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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.

AbstractTo aid understanding of physics related to fast ions in the Large Helical Device (LHD) with a three-dimensional shape, a comprehensive set of fast-particle diagnostics has been developed. Fast ions (H+) have been created by neutral beam injection and ion cyclotron resonance heating in LHD discharges. Intense fast-ion populations not only heat the bulk plasma but also drive collective instabilities under certain experimental conditions. Fast-ion experiments on LHD have been conducted to investigate neoclassical confinement of fast ions and issues on anomalous transport of fast ions induced by fast-ion-driven instabilities. This paper reviews fast-particle diagnostics that are essential for study of fast-ion physics on LHD.

AbstractVarious ion cyclotron range of frequencies (ICRF) heating experiments have been conducted in the Large Helical Device (LHD) by changing the magnetic field strength and the wave frequency using hydrogen and helium. When the resonance layer of hydrogen was located in the peripheral region on the lower-magnetic field side, efficient electron heating, i.e., mode conversion heating, was realized. When the ion cyclotron resonance layer was located near the “saddle point” of magnetic field strength, where the gradient of the magnetic field strength is zero, hydrogen ions were efficiently heated by the minority ion heating. The second-harmonic ion cyclotron heating experiments were also conducted by decreasing the magnetic field strength, and the plasma was successfully sustained for 1 s. Ion tails were observed in the ion heating modes. High-energy ions were well confined by the inward-shifted magnetic configuration. The ion tail formed by the second-harmonic heating was enhanced by the injection of a perpendicular neutral beam.

AbstractThe 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 and minor radii of the plasma are 3.6 to 3.9 and 0.6 to 0.65 m, respectively. A plasma with an elliptic cross section confined in the helical magnetic field rotates poloidally along the magnetic axis and has no axial symmetry. For the installation of various kinds of diagnostic instruments, large-sized ports are equipped. The diameter of the largest horizontal ports is 2410 mm, which enables us to easily access the full plasma cross section with multichannel viewing chords aligned parallel to one another. For the precise measurement of plasma quantities in a three-dimensional helical plasma, an extensive set of diagnostics has been developed with national and international collaborators and is routinely operated in LHD. The diagnostic system now consists of [approximately]50 measuring instruments and includes many challenging diagnostics that have been developed and operated for the study of LHD plasma confinement. These are classified as profile diagnostics, fluctuation diagnostics, and advanced diagnostics, some of which are selected for introduction in this article. In addition, diagnostics for the divertor and for energetic particles are discussed, along with topics that are somewhat unique to helical devices such as diagnosing three-dimensional phenomena and flux surface mapping. This large number of diagnostics in LHD rely on a data acquisition system that has broken world records for the amount of information accumulated in one shot. Finally, looking to the near future, countermeasures have been taken to protect diagnostics from the neutrons and gamma fluxes anticipated during deuterium-deuterium experiments, such as placing much of the diagnostic instrumentation behind a 2-m-thick concrete biological shield encompassing the LHD test cell.

AbstractOn 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-ion tail formation during the experiments of ion cyclotron range of frequency heating. The fast-ion loss effect on the tail formation is 10 times more significant in the standard configuration than in the inwardly shifted configurations. The superiority is also confirmed by the comparison of steady-state fast-ion spectra for the neutral beam (NB)–heated plasmas and by the flux decay of fast neutrals after the short-pulse injection of radial NB.

AbstractFusion product diagnostics can be used to determine a fusion reaction rate, which indicates how close the plasma is to the ultimate goal of making a power plant based on nuclear fusion. However, these diagnostics can also provide large amounts of additional information, such as ion temperatures, the thermonuclear fraction in the fusion reaction rate, degree of fast ion confinement, fast ion loss mechanism, etc. Measurement systems for fusion product diagnostics are usually designed and optimized to a specific performance so that they play different roles in the experiment. The neutron emission rate, which is directly related to the fusion output, can be determined by (a) time-resolved emission monitors, which are well calibrated onsite, in combination with (b) activation systems and (c) profile monitors with accuracy up to several percent. The time-resolved neutron profiles also provide useful information for transport analysis. Velocity distributions and confinement properties of fast ions can be obtained from (d) the neutron spectrometers and (e) gamma-ray measurement. The interaction between plasma dynamics and fast ions can be studied with most fusion product diagnostic systems, especially with (f) escaping charged fusion product detectors. Each section of this chapter contains a general explanation of these systems, showing some experimental results obtained on present devices. A lot of interesting and useful information on the behavior of energetic particles and their degree of confinement are provided by them because interaction between thermal and nonthermal energetic ions and that among nonthermal ions contribute dominantly to the fusion reaction rate in present deuterium-deuterium experiments. In future deuterium-tritium fusion experiments on ITER, the contribution of thermonuclear fraction will be increased, and the combination of several neutron measurement systems will provide the absolute fusion output and neutron fluence on the first wall. Together with neutron measurement, alpha particle and gamma-ray measurement play important roles in research on self-heating burning plasma physics and hence in the burning control of the device.

