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Simultaneously oscillated 48-μm and 57-μm CH3OD lasers pumped by a 9R(8) CO2 laser have been developed for fusion plasma diagnostics. A back reflection from an optically pumped far-infrared (FIR) laser cavity affects not only the FIR laser stability but also the pump CO2 laser stability. The back reflection has been reduced by an optical isolator using a quarter-wave plate and an absorbing thin-film reflector. As a result, we have achieved output stabilities of ±1.4 %/h at the 48-μm laser and ±1.3 %/h at the 57-μm laser.

“Fusion virtual laboratory (FVL)” is the experiments’ collaboration platform covering multiple fusion projects in Japan. Major Japanese fusion laboratories and universities are mutually connected through the dedicated virtual private network, named SNET, on SINET4. It has 3 different categories; (i) LHD remote participation, (ii) bilateral experiments’ collaboration, and (iii) remote use of supercomputer. By extending the LABCOM data system developed at LHD, FVL supports (i) and (ii) so that it can deal with not only LHD data but also the data of two remote experiments: QUEST at Kyushu University and GAMMA10 at University of Tsukuba. FVL has applied the latest “cloud” technology for both data acquisition and storage architecture. It can provide us high availability and performance scalability of the whole system. With a well optimized TCP data transferring method, the unified data access platform for both experimental data and numerical computation results could become realistic on FVL. The FVL project will continue demonstrating the ITER-era international collaboration schemes and the necessary technology.

This paper presents a system called “TSMAP” that maps electron temperature profiles to flux coordinates for the Large Helical Device (LHD). Considering the flux surface is isothermal, TSMAP searches an equilibrium database for the LHD equilibrium that fits the electron temperature profile. The equilibrium database is built through many VMEC computations of the helical equilibria. Because the number of equilibria is large, the most important technical issue for realizing the TSMAP system is computational performance. Therefore, we use multiple personal computers to enhance performance when building the database for TSMAP. We use virtual machines on multiple Linux computers to run the TSMAP program. Virtual machine technology is flexible, allowing the number of computers to be easily increased. This paper discusses how the use of virtual machine technology enhances the performance of TSMAP calculations when multiple CPU cores are used.

This paper describes the interferometer systems on the Large Helical Device (LHD). LHD is equipped with five interferometer systems, each of which has a different operational purpose and me...

Particle confinement processes are studied in detail on the Large Helical Device (LHD). Diffusion coefficients (D) and convection velocities (V) are estimated from density modulation experiments. T...

The observation of a significant rise of the core electron temperature Te in response to edge cooling in a helical plasma was first made on the Large Helical Device (LHD). When the phenomenon takes...

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 ...

This paper reviews progress in the study of pressure-driven interchange stability in the Large Helical Device (LHD) for 10 years. When the plasma approaches the boundary of ideal interchange mode, a strong magnetohydrodynamic (MHD) mode appears, leading to a distortion of pressure profile, although no major disruption is caused. The experiments for investigating magnetic shear effects in the magnetic hill configuration indicate that the reduction of magnetic shear leads to a minor collapse due to an excitation of low-order MHD mode. In the high-beta regime of more than 4%, MHD modes excited in the periphery with magnetic hill are observed to dominate, and it was found that the amplitude depends on the magnetic Reynolds number as well as the pressure gradient, which is qualitatively consistent with the prediction of resistive interchange mode. Also, experiments and theory for finding parameter dependence of the onset of the mode indicate that the onset is related to both the magnetic Reynolds numbe...

The Large Helical Device (LHD) is the world’s largest heliotron-type device with l = 2, m = 10 continuous superconducting helical coils and three pairs of superconducting poloidal coils. The major ...

The magnetic configuration of the Large Helical Device (LHD) changes with the increment in beta. To distinguish between the beta effect and the configuration effect on the gradual degradation of th...

A broadband frequency tunable microwave reflectometer system, which has the ability of fast and stable frequency sweeping operation, is applied in the large helical device (LHD) to measure the boundary of the high-temperature plasma. When the microwave is launched, with extraordinary polarization, from the low field side of the magnetic field with its frequency swept from low to high, the microwave of minimum right-hand cut-off frequency is reflected most outside. We can then estimate the plasma boundary position by measuring the change of the reflected power from the cut-off layer.

