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Edge MHD instabilities with pressure collapse are found in relatively high beta and low magnetic Reynolds number regions with a magnetic axis torus outward-shifted configuration of the large helical device (LHD), and characteristics and onset conditions of the instability are investigated. The instability has a radial structure with an odd parity around the resonant surface, which is different from that of the interchange instability typically observed in the LHD. The onset condition dependence on the magnetic axis location shows that the onset beta increases as the magnetic axis location moves more torus inwardly, and the instability appears only in limited configurations where the magnetic axis is located between 3.65 and 3.775 m. In such configurations, the resonant surface location is close to an index of the plasma boundary. This fact suggests that the distance between the resonant surface location and the plasma boundary plays an important role in the onset, and a possibility that the instability is driven by an external mode.

The Large Helical Device (LHD) successfully started the deuterium experiment in March 2017, in which further plasma performance improvement is envisaged to provide a firm basis for the helical reactor design. Some major upgrades of facilities have been made for safe and productive deuterium experiments. For radiation safety, the tritium removal system, the integrated radiation monitoring system, and the access control system have been newly installed. Each system has new interlock signals that will prevent any unsafe plasma operation or plant condition. Major interlock extensions have been implemented as a part of the integrated radiation monitoring system, which also has an inter-connection to the LHD central operation and control system. The radiation monitoring system RMSAFE (Radiation Monitoring System Applicable to Fusion Experiments) is already operating for monitoring γ(X)-rays in LHD. Some neutron measurements have been additionally applied for the deuterium experiments. The LHD data acquisition system LABCOM can acquire and process 24 h every day continuous data streams. Since γ(X)-ray and neutron measurements require higher availability, the sensors, controllers, data acquisition computers, network connections, and visualization servers have been designed to be duplicated or multiplexed for redundancy. The radiation monitoring displays in the LHD control room have been carefully designed to have excellent visual recognition, and to make users immediately aware of several alerts regarding the dose limits. The radiation safety web pages have been also upgraded to always show both dose rates of γ(X)-rays and neutrons in real time.

A transition from an interchange mode to a non-resonant mode is found in the nonlinear magnetohydrodynamic simulation for the partial collapse in a large helical device (LHD) plasma with a net toroidal current. This transition can occur when the magnetic shear is weak and the rotational transform is close to unity in the core region. In this transition, the mode number of the dominant Fourier component is reduced. As a result of the nonlinear evolution, the (m, n) = (1, 1) component can be dominant, where m and n are the poloidal and the toroidal mode numbers, respectively. This transition is considered to be a candidate to explain the observation in the LHD experiments with the net toroidal current that show partial collapses are caused by the (1, 1) mode.

The internal mode structure of a precursor with the tearing-parity structure of the locked-mode-like instability is investigated. For the first time, the Thomson scattering system with high temporal and spatial resolution enables us to find the non-rotating temperature flattening region in the torus outboard side in addition to the rotating island during the slowing-down phase. Additionally, the width of the flattening region is 10% normalized by the plasma minor radius. Furthermore, the radial profiles of the pressure degradation during the slowing-down phase are evaluated. At the beginning of the slowing-down phase, the pressure degradation area is located slightly inside the resonant surface. After that, the peak location moves to the core region. Finally, the degradation area peaks at the plasma center.