Several unexplored physics issues will be studied in burning plasma experiments, planned on ITER in the near future. To carry out these studies, it is essential to develop novel diagnostic systems, such as fusion product measurement systems and high-resolution, high-reliability profile measurement of various plasma parameters. Scientific research on “Advanced Diagnostics for Burning Plasma Experiment” was assigned as a priority area by MEXT in 2004, and more than 20 studies are currently underway under this program. The feasibility of several new diagnostic concepts applicable to ITER is examined and diagnostic components for these systems are now under development.

AbstractElectrode biasing experiments under electrode current control were carried out in the Tohoku University Heliac and the Compact Helical System to examine the role of an ion viscosity on a transition to a high-confinement regime and to investigate the dependence of the ion viscosity on magnetic structure. Observations included (a) an increase of electron density, (b) an increase of electron stored energy, (c) a formation of the steep gradient of electron density, and (d) a formation of a negative electric field in both devices during electrode biasing negatively. The dependence of the ion viscosity normalized by the ion pressure on the poloidal Mach number qualitatively agreed with the neoclassical theory based on the Shaing model. This result supported the transition mechanism of the neoclassical theory based on ion viscosity, which advocates that the transition to a high-confinement mode is the bifurcation phenomenon resulting from the existence of local maximum in ion viscosity.

AbstractIn the Tohoku University Heliac, a high-density plasma is produced by a vanadium electrode. The vanadium electrode is pretreated for hydrogen storage. In biasing experiments using the vanadium electrode, a high-density plasma is observed in not only argon plasmas but also helium plasmas. When the vanadium electrode is biased negatively, the radial distribution of the electron density steepens at the electrode position, and a strong negative radial electric field is formed between the electrode and the last closed flux surface. The E× Bdrift velocity is 30 km/s, and the estimated poloidal Mach number Mpis -20. The measured beta value exceeded 0.5% in the low-field discharges.

The purpose of this work is to reveal the effects of the energetic particle mode (EPM) on fast-ion transport and consequent fast-ion loss in the compact helical system (CHS). For this purpose, fast particle diagnostics capable of following fast events originating from the EPM (f< 100?kHz) and from the toroidicity-induced Alfvén eigenmode (TAE) (f= 100-200?kHz) are employed in CHS. Experiments show that the EPM excited by co-circulating fast ions in an outward-shifted configuration is identified as a mode of m/n= 3/2 and can enhance fast-ion loss when its magnetic fluctuation amplitude exceeds ~4 × 10?5?T at the magnetic probe position. The lost fast-ion probe (LIP) located at the outboard side of the torus indicates that bursting EPMs lead to periodically enhanced losses of co-going fast ions having smaller pitch angles in addition to losses of marginally co-passing fast ions. Coinciding with EPM bursts, the H? light detector viewing the peripheral region at the outboard side also shows large pulsed increases similar to that of the LIP whereas the detector viewing the peripheral region at the inboard side does not. This is also evidence that fast ions are expelled to the outboard side due to the EPM. The charge-exchange neutral particle analyser indicates that only fast ions whose energy is close to the beam injection energy Ebare strongly affected by EPM, suggesting in turn that observed EPMs are excited by fast ions having energy close to Eb.