Measurement of an internal magnetic field distribution in magnetically confined fusion devices is indispensable for both understanding the plasma physics and controlling plasmas. A polari-interferometer based on the Faraday effect has been used for such a purpose. This paper describes performance evaluations of a part of the polarimeter with photoelastic modulators (PEMs) of the short-wavelength far-infrared (FIR) laser polari-interferometer. The wavelengths of the light source are 57.2 and 47.7 μ m, which are suitable for a high-density operation and large fusion devices. The PEM for the FIR region is newly developed with high-resistive silicon in which absorption of the FIR laser is small. The polarization angle is successfully measured and an angle resolution of 0.01° with a time resolution of 1 ms is achieved. A drift of the baseline of about 0.1° for 1000 s is observed and is found to be caused by changes in the room temperature.

This article reviews 10 years of engineering and physics achievements by the Large Helical Device (LHD) project with emphasis on the latest results. The LHD is the largest magnetic confinement device among diversified helical systems and employs the world's largest superconducting coils. The cryogenic system has been operated for 50,000 h in total without any serious trouble and routinely provides a confining magnetic field up to 2.96 T in steady state. The heating capability to date is 23 MW of NBI, 2.9 MW of ICRF and 2.1 MW of ECH. Negative-ion-based ion sources with the accelerating voltage of 180 keV are used for a tangential NBI with the power of 16 MW. The ICRF system has full steady-state operational capability with 1.6 MW. In these 10 years, operational experience as well as a physics database have been accumulated and the advantages of stable and steady-state features have been demonstrated by the combination of advanced engineering and the intrinsic physical advantage of helical systems in LHD. Highlighted physical achievements are high beta (5% at the magnetic field of 0.425 T), high density (1.1 × 1021 m−3 at the central temperature of 0.4 keV), high ion temperature (Ti of 5.2 keV at 1.5 × 1019 m−3), and steady-state operation (3200 s with 490 kW). These physical parameters have elucidated the potential of net-current free helical plasmas for an attractive fusion reactor. It also should be pointed out that a major part of these engineering and physics achievements is complementary to the tokamak approach and even contributes directly to ITER.

Microwave to millimeter-wave diagnostic techniques such as interferometry, reflectometry, scattering and radiometry have been powerful tools for diagnosing magnetically confined plasmas. Recent advances in electronic devices and components together with computer technology have enabled the development of advanced diagnostics in microwave to millimeter-wave region. The two-dimensional and three-dimensional profiles of plasma density/temperature and dynamic behaviors of their fluctuations are measured by using the advanced diagnostic systems. The measurements have clarified the physics issues such as stability, wave phenomena and fluctuation-induced transport. We report here on the representative microwave diagnostics and their contribution to plasma confinement physics.

Simultaneously oscillated multi-line far infrared (FIR) lasers from a single FIR laser cavity have been developed as optical sources of a two-color interferometer for plasma diagnostics. Using two pump CO2 lasers which oscillated at the different wavelength, we have obtained simultaneously oscillated 119-μm laser pumped by 9P(36) CO2 laser and 57-μm laser pumped by 9R(8) CO2 laser from mixed gases of CH3OH and CH3OD. Common beam splitters of the two-color FIR laser interferometer have been designed by using precise optical constants (refractive index and absorption coefficient) of silicon.

"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."

The LABCOM data acquisition system (DAQ) has set a new world record of acquired data amount of 90 GB/shot in 2005?2006 campaign. Its CompactPCI-based new DAQ enables about 80 MB/s real-time acquisition for each plasma measurement. The basic performance of a DAQ unit has already good prospects for the next-generation fusion experiments, however, some surrounding utilities are not necessarily developed to construct 10 times larger DAQ system yet. The increasing number of parallel DAQ units has made the operational and maintenance burden very heavy. So, “more distributed acquisition and centralised operations” would become expected. Here, DAQ front-end consisting of some digitizer chassis, a timing module, and an acquisition computer has been entirely re-designed to realize a lower-cost and maintenance-free “DAQ box” for fusion plasma measurements. OS and softwares have been also remodelled to use free drivers on Linux. Network-bootable diskless computers are also desirable for reducing the possibility of hardware troubles, and it simultaneously realizes to omit first-tier data storage and reduce the migration operational cost. As a result, total hardware cost could be reduced almost one-fifth than the previous. This could be the way to realize 10 times larger fusion experiment in the near future.

