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Electron cyclotron current drive (ECCD) experiments were conducted in the stellarator/heliotron (S/H) devices, such as Heliotron J, TJ-II, CHS, and LHD. Experimental results show that ECCD can be controlled by the power injection angle, absorption position and magnetic field structure. The current drive efficiencies are similar, γ = neIECR/PEC = 8 - 16×1016 A/Wm2 , ζ = 32.7neIECR/PWTe = 0.03-0.05, when the magnetic field ripple ratio is 0.93 < Bmin/Bmax < 1.0. The reversal of driven current direction is observed depending on the magnetic field ripple structure, indicating that the amplitude and direction of EC current is determined by the balance between the Fisch-Boozer effect and the Ohkawa effect, and that the Ohkawa effect is stronger in S/H devices compared with tokamaks. Control of net toroidal current by using ECCD is demonstrated; a net zero current state is attained by cancelling the bootstrap current.

An active current drive is a promising technique for improving plasma performances by controlling rotational transform and/or magnetic shear profiles in helical devices. A current drive based on electron cyclotron resonance heating is the most appropriate scheme for this purpose in terms of locality of driven current. Optimum conditions for an efficient electron cyclotron current drive (ECCD) in Large Helical Device (LHD) are being investigated using three-dimensional ray-tracing code, which can simulate propagation and power dissipation of electron cyclotron waves with large parallel refractive index. In the present experiment, inversion of directions of driven plasma current corresponding to injected ECCD modes was demonstrated successfully, and the results could be elucidated by the Fisch-Boozer theory. In addition, clear shifts of rotational transform were observed by motional Stark effect polarimetry. Our findings verified that the ECCD can be used as an effective actuator for controlling the rotational transform and magnetic shear profile in LHD.

Density fluctuations are analyzed in high volume average beta and high core density discharges in the Large Helical Device (LHD) using a 2D phase contrast imaging system and far infra-red interferometer. Both these regimes share similarly high beta gradients and evidence of pressure driven MHD modes is presented. In high volume average beta plasmas, both large and ion-gyro scale density fluctuation levels increase with beta and, in the edge, compare favorably with growth rate of resistive interchange modes, showing additional dependence on density at fixed β. In high core density plasmas with internal diffusion barrier, intermittent fluctuation bursts around mid radius are observed which are triggered when the normalized density gradient exceeds a certain threshold. The intermittent character is stronger for outward shifted plasmas and there appears to be a fluctuation suppression mechanism, possibly related to temperature gradient.

High ion temperature (Ti) hydrogen plasmas were successfully demonstrated in the LHD in an experimental campaign (FY2006). The power increase of the perpendicular neutral beam injections (NBIs) has mainly contributed to this achievement. Ti exceeded 5 keV at the average plasma density (ne) of 1.2 × 1019 m-3 and also achieved 3 keV at ne > 3 × 1019 m-3. Neoclassical (NC) analysis was performed for such plasmas. We confirmed that even if Ti or the plasma density became higher, NC transport was remained to a certain value because of the existence of the ambipolar radial electric field (Er), which was roughly two orders of magnitude smaller for cases not involving the existence of ambipolar Er. Besides, a Ti and ne scan of discharge 75235 was also performed to consider the plasma parameter dependence of NC transport. Using these calculations, it is shown that NC ion thermal diffusivities are reduced to a small percentage of NC electron thermal diffusivites even at the fusion reactor relevant parameters such as ne ∼ 1 × 1020 m-3 and Ti ∼ Te ∼ 10 keV.

Three states of a magnetic island are observed when the magnetic shear at the rational surface is modified using inductive current associated with the neutral beam current drive in the Large Helical Device. One state is the healed magnetic island with a zero island width. The second state is the saturated magnetic island with partial flattening of the T(e) profile. The third state is characterized by the global flattening of the T(e) profile in the core region. As the plasma assumes each of the three states consecutively through a bifurcation process a clear hysteresis in the relation between the size of the magnetic island and the magnetic shear is observed.

In a circular cross-section plasma bounded by a limiter, the {ital H}-mode transition is triggered by a rapid rampdown in plasma current during auxiliary heating, even in the case when the edge electron temperature gradually decreases prior to the transition. This result suggests that the transition is governed by the enhancement of the magnetic shear near the plasma edge, associated with the radial modification of the edge current-density profile.