Aims The use of new psychoactive substances (NPS) has become increasingly widespread over the last decade, in Japan and internationally. NPS are associated with a range of increasingly serious clinical, public, and social issues. Political measures to ameliorate the effects of NPS in Japan have focused on tightening regulation rather than establishing treatment methods. The current study sought to compare the neuropsychiatric symptoms of patients with NPS‐related disorders across several years. We examined patients who attended specialized hospitals for treating addiction, to elucidate the impacts of legal measures to control NPS. Methods Subjects (n = 864) were patients with NPS‐related disorders who received medical treatment at eight specialized hospitals for treating addiction in Japan between April 2012 and March 2015. Clinical information was collected retrospectively from medical records. Results Among psychiatric symptoms, the ratio of hallucinations/delusions decreased over time across 3 years of study (first year vs second year vs third year: 40.1% vs 30.9% vs 31.7%, P = 0.037). Among neurological symptoms, the ratio of coma/syncope increased over the 3‐year period (7.8% vs 11.0% vs 17.0%, P = 0.002), as did the ratio of convulsions (2.8% vs 4.3% vs 9.7%, P = 0.001). Conclusion The symptoms associated with NPS were primarily psychiatric in the first year, while the prevalence of neurological symptoms increased each year. The risk of death and the severity of symptoms were greater in the third year compared with the first year, as regulation of NPS increased., We collected information retrospectively based on medical records and compared with psychiatric and neurological symptoms of NPS users observed in the fiscal year of their first medical examination. We found that adverse health effects resulting from the use of NPS were changed from mainly psychiatric symptoms to neurological symptoms. These findings suggest the possibility that contained more dangerous and toxic compounds after drug‐related regulations had been tightened.

The slowing-down mechanism of the locked-mode-like instabilities with and without an island structure is investigated through the effects of an external RMP (resonant magnetic perturbation) on the instabilities. For both instabilities, the slowing-down duration decreases with the increase in the external RMP, and the RMP dependence is consistent with the braking model of the j × B force due to the interaction between the instabilities and the external RMP. Moreover, the relationship between the amplitude and the frequency of both locked-mode-like instabilities during the slowing down is consistent with the force balance model between the j × B force due to the external RMP and a viscous force. These results suggest that the slowing down of both locked-mode-like instabilities with finite external RMP occurs due to the j × B force driven by the external RMP.

Radial profile of externally applied resonant magnetic perturbation (RMP) field with mode numbers of m = 6 and n = 2 in a small tokamak device HYBTOK-II have been investigated using a magnetic probe array, which is able to measure the radial profile of magnetic field perturbation induced by applying RMP. Results of RMP penetration into the plasma show that the RMP decreased toward the plasma center, while they were amplified around the resonant surface with a safety factor q = 3 due to the formation of magnetic islands. This suggests that RMP fields for controlling edge plasmas may trigger some kind of MHD instabilities. In addition, simulation results, based on a linearized four-field model, which agrees with the experimental ones, indicates that the penetration and amplification process of RMP strongly depend on a Doppler-shifted frequency between the RMP and plasma rotation.

The resistive interchange mode destabilized by the resonant interaction with the trapped energetic ions is fully suppressed when the injected power of electron cyclotron heating exceeds a certain threshold. It is shown for the first time that the complete stabilization of the energetic-particle-driven mode without relaxing the energetic particle (EP) pressure gradient is possible by reducing the radial width of the eigenmodes δ_{w}, especially when δ_{w} narrows to a small enough value relative to the finite orbit width of EP.

We investigate the penetration threshold of resonant magnetic perturbation (RMP) by the external coils in the Large Helical Device (LHD) for various plasma aspect ratio configurations. The qualitative dependence on the collisionality is opposite to that in a high plasma aspect configuration; this is a quite unique property first found in the LHD. We also investigate the threshold dependence on the ion species, and find that the threshold in deuterium discharges is much smaller than that in hydrogen discharges. In all cases the thresholds are higher as the poloidal rotation becomes faster, which shows that poloidal rotation is the dominant driver to the RMP shielding. This is qualitatively consistent with the torque balance model between the electromagnetic and poloidal viscous torques. In a configuration of the LHD, the dependence of the threshold on the density is qualitatively similar to that in Ohmic tokamak plasmas, but the beta dependence is opposite to that of tokamaks. The difference arises from the cause of the viscous torque.

The mode slowing down mechanism of the locked-mode-like instability without a large magnetic island is investigated, based on the LHD experimental analysis. The mode frequency coincides with E × B rotation frequency at the resonant surface and the slowing down is caused by two processes. One is the resonant surface moving to the small E × B rotation frequency region and the other is the slowing down of the E × B rotation frequency near the resonant surface. Both processes are almost the same as those of the locked-mode-like instability with a large magnetic island. The results suggest that the slowing down occurs even though the precursor does not have a large magnetic island. However, when the external RMP is imposed, the mode frequency in the slowing down phase sometimes does not coincide with the E × B rotation frequency. Moreover, the mode amplitude during the slowing down phase increases with the decrease of the mode frequency both with and without the imposed external RMP, which suggests that the instability growth in the slowing down phase is more strongly related to the mode frequency than the E × B rotation frequency because the mode sometimes does not rotate with the E × B rotation.

