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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.

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.

Large helical device, the largest superconducting stellarator, has been operated for the research of fusion plasma since 1998. The toroidal field of almost 3 T is produced by a pair of pool-cooled helical coils, in the innermost layers of which a normal-zone had been induced several times at the bottom of the coil at higher currents than 11.0 kA. Since the field is not the highest there, the local cooling conditions are probably deteriorated by bubbles gathered by buoyancy. In order to improve the cryogenic stability by subcooling, an additional cooler with two-stage cold compressors was installed at the inlet of the coil in 2006. The inlet and outlet temperatures of the coils were successfully lowered to 3.2 K and 3.8 K, respectively, with a mass flow of 50 g/s. In spite of a half charging rate to reduce AC losses, a normal-zone was induced near the top of the coil at 11.45 kA. It propagated to one side and stopped near the inner equator, where the field is the highest. In comparison with the stability tests with a model coil, the local temperatures of the innermost layers near the top is considered to have been raised up to almost the saturated temperature of 4.4 K by charging. The excitation method was revised to waiting cool-down at 11.0 kA, and the excitations up to 11.5 kA have been attained.

The impurity transport properties in the ergodic layer of LHD are analyzed by the 3D edge transport code as well as 1D simple model, which is used to illustrate the essential transport terms in the analysis. It is found that as the plasma density increases the edge surface layers, very edge region of the ergodic layer, enters friction dominant regime, resulting in impurity retention. It is considered that the cause for the retention is both temperature drop and the flow acceleration in the edge surface layers. The edge surface layers can provide effective retention of impurities coming from divertor as well as the first wall, because of the geometrical advantage of the edge region of LHD. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

"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.

The divertor performance of LHD is studied for the two configurations, LID and HD. It is shown that the both divertor configurations play important roles for obtaining high performance plasmas in LHD : the large pumping capability of the LID to keep the low edge density in the IDB-SDC plasma, the large wetted area and the flexibility of strike point sweep of HD to reduce the power load on the divertor plates in long pulse operations. The possible effect of the ergodic layer on impurity retention in divertor is discussed by using the 3D edge transport modelling. It is found that the drag force exerted by the plasma flow can dominate over the thermal force, providing the impurity retention effect. The further changes needed to improve the current divertor configurations are discussed. New divertor designs for the future upgrade of LHD and for a LHD-type reactor are presented.

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 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.

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.

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.

This paper is devoted to the analysis of the new physics mechanism of the removal of cold alpha particles from the fusion helical reactor. Earlier, we have proposed some mechanisms, such as the drift island motion and estafette of drift resonances to remove the test particles, namely the cold alpha particles from the helical device of a LHD type. In that time, we kept in mind the cold alpha particles with the energy W = 350 keV. Now, we concentrate our study on the removal of the helium ash, i.e. cold alpha particles with the energy W = 35 keV while the main plasma ions are confined.

One of the characteristics of the heliotron-type magnetic configuration is that it has an intrinsic divertor structure (helical divertor). Particle control using a helical divertor configuration, to achieve improved confinement and sustainment of steady-state high-performance plasmas, is a major experimental goal in the Large Helical Device (LHD), the largest heliotron-type superconducting device, and it needs to be demonstrated on the route to the design of the heliotron-type fusion reactor. The LHD scrape-off layer (SOL) in the intrinsic helical divertor configuration has a unique magnetic field line structure consisting of stochastic regions, residual islands, whisker structures, and laminar layers contrasting with the "onion-skin"-like magnetic field line structure in poloidal divertor tokamak SOLs. Since the first experimental campaign in LHD in 1998, studies aiming at understanding the edge plasma properties in the "open" helical divertor configurations have been conducted experimentally and theoretically. In this paper, the helical divertor studies in the LHD are reviewed, and the future experimental plan is shown.

A repetitive pellet injector has been developed and pellet refuelling experiments have been launched on the Large Helical Device (LHD) in a new regime characterized by high density steady state operation. In order to suppress the disturbance to the core plasma as much as possible, we employ relatively small and slow pellets. The size and velocity of the pellets are typically 2.5?mm in diameter and 250?m?s?1, respectively. The pellets were injected into neutral beam heated plasmas with repetition frequency of 10?Hz. The density rise per pellet is 1.5 ? 1019?m?3 and the pellet penetration depth, which is normalized by the plasma minor radius, is typically less than 0.5. Quasi-steady-state operation was achieved at plasma parameters of , Te = Ti = 1.0?keV. Repetitive pellet fuelling showed improved energy confinement compared with massive gas puff fuelling in the high density regime in common with previous transient pellet injection experiments, and it can sustain such an improved confinement property.