In charge exchange spectroscopy (CXS), a simultaneous observation in different plasma toroidal cross sections and/or viewing ports is required to investigate radial distributions of ion temperatures Ti (r), and poloidal rotation velocities Vp (r) in magnetically confined toroidal plasmas. In recent studies of the edge transport barrier (ETB) in the Compact Helical System (CHS), a simultaneous viewing of the vertically elongated and the horizontally elongated plasma cross sections is used to improve the spatial resolution at the edge region. The 90 fibers used for this purpose are connected to one spectrometer, and a 256 ×243 pixel sampling CCD is used to detect the diffraction image. It is found that there is a localized edge ion temperature pedestal region with ΔTi ? 100 eV and Δr/a ? 0.1, where r and a are flux surface averaged minor radii of measured surfaces and the outermost flux surface, respectively. The negative radial electric field at the edge is increased in the high confinement phase because of the increased ion pressure.

The ergodization of the magnetic field lines imposed by the dynamic ergodic diverter (DED) in TEXTOR can lead both to confinement improvement and to confinement deterioration. The cases of substantial improvement are in resonant ways related to particular conditions in which magnetic flux tubes starting at the X points of induced islands are connected with the wall. This opening process is connected with a characteristic modification of the heat deposition pattern at the divertor target plate and leads to a substantial increase and steepening of the core plasma density and pressure. The improvement is tentatively attributed to a modification of the electric potential in the plasma carried by the open field lines. The confinement improvement bases on a spontaneous density built up due to the application of the DED and is primarily a particle confinement improvement.

The role of zonal flows (ZFs) in the formation of an internal transport barrier in a toroidal helical plasma is analyzed. The turbulent transport coefficient is shown to be suppressed when the plasma state changes from the branch of a weak negative radial electric field to the strong positive one. This new transition of turbulent transport is caused by the change of the damping rate of the ZFs. It is clearly demonstrated, theoretically and experimentally, that the damping rate of the ZFs governs the global confinement of toroidal plasmas. © 2007 American Institute of Physics.

In the present study, a coherent density fluctuation similar to the edge harmonic oscillation (EHO) in tokamaks was observed in the edge transport barrier discharge in the Compact Helical System (CHS) [K. Matsuoka et al., Plasma Physics and Controlled Nuclear Fusion Research, 1988 (International Atomic Energy Agency, Vienna, 1989), Vol. 2, pp. 441] using beam emission spectroscopy (BES). The fluctuation had both fundamental (f=4.5 kHz) and second-harmonic (2f=9 kHz) frequencies. EHO in CHS had a peak amplitude at approximately ρ=0.95. The mode has a continuous phase shift in the radial direction. If this is interpreted as the radial propagation, the mode propagates in the outer radial direction at an apparent phase velocity of several hundreds of meters per second, which is a characteristic similar to the radial phase shift of EHO in tokamaks.

An overview is given of the National Centralized Tokamak (NCT) programme as a research programme for advanced tokamak research to succeed JT-60U. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility is pursued in aspect ratio and shape controllability for the demonstration of the high-β steady-state, feedback control of resistive wall modes, wide current and pressure profile control capability and also very long pulse steady-state operation. Existing JT-60 infrastructure such as the heating and current drive system, power supplies and cooling systems will be best utilized for this modification.

An edge transport barrier (ETB) similar to the tokamak H-mode has been observed for beam-heated plasma with two co-injected Neutral Beam Injectors (NBIs) in CHS. The Hα emission showed a clear spontaneous drop followed by an increase of line-averaged electron density at the L-H transition. Stored energy increased by ˜ 40% with H-factor improvement up to ˜ 30% compared to the international stellarator scaling (ISS04v03). A rapid density increase in the edge region to two-times level of the L-mode was observed to accompany a ˜ 15% increase in the density gradient. An ETB was formed when the plasma density exceeded the lower density limit by ˜ 1.5 × 1013 cm-3 and the total NBI power exceeds the threshold level (Pdeposit / ne ˜ 200 × 10-13 kW/cm-3 at BT = 0.95 T, Rax = 92.1 cm, where Pdeposit is the deposited NBI power, ne is electron density, BT is troidal magnetic field strength and Rax is location of magnetic axis). The power threshold increased with the magnetic field strength, as in tokamak scaling, and depends on the magnetic field configuration of the helical plasma.