The ion cyclotron range of frequencies (ICRF) heating power injection in the hydrogen experiment in LHD was demonstrated after the upgrade of ICRF antennas. The ICRF wave couples and accelerates the energetic particles injected by perpendicular-NBIs with 40 keV. The simulation by the MORH code shows the existence of energetic particles around the ICRF third harmonic resonance layers. As the result of ICRF heating power deposition, the beta value increased by 0.2% in absolute beta mainly due to the increased energetic particle content. The increase of energetic ions particularly around 60 keV, which should be accelerated by the ICRF heating, is observed. The ICRF heating efficiency was approximately 30%–50%, estimated from the break-in-slope analysis at the turn off timing of ICRF power from the stored energy measured by diamagnetic loops. This increase of the stored energy is mostly the contribution of the increased energetic particles. The heating efficiency increases as the density increases.

Magnetic flux surface measurements have been carried out in the Large Helical Device (LHD) in the standard magnetic field configuration with toroidal magnetic field strength up to 2.75 T. An electr...

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

Trials of steady-state operation (SSO) in the Large Helical Device (LHD) were started when a continuous wave (cw) gyrotron with the output power up to 0.2 MW was introduced to the electron cyclotro...

This paper describes present status of magnetic measurements in the Large Helical Device (LHD). The magnetic measurements have been mainly applied for estimation of global parameters and fo...

Energetic ion-driven magnetohydrodynamic instabilities such as Alfven eigenmodes (AEs), energetic particle modes (EPMs), and their impacts on energetic ion confinement are being studied on the Larg...

In the vacuum of the Large Helical Device (LHD), we can change the plasma volume, the aspect ratio, the ellipticity, the rotational transform, and the height of the magnetic hill through th...

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

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

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

In recent years, fusion research has become exceedingly popular in the Asian region. Many toroidal devices are currently operated for fusion research in universities and institutes. The long pulse operation of a high-performance plasma is also possible in four superconducting devices, which are EAST (Institute of Plasma Physics Chinese Academy of Sciences (ASIPP), China), KSTAR (National Fusion Research Institute (NFRI), Korea), LHD (National Institute for Fusion Science (NIFS), Japan) and SST-1 (Institute for Plasma Research (IPR), India). The world's largest superconducting tokamak, JT-60SA (National Institutes for Quantum and Radiological Science and Technology (QST), Japan), and middle-size tokamak, HL-2M (Southwestern Institute of Physics (SWIP), China), have now been constructed for the upcoming operation. Based on the progressive development of fusion research in Asian region an international collaboration on critical physics for the steady state operation of high-performance plasmas has been carried out among China, Japan and Korea during past five years through the A3 foresight program. Recent fusion research activities in the Asian region and the results of the collaboration are briefly reported with future prospects for international collaboration.

In order to clarify the mass effect and the atomic number effect on the rotation of the interchange MHD mode, the relationship between the mode rotation and the plasma rotation in hydrogen-rich plasmas is compared with that in helium-rich plasmas and deuterium-rich plasmas. In all cases, the frequency of the m/n = 1/1 resistive interchange mode is almost the same as the sum of the E × B rotational frequency and the electron diamagnetic rotational frequency at the ι/2π = 1 resonant surface. This result indicates that the rotation of the resistive interchange mode does not depend on the ion species of which the plasma consists.

The numerical method for analysis of the “MHD stability beta limit” based on a hierarchy integrated simulation code TASK3D has been developed. The numerical model for the effect of the MHD instabilities is introduced such that the pressure profile is flattened around the rational surface due to the MHD instabilities. The width of the flattening of the pressure gradient is determined from the width of the eigenmode structure of the MHD instabilities (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Global energy confinement in the Large Helical Device has been proved to be comparable to those of tokamaks in ELMy H-mode. It shows a gyro-Bohm-like property as seen in international stellarator s...

