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The activities carried out to obtain public consent for deuterium experiments in LHD, which began in 2017, are reviewed in this paper. In addition, the upgrades and the safety management of LHD for deuterium experiments, including neutron yield measurement system, exhaust detritiation system, institutional regulation for radiation control, and other issues, are briefly presented. During the first campaign of the deuterium experiments in LHD, the shielding of gamma-ray and neutron by the concrete wall of the LHD torus hall was evaluated. Also, the confinement of radioactive isotopes in air inside the torus hall was investigated. No increase of radiation dose was measured outside the torus hall, although the high radiation dose field inside the torus hall was found during deuterium experiments. Therefore, almost all gamma-rays and neutrons were shielded by the concrete wall of the torus hall due to its sufficient thickness of 2 m. The radioactive isotopes in air as well as in other components were well confined in the torus hall. In particular, the pressure control inside the torus hall being lower than outside the torus hall effectively prevented the radioactive isotopes in air from diffusing to the unprescribed area.

Spherical tokamak (ST) research in Japan has produced many innovative results: (i) plasma start-up to I p > 70 kA was achieved by electron cyclotron wave (ECW) with N ∥ = 0.75, while electron heating to T e > 500 eV was achieved with N ∥ = 0.26 on QUEST. (ii) The radiofrequency (RF)-induced transport model was combined with the x-ray emission model, and extended magnetohydrodynamics equilibrium with kinetic electrons was developed to interpret fast-electron-dominated lower hybrid wave sustained plasmas on TST-2. (iii) Density as high as 30 times the cutoff density was achieved by electron Berstein wave current drive combined with electron beam injection on LATE. (iv) Multiple plasmoids formed by tearing instability in the elongated current sheet were observed, and flux closure and ion heating by plasmoid-mediated fast magnetic reconnection were observed on HIST. (v) Optimization of ECW-assisted inductive start-up with a vertical field with positive decay index was performed on TST-2. (vi) Stabilization of the vertical displacement event by a set of upper and lower helical field coils was demonstrated on TOKASTAR-2. (vii) A 6 h discharge was achieved by cool-down of the center stack cover on QUEST, where the plasma duration limit was consistent with the wall saturation time estimated by modeling. (viii) Extension of ion heating by plasma merging was achieved on TS-3U, TS-4U, UTST, MAST, and ST40.

The large helical device (LHD) is one of the world’s largest superconducting helical system fusion-experiment devices. Since the start of experiments in 1998, LHD has extended its parameter regime, aiming at achievement of the reactor-relevant plasma conditions and the exploration of related plasma physics in helical-type magnetic configurations. The LHD has also demonstrated its inherent advantage for steady-state operation. Based on these leading developments of helical plasma research, LHD has progressed to the advanced research phase, that is, the deuterium experiment that started in March 2017. It is expected that plasma parameters should be extended toward more reactor-relevant regime, and the related physics research is allowed in such extended regime. Taking this opportunity, parameter extensions such as density, temperature, and steady-state operation achieved in the hydrogen experiment phase are overviewed, along with the initial highlighted results in the very first deuterium experiment campaign in 2017. The design activity of LHD-type steady-state helical fusion reactor is also briefly introduced.

The deuterium experiment started from March 2017 on the large helical device (LHD) as a part of the LHD high-performance upgrade project. The objectives of the deuterium experiment are: 1) to realize the high-performance operation; 2) to explore the physics of isotope effect; 3) to demonstrate the confinement capability of energetic particles in helical devices; and 4) to explore the research on plasma–material interactions using the benefit of stable steady-state operation ability of LHD. As preparations for deuterium experiment, the positive-ion-based neutral beam injectors (NBIs) are upgraded to increase their injection power in deuterium operation. On the other hand, the injection power of negative-ion-based NBI is deteriorated due to the isotope effect of negative-ion source. The neutron diagnostic and the exhaust detritiation system are newly installed for the deuterium experiment. The commissioning of LHD for the deuterium experiment is quite successful. The high ion temperature operation region is extended by the deuterium experiment. Ion temperature exceeding 9 keV is achieved with the deuterium operation of LHD.