The effects of resonant magnetic fluctuations on plasma confinement have been investigated in current carrying high-β plasmas in magnetic hill configurations of the Large Helical Device. The m/n= 2/1 mode excited in core region was dominantly observed in the unstable configuration as predicted by linear theory on ideal instability, and the plasma current decreasing magnetic shear enhances the mode activity. The disappearance of the m/n = 2/1 mode was observed with an increase in plasma current, and then the plasma confinement and the beta value increase by less than 10% with the recovery of core pressure. When the plasma current exceeds a certain value, the plasma confinement is improved by 20% with the reduction of the amplitudes of the modes in the peripheral region in addition to the disappearance of the m/n = 2/1 mode. Then the beta value increases by 38%. MHD activities led to no disruption even in such an unstable configuration, and the moderate plasma current mitigated the effect of MHD activity on the plasma confinement.

A minor collapse observed in current-carrying plasma has been investigated in Large Helical Device (LHD). The magnetic configuration with high central rotational transform has ι/2π = 1 surface at the core region and is relatively unstable for the m/n = 1/1 mode (here, m and n are the poloidal and toroidal mode number, respectively). When the beam-driven current exceeds a certain value, the m/n = 1/1 mode grows with a growth time of ˜30 ms and causes a sudden drop of the plasma stored energy and the electron temperature, and it also limits the plasma current itself. A local flattening in an electron temperature profile appears just after the minor collapse. The mode does not rotate and stays at the same spatial location. The possibility of pressure- and current-driven magneto-hydro dynamics (MHD) instabilities is discussed.

This paper presents experimental studies of nonlinear processes between meso-scale structures and turbulent fluctuations, by use of bispectral analysis, in the edge region of toroidal plasmas. In ohmically heated plasmas of the JFT-2M tokamak, the drift wave-zonal flow system is investigated by observation of the potential fluctuations. The total bicoherence is composed of two components, i.e., the peak at the geodesic acoustic mode (GAM) frequency and broadband distribution. The biphase is also obtained, which indicates the phase angle of coherent interaction between GAM oscillations and broadband fluctuations. Bispectral functions (e.g., magnitude and phase) give an order-of-estimate agreement for the drift wave-zonal flow systems. Bicoherence analysis is also applied to the edge plasmas of the high-confinement mode (H-mode) of the Compact Helical System (CHS). In the case of CHS plasmas, nonlinear couplings in a system composed of coherent global magneto-hydrodynamic (MHD) oscillations and broadband fluctuations (up to hundreds kHz range) have been observed. Bispectral analysis enhances our understanding of the self-regulation systems of various toroidal plasmas in which strong turbulence and meso-scale structures coexist.

An improved confinement mode with reheat and edge transport barrier is observed in the Compact Helical System (CHS). This mode provides enhanced confinement in high density region (ne ˜ 1.2 × 1020 m-3) due to the increase in temperature while high density is maintained in the peripheral region.

The change of rotational transform due to noninductive plasma current driven by a negative neutral beam, which has a value typically less than 10% of total rotational transform determined by the external current in the helical coils, is measured in the Large Helical Device with motional Stark effect spectroscopy. Radial profiles of rotational transform are derived from the radial profiles of the polarization angle of the sigma component in the Halpha line emitted from the high energy hydrogen atom of the beam with four sets of linear polarizers, spectrometers, and CCD detectors. The radial profile of the change in the rotational transform due to the noninductive toroidal current driven by the neutral beam is measured.

A neutral lithium beam probe for two-dimensional diagnosis of edge plasmas has been designed and installed on the compact helical system. A lithium beam with an energy of 15keV and a current of 0.1mA is used. The spatial resolution is about 10mm, and the time response is about 10ms. The beam penetration depth is expressed in terms of the line integral density ⟨nel⟩, which is about 2×1018m−2. The beam injection angle can be varied and the observation point covers the edge and separatrix region of the helical diverter configuration. Two-dimensional electron density profiles for electron cyclotron heating and neutral beam injection (NBI) heated plasmas are obtained near and outside the last closed flux surface (LCFS). Analysis for two-dimensional density profile reconstruction indicates that significant amounts of surface plasma are confined outside the LCFS for NBI plasmas even though the ergodic layer is cut by the vacuum chamber wall (inboard limiter configuration). The usefullness of this new two-dimensi...