The LABCOM data acquisition system (DAQ) has set a new world record of acquired data amount of 90 GB/shot in 2005–2006 campaign. Its CompactPCI-based new DAQ enables about 80 MB/s real-time acquisition for each plasma measurement. The basic performance of a DAQ unit has already good prospects for the next-generation fusion experiments, however, some surrounding utilities are not necessarily developed to construct 10 times larger DAQ system yet. The increasing number of parallel DAQ units has made the operational and maintenance burden very heavy. So, “more distributed acquisition and centralised operations” would become expected. Here, DAQ front-end consisting of some digitizer chassis, a timing module, and an acquisition computer has been entirely re-designed to realize a lower-cost and maintenance-free “DAQ box” for fusion plasma measurements. OS and softwares have been also remodelled to use free drivers on Linux. Network-bootable diskless computers are also desirable for reducing the possibility of hardware troubles, and it simultaneously realizes to omit first-tier data storage and reduce the migration operational cost. As a result, total hardware cost could be reduced almost one-fifth than the previous. This could be the way to realize 10 times larger fusion experiment in the near future.

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.

The primary facility for magnetic fusion energy (MFE) research in Japan is JT-60, in which DD plasma is heated up to an electron temperature of T e >10 keV with neutral beam injection (NBI) and radio frequency (RF) powers. It is noted that a normalized β N of 2.3 was maintained for 22 s. The JT-60 team has discovered internal transport barrier (ITB) in high-beta-poloidal discharges as an additional improved confinement mechanism like a previously known edge transport barrier (ETB). The operation forming the ITB has remarkably improved the plasma confinement. The second major MFE facility is a large helical device (LHD) at National Institute for Fusion Science (NIFS). The world's largest superconducting coil system of LHD generates helical magnetic field of 3 T where the stored magnetic energy reaches 1 GJ. A quasi-stationary plasma with an electron density of 4×10 18 m −3 at a temperature of about 1 keV was sustained for 54 min with the help of ion cyclotron range of frequency (ICRF) and electron cyclotron range of frequency (ECRF). The maximum electron density so far achieved is 5×10 20 m −3 with T e ∼1 keV. We report MFE programs related to International Thermonuclear Experimental Reactor (ITER).

The characteristics of particle transport in three different magnetic configurations are studied from density modulation experiments in the Large Helical Device (LHD). These three configurations are represented as different magnetic axis positions (R ax ) of the vacuum field. Experiments were carried out in a range of different heating powers for each configuration with almost constant density. The experimental values of particle diffusion coefficients (D) and particle convection velocities (V) are compared with neoclassical estimates. The value of D is found to be anomalously large compared to neoclassical values in both the core and edge in all configurations. At low collisionality, this anomaly tends downward. The core convection velocities are comparable with neoclassical estimates. In more-outward-shifted configurations, particle transport is enhanced. The electron temperature and electron temperature gradient are the determinate parameters for D and V, respectively, in each configuration. The effective helical ripple is one of the important parameters for particle transport in the LHD; however, other hidden parameters exist. The role of fluctuations in particle transport is investigated from turbulence measurements using a two-dimensional phase contrast interferometer. Three kinds of fluctuation having different locations, propagation direction, and peak wave number are observed. One of these, which exists in the outermost edge region and propagates in the ion diamagnetic direction in the laboratory frame, plays a possible role in edge anomalous diffusion. The amplitudes of ion diamagnetic fluctuation components are compared with the linear growth rate of the ion temperature gradient mode.

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.

Local transport properties of high-beta plasmas in the Large Helical Device are studied by comparing the beta dependence of the experimental results with that of the gyro-reduced Bohm-type transpor...

In local island (m/n = 1/1) divertor discharges in the large helical device a stable super dense core plasma develops when a series of pellets are injected. A core region with a density as high as 4.6 × 1020 m−3 and a temperature of 0.85 keV is maintained by an internal diffusion barrier with a very high density gradient. In a study of island dynamics, we find that an externally imposed large island (m/n = 1/1) as large as 15% of the minor radius is healed when beta at the island exceeds a critical value.

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.

The diagnostic system of JFT-2M has consisted of about 30 individual diagnostic instruments, which were used to study plasma production, control, equilibrium, stability, confinement, plasma heating...