In order to investigated the dynamics of H− ions and understand the extraction process inside filament-arc-driven plasmas in a Cs-seeded negative ion source, diagnostic experiments using a directional Langmuir probe combined with photodetachment measurement have been conducted. Two-dimensional flow pattern of H− ion flow has been obtained as well as the profile of H− ion density. The result shows that H− ions come from the direction of plasma grid surface, flow to the plasma in the extraction region. During beam extraction, H− ion flow turns to the beam direction. The stagnation point of H− ion flow is at ∼20 mm apart from the plasma grid. Around the stagnation point, the maximum decrement of H− density caused by beam extraction appears. The results indicate that H− ions are extracted from the plasma volume. The extraction process occurs in a region near the plasma grid aperture with boundary of ∼20 mm apart from the plasma gird.

Large Helical Device (LHD) is one of the world largest superconducting fusion experiment devices, having demonstrated its inherent advantage for steady-state operation since the start of experiments in 1998. LHD has also demonstrated reliable operation of the large-scale superconducting magnet system for almost two decades. Development of the challenging heating systems, such as negative-ion-based neutral beam injection (NBI), high-power and high-frequency electron cyclotron heating, and steady-state ion cyclotron heating, have led to wide-ranging physics and engineering achievements. LHD has progressed to the next stage, that is, the deuterium experiment starting in March 2017, which should further extend plasma parameters toward reactor-relevant regime. For establishing firm basis for designing steady-state helical fusion reactor, advanced physics research, such as on isotope effect, energetic particle confinement, and plasma–wall interaction, will be intensively performed in the deuterium experiments. In an engineering aspect, the upgrade of NBI system has been carried out in preparation to the deuterium experiment, and it should contribute to future NBI development for fusion reactors including ITER. For enhancement of the particle control, the closed divertor system has been installed with pumping capability. Diagnostics for neutron measurements are newly developed and installed for the deuterium experiment. Aligned with all the progress of LHD project in terms of engineering and physics aspects, the conceptual design activity of the LHD-type helical fusion reactor, FFHR-d1, has been programmatically conducted. In parallel to the design study, engineering research and development for the component development have been performed, including those based on employing challenging ideas such as high-temperature superconductor, liquid metal ergodic divertor, and molten-salt breeder blanket. The present status of LHD project entering the deuterium experiment phase is overviewed with putting emphasis on the engineering aspects, and then the engineering research and development activities toward steady-state helical fusion reactor are described.

Divertor pumping was applied to plasma discharges for superior fuel particle control in the Large Helical Device (LHD). The LHD is equipped with two different pumping systems. One is the main pumping system, in which the pumping speed is 260 m3/s in hydrogen. The other pumping system is the divertor pumping system in which the pumping speed is 70 m3/s in hydrogen. Divertor pumping was applied to 40-second long pulse Electron Cyclotron Heating (ECH) discharges to assess the improvement in particle control provided by divertor pumping. The results show that without divertor pumping, the electron density was not controlled by gas puffing using the feedback signal of line-averaged electron density. Then, the plasma confinement deteriorated, finally leading to radiation collapse. On the other hand, with divertor pumping, the density was well-controlled by gas puffing using the feedback signal. The results indicate that divertor pumping is one of the key tools for controlling the particles in fusion plasmas.