In the large helical device (LHD) having three dimensional configuration, Alfven eigenmodes (AEs) destabilized by energetic ions are widely investigated using neutral beam heated plasmas with monotonic and non-monotonic rotational transform (ι/2π) profiles. In a plasma with monotonic ι/2π-profile, core-localized toroidicity-induced Alfven eigenmode (TAE) as well as global one are often observed. With the increase in the averaged toroidal beta value, defined as the ratio of total plasma pressure to toroidal magnetic pressure, core-localized TAE with low toroidal mode number becomes global. In a relatively high beta plasma with monotonic ι/2π-profile, two TAEs with different toroidal mode number often interact nonlinearly and generate another modes through three wave coupling. In a plasma with non-monotonic ι/2π-profile generated by intense counter neutral beam current drive, reversed shear Alfven eigenmode (RSAE) and geodesic acoustic mode (GAM) excited by energetic ions were observed for the first time in a helical plasma. Nonlinear coupling was also observed between RSAE and GAM.

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.

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

The 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 H2 discharges of large helical device (LHD), H-mode-like transition has been obtained with edge localized mode (ELM)-like bursts in an outwardly shifted configuration of R ax = 4.00 m with an extremely thick ergodic layer, where the ι/2π = 1 position is located in the middle of the ergodic layer. This H-mode-like transition can be triggered by changing P NBI (13 cm−3. The spontaneous H-mode-like transition is also observed during a stable discharge with constant heating power and density, where a steep edge pressure gradient is formed at the ι/2π = 1 position. The ELM-like bursts occurred in the ergodic layer accompanied with density bursts. It is found that the density bursts triggered at the ι/2π = 1 position in the ergodic layer instantaneously propagate outward beyond the ergodic layer with positive density pulse, and a few ms later also begin to propagate inward from the ergodic layer with negative density pulse to recover the previous edge density profile. In the case of He discharges no H-mode-like transition and ELM-like bursts are observed. The frequency of the ELM-like bursts increases with heating power, e.g. 50 Hz (3 MW)–100 Hz (5 MW) at n e = 4 × 1013 cm−3. The external resonant magnetic field to generate a static m/n = 1/1 island inside the ergodic layer is supplied to the ELM-like bursts. It is also found that the externally supplied magnetic field makes a clear response even in the ergodic layer. As a result, the frequency of the ELM-like bursts decreases, e.g. from 60 to 45 Hz. The present study strongly indicates that those phenomena observed in LHD are quite different from the ones which have been observed in tokamaks until now.

Electron heat transport in the low-collisonality electron cyclotron heating plasma is investigated to clarify the effect of neoclassical transport optimization on the thermal plasma transport in the Large Helical Device (LHD). Five configurations are realized by shifting the magnetic axis position in major radius: 3.45, 3.53, 3.6, 3.75, and 3.9 m. A clear effective helical ripple (which is a quantitative measure of the neoclassical transport optimization) dependency on the enhancement factor of the global energy confinement relative to ISS95 is observed. Local heat transport analyses show a higher electron temperature and a lower heat transport in the neoclassical transport optimized configuration at half the minor radius. The comparisons of the experimental total heat fluxes with that of the neoclassical transport by DCOM/ NNW suggest that the neoclassical transport plays a significant role in the heat transport and that the neoclassical transport optimization is effective in improving the plasma confinement in the low-collisionality LHD plasma.

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

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

The interaction between pressure driven modes and magnetic islands generated by a resonant magnetic perturbation (RMP) in the large helical device (LHD) is numerically analyzed. In this analysis, three-dimensional treatment is essential in the equilibrium and dynamics calculations, because the equilibrium pressure profile is deformed by the RMP. The deformation changes the linear mode structure from the interchange type to the ballooning-like type that is localized around the X-point of the island in the equilibrium magnetic field including the RMP. This mode causes a pressure collapse in the nonlinear evolution, which spreads from the X-point to the core. Therefore, the spatial phase of the collapse is fixed to the island geometry. The fixed phase agrees with the LHD experimental results with a natural error field.