Confinement degradation of energetic ions due to Alfvén eigenmodes (AEs) has been investigated with negative ion based neutral beam injection at ~370?keV into JT-60U weak shear plasmas. AEs with a rapid frequency sweeping and then saturation as the minimum value of the safety factor decreases have been observed. This frequency behaviour can be explained by the reversed-shear induced AE (RSAE) and the transition from RSAEs to the toroidicity-induced AEs (TAEs). The associated change in the charge exchange neutral particle flux suggests energetic ion transport due to these AEs. The total neutron emission reduction rate in the presence of these AEs is evaluated by calculation using the orbit following Monte-Carlo code taking into account the change in bulk plasmas. The evaluation indicates confinement of energetic ions is degraded due to these AEs. In particular, it is found that the confinement degradation of energetic ions is maximum during the transition from RSAEs to TAEs and the maximum reduction rate (?Sn/Sn)MAXis estimated to be ~45%.

AbstractThe influence of the low-order rational surface (n,m) = (5, 3) on ion viscosity and radial electric field formation were investigated by electrode current sweep biasing with a hot cathode at the Tohoku University Heliac. In the improved mode, the position of the maximum electric field remained on the n/m= 5/3 rational surface. After the H-L transition, the local maxima shifted outward to the plasma periphery of ρ ~ 0.8. The low-order magnetic islands were formed resonating the magnetic Fourier components of (n,m) = (5, 3) by external perturbation coils. As the widths of the magnetic islands were increased, the biasing electrode current required for the improved mode transition increased. It was suggested that the ion viscosity increased according to the increase in magnetic island width. The increase in the biasing electrode current is equivalent to the increase in the driving force for the poloidal rotation. This suggests the possibility of an active viscosity control assisted by externally controlled island width and magnetic island rotation.

AbstractRemarkable progress to access the reactor-relevant regime has been made in a recent experiment in the Large Helical Device. Optimizing the rotational transform, the average beta value of 4.3%, which is the highest record among helical devices, was achieved. The high-performance plasma with a fusion triple product up to ~2.2 × 1019m−3·keV·s was sustained for >7 s by repetitive hydrogen pellet injection. With regard to steady-state operation, which is one of the key issues to realize a fusion reactor, discharges for >30 min were successfully sustained by ion cyclotron range of frequency heating with the aid of the magnetic axis swing technique to reduce the heat load to the plasma-facing component. In the discharge, the total input energy to the plasma reached 1.3 GJ, which also established a new record.

ICRF heating in the Large Helical Device is studied applying two global simulation codes; a drift kinetic equation solver, GNET, and a wave field solver, TASK/WM. Characteristics of energetic ion distributions in the phase space are investigated changing the resonance heating position; i.e. the on-axis and off-axis heating cases. A clear peak of the energetic ion distribution can be seen in the off-axis heating case because of the stable orbit of resonant energetic ions. The simulation results are also compared with experimental results evaluating the count number of the neutral particle analyzer and a relatively good agreement is obtained.

The density and potential fluctuations were measured in hot-cathode biasing plasma at the Tohoku University Heliac. In the improved mode, high-frequency fluctuations (>100?kHz) appeared in the density signal. On the other hand, low-frequency fluctuations (<100?kHz) in the density and potential signals were suppressed. The characteristics of high-frequency fluctuation were compared with three kinds of instability, and they were consistent with those of the flute instability driven by the supersonic poloidal rotation. The suppression of low-frequency fluctuations in improved mode is considered the effect of E× Bpoloidal rotation or its shear. The profile of the anomalous particle flux was estimated by analysing the low-frequency fluctuation signals. The flux decreased in the improved mode in most of the region, although the decrease in flux was small near the rational surface (n/m= 5/3).

A series of hot cathode biasing experiments with marginal conditions for improved mode transition were carried out in the Tohoku University Heliac (TU-Heliac). Spontaneous transitions were observed accompanied by a delay of a few milliseconds. Transition conditions were explored over a wide operation range. The transition points can be identified clearly and easily in the operation range, because the plasma parameters changed slowly until the spontaneous transition. Although operation conditions were spread over a wide range, poloidal Mach numbers for transitions were concentrated in the range of ?Mp= 1-2 and normalized driving forces for poloidal rotation agreed well with the local maximum value of ion viscosity predicted by neoclassical theory. The local maximum of ion viscosity against the poloidal Mach number was found to play a key role in the L-H transition. Marginal hot cathode biasing is suitable to determine the threshold conditions for the L-H transition.