The deuterium (D) operation was initiated in the LHD in 2017. In the first campaign of the D experiments, we successfully extended the high temperature regime in the LHD. The new record of the ion temperature (Ti) of 10 keV associated with the ion internal transport barrier (ITB) was achieved due to several operational optimization. The thermal confinement characteristics of ITB plasmas were compared between hydrogen and D discharges. The effective ion thermal diffusivity of the ion-ITB plasmas was found to be smaller in the D discharges than that in the H discharges. The profiles of the Ti, the electron density, and the impurity of the high Ti plasmas strongly depended on the magnetic configuration and these profiles tended to peaked in the inward-shifted configuration. It was also found that the electron thermal confinement of the electron-ITB plasmas was clearly improved in the deuterium case. The GKV simulation showed the linear growth rate of TEM/ITG reduced in the plasmas with D both for the ion ITB and the electron ITB plasmas and qualitatively agreed with the tendency of the change in the thermal diffusivity obtained from the power balance analysis.

The long pulse experiments in the Large Helical Device has made progress in sustainment of improved confinement states. It was found that steady-state sustainment of the plasmas with improved confinement at the core region, that is, electron internal transport barrier (e-ITB), was achieved with no significant difficulty. Sustainment of a plasma having e-ITB with the line average electron density n e_ave of 1.1 × 1019 m−3 and the central electron temperature T e0 of ~3.5 keV for longer than 5 min only with 340 kW ECH power was successfully demonstrated.

Deuterium beam operations on neutral beam injectors (NBI) have been carried out successfully in the Large Helical Device (LHD) in 2017. We injected a total of 31 MW neutral beam power, which gave us a new operation region for high ion temperature plasma. Upgrading the beam energy from 40 keV to 60 keV and then to 80 keV for deuterium operation in a positive ion based NBI (P-NBI) of BL4 and BL5, respectively, we have achieved 10.6 MW (75 keV) beam injection in the P-NBI. We confirmed a good arc efficiency (extracted beam current/arc power) with 83% high deuteron ratio which is the same as the hydrogen operation in the positive source. We also extracted deuterium ion beam from three negative-ion based NBIs (N-NBI) without increase beam energy, although the injection power decreased by less than half. The averaged arc efficiency of the deuterium negative ion is approximately 66% of the hydrogen case, and the current density of the accelerated beam current reached 190 A/m2. We have clearly found the lower limit of electron current ratio depending on the arc discharge power. Certainly, the extracted electron current ratio increased in deuterium operation. However, we have had no trouble on the beam line components.Deuterium beam operations on neutral beam injectors (NBI) have been carried out successfully in the Large Helical Device (LHD) in 2017. We injected a total of 31 MW neutral beam power, which gave us a new operation region for high ion temperature plasma. Upgrading the beam energy from 40 keV to 60 keV and then to 80 keV for deuterium operation in a positive ion based NBI (P-NBI) of BL4 and BL5, respectively, we have achieved 10.6 MW (75 keV) beam injection in the P-NBI. We confirmed a good arc efficiency (extracted beam current/arc power) with 83% high deuteron ratio which is the same as the hydrogen operation in the positive source. We also extracted deuterium ion beam from three negative-ion based NBIs (N-NBI) without increase beam energy, although the injection power decreased by less than half. The averaged arc efficiency of the deuterium negative ion is approximately 66% of the hydrogen case, and the current density of the accelerated beam current reached 190 A/m2. We have clearly found the lower lim...

The properties of cesium (Cs) recycling due to back-streaming ions have been investigated using an optical emission spec-troscopy in the hydrogen negative ion (H−) source with a slot-type grounded grid (GG). The slot-type GG performed well to enhance the beam performance, and to reduce the thermal loading on GG by high transparency. We clearly observed increase of Cs optical emission intensity during beam extraction owing to the increase of the Cs ions sputtered from the back plate of the source due to the back-streaming positive hydrogen ions. Increase of Cs is closely related with the extracted H− current, but it does not depend on the beam energy. Recycling Cs from the back plate is deeply relevant to the perveance condition of the H− beam, and is minimized at the optimum perveance. Strong Cs recycling from the back plate owing to the back-streaming ions is promoted in the high perveance condition with large divergence, which is consistent with the numerical calculation for the distribution of back str...