The divertor/SOL transport of the Local Island Divertor (LID) configuration in the Large Helical Device (LHD) is studied. Since the LID configuration has an inherent three dimensional magnetic structure, the 3D edge transport code, EMC3-EIRENE, is applied in order to elucidate the 3D divertor transport physics. The main observations reported here are: 1. The effective temperature cooling due to the integrated perpendicular energy loss along the long connection length. 2. The significant plasma flow acceleration caused by the upstream source that is attributed to the geometrical significance of the LID configuration. 3. The effect of the flow acceleration on the impurity screening considering a parallel transport along the SOL.

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.

Steady state divertor operation with high performance plasmas (ne ~ 0.7 × 1019 cm−3, Ti ~ 2 keV) was demonstrated for half an hour in the Large Helical Device (LHD), the superconducting helical device (R = 3.6–3.9 m, a = 0.6 m, B = 3 T, l/m = 2/10). The high performance plasmas have been sustained with an averaged heating power of 680 kW and achieved an injected energy of 1.3 GJ. This required both advanced technological integration of heating systems and divertor heat flux control. In particular, optimization of divertor heat flux distribution along the divertor leg trace on divertor plates and real-time magnetic axis sweeping (R = 3.67–3.7 m) have allowed LHD to access a steady state regime with a margin of safety for the actively cooled divertor plates. The distribution of divertor heat load along the traces was investigated with calorimetric measurements and it was found that there was a localized heat load connected with the loss of high-energy ions produced by ion cyclotron radio frequency near-fields. Orbit analysis shows that the behaviour of high-energy ions is qualitatively in good agreement with the experimental result. Long-pulse discharges were terminated by radiation collapse due to penetration of metallic flakes into the plasma.

Self-sustained detachment has been obtained in the Large Helical Device (LHD). Strong hydrogen gas puffing of ~200 Pa m3 s−1 after a density feedback phase detaches the plasma from the divertor plates with high reproducibility. High electron density of over 1 × 1020 m−3 is sustained without gas puffing until the heating beam stops and a high-density flat top for 2 s has been demonstrated. Throughout the self-sustained detachment phase, the minor radius of the hot plasma column shrinks to ~90% of the last-closed-flux-surface, which corresponds to the rational surface. This new state has been named the 'Serpens mode', for self-regulated plasma edge 'neath the last-closed-flux-surface. Global energy confinement of the Serpens mode is compared with the international stellarator scaling 1995 (ISS95) and the recently established scaling for high-density LHD plasmas (HD scaling), where shrinking confinement volume and shallow penetration of the heating beams are taken into account. Although the energy confinement of the Serpens mode seems deteriorated compared with ISS95, as in the case of high-density attached plasmas, it is consistent with the HD scaling. This suggests that the energy confinement properties of detached plasmas in LHD are similar to those in high-density attached plasmas.

In the Large Helical Device, the structure of localized oscillations at the m/n = 2/1 rational surface during neutral beam (NB) injection in the counter direction (counter-NB injection) has been investigated in detail using an electron cyclotron emission (ECE) diagnostic. Here m and n are poloidal and toroidal mode numbers, respectively. Under the condition of increasing plasma current in the counter direction and nearly constant pressure, the oscillations are observed at larger magnitudes of plasma current (−20 kA T−1). In the NB reversal experiments, where the direction of the NB is changed from the co- to the counter-direction and vice versa, the oscillations were observed only for the former case. These results suggest that the profile and magnitude of the plasma current play an important role in the mode onset. Analysis based on the Mercier criteria shows that for a flat or a hollow current profile an interchange mode becomes more unstable at higher negative plasma current, showing a tendency similar to the experimental observations. In addition, radial displacement due to the localized oscillations evaluated from electron temperature perturbation profiles with the ECE diagnostic agrees well with that predicted by a three-dimensional ideal magnetohydrodynamic stability code.

The hydrogen particle balance of the plasma-wall system in the large helical device (LHD) is analyzed, using a zero dimensional model for plasma particles, neutrals in vessel and hydrogen inventory in wall. Based on the measurement of neutral gas pressure, plasma density and the pumping speed of the cryo-pumping system, it is found that the hydrogen retained in the wall desorbes with short and long time constant. The short term desorption is of order of 10 21 atoms with a time constant of a few minutes, which is much smaller than the wall pumping for one shot, 10 22 atoms. In a long time scale of about one experimental day, the wall absorbs significantly large amounts of hydrogen, up to 10 24 atoms. One of the possible reasons for the large wall pumping is a carbon deposition layer on the first wall surface. The effect of hydrogen retention on density control is also discussed.