The radial electric field profiles are derived from poloidal rotation velocity for a plateau regime plasma heated by a neutral beam in a CHS device. The measured electric field is more negative than the neoclassical prediction especially near the plasma periphery. However, the negative radial electric field is not directly related to the reduction of thermal diffusivity. Physics of Fluids B: Plasma Physics is copyrighted by The American Institute of Physics.

Two-dimensional profiles (32×16 spatial channels) of energy spectra of x-ray emission are measured with an energy resolution of 0.16 keV using the photon counting soft x-ray charge coupled device (CCD) camera by optimizing the intensity of x ray with attenuation Be filter in the compact helical system. The energy spectra measured with the CCD camera agrees with that estimated from electron temperature and density measured with YAG Thomson scattering.

Boronization with decaborane which is solid at room temperature has been successfully applied to the Compact Helical System (CHS) experiment. The effect of boronization on neutral-beam-injection (NBI)-heated plasma has been compared with the optimized case of titanium gettering. The concentration of oxygen has been reduced by a factor of 2 with boronization. Metal impurities can also be completely suppressed. Plasma performance has been improved in the high-density region with \barn e> 5×1019 m-3. The effective lifetime covers at least 200 NBI discharges while the favorable condition with titanium gettering has an effective lifetime of only 20∼30 discharges.

In an effort to understand the role of neutral beam orientation and plasma geometry in stabilizing fishbone activity in PBX, comparisons were made between near-perpendicular and tangential injection, and discharges with and without significant (>10%) indentation. Periodic spikes in the fast neutral flux and drops in the neutron production, reminiscent of sawtooth oscillations observed during tangential injection in PLT, were observed during tangential injection in PBX. Changes in the fast neutral signal and neutron emissivity were also measured in association with high frequency (120-210 kHz) instabilities in these discharges. Although neither phenomenon has been correlated yet with a mechanism like fishbones for expelling beam ions from the plasma, their depletion from the center of the discharge suggests the need for controlling them to ensure efficient heating of the plasma center with tangential beams.

The effect of the current density profile on the limit of the safety factor q (a) in the non-circular tokamak TNT-A is described in this paper. The safety factor q (a) is measured by magnetic probes and one-turn loops. The current density profile is estimated from the shape of the plasma boundary with an equilibrium code. A gas puffing system and pulse forming network (PFN) circuit for the Joule current are used to control the profile. The limit of q (a) increases from 2.6 to 4.0 as the current density profile becomes flatter and the current density profile parameter enters the unstable region for the m =3 and 4 external kink modes. This limit of q (a) determines one of the boundary of the operational stable region in TNT-A.

Impurity studies in L- and H-mode discharges in PBX have shown that both types of discharges can evolve into either an impurity accumulative or nonaccumulative case. In a typical accumulative discharge, Z eff peaks in the center to values of about 5. The central metallic densities can be high, n met / n e ≈ 0.01, resulting in central radiated power densities in excess of 1 W/cm 3 , consistent with bolometric estimates. The radial profiles of metals obtained independently from the line radiation in the soft X-ray and the VUV regions are very peaked. Concurrent with the peaking, an increase in the impurity influx coming from the edge of the plasma is observed. At the beginning of the accumulation phase the inward particle flux for titanium has values of 6 × 10 10 and 10 × 10 10 particles / cm 2 s at minor radii of 6 and 17 cm. At the end of the accumulation phase, this particle flux is strongly increased to values of 3 × 10 12 and 1 × 10 12 particles / cm 2 s. This increased flux is mainly due to influx from the edge of the plasma and to a lesser extent due to increased convective transport. Using the measured particle flux, an estimate of the diffusion coefficient D and the convective velocity ν is obtained.