Total power of 16 MW has been successfully delivered to the plasma confined in the Large Helical Device (LHD) from three Neutral Beam Injectors (NBIs) equipped with negative hydrogen (H−) ion sources. However, the detailed mechanisms from production through extraction of H− ions are still yet to be clarified and a similar size ion source on an independent acceleration test bench called Research and development Negative Ion Source (RNIS) serves as the facility to study physics related to H− production and transport for further improvement of NBI. The production of negative-ion-rich plasma and the H− ions behavior in the beam extraction region in RNIS is being investigated by employing an integrated diagnostic system. Flow patterns of electrons, positive ions and H− ions in the extraction region are described in a two-dimensional map. The measured flow patterns indicate the existence a stagnation region, where the H− flow changes the direction at a distance about 20 mm from the plasma grid. The pattern also suggested the H- flow originated from plasma grid (PG) surface that turned back toward extraction apertures. The turning region seems formed by a layer of combined magnetic field produced by the magnetic filter field and the Electron-Deflection Magnetic (EDM) field created by magnets installed in the extraction electrode.

To improve the performance of negative-ion based neutral beam injection on the Large Helical Device, the accelerator was modified on the basis of numerical investigations. A field limiting ring was installed on the upper side of a grounded grid (GG) support and a multi-slot GG was adopted instead of a multi-aperture GG. As a result, the voltage holding capability is improved and the heat load on the GG decreases by 40%. In addition, the arc efficiency is improved significantly only by replacing the GG.

At National Institute for Fusion Science (NIFS), a multi-ampere negative ion source is used to support the RandD on H- production, extraction, and acceleration. In this contribution, we study the characteristics of the acceleration system of this source, in order to characterize the beam optics at different operational conditions. A dedicated experimental campaign was carried out at NIFS, using as main diagnostic the infra-red imaging of the beam profiles. The experimental measurements are also compared with 3D numerical simulations, in order to validate the codes and to assess their degree of reliability. The simulations show a satisfactory agreement with the experimental results.

A cylindrical carbon pellet with a size of 1.2L × 1.2 mm to 1.8L × 1.8 mm and a velocity of 100 m/s to 300 m/s was injected into large helical device (LHD) for an efficient fueling based on its deeper deposition instead of hydrogen gas puffing and ice pellet injection. Electron density increment of Δne = 1014 cm−3 is successfully obtained by single carbon pellet injection without plasma collapse. Typical density and temperature of the ablation plasma of the carbon pellet, e.g., 6.5 × 1016 cm−3 and 2.5 eV for CII, are examined respectively by spectroscopic method. A confinement improvement up to 50% compared to ISS-95 stellarator scaling is clearly observed in a relatively low-density regime of ne = 2×1013 cm−3 to 4×1013 cm−3, and high ion temperature Ti(0) of about 6 keV is also observed with an internal transport barrier at ne = 1.2×1013 cm−3. In particular, the improvement in the ion temperature largely exceeds that observed in hydrogen gas-puffed discharges, which typically ranges below 3 keV.

On the Large Helical Device (LHD), improved confinement of fast ions by moving the magnetic axis location inward is theoretically predicted. This improvement is observed significantly in the fast-i...

Ion heating experiments have been intensively carried out in high- and low-Zeff conditions of Large Helical Device plasmas. In high-Zeff plasmas utilizing neon or argon gus puffing, the ion heating...

A new approach for an estimation of a static ferromagnetic shielding effectiveness is proposed and verified considering the magnetic hysteresis of shielding materials. The measured leakage field of the shielding enclosure corresponds to the results of the finite element method (FEM) analysis when the permeability used for the calculation takes into consideration the effect of coercivity as a parameter representing the hysteresis of the shielding material. The effects of both permeability and coercivity on the leakage fields are discussed with regard to the magnetic properties of the shielding materials by using FEM analysis. The maximum leakage field, a figure of merit of the shielding efficacy, is determined by the coercivity of the material used for the shield. An increase in the leakage field caused by the deterioration of the soft magnetic properties is also discussed from the viewpoint of the effect of the residual strain induced during the assembly and installation of the shielding system. A double-layer shielding technique is effective in reducing the leakage field especially when it is feared that the shields are strained.