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).

Density and power scan experiments have been carried out at various magnetic field strengths on the Large Helical Device (LHD), to investigate the temperature and magnetic field dependences of the thermal diffusivity. In the moderate density regime, the thermal diffusivity shows gyro-Bohm-like parameter dependences. In the high-density and low-temperature regime, on the other hand, the thermal diffusivity increases with the square root of the local electron temperature and decreases with a 1.2 power of the magnetic field strength. The turning point temperature where the weak temperature dependence changes to the gyro-Bohm type is also found to increase with a 0.8 power of the magnetic field strength. Based on the experimental results, dimensionally correct scalings for the energy confinement time, τE, and the plasma stored energy for the high-density LHD plasmas have been derived. Compared with the gyro-Bohm model (τE ∝ (n/P)3/5, where n and P denote the density and the heating power, respectively) and the international stellarator scaling 1995 (τE ∝ n0.51P−0.59), our scaling has a weaker positive dependence on the density together with the mitigated power degradation as τE ∝ (n/P)1/3, owing to the weak temperature dependence of the thermal diffusivity.

In the Large Helical Device (LHD), two different divertor configurations, i.e. helical divertor (HD) and local island divertor (LID), are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.

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.

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."

A local island divertor (LID) experiment has begun on LHD, with the aims of controlling edge recycling and improving the plasma confinement. The fundamental divertor functions of the LID have been demonstrated in the recent experiments. From the particle flux profile measurements on the LID head it was found that the particles diffusing out from the core region are well guided along the island separatrix to the LID head. Owing to the closed configuration around the LID head, evidence of the high efficient pumping was observed, together with a strong capacity to screen impurities. The first results of edge modeling using the EMC3-EIRENE code are also presented.

Helium retention behavior of 316L stainless steel (316L SS) used for the first wall of the large helical device (LHD) was investigated by using an ECR plasma/ion apparatus and thermal desorption spectroscopy. Samples of 316L SS were irradiated by helium ions with different energy. After helium ion irradiation with 100s eV energy, the desorption temperature of the retained helium was below 800 K. As increasing ion energy, the fraction of helium desorbed temperatures >800 K gradually increased. The retained amount of helium saturated for ion fluences above 1.0 × 1017 He/cm2. The retained amount of He was similar to that observed in other plasma facing materials. For irradiation fluences >1.0 × 1017 He/cm2, blister formation was observed. The present results are consistent with the desorption behavior of 316L SS samples exposed to LHD plasmas.

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.

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.

Confinement of energetic ions from neutral beam injection heating is investigated by changing the magnetic field configuration of the Large Helical Device from a classical heliotron configuration t...

Impurity deposition and gas retention were investigated for material probes which were installed along toroidal and poloidal directions on plasma facing walls during the 4th experimental campaign in the large helical device (LHD). The carbon deposition was observed to be dominant on the surface of probes located far from the plasma. The retained amount of helium was also large in the probes located near the plasma, although that of hydrogen was large in the probe far from the plasma. The retained amounts of discharge gases and impurity gases were large near the anodes used for glow discharges, since the current density near the anode was large.

Helical coils of the Large Helical Device are pool-cooled superconducting magnets. The operating current is restricted below about 90% of the design current because a normal-zone has propagated dynamically at several times at almost the same current. In order to estimate the effect of lowering temperatures on the cryogenic stability, an R&D coil was made of the same conductor. The cryogenic stability of the R&D coil was examined in saturated and subcooled helium. A normal-zone was initiated by a heater inserted between the conductor and the layer to layer spacer. The propagation was detected by voltage taps. In saturated helium of 4.4 K and 0.12 MPa, the minimum current to begin propagation is 10.7 to 10.8 kA. It becomes higher at the lower temperature, and it exceeds 11.7 kA in subcooled helium of 3.5 K as a temperature inside the R&D coil.

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.

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.

The big asymmetry of divertor flows to symmetrically located collectors was observed when divertor plasma flow distributions (DPFD) were measured in Heliotron-E device. Several possible reasons of DPF asymmetry were discussed in previous publications but no one was singled out. In the present work, to clear up the plasma heating effects on DPFD asymmetry, the difference of the DPFDs at plasma heating phase and at an ‘afterglow’ phase of plasma discharge is analyzed for different experimental conditions. During the afterglow phase the plasma is in the regime of Pfirsch–Schluter diffusion when effects connected with locally trapped particles would be negligible. Besides, the inhomogeneity connected with the locality of plasma heating would disappear and all plasma parameters should be constant along the torus. Therefore, the asymmetry of DPFD observed at this phase of discharge was taken as a base for evaluation of the effects stipulated by the trapped particles.

