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Three-dimensional (3D) effects on the stochasticity of magnetic field lines in helical devices are investigated. The stochastization due to increased plasma pressure is an intrinsic property of the stellarator/heliotron because the pressure-gradient driven currents are the source of perturbation fields. The stochastization depends on the plasma pressure profile. In an advanced stellarator, the stochastization due to the 3D effects is smaller. However, magnetic islands in the edge are affected by 3D effects. 3D effects in a tokamak plasma are studied,too, and differences from the vacuum approximation are found (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

To understand the edge plasma transport in the complex magnetic field lines structure in the LHD heliotoron, the heat and particle fluxes profiles on divertor plates were investigated experimentally and theoretically by using EMC3-EIRENE code. The heat flux profiles on a divertor plate in different magnetic configuration were measured and the main channel of the heat transport was revealed to be the field lines with long connection lengths which connect the divertor plate and the stochastic field lines layer. A comparison of the heat and the ion fluxes profiles on the divertor plates between the experimental data and the results of the calculation using EMC3-EIRENE code was conducted. It was revealed that the change of the cross-field transport in the edge plasma region causes not only the change of the heat and particle flux profiles on divertor plates, but also the changes of the intrinsically non-uniform global heat and particle deposition profiles on the helical diveror. The cross-field transport seems to be larger in the higher edge electron temperature discharge in the experimental conditions in this paper (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

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)

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.

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.

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.

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.

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.

A rapid bifurcation of the equilibria with and without a large island (n/m = 1/1) has been observed in the medium to high beta large helical device discharges. A large island imposed by an external resonant field is suddenly suppressed nearly perfectly by the plasma effects when the beta at the ι/2π = 1 surface exceeds a critical value. The critical beta value is nearly proportional to the externally imposed resonant field normalized by the main field strength.

Magnetic island formation is experimentally investigated in the Large Helical Device. The (m, n) = (1, 1) vacuum magnetic island is generated by using the local island diverter field, where m and n are the poloidal and toroidal mode numbers, respectively. The island width depends on plasma parameters (the electron temperature and the β) and the magnetic axis position. In the case of Rax = 3.53 m the magnetic island in the plasma is larger than that in the vacuum field. Here, Rax is the major radius of the magnetic axis. In the case of Rax = 3.6 m the magnetic island is not generated when the error field is less than the threshold which is increased as the β is increased. Evidence of island current is obtained when the magnetic island is formed due to a small error field. However, the mechanism that generates the island is not yet known.

In the Large Helical Device (LHD) programme, one of the key research issues is to enhance helical plasma performance through edge plasma control. For the first time in the LHD programme, the edge plasma control was performed with a local island divertor (LID) that is a closed divertor, utilizing an m/n = 1/1 island generated externally by which 20 small perturbation coils, and fundamental LID functions were demonstrated experimentally. It was found that the outward heat and particle fluxes crossing the island separatrix flow along the field lines to the rear of the divertor head, where carbon plates are placed to receive the heat and particle loads. Accordingly, high efficient pumping was demonstrated, which is considered to be the key in realizing high temperature divertor operations, resulting in an improvement in energy confinement. In this experiment, relatively good energy confinement is achieved in the high density regime at a magnetic axis position, Rax, of 3.75?m. Results of edge modelling are also presented by using the EMC3-EIRENE code.

Conditions for the excitation of Alfven eigenmodes (AEs) by energetic ions are investigated in neutral-beam-injection (NBI) heated plasmas of the Large Helical Device (LHD). This study is carried out in a wide parameter range of the beta values of the energetic ion components and the ratio of the energetic ion velocity to the Alfven velocity (up to with the assumption of classical slowing down and ). These ranges of parameters cover those predicted for the International Thermonuclear Experimental Reactor (ITER). During this experimental campaign of LHD, toroidicity-induced AEs (TAEs) with n = 1–5 (n being the toroidal mode number), global AEs (GAEs) with n = 0 and 1, and energetic particle modes (EPMs) were observed. The effect of the magnetic configuration on the TAE spectrum was also investigated. In magnetic configurations with relatively high magnetic shear, only TAEs with n = 1 and 2 were observed. On the other hand, TAEs with n up to 5 were observed in magnetic configurations with low magnetic shear. For two typical shots obtained in magnetic configurations characterized by different values of the magnetic shear, eigenfunctions of TAEs were calculated by using a global mode analysis code CAS3D3. The calculated results indicate that the eigenfunctions tend to be localized around the relevant TAE gaps. When the gap is located in the plasma core region (normalized minor radius ρ ≤ 0.4), the TAE tends to become a core-localized type. When the gap is in the outer region (typically 0.5 ≤ ρ ≤ 0.9) of the plasma, the TAE tends to (a) either become a global type having a radially extended structure if the magnetic shear is very weak in the core region inside the gap, (b) or become a gap localized type in the case of finite central magnetic shear. Transition of the eigenmode from the core-localized type with m ~2/n = 1 TAEs (m being the poloidal mode number) to the n = 1 GAEs (or cylindrical AEs) has been observed when the rotational transform at the core ι (0)/2π exceeds the specific value of ι(0)/2π = 0.4.