An increase in the ion temperature due to transport improvement has been observed in plasmas heated with high-energy negative-ion-based neutral beam injection (NBI), in which the electrons are dominantly heated, in the Large Helical Device. When centrally focused electron-cyclotron-resonance heating is superposed on the NBI plasma, the ion temperature is observed to rise, accompanied by the formation of the electron internal transport barrier (electron-ITB). In the electron-ITB plasmas, an increase in a positive radial electric field is observed, and the transport analysis indicates that the ion transport in the half-radius region is improved with a reduction of the anomalous transport. Thus, this ion temperature rise is ascribed to the ion transport improvement by the transition to the neoclassical electron root. In high-Z plasmas, the ion temperature is increased with an increase in the ion heating power and reaches 13.5 keV. The central ion temperature increases with an increase in a gradient of the electron temperature in an outer plasma region of ρ = 0.8, suggesting the ion transport improvement in the outer plasma region induced by the neoclassical electron root. These results indicate the effectiveness of the electron-root scenario for obtaining high-ion temperature plasmas in helical systems.

Characteristics of ionic plasmas, observed in a high-density hydrogen negative ion source, are investigated with a multi-diagnostics system. The ionic plasma, which consists of hydrogen positive- and negative-ions with a significantly low-density of electrons, is generated in the ion extraction region, from which the negative ions are extracted through the plasma grid. The negative ion density, i.e., the ionic plasma density, as high as the order of 1×1017m−3, is measured with cavity ring-down spectroscopy, while the electron density is lower than 1×1016m−3, which is confirmed with millimeter-wave interferometer. Reduction of the negative ion density is observed at the negative ion extraction, and at that time the electron flow into the ionic plasma region is observed to conserve the charge neutrality. Distribution of the plasma potential is measured in the extraction region in the direction normal to the plasma grid surface with a Langmuir probe, and the results suggest that the sheath is formed at the plasma boundary to the plasma grid to which the bias voltage is applied. The beam extraction should drive the negative ion transport in the ionic plasma across the sheath formed on the extraction surface. Larger reduction of the negative ions at the beam extraction is observed in a region above the extraction aperture on the plasma grid, which is confirmed with 2D image measurement of the Hα emission and cavity ring-down spectroscopy. The electron distribution is also measured near the plasma grid surface. These various properties observed in the ionic plasma are discussed.

Distributions of hydrogen Balmer-α (Hα) intensity and its reduction behavior close to a plasma grid (PG) surface have been observed by a spectrally selective imaging system in an arc discharge type negative hydrogen ion source in National Institute for Fusion Science. Hα reduction indicates a reduction of negative hydrogen ions because the mutual neutralization process between H+ and H− ions causes the dominant excitation process for Hα emission in the rich H− condition such as in ionic plasma. We observed a significant change in Hα reduction distribution due to change in the bias voltage, which is used to suppress the electron influx. Small Hα reduction in higher bias is likely because the production of negative ions is suppressed by the potential difference between the plasma and PG surface.

In the framework of the accompanying activity for the development of the two neutral beam injectors for the ITER fusion experiment, an instrumented beam calorimeter is being designed at Consorzio RFX, to be used in the SPIDER test facility (particle energy 100keV; beam current 50A), with the aim of testing beam characteristics and to verify the source proper operation. The main components of the instrumented calorimeter are one-directional carbon-fibre-carbon composite tiles. Some prototype tiles have been used as a small-scale version of the entire calorimeter in the test stand of the neutral beam injectors of the LHD experiment, with the aim of characterising the beam features in various operating conditions. The extraction system of the NIFS test stand source was modified, by applying a mask to the first gridded electrode, in order to isolate only a subset of the beamlets, arranged in two 3x5 matrices, resembling the beamlet groups of the ITER beam sources. The present contribution gives a description of the design of the diagnostic system, including the numerical simulations of the expected thermal pattern. Moreover the dedicated thermocouple measurement system is presented. The beamlet monitor was successfully used for a full experimental campaign, during which the main parameters of the source, mainly the arc power and the grid voltages, were varied. This contribution describes the methods of fitting and data analysis applied to the infrared images of the camera to recover the beamlet optics characteristics, in order to quantify the response of the system to different operational conditions. Some results concerning the beamlet features are presented as a function of the source parameters.