For the third experimental campaign of LHD, graphite tiles were installed in entire region of divertor strike region. Several material probes of SS 316L and graphite were placed along the poloidal direction at #7 toroidal sector. These probes were extracted after the campaign, and impurity deposition on the probe and retention of discharge and impurity gases were analyzed by AES and TDS, respectively. Major impurity species deposited on the probe were C, Fe and O. The iron deposition rate per one main discharge was approximately a half of that in the second campaign. The carbon deposition was significantly large, and the carbon concentration was as high as 90 at.%. These results correspond to the reduction of metal impurity level in the plasma. The retention of discharge gas and impurities doubled compared to the case of the second campaign, due to the carbonized wall. Significant retention of helium was observed, which is one of the characteristics of the LHD wall. The retention of helium is due to charge exchanged He during He main discharge shots and implantation of He ion during He glow discharges.

A divertor head of local island divertor (LID) that utilizes an m/n=1/1 island was installed on the large helical device (LHD) after the fifth experimental campaign in 2001–2002. Up to the fifth campaign, the effect of the LID magnetic configuration on plasma performance was studied without the divertor head. It was found that the island itself deteriorates plasma parameters such as electron temperature, stored energy, etc., and that the ergodic layer around the natural helical separatrix plays a role on preventing impurities from penetrating into the core plasma.

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 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.

A 30 keV lithium beam probe system to measure edge density and its fluctuation has been developed on the large helical device (LHD) and has just started its operation. On a test setup relevant to LHD conditions, the equivalent beam current ∼0.1 mA was obtained without the unfavorable beam divergence, in spite of a long distance more than 6 m from the injector. In the preliminary experiments in LHD, it was found that the beam penetrates up to ρ∼0.8 (ρ:normalized minor radius). By using the electrostatic deflection plates coupled with an in situ beam position monitor, the probe system operation is available, without the magnetic field shield, under the flexible experimental conditions with various magnetic field configurations.

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

Intrinsic impurity behaviour in neutral beam heated LHD long pulse discharges with a discharge duration of 10 s was investigated. Spectroscopic and bolometric measurements showed a remarkable temporal increase of core radiation due to metallic impurities. Central impurity accumulation was found in a narrow plasma density region (e = 1–3×1019 m−3) for both density ramp-up discharges and constant density discharges. In the density ramp-up discharges, high-Z impurities were accumulated in the low-density region and diffused out from the core plasma in the high-density region. The impurity behaviour was compared with neoclassical predictions.

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.

Divertor plasma characteristics in the Large Helical Device (LHD) have been investigated mainly by using Langmuir probes. The three-dimensional structure of the helical divertor, which is naturally produced in the heliotron-type magnetic configuration, is clearly seen in the measured particle and power deposition profiles on the divertor plates. These observations are consistent with the numerical results of field line tracing. The particle flux to the divertor plates increases almost linearly with the line averaged density. The high-recycling regime and divertor detachment, which are observed in tokamaks, have not been observed even during high density discharges with low input power. Both electron density and temperature decrease with increasing radius in the stochastic layer with open field lines, and at the divertor plate they become fairly low compared with those at the last closed flux surface. This means the reduction of pressure along the magnetic field lines occurs in the open field line region in LHD.

Energetic tail formation by ion-cyclotron-range-of-frequencies (ICRF) heating was studied in the large helical device (LHD), by measuring high-energy charge-exchange atom spectra with a specially developed natural diamond detector, which views the LHD plasma perpendicularly. Comparison of measured effective perpendicular temperature of ICRF-driven H+ minority ions with the classical Stix model indicates that perpendicular ions with energy at least up to 150 keV are well confined in the centre region of the LHD. However, deviation from the classical prediction was observed for high-energy perpendicular protons in the outer region of the plasma, indicating ripple-induced transport and charge-exchange losses. With perpendicular fast atom spectrum and flux measurements, small yet notable differences were detected co-injected and counter-injected ICRF-driven beam ion confinement during combined neutral beam injection (NBI) and ICRF heating. These results can be explained by the differences in the orbit topology and ICRF-induced drift.

In the Large Helical Device (LHD), active edge plasma control using two types of divertor, helical divertor (HD) and the local island divertor (LID), is planned, respectively. The former is intrinsic in the heliotron-type magnetic configuration. In the latter case, large pumping efficiency is expected by making particle recycling localized toroidally using a divertor head and an externally produced m = 1, n = 1 island. In this paper, the divertor properties in the HD configuration are reviewed. The numerical result for design of LID is described. The results of the LID experiment performed in the Compact Helical System are also briefly reviewed.