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.

Hydrogen ice pellets have been injected using two different location configurations, namely outer port injection and coil side injection, in the large helical device. The behaviour of the pellet ablation has been observed using a fast camera, which possesses high spatial and time resolution. Striking toroidal deflection of the pellet trajectory is observed. The deflection is in the direction of tangential neutral beam injection. The toroidal velocity ultimately reaches 1000 m s−1 or more. The possibility of a rocket effect due to a unilateral ablation by the fast ions is discussed. The effective penetration depth of the pellet, which is measured by images, agrees with the prediction from the neutral-gas-shielding model. The penetration depth is compared with the measured deposition profile of the pellet.

Material probes have been installed at the inner walls along the poloidal direction in large helical device (LHD) from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to study the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was thoroughly cleaned by glow discharge conditioning and the number of main discharge shots increased. For the 3rd and 4th campaigns, graphite tiles were installed over the entire divertor strike region, and then the wall condition was significantly changed compared with the case of a stainless steel (SS) wall. It was seen that graphite tiles in the divertor were eroded mainly during main discharges, and the SS first wall mainly during glow discharges. During main discharges the eroded carbon was deposited on the entire wall. A reduction of metal impurities in the plasma was observed, which corresponds to the carbonized wall. The deposition thickness was great at the wall far from the plasma. Since the entire wall was carbonized, the amount of discharge gases retained such as H and He became large. In particular, helium retention was large at a position close to the anodes used for helium glow discharge cleanings. One characteristic of the LHD wall is a large retention of helium since the wall temperature is limited to below 368 K. In order to reduce the recycling of the discharge gas, wall heating is necessary.

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.

Divertor footprints have been identified within a few mm in accuracy after 10 000 shots. This is the large merit of the large helical device divertor for erosion/deposition studies due to high reproducibility with an external superconducting coils system. Helical distribution of divertor erosion is compared with the prediction from magnetic field characteristics. The measured net erosion depth is found to be about a factor 3 less than the estimated one. Numerical simulations have revealed the net erosion to be very sensitive to deposition of C impurity in the plasma. Eroded carbon atoms are mainly redeposited near the divertor tiles, and partly deposited near the divertor strike point, forming a mixed layer with promptly deposited metals. Deposited metals accumulate locally at the edge of microscale open pores and around grains of graphite. This kind of metals sink possibly plays an important role as an impurity source after the tiles installation. This aspect of ‘microscopic-PSI study’ is very informative for understanding macroscopic-PSI.

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 Shafranov shift is derived from the two-dimensional profile of x-ray intensity measured with a soft x-ray CCD camera in large helical device. The accuracy of the measurement of the magnetic axis is 3% of the Shafranov shift at high β. The measured Shafranov shift increases linearly up to 280±3 mm, which is 47% of the minor radius as the volume averaged β βdia measured with a diamagnetic loop is increased up to 2.6%. The Shafranov shift measured with a soft x-ray CCD camera agrees with that calculated with the three-dimensional equilibrium code VMEC. The reduction of the Shafranov shift due to the higher harmonics of the Fourier components of the magnetic field is also observed.

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.

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.

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

The energy confinement time of net current-free plasmas in the Large Helical Device (LHD) is comparable to ELMy H-mode with q = 4.5 on the scale of IPB98(y,2). Broad temperature with high edge pressure, which is of robust nature in LHD, recalls the pedestal structure in tokamak H-mode. Basic parameter dependence is, however, weak gyro-Bohm in both edge and core regions. This suggests that one physical model could explain the observed profile. The edge temperature is proportional to the square root of heating power and inversely proportional to the square root of edge density. This means that edge pressure and its gradient are not limited and can be increased by heating power.