To investigate the flows of charged particles in front of the plasma grid (PG) in a negative hydrogen ion source, the information of the local densities of electrons and negative hydrogen ions (H-) are necessary. For this purpose, the laser photodetachment is applied for pure hydrogen plasmas and Cs-seeded plasma in a 1/3-size negative hydrogen ion source in NIFS-NBI test stand. The H- density obtained by photodetachment is calibrated by the results from cavity ring-down (CRD). The pressure dependence and PG bias dependence of the local H- density are presented and discussed. The results show that H- density increases significantly by seeding Cs into the plasma. In Cs-seeded plasma, relativity exists between the H- ion density and plasma potential.

The velocity distribution function of hydrogen atoms is one of the useful parameters to understand particle dynamics from negative hydrogen production to extraction in a negative hydrogen ion source. Hydrogen atom temperature is one of the indicators of the velocity distribution function. To find a feasibility of hydrogen atom temperature measurement in large scale filament arc negative hydrogen ion source for fusion, a model calculation of wavelength-modulated laser absorption spectroscopy of the hydrogen Balmer alpha line was performed. By utilizing a wide range tunable diode laser, we successfully obtained the hydrogen atom temperature of ∼3000 K in the vicinity of the plasma grid electrode. The hydrogen atom temperature increases as well as the arc power, and becomes constant after decreasing with the filling of hydrogen gas pressure.

A negative-ion-based neutral beam injection (NBI) system, in which large caesium-seeded negative-ion sources are utilized, has operated reliably in the Large Helical Device (LHD) since it was operational in 1998. The injection power of the hydrogen beam has been increased up to 13.1 MW with three injectors. In one injector with modified ion sources utilizing a multi-slotted grounded grid, the injection power reached 5.7 MW with an energy of 184 keV, which exceeds the designed value of 180 keV and 5 MW. Individual control of the local arc-discharge with independently divided arc and filament power supplies is effective in improving beam uniformity in a large negative ion source. The injection duration has been extended to 120 s, during which the LHD plasma is sustained by the NBI alone with a reduced power of 0.2–0.3 MW using one ion source with a cooled plasma grid. These results demonstrate that the injection performance of the negative-ion-based NBI is comparable to that of the conventional positive-ion-based NBI. The progress of the negative-NBI in the LHD is reviewed in detail from the point of view of the improvement of the negative ion sources.

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

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.

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.

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

In this article we describe the first results of beam injection using an improved negative ion source, which consists of a set of steering and multislot grounded grids. The set of the grids is designed to achieve high power beam injection for the large helical device (LHD). The main purpose in replacing the conventional multiaperture with multislot grids is to reduce the heat load onto the grid. By adopting the multislot grid, the frequency of voltage breakdown at the acceleration gap decreased considerably and the maximum beam energy attained was 180 keV, which is one of the target values of the LHD neutral beam injector (NBI). The maximum injection power reached 4.4 MW, consequently. The heat load onto the multislot grid was water calorimetrically measured and the load is suppressed by 50% compared to the multiaperture case. In the condition for maximum injection power, the beam profile obtained in the accelerator with the multislot grounded grid becomes vertically elongated due to the asymmetric electric field close to the grid slot. Although some part of this elongated beam is lost at the injection port, the multislot grid has enough performance for beam injection.