In Large Helical Device (LHD), the discharges of high Z are often used in order to obtain high ion temperature. The ion density is relatively smaller than the electron density in high Z plasma. Therefore the high ion temperature can be obtained since the input energy per ion atom is large. In the charge exchange neutral diagnostic, the ion temperature can be obtained by observing the spectra of neutral particles, which are generated by the charge exchange between the background neutral and the plasma ion, and assuming theMaxwellian distribution of the spectra. In calculation, we also consider the charge exchange between the partially ionized high Z ion and proton. The contribution of neutral particle from the charge exchange between argon and proton is small near the center but cannot be neglected near the peripheral region comparing with that of hydrogen charge exchange

Phase counters, which are used with heterodyne interferometers for plasma density measurements, frequently suffer from phase jumping and cause difficulties for data interpretation. An automatic fringe jump corrector (AFJC) circuit has been developed to compensate for fringe jumps. The AFJC can correct phase jumps automatically in real-time. The AFJC, which is integrated on one chip, is installed on the presently working phase counter circuit. As for the specification of this phase counter the intermediate beat signal is 1 MHz, the phase detection range is 31 fringes with phase resolution of 1/80 of a fringe and the response time of 10 μs. The circuit has been tested on the far infrared (FIR) laser interferometer on LHD. The AFJC works fine to correct fringe jumps, when fringe jumps occurred due to the strong density gradient produced by the hydrogen pellet injection.

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.

Experiments in the large helical device have been developing since the first discharge in 1998. Baking at 95 °C, electron cyclotron resonance discharge cleaning, glow discharge cleaning, titanium gettering and boronization were attempted for wall conditioning. Using these conditioning techniques, the partial pressures of the oxidized gases, such as H2O, CO and CO2, were reduced gradually and the plasma operational regime enlarged. The glow discharge cleaning with the various working gases, such as hydrogen, helium, neon and argon, was effective in increasing the plasma purity. By this method, we obtained a central ion temperature of 10 keV. Boronization, which was started from FY2001, was also effective in reducing the radiation losses from impurities and in enlarging the density operational regime. We obtained a plasma stored energy of 1.31 MJ and an electron density of 2.4 × 1020 m−3.

Two modifications of imaging interferometry: heterodyne (HI) and phase contrast interferometers (PCI) are designed for observation of plasma density profiles and density fluctuations respectively. Besides, spatial distributions of plasma velocities, velocities fluctuations and related electrical fields can be obtained from the analysis of HI and PCI data. New sensitive phase counters, developed at Budker Institute of Nuclear Physics, enable HI to include some capabilities of the PCI. In addition to well recognized transversal spatial resolution of imaging technique, progress in deconvolution of line-of-sight-integrated data was recently made. Computer simulation, bench-test experiments and recent experimental results from the Large Helical Device illustrate the potentials of the imaging interferometry for investigation of plasma. Application of the imaging interferometry with spatial resolution along the viewing line to mirror machines is finally considered.

A powerful short-wavelength far-infrared (FIR) laser from 40µm to 100µm in wavelength is required for the optical source of diagnostics of high-density and large-volume plasmas, and a production of γ-rays by inverse Compton scattering. In order to design the optical system, we have measured the optical constants (refractive index and absorption coefficient) of crystal quartz, CVD-diamond, and silicon for the short-wavelength FIR lasers. © 2005 Wiley Periodicals, Inc. Electr Eng Jpn, 153(3): 1–8, 2005; Published online in Wiley InterScience (www.interscience.wiley.com). DOI 10.1002/eej.20082

Gyrotron FU II has been successfully applied as a submillimeter wave radiation source to plasma scattering measurements on the Compact Helical System (CHS) in National Institute for Fusion Science (NIFS) in Japan. The gyrotron operates in a long pulse mode (the pulse width is about 600 ms) at a frequency of about 350 GHz (the corresponding wavelength is 0.85 mm). The output power is about 110 W. The output power is transmitted along a circular waveguide system and converted to a Gaussian-like beam by a quasi-optical antenna. After that, the beam is directed onto the CHS plasma and the scattered signal is detected by a homodyne detection system. The frequency and the wave number of the scattered signal are analyzed. The results suggest that a broad band low frequency (several tens to several hundreds kHz) density fluctuation is excited in the CHS plasma only during neutral beam injection (NBI) or ion cyclotron resonance (ICR) heating.

