron J experiments. The confinement data in helical devices are scaled empirically as ISS-95 (International Stellarator Scaling), and plasma performance is comparable with that in tokamaks.« less