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

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

We proposed a new method for measuring plasma energy using superconducting helical field coil (HFC) signals. The change in HFC flux is experimentally confirmed due to the increase in plasma energy. The VMEC-DIAGNO equilibrium analysis code clarified the relation between the HFC flux change and the plasma energy increase, and the scaling law for the plasma energy measurement is obtained. The comparison between this proposed method and the conventional diamagnetic coil measurement has been carried out, and the effectiveness of this proposed technique using HFC is confirmed.

Development of an integrated simulation system for helical plasma is described that draws new experimental plans, including those in new devices, and does experimental data analysis from the viewpoint of integrated physics. The integrated simulation system to be developed has a modular structure that consists ofmodules for calculating magnetohydrodynamic (MHD) equilibrium/ stability, transport, and heating. Each module can be selected in accordance with a user's request and can be combined with other modules. When we want to perform the integrated simulation during the entire plasma duration, a transport module is to be a core module. An integrated tokamak transport code will be extended for the helical configuration and used as a transport module. As the first step of the extension, time evolution of the plasma net current, which is consistent with the three-dimensional MHD equilibrium, is planned to be solved for Large Helical Device plasmas by taking into account the bootstrap current and the beam-driven current.

This paper describes recent progress in magnetohydrodynamics (MHD) study of high-beta plasmas of the Large Helical Device. Control of the plasma aspect ratio (Ap) in the range of 6.3 to 8.3 was done in order to optimize the configuration for high-beta plasma production and to investigate the MHD characteristics. The experiments brought a maximum average beta of 4.3% at the Ap = 6.6 configuration. MHD activities in the periphery are dominantly observed in such a high-beta region, and their amplitudes increase with decreasing magnetic Reynolds number (S) and have clear dependence on the S parameter. When the plasma aspect ratio is increased, minor collapse due to the m/n = 1/1 mode without rotation occurs. It is enhanced further by the plasma current reducing magnetic shear and degrades the beta value by >50%. The results are expected to give important information on the operation regime and the future design of the helical fusion reactor and to contribute to experimental knowledge of ideal and resistive instability.

The Large Helical Device (LHD) is an experimental facility for fusion plasma using a large scale mutually coupled superconducting coil system and it is now operating at the National Institute for Fusion Science. This paper introduces the design and the operation results of the dc power systems to drive these coils. First, the outline of the LHD and the power supplies are described. Next, a control system and the current controller for the power supplies are introduced. Finally, some experimental results are introduced.

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.

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

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

Fixed boundary three-dimensional anisotropic pressure equilibria are computed that are consistent with a family of distribution functions that load particles preferentially towards v parallel to similar or equal to v and model the effects of balanced tangential neutral beam injection. The total pressure surfaces remain coupled with the magnetic surfaces even when the parallel beam pressure greatly exceeds its perpendicular counterpart. The anisotropy factors and the Shafranov shift, which increases with the anisotropy, are calculated in a currentless two-field period quasiaxisymmetric stellarator device.

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

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

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

Recently, dramatic progress has been achieved in the study of helical systems with high-beta experiments. Discharges with more than 3% beta plasmas have been achieved in Large Helical Device (LHD) and Wendelstein 7-AS (W7-AS). Although magnetohydrodynamic (MHD) instabilities affect local pressure gradients, the global transport property does not seem to limit the achieved beta value in either device. We summarize the LHD high-beta properties in MHD stability, equilibrium, and transport, and we show the relationship between the experimentally achieved parameters and theoretical predictions. We contrast the LHD results with the W7-AS high-beta properties. In both devices, stationary discharges in the definitely MHD unstable region have not been observed. We mention the key issue for achievement of the beta values >5%.

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

Nonlinear absorption in carbon disulfide was investigated at the wavelengths of 795 and 1054 nm. Nonlinear absorption coefficients were measured to be 5 × 10 −11 cm W −1 (due to two-photon absorption, λ =795 nm) and 2.8 × 10 −23 cm 3 W −2 (due to three-photon absorption, λ =1054 nm). The influence of two-photon absorption on the variations of nonlinear refractive index of CS 2 at high pulse repetition rates was analyzed.