The flattening of the electron density profiles due to the static magnetic island is observed when the n/m = 1/1 perturbation field is applied by the external coil. The iso-density widths are estimated from the model fitting analysis. The temporal history of the iso-density width and its dependence on normalized collision frequency and plasma β are estimated from a single shot. The iso-density width increases as the external perturbation field increases. Although determination errors are not small compared with the dynamic range of the estimated iso-density width, the following tendencies are observed. The iso-density width shows positive dependence on logarithm of normalized collision frequency and negative dependence on β.

A mild pedestal has been observed in the large helical device (LHD) discharges. Broad profile pedestal leads to enhancement of the energy confinement. The shapes of the electron temperature profile in the gas puff divertor discharges are very close to parabolic profile, nearly being independent of density, input power and magnetic field. The plasma tends to preserve the temperature scale length (or local shape), ∇T/T even with a large island or during radiative collapse.

Positive radial electric field, Er, was observed at the edge region in the low-density NBI plasmas of the Large Helical Device (LHD). Ion thermal diffusivities, χi, are evaluated by the local transport analysis for two different density plasmas and for two cases of the magnetic axis position. In the standard configuration case, χi was reduced by large positive Er, while no reduction in χi was found in the inward shifted case. A certain level of suppression of the neoclassical ripple transport by Er seems to be confirmed since the neoclassical transport is considered to be more dominant in the standard magnetic configuration plasmas.

The improvement in the particle confinement of pellet shots on the Large Helical Device (LHD) was found after methane (CH4) mixed hydrogen gas-puff discharges. Only four discharges introducing ∼20 Pa m 3 of CH4 caused the reduction in the radiation loss and the level of metal impurities, together with the enhanced recycling, which is expected as the real time carbonization effect. The decay rate of the electron density was mitigated in the pellet shot after CH4 discharges. Transport analysis shows 60% reduction in the particle transport coefficient at half the averaged minor radius.

Strong nonuniformity in the helical divertor flux distribution is seen both in numerical calculations and in experiments. Flux profiles along the divertor trace on the vacuum vessel change according to vacuum magnetic configurations. Diverted particles choose one of two X-points as the escaping channel to divertor legs, which is determined by the local magnetic shear just inside the X-point. Ergodicity inside both X-points is also estimated quantitatively by a measure of Kolmogorov length for each magnetic configuration. The relationship between ergodicity and the choice of X-point for diverted particles is also discussed.

The radial electric field (Er) properties in LHD have been investigated to give guidance towards improved confinement with a possible Er transition and bifurcation. The ambipolar Er is calculated from the neoclassical flux on the basis of analytical formulas. This approach is appropriate to clarifying the Er properties over a wide parameter range in a more transparent way. A comparison between calculated Er and the experimental values has shown a qualitatively good agreement, for example, for the threshold density for the transition from the ion root to the electron root. The calculations also reproduce well the experimentally observed tendency that the electron root is possible by increasing temperatures even for higher density and the ion root is enhanced for higher density. On the basis of the usefulness of this approach, calculations over a wide range have been performed to clarify the parameter region where the Er transition and bifurcation are possible. This gives a comprehensive understanding of the parameter region of interest, which is valuable for the promotion of experiments towards improved confinement.

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

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

Two large helical devices/ stellarators, LHD and W7X are under construction, equipped with divertor system. The divertor concepts in these devices are briefly reviewed.

The generation and transport of impurity particles in a boundary plasma are analysed on the basis of a single particle motion together with a Monte Carlo method, where the Langevin-type equation of an impurity particle is used. The impurity generation and transport in a scrape-off layer are modelled after the Large Helical Device (LHD). The screening efficiency of the SOL for the impurity penetration into a core plasma, impurity redeposition on the first wall and the other effects are analysed from the statistics of traces of many impurites.

The first design of Carbon Sheet Pump (CSP) proposed in the Large Helical Device (LHD) project to achieve high plasma performances with low overall recycling ratio is presented. The CSP sheets of C/C composite are designed to be mounted on the vessel wall near the divertor plate, and operated in the cycle of (1) pumping of charge-exchanged (CX) fast H neutrals by trapping them in the sub-surface layers of the carbon at temperatures below 200°C in the total duration time of about 100 s, (2) degassing of trapped hydrogen by direct Joule heating of the carbon sheet at a temperature above 800°C in a period between main discharges, and then (3) cooling down to the operation temperature below 200°C. The first demonstration of CSP carried out successfully in a hydrogen glow discharge chamber is presented.