In the Tohoku University Heliac (TU-Heliac), a helical axis stellarator, an electron injection electrode using a hot cathode made of LaB6was developed and the transition mechanism to an improved mode has been intensively studied. In the electrode current-sweepmode of biasing experiments, a bifurcation phenomenon, i.e. a negative resistance feature in the electrode characteristics was observed accompanied with transition to an improved mode (H-mode) or transition from H-mode to L-mode, in helium plasma discharges of a wide range of collisionality. The ion viscous damping force was estimated from the J× Bdriving force for poloidal rotation. The local maxima in viscosity were found at the poloidal Mach number around ?Mp~ 1-3, as predicted by neoclassical theory. It was also found that the negative resistance (a bifurcation phenomenon) was observed when the poloidal viscosity showed a local maximum.

Electrode bias experiments were carried out in Tohoku University Heliac (TU-Heliac). This paper presents a comparison of the plasma behaviour in discharge of the electrode current ramped up with that ramped down to investigate the bifurcation and hysteresis phenomena. Bifurcation of the electrode current against the bias voltage, i.e. negative resistance and hysteresis between (1) electron stored energy and input power from the electrode, (2) fluctuation level of ion saturation current and radial electric field and (3) poloidal momentum damping force and poloidal Mach number were observed. The driving force in the region where the hysteresis was observed corresponded to those in the negative resistance phase and those were located at local maximum of the ion viscosity. Hysteresis between the poloidal Mach number and the poloidal damping force was observed. However, the difference in ion viscosity, which was obtained by subtracting the friction term from the total poloidal damping force, between the electrode current ramped up/down in the transition region was rather small. The data in the negative resistance region corresponded to those in the regime where the viscosity decreased against the poloidal Mach number.

Due to the high neutron yield and the large plasma size many ITER plasma parameters such as fusion power, power density, ion temperature, fast ion energy and their spatial distributions in the plasma core can be measured well by various neutron diagnostics. Neutron diagnostic systems under consideration and development for ITER include radial and vertical neutron cameras (RNC and VNC), internal and external neutron flux monitors (NFMs), neutron activation systems and neutron spectrometers. The two-dimensional neutron source strength and spectral measurements can be provided by the combined RNC and VNC. The NFMs need to meet the ITER requirement of time-resolved measurements of the neutron source strength and can provide the signals necessary for real-time control of the ITER fusion power. Compact and high throughput neutron spectrometers are under development. A concept for the absolute calibration of neutron diagnostic systems is proposed. The development, testing in existing experiments and the engineering integration of all neutron diagnostic systems into ITER are in progress and the main results are presented.

To investigate energetic ion transport induced by bursting modes in the frequency range of Alfvén eigenmodes, which is called abrupt large-amplitude events (ALEs) driven by negative-ion-based neutral beam (N-NB) injection, neutron emission profile measurement and charge exchange (CX) neutral particle (flux) measurement using a natural diamond detector have been performed simultaneously in JT-60U. It is found from the CX neutral particle (flux) measurement that energetic neutral particles in a limited energy range (100–370 keV) are enhanced due to ALEs, and the neutron radial profile is flattened. The change in the energetic ion density profile inferred from these measurements indicates that ALEs expel energetic ions from the core region of the plasma and induce both redistribution and loss of energetic ions. It has been shown that the energy range of transported energetic ions is consistent with a resonance condition between energetic ions and ALEs, and the energetic ion transport results from the resonant interaction between energetic ions and ALEs. Further, a fraction of the energetic ion loss has been quantitatively estimated to be ∼4%.

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.

AbstractConfinement of energetic ions from neutral beam injection heating is investigated by changing the magnetic field configuration of the Large Helical Device from a classical heliotron configuration to an optimized neoclassical transport configuration to a level typical of “advanced stellarators.” The experimental results show the highest count rate of fast neutral particles not in the optimized configuration but in the inward-shifted one. The GNET simulation results show a relatively good agreement with the experimental results, and they also show a lower energy loss rate in the optimized configuration. This contradiction can be explained by the radial profile of the energetic ions. The relatively good agreement between experimental and simulation results suggest that ripple transport (neoclassical) dominates the energetic ion confinement and that the optimization process is effective in improving confinement in helical systems.