Intrinsic impurity behaviour and transport properties in neutral beam heated L- and H-mode PBX tokamak plasmas were studied with a variety of impurity diagnostics. Central impurity accumulation was most often observed in H-mode discharges; sometimes it resulted in a thermal collapse due to high central metallic radiation (~1.5 W • cm−3). The impurity accumulation was evident from peaked Zeff and radiated power profiles and was further substantiated by specific vacuum ultraviolet and X-ray spectroscopy measurements. It is shown that impurity accumulation was neither unique nor inevitable in H-mode discharges and it could be suppressed by sufficient gas puffing. Central impurity accumulation was also seen in L-mode plasmas even with co-injected neutral beams. This usually occurred at high beam power and relatively low density. While there was no significant difference in the degree of accumulation between L-mode and H-mode discharges, the Zeff profile itself was more peaked in the H-mode because the electron density profiles are flatter in H-mode plasmas than in L-mode plasmas. The degree of accumulation increased as Zeff(0) itself increased, which suggests that neoclassical convection and diffusion driven by both impurity-impurity and impurity-plasma ion collisions contribute to the central plasma particle transport.

The effects of MHD activity on intrinsic impurity transport are studied in ohmic discharges of the Princeton Beta Experiment (PBX) by measuring the Zeff profile from visible bremsstrahlung radiation and spectral line intensities from ultraviolet spectroscopy. A diffusive/convective transport model, including an internal disruption model, is used to simulate the data. The Zeff profile with no MHD activity is fitted with a strong inward convection, characterized by a peaking parameter cv(=-a2 nu /2rD)=11 (-3.5, +4.5). At the onset of MHD activity (a large m=1 n=1 oscillation is followed by sawteeth), this strongly peaked profile is flattened and subsequently reaches a new quasi-equilibrium shape. This new profile is characterized by reduced convection (cv=3.6 (-1.1, +1.6), D=1.4 (-0.7, +5.6)*104 cm2 s-1) in addition to the particle redistribution which accompanies the sawtooth internal disruptions, indicating that the transport processes themselves change with the onset of sawtooth activity.

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 k1 spectrum.

An ion temperature profile has been obtained with the CVI 5292-Å (n=8 7) charge-exchange recombination (CXR) line using a space- and wavelength-resolving visible spectrometer installed on the JIPP TII-U tokamak. Two sets of 50-channel optical-fiber arrays, one viewing a fast neutral hydrogen beam (CXR channels) and the other viewing off the neutral beamline (background channels), are arranged on the entrance slit of the spectrometer. This spectrometer is coupled to an image intensifier and CCD detector at the focal plane, and provides a temperature profile every 1/60 s. An ion temperature is derived from the Doppler-broadened line profile after subtracting the simultaneously measured cold component (background channels), which is due to electron excitation and/or charge-exchange recombination in the plasma periphery. An alternative approach to obtain the ion temperature profile without CXR is also demonstrated. This method is based on an Abel inversion technique for each wavelength separately.

A series of initial resonant helical divertor (RHD) experiments have been carried out in ohmically and auxiliary heated JIPP T-IIU plasmas. Disruptive and MHD instabilities make the interpretation of the RHD results difficult but an apparent increase in the energy confinement time is observed when the helical magnetic perturbation is applied. This may be due to the suppression of MHD activity or to a reduction in the edge convective heat losses. Magnetic island effects have been observed on the floating potential of a Langmuir probe array and energy scrape-off layer widths have been measured with and without helical perturbations during ICRF operation. Basic pump limiter data is presented including ion temperatures and C4+ impurity profiles. Energy confinement times are reported in ohmically and NBI heated discharges.

Bean-shaped configurations favorable for high ..beta.. discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present ..beta.. limit.

The observation of mode conversion during ICRF heating in a deuterium-hydrogen plasma is described. The mode-converted ion Bernstein wave is detected near the ion-ion hybrid resonance layer by 2-mm microwave scattering. In addition, a fast magnetosonic wave is also observed by magnetic probes inserted into the plasma. The plasma dispersion of both waves as determined experimentally agrees well with the results of calculations for a hot plasma.

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, Δ. In the experiment, Δ is scanned in a close pitch and it is found that the electron heating efficiency is drastically improved with Δ = π and the lowest efficiency is obtained with Δ = 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.

Charge-dependent convective impurity transport is observed in the central core of PBX H-mode discharges by measurement of radial profiles for both low- and high-Z intrinsic impurities in the presence of strong sawtooth activity. The time-evolving ${Z}_{\mathrm{eff}}$ profile and central metallic densities are modeled to yield a diffuison coefficient and a Z-dependent inward convective pinch which are in general agreement with estimates of neoclassical transport in the central plasma region.