A divertor magnetic configuration is proposed that significantly reduces heat load on the divertor plates in heliotron/torsatron devices. The proposed configuration utilizes an octupole field for guiding the divertor channels to a remote area outside the main coil system, where the magnetic field is weak. This allows a significant reduction of the heat load due to expansion of the divertor channels as well as substantially easier access to the divertor plates for maintenance, the key requirements for toroidal fusion reactor designs

The Large Helical Device (LHD) is a heliotron/torsatron-type confinement device (B = 4 T, R = 3.9 m) equipped with a helical divertor. In the LHD configuration, the plasma region is shifted inwards approximately a third of the plasma minor radius relative to the center of the two pairing helical coils, thus making the distance between the coil center and the edge plasma small, 328 mm on the small major radius side of the torus. Within this space, many components must be installed, such as the superconducting helical coil (163), its coil can (45), the thermal shield (30), the vacuum gap (35), the plasma vacuum vessel (15), the first wall (25) (the numbers in the parentheses are allocated radial space in mm for the respective components). Effective edge plasma control by the divertor requires a space of more than 15 mm between the plasma and the first wall. To meet the above space requirement, the thickness of the helical coil is designed to be as small as possible, yet the coil current density and maximum field strength are within the limits of reliable coil operation.

Irregular magnetic fields resonate with the rational surface of the magnetic confinement systems, form magnetic islands and ergodic layers, and destruct the plasma confinement. To avoid this confinement destruction the requirement of an accuracy of 10 −4 in the magnetic field is adopted as the magnetic-accuracy design criterion for the LHD machine. Following this criterion the width of the undesirable magnetic island is kept less than one tenth of the plasma radius. The irregular magnetic field from the superconducting (SC) helical and poloidal coils is produced by winding irregularity, installing irregularity, cooling-down deformations and electromagnetic deformations. The local irregularities such as feeders, layer connections, adjacent-conductor connections of the coils also produce an error field. The eddy currents on the supporting shell structure of SC coils, the cryostat, etc. are also evaluated. All irregular effects are analyzed using Fourier decomposition and field mapping methods for the LHD design, and it is confirmed that the present design of the superconducting coil system satisfies the design criterion for these field irregularities.

A steady-state H-mode plasma with frequent ELMs (edge localized modes) was obtained by applying magnetic fields with high poloidal mode numbers resonant in the plasma edge. The induced ELMs limit the density and impurity accumulation normally observed during ELM-free H-mode. An EML coil with numerous resonant modes ( m 10) in the plasma edge appears the most effective. This method of control is extended to higher auxiliary heating power by increasing the magnitude of the magnetic perturbations.

Neoclassical transport in the banana regime has been analysed with the DKES (Drift Kinetic Equation Solver) code for the Large Helical Device (LHD). It is found that in the 1/ν regime the diffusion coefficients change by one order of magnitude for various LHD configurations, depending on the structure of the helical magnetic ripple. The neoclassical transport calculated with the DKES code is quantitatively in good agreement with a multi-helicity theory formulated by Shaing and Hokin (1983). Incorporating the multi-helicity effect into the diffusion coefficient, the authors propose an interpolation formula for the diffusion coefficient in the 1/ν and ν regimes. When the ion temperature is increased at a fixed density of n = 1020 m-3, the ions undergo a transition from 1/ nu neoclassical transport to the nu regime when Ti > 3 keV, with a radial electric potential e comparable to the ion temperature (e/Ti ≈ 1). For the optimized configuration, the ion thermal diffusivity has a maximum value of ≈ 3.5 m2/s at a minor radius r/a ≈ 0.5. The bootstrap current has also been studied and the results have been compared with the theory. At the collisionless limit with a moderate radial electric potential of e/Ti ≈ 1, the DKES calculations evaluated for various LHD configurations support the theoretical formula given by Shaing and Callen (1983). For collision frequencies between the plateau regime and the banana regime, where the analytical theory is not applicable, the bootstrap current might become larger (by a factor of about two) than in the collisionless limit, depending on the radial electric field