Transition property of the radial electric field (Er) in LHD have been theoretically investigated and also applied to explain experimental results. Especially, effects of the helicity of the magnetic configuration on the condition to realize the electron root are examined. Larger helicity makes the threshold collisionality higher. This is attributed to the nonlinear dependence of Γe(Er) in a low collisional regime. This interesting feature predicts that the threshold temperature becomes higher for a case of smaller helicity. The variation of the threshold density anticipated from the analysis for cases with different magnetic axis position is qualitatively verified in the density scan experiment.

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

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

The five-year operation of the negative-ion-based neutral beam injector (N-NBI) system on the large helical device (LHD) is reviewed, and the prospects of negative-ion technology for applying it to future helical fusion reactors are discussed from these experiences. The N-NBI system was designed and constructed based on the results of R&D at NIFS. Using three beamlines, the total port-through injection power of 10.3 MW for 2 s has been achieved. Each beamline attained almost the same performance: ~3.5 MW of the port-through power with the beam energy of ~165 keV for hydrogen, which corresponds to the averaged negative ion current density of 25 mA cm−2 from each ion source. Recent modification of accelerators in one of the beamlines was very successful to increase beam energy up to 180 keV and port-through power, 4.4 MW. Long pulse injection heating was also pursued under the steady state LHD operation, and 110 s for 0.1 MW by one ion source, and 80 s for 0.5 MW by two ion sources (one beamline) were achieved. These facts show that the negative ion technology has been established to the same power level of conventional positive ion systems. Several technical improvements needed for obtaining these successful results are described, and the problems that limit the present performance are clarified. Some results of recent R&D to solve these problems are shown.

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

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

Superior control of particle recycling and hence full governance of plasma density has been established in the Large Helical Device (LHD) using largely enhanced active pumping of the closed helical divertor (CHD). In-vessel cryo-sorption pumping systems inside the CHD in five out of ten inner toroidal divertor sections have been developed and installed step by step in the LHD. The total effective pumping speed obtained was 67?????5 m3 s?1 in hydrogen, which is approximately seven times larger than previously obtained. As a result, a low recycling state was observed with CHD pumping for the first time in LHD featuring excellent density control even under intense pellet fueling conditions. A global particle confinement time () is used for comparison of operation with and without the CHD pumping. The was evaluated from the density decay after the fueling of hydrogen pellet injection or gas puffing in NBI plasmas. A reliably low base density before the fueling and short after the fueling were obtained during the CHD pumping, demonstrating for the first time full control of the particle balance with active pumping in the CHD.

As the finalization of a hydrogen experiment towards the deuterium phase, the exploration of the best performance of hydrogen plasma was intensively performed in the large helical device. High ion and electron temperatures, Ti and Te, of more than 6 keV were simultaneously achieved by superimposing high-power electron cyclotron resonance heating onneutral beam injection (NBI) heated plasma. Although flattening of the ion temperature profile in the core region was observed during the discharges, one could avoid degradation by increasing the electron density. Another key parameter to present plasma performance is an averaged beta value $\left\langle \beta \right\rangle $ . The high $\left\langle \beta \right\rangle $ regime around 4% was extended to an order of magnitude lower than the earlier collisional regime. Impurity behaviour in hydrogen discharges with NBI heating was also classified with a wide range of edge plasma parameters. The existence of a no impurity accumulation regime, where the high performance plasma is maintained with high power heating >10 MW, was identified. Wide parameter scan experiments suggest that the toroidal rotation and the turbulence are the candidates for expelling impurities from the core region.

A simultaneous high ion temperature (Ti) and high electron temperature (Te) regime was successfully extended due to an optimized heating scenario in the LHD. Such high-temperature plasmas were realized by the simultaneous formation of an electron internal transport barrier (ITB) and an ion ITB by the combination of high power NBI and ECRH. Although the ion thermal confinement was degraded in the plasma core with an increase of Te/Ti by the on-axis ECRH, it was found that the ion thermal confinement was improved at the plasma edge. The normalized ion thermal diffusivity ${{\chi}_{\text{i}}}/T_{\text{i}}^{1.5}$ at the plasma edge was reduced by 70%. The improvement of the ion thermal confinement at the edge led to an increase in Ti in the entire plasma region, even though the core transport was degraded.