AbstractRecent progress in the Large Helical Device (LHD) experiment during the last 2 yr is reviewed. The LHD has been extending its operational regime toward fusion-relevant conditions while taking advantage of the net-current-free heliotron concept employing a superconducting coil system. Heating capability has exceeded 10 MW, and the central ion and electron temperatures have reached 7 and 10 keV, respectively. The maximum values of beta and pulse length have been extended to 3.2% and 150 s, respectively. Several encouraging physics observations have been obtained, i.e., simultaneous achievement of the mitigation of the magnetohydrodynamic instability criteria and good confinement, and formation of an internal transport barrier. The initial results have been obtained using a local island divertor, which shows the possibility of particle control at the plasma edge.

Fusion product measurements planned for ITER are reviewed from the viewpoint of alpha particle-related physics studies. Recent advances in fusion plasma physics have extended the desirable measurement requirements to the megahertz region for neutron emission rate, better resolution of neutron profiles for the study of internal transport barriers (ITBs), etc. Employing threshold counters and/or scintillation detectors confers megahertz capability on neutron emission rate measurement. The changes in the neutron/alpha particle birth profile due to the formation of ITB and its deviation from uniformity on the magnetic flux surface can be measured by addition of eight viewing chords in an equatorial port plug and seven viewing chords from the divertor to the original radial neutron camera. On the other hand, it is still difficult to measure the distributions of confined and escaping alpha particles. Several proposals to resolve these difficulties are currently under investigation.

The behaviour of helical ripple-trapped particles in large helical device (LHD) have been investigated by two line-of-sights natural diamond detector measurements in plasma sustained with ICRF for two types of resonance configurations, i.e. on-axis and saddle-shaped resonances. The tail temperature increases linearly as the Stix-temperature increases and then saturation follows, indicating the effect of the finite confinement time of the energetic ions. In the linear region, the tail readily increases when the line-of-sight is focused on the larger resonance area. The confinement time of the energetic ions is obtained from the empirical relation based on the Stix theory. The configuration with a saddle-shaped resonance has a longer confinement time than the on-axis resonance. Analyses of experimental measurements are consistent with the theoretical calculations including collisional ripple-induced transport.

Energy confinement and heat transport of net-current-free NBI heated plasmas in the large helical device (LHD) are discussed with emphasis on density and power deposition profile dependences. Although the apparent density dependence of the energy confinement time has been demonstrated in a wide parameter range in LHD, the loss of this dependence has been observed in the high density regime under specific conditions. Broad heat deposition due to off-axis alignment and shallow penetration of neutral beams degrades the global energy confinement while the local heat transport maintains a clear temperature dependence, lying between Bohm and gyro-Bohm characteristics. The central heat deposition tends towards an intrinsic density dependence like τE∝(n¯e/P)0.6 from the state where density dependence is lost. The broadening of the temperature profile due to the broad heat deposition profile contrasts with the invariant property that has been observed widely as profile resilience or stiffness in tokamak experiments. The confinement improvement as a result of the inward shift of the magnetic axis is obvious in the core region, which emphasizes the improvement of transport because of the geometry being unfavourable for the central heating of NBI in this configuration. The edge pressure, clearly, does not depend on the magnetic axis position. Unlike a tokamak H-mode, the edge pressure is determined by transport and can be increased by increasing the heating power.

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.

Ion temperature at the plasma centre has been measured from Doppler broadening of Ti  (2.61 &Angstrom;) and Ar  (3.95 &Angstrom;) 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 T_i>T_e was found in low-density (n_e&lt;10¹³ cm⁻³) H-minority ICRF discharges operated in the ion-heating regime (P_i&geq;P_e), whereas the ion temperature was roughly equal to the electron temperature in NBI discharges operated in the electron-heating regime (P_i&lt;P_e). The ion temperature in the ICRF and ICRF + NBI discharges was correlated with the input power to bulk ions over the ion density (P_i/n_i). As a result, a good correlation was obtained between them and it also indicated a clear rise of the ion temperature for the increasing P_i in a range of P_i/n_i&lt;5 (MW/10¹³ cm⁻³).

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.