A multichannel imaging heterodyne interferometer is installed on Large Helical Device. Two branches of the interferometer: one with CO2 laser (wavelength λi=10.6 μm) and other with diode pumped yttritium–aluminium–garnet (YAG) laser (wavelength λi=1.06 μm) are used for electron density measurements and vibration compensation. Two slab beams and one circular beam are injected vertically. Liquid nitrogen cooled linear detector arrays are used for detection of slab beams and room temperature detector is employed for single circular beam for CO2 beams. Avalanche photodiodes record signals for YAG beams. Chordal resolution is from 15 to 22.5 mm, phase resolution is 10−3 of CO2 fringe, which is determined by the electrical noise of phase counter. About 5×10−3∼10−2 of CO2 fringe uncompensated signal remains after vibration compensation. Combining existing 13-channels far infrared (wavelength λi=119 μm) laser interferometer, profile measurements at high density (>1020 m−3) become possible.

Electron cyclotron emission (ECE) diagnostics have been developed in the large helical device (LHD). The ECE is transmitted over 80 m from the antenna to the spectrometer with the corrugated waveguide system, which has a low transmission loss (/spl sim/30%). The polarization of ECE is identical for all frequencies, while the field angle varies from -40/spl deg/ to +30/spl deg/ on the ECE sight line. The ECE temperature measured by the Michelson, which is calibrated with a hot source, agrees well with Thomson scattering measurement in a wide range of the electron density. Magnetic island formation is investigated using the electron temperature profile measured with the Michelson. The radiometer is cross-calibrated to the Michelson and is employed for the study of fast phenomena, such as the sawtooth oscillation. Besides the normal ECE, tangential ECE is investigated in LHD in order to explore the potential of ECE diagnostics.

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.

Powerful lasers in the far-infrared wavelength range (47.6 and 57.2 μm) have been developed to measure the plasma density in the Large Helical Device at National Institute for Fusion Science and future plasma devices such as the International Thermonuclear Experimental Reactor. The intensification of these lasers has been done by cooling the laser tube wall, adding He as the buffer gas, and using a chemical-vapor-deposited diamond output window. The output powers for the 57.2 and 47.6 μm lasers have been found to be 1.6 and 0.8 W, respectively.

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.

Characteristics of MHD instabilities and their impacts on plasma confinement are studied in current free plasmas of the Large Helical Device. Spontaneous L?H transition is often observed in high beta plasmas close to 2% at low toroidal fields (Bt ? 0.75?T). The stored energy starts to rise rapidly just after the transition accompanying the clear rise in the electron density but quickly saturates due to the growth of the m = 2/n = 3 mode (m and n: poloidal and toroidal mode numbers), the rational surface of which is located in the edge barrier region, and edge localized mode (ELM) like activities having fairly small amplitude but high repetition frequency. Even in low beta plasmas without L?H transitions, ELM-like activities are sometimes induced in high performance plasmas with a steep edge pressure gradient and transiently reduce the stored energy up to 10%. Energetic ion driven MHD modes such as Alfv?n eigenmodes (AEs) are studied in a very wide range of characteristic parameters (the averaged beta of energetic ions, ?b?, and the ratio of energetic ion velocity to the Alfv?n velocity, Vb?/VA), of which range includes all tokamak data. In addition to the observation of toroidicity induced AEs (TAEs), coherent magnetic fluctuations of helicity induced AEs (HAEs) have been detected for the first time in NBI heated plasmas. The transition of a core-localized TAE to a global AE (GAE) is also observed in a discharge with temporal evolution of the rotational transform profile, having a similarity to the phenomenon observed in a reversed shear tokamak. At low magnetic fields, bursting TAEs transiently induce a significant loss of energetic ions, up to 40% of injected beams, but on the other hand play an important role in triggering the formation of transport barriers in the core and edge regions.

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.

Significant advances in microwave and millimeter wave technology over the past decade have enabled the development of a new generation of imaging diagnostics for current and envisioned magnetic fusion devices. Prominent among these are microwave electron cyclotron emission imaging, microwave phase imaging interferometers, imaging microwave scattering, and microwave imaging reflectometer systems for imaging Te and ne fluctuations (both turbulent and coherent) and profiles (including transport barriers) on toroidal devices such as tokamaks, spherical tori, and stellarators. The diagnostic technology is reviewed, and typical diagnostic systems are analyzed. Representative experimental results obtained with these novel diagnostic systems are also presented.