In the large helical device, plasma with more than 4% average beta was successfully produced by multi-pellet injections in a regime with a collisionality one order of magnitude lower than that in previous high-beta operations. An improvement in particle confinement was observed during a high-beta discharge produced by a gas puff, and particle flux to the divertor was reduced by more than 40%. High instabilities at the plasma edge occurred and suppressed the increment of the average beta to 3.4%. A spontaneous change in the magnetic topology contributes to an increase in the average beta value while triggering the excitation of edge MHD instabilities.

A fast time response, wide dynamic range neutron flux monitor has been developed toward the LHD deuterium operation by using leading-edge signal processing technologies providing maxi- mum counting rate up to ?5 × 109 counts/s. Because a maximum total neutron emission rate over 1 × 1016 n/s is predicted in neutral beam-heated LHD plasmas, fast response and wide dy- namic range capabilities of the system are essential. Preliminary tests have demonstrated success- ful performance as a wide dynamic range monitor along the design.

The ICRF heating experiments conducted in 1999 in the third experimental campaign on LHD are reported, with an emphasis on the optimization of the heating regime. Specifically, an exhaustive study of seven different heating regimes was carried out by changing the radiofrequency relative to the magnetic field intensity, and the dependence of the heating efficiency on H minority concentration was investigated. It was found in the experiment that both ion and electron heating are attainable with the same experimental set-up by properly choosing the frequency relative to the magnetic field intensity. In the cases of both electron heating and ion heating, the power absorption efficiency depends on the minority ion concentration. An optimum minority concentration exists in the ion heating case while, in the electron heating case, the efficiency increases with concentration monotonically. A simple model calculation is introduced to provide a heuristic understanding of these experimental results. Among the heating regimes examined in this experiment, one of the ion heating regimes was finally chosen as the optimized heating regime and various high performance discharges were realized with it.

Recent large helical device experiments revealed that the transition from ion root to electron root occurred for the first time in neutral-beam-heated discharges, where no nonthermal electrons exist. The measured values of the radial electric field were found to be in qualitative agreement with those estimated by neoclassical theory. A clear reduction of ion thermal diffusivity was observed after the mode transition from ion root to electron root as predicted by neoclassical theory when the neoclassical ion loss is more dominant than the anomalous ion loss.

During experiments in LHD using stainless steel divertor plates, a slow (~1 s) cyclic oscillation in the plasma parameters known as `breathing' plasma was observed during NBI heated long pulse discharges. Using an average ion, corona equilibrium model for the iron impurity cooling rate, the iron impurity density profile is calculated for 0.0

Since the start of the Large Helical Device (LHD) experiment, various attempts have been made to achieve improved plasma performance in LHD [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)]. Recently, an inward-shifted configuration with a magnetic axis position R_ax of 3.6 m has been found to exhibit much better plasma performance than the standard configuration with R_ax of 3.75 m. A factor of 1.6 enhancement of energy confinement time was achieved over the International Stellarator Scaling 95. This configuration has been predicted to have unfavorable magnetohydrodynamic (MHD) properties, based on linear theory, even though it has significantly better particle-orbit properties, and hence lower neoclassical transport loss. However, no serious confinement degradation due to the MHD activities was observed, resolving favorably the potential conflict between stability and confinement at least up to the realized volume-averaged beta of 2.4%. An improved radial profile of electron temperature was also achieved in the configuration with magnetic islands, minimized by an external perturbation coil system for the Local Island Divertor (LID). The LID has been proposed for remarkable improvement of plasma confinement like the high (H) mode in tokamaks, and the LID function was suggested in limiter experiments.