Charge transfer cross sections of the ground state O⁺(⁴S) ions with H₂, CO, CO₂, CH₄, C₂H₂, C₂H₆ and C₃H₈ molecules have been measured using an initial growth rate method in the energy range 0.2-4.5 keV. In the present experiment, the ground state O⁺(⁴S) ions have been produced by carefully energy-controlled electron impact to avoid the influence of the metastable O^+*(²D, ²P) state ions. The observed cross sections are compared with calculations based on a formula developed by Olson. The energy dependence of the present cross sections for H₂ molecules shows a minimum at around 2 keV. The present experimental results for CO and CO₂ molecules are slightly different from previous measurements. For hydrocarbon molecules, where no experimental data are available so far, both the present experimental and calculated results are in good agreement with each other, if product molecular ions are assumed to be in some vibrationally excited states.

The experimental results from the Large Helical Device (LHD) heliotron / torsatron of the first two years are reviewed. The world's largest superconducting helical coils have been driven up to 2.9 Tesla on the axis which is close to the designed value (3 T). The obtained plasma performances are better than those predicted by the database from the medium-size helical devices. These improvements are attributed mainly to the optimization of a magnetic field configuration which can be controlled by shifting the magnetic axis inward than that of standard case. This configuration improves particle orbits of trapped high energy ions resulting in success of ICRF heating in LHD. Efforts have also been made on steady state plasma operation, and long pulse discharges more than one minute have been achieved both by ICRF and NBI. It should be noted that the feature of no current-disruption in helical plasma makes the discharges easy.

A high ion temperature (T_i) mode is observed for neutral beam heated plasmas in the Compact Helical System (CHS) heliotron/torsatron. The high T_i mode plasma is characterized by a high central ion temperature, T_i(0), and is associated with a peaked electron density profile produced by neutral beam fuelling with low wall recycling. Transition from L mode to high T_i mode has been studied in CHS. T_i(0) in the high T_i mode discharges reaches 1 keV, which is 2.5 times higher than that in the L mode discharges. The ion thermal diffusivity is significantly reduced by a factor of more than 2-3 in the high T_i mode plasma. The ion loss cone is observed in neutral particle flux in the energy range 1-6 keV with a narrow range of pitch angle (90° ±20°) in the high T_i mode. However, the degradation of ion energy confinement due to this loss cone is negligible.

Abstract: The largest superconducting fusion machine, Large Helical Device (LHD), is now under construction in Japan and will begin operation in 1997. Design and construction of related R&D programs are now being carried out. The major radius of this machine is 3.9 m and the magnetic field on the plasma center is 3 T. The NbTi superconducting conductors are used in both helical coils and poloidal coils to produce this field. This will be upgraded in the second phase a using superfluid coil cooling technique. A negative ion source is being successfully developed for the NBI heating of LHD. This paper describes the present status and progress in its experimental planning and theoretical analysis on LHD, and the design and construction of LHD torus, heating, and diagnostics equipments.

An ion cyclotron range of frequencies (ICRF) heating experiment was performed on the Compact Helical System (CHS) device in order to identify the problems in establishing this as a reliable heating method. Radiofrequency heating was applied to plasmas with two ion species produced by ECH. Stored energy increased upto 2.2 kJ with the application of a 590 kW ICRF heating pulse, giving a heating efficiency comparable to that of NBI heating. More importantly, the `flat-top' of the stored energy was maintained during the pulse. Good heating was realized with 30% proton and 70% deuteron plasmas at high density (4*10¹⁹ m^-3). It was found that the best discharges are dominated by electron ICRF heating, which seems to be the reason for the successful results. On the contrary, a flat-top in stored energy has never been obtained in discharges where ion heating dominates. The loss of high energy ions with large pitch angles produced by ion heating appears to degrade the performance and limit the duration of the heating. A similar degradation was encountered in previous ICRF heating experiments in helical systems. A newly designed detector was used to study the loss of the helically trapped particles

The effect of up-down asymmetric toroidal field ripple on fast ion loss was investigated experimentally in JT-60U deuterium discharges. The experiment was carried out for neutral beam (NB) injected fast ions and tritons from the d(d,p)t reaction. The loss of the NB injected fast ions was estimated from the decay of 2.5 MeV neutrons following a short pulse D⁰ beam injection of 30 ms, and the triton loss was investigated from the decay of 14 MeV neutrons following 25 MW neutral beam injection (NBI) with a duration of 2 s. It was found that up-down asymmetric ripple did not enhance the NB injected ion loss compared with the ion loss in up-down symmetric ripple. The effect of the ion Del B drift direction on the beam ion and triton losses was studied in up-down asymmetric ripple. No significant difference was observed in the loss for energetic ions between upward and downward Del B drift discharges. The loss fraction of NB injected ions predicted by an orbit following Monte Carlo code agrees with the experimental fraction

A large, rapid change in the local plasma potential is found to be induced by off-axis hydrogen ice pellet injection into a tokamak plasma. The polarity of the change during the ablation is reversed when the pellet is injected into the upper and lower halves of the poloidal plasma cross-section. This change can be interpreted as being due to the Del B drift of particles in the high density plasmas of the pellet cloud, before the increase of the plasma density due to the ablation becomes uniform on the magnetic surface

Fast changes of electric potential with different polarities are observed during sawtooth oscillations in the core region of a tokamak plasma using a heavy ion beam probe. Near the inversion radius the polarity of the observed change of the potential is found to be dependent on the swift movement of the hot core at the crash and is consistent with the prediction of one-fluid MHD theory. Near the magnetic axis the change of the potentials is positive and outside the inversion radius the change is negative. This is in contradiction with the MHD prediction

In the JIPP T-IIU tokamak, a high power ICRF heating experiment has been conducted, up to an extremely high power density (~2 MW·m?3), with a total RF power of PRF= 2 MW. Great attention has initially been paid to the problem of impurities, and it has been found that (a) the adoption of low Z materials for the limiter, (b) in situ carbon coating (i.e. carbonization) and (c) adequate gas puffing synchronized to the RF pulse are very effective in suppressing radiation loss. With these methods, a remarkable reduction in metal impurities (especially in iron impurity) has been achieved; the total radiation loss has been reduced to less than 30-40% of the input power. In these reduced radiation loss plasmas, the characteristics of ICRF heated plasmas have been studied intensively. With an increase in the ICRF heating power, a deterioration of the energy confinement time has been observed, indicating quantitative agreement with the Kaye-Goldston L-mode scaling. It is shown that the so-called profile consistency, which is the leading feature in neutral beam heated plasmas, also holds in ICRF heated plasma. It has been observed that the electron temperature profile only responds to the safety factor q(a) and does not change when the deposition profile is controlled by tailoring the k1spectrum.

Data of JIPP T-II ICRF heating experiments are presented. The experiment covers three typical cases: the low-concentration hydrogen minority case, the high-concentration hydrogen minority case, and the 3He-minority case. The best heating efficiency is obtained for the 3He-minority case. It is shown through power balance analysis that the two H-minority cases are different in the wave energy deposition profile. The difference is explained by the presence of a local cavity mode in the high-concentration minority case. The ion temperature stops rising at a power density level of 0.65 W-cm?3for the hydrogen minority experiments. No such deterioration is found in the case of the 3He minority experiment up to a power density level of 0.45 kW-cm?3. This is the maximum possible to attain within the maximum power injection (180 kW) up to which the experiment is conducted. An analytic solution of the Fokker-Planck equation is derived in order to interpret the deterioration of the heating efficiency.

A confined alpha particle diagnostic using pellet injection is proposed. The alpha particle density can be obtained from the number of decaying photons emitted from the radiative decay of He+*(n ? 2) ions produced through the one-electron transfer reaction of an alpha particle by a probing ion ablated from the pellet before it is fully ionized. The counting rate, estimated using theoretical cross-sections, and the calculated pellet ablation rate are sufficient for observation against the bremsstrahlung background. The alpha particle velocity distribution can also be obtained from the Doppler broadening of the emission lines.

The paper discusses the results of a heating experiment in the ion cyclotron range of frequencies (ICRF) conducted on the JIPP T-IIU device with a five-element antenna array. This antenna system provides precisely shaped k?, spectra which vary according to the relative phasing angle of the neighbouring antennas, ?. In the experiment, ?is scanned in a close pitch and it is found that the electron heating efficiency is drastically improved with ?= ? and the lowest efficiency is obtained with ?= 0. This observation is analysed by calculating the k?spectra and then the power deposition profiles using a ray tracing code. It is also found that the effect of the phasing on the impurity problem is unexpectedly small. The reasons for the discrepancies between these data and theoretical expectations are also discussed.