Fueling requirements and impurity levels in neutral-beam-heated discharges in the PDX tokamak have been compared for plasmas formed with conventional graphite rail limiters, a particle scoop limiter, and an open or closed poloidal divertor. Gas flows necessary to obtain a given density are highest for diverted discharges and lowest for the scoop limiter. Hydrogen pellet injection provides an efficient alternative fueling technique, and a multiple pellet injector has produced high density discharges for an absorbed neutral beam power of up to 600 kW, above which higher speeds or more massive pellets are required for penetration to the plasma core. Power balance studies indicate that 30–40% of the total input power is radiated while ~15% is absorbed by the limiting surface, except in the open divertor case, where 60% flows to the neutralizer plate. In all operating configurations, Z eff usually rises at the onset of neutral beam injection. Both open divertor pl;asmas and those formed on a well conditioned water-cooled limiter have Z eff ⪅ 2 at the end of neutral injection. A definitive comparison of divertors and limiters for impurity control purposes requires longer beam pulses or higher power levels than available on present machines.

The energy confinement characteristics of ICRF-heated Tokamak plasmas are studied at high input power density approximately 2 MW m-3 volume averaged, on the JIPP T-IIU device (R=0.91 m, a=0.23 m). High electron and ion temperatures (Te approximately 2.5 keV, Ti approximately 2.0 keV, at each maximum) have been achieved by the operation at a plasma current IP of 280 kA, plasma line-averaged electron density ne of 7*1013 cm-3 and input power of 2 MW, with a suppression of total radiation loss (30-40% of the total input power) by a carbon coating on the vacuum vessel. The fraction of ICRF power absorbed by the plasma, alpha , is determined experimentally from the decay of the stored plasma energy just after the ICRF pulse is terminated. The value of alpha increases slightly with increasing electron density and decreases from 90 to 70% as the ICRF power is increased from 1 to 2 MW m-3 volume averaged. The global energy confinement time tau epsilon , defined by WP/(POH+ alpha Pr.f), decreases by a factor of 2 approximately 3 from that in Ohmic plasmas as the heating power increases up to 2 MW. It is found that the energy confinement time has a strong line-averaged electron density dependence as tau E varies as ne0.6, which is obtained by the use of the measured absorbed power, while the Kaye-Goldston scaling predicts tau E varies as ne0.26.

The primary diagnostic on PBX for ion temperature measurements is charge-exchange recombination spectroscopy of low Z ions, wherein fast neutrals from the heating neutral beams excite spectral lines from highly excited states (n greater than or equal to 4) of hydrogenic 0, C, and He via charge-exchange collisions with the respective fully stripped ions. Since the neutral beams on PBX provide relatively low velocity neutrals (i.e., D/sup 0/ beams at 44 keV), the best signals are obtained using the near-uv lines of 0/sup 7 +/ (e.g., n = 8-7, 2976 A). Off-line analysis of the Doppler broadened and shifted line profiles includes non-linear least squares fitting to a model line profile, while a simplified on-line fast analysis code permits between-shot data analysis.

Two improvements to the SPRED multichannel VUV spectrographs used on the TFTR and PBX tokamaks have been made: (1) a new 2100‐g/mm grating covering the 100–320‐A region with 0.4‐A resolution (FWHM) has been added to the existing 450‐g/mm grating (100–1100 A with 2‐A resolution), and (2) the TFTR SPRED has been absolutely calibrated using synchrotron radiation from the NBS SURF II facility, while the PBX system has been calibrated using conventional branching ratios along with line ratios from charge‐exchange‐recombination excited lines. The availability of high‐resolution spectra in the 100–320‐A range provides improved measurements of metallic ion emissions and, when the instrument views across a neutral beam as in PBX, allows carbon and oxygen densities to be measured via charge‐exchange‐recombination spectroscopy.

One of the most important things to design a helical system is an accuracy of the coil windings. To get the meaningful experimental results, we designed carefully and, as a result, the expected error field was reduced less than 1–2 gauss at the plasma surface when the field strength on the magnetic axis is 1.5 Tesla. The influence of this error field is negligible, because this value is very small and has no dangerous resonance. To ensure this, we measured the magnetic field components by the Hall device and the magnetic surfaces by the fluorescent mesh.

Experimental activities of the PBX project during its first seven months of operation are summarized.

Charge-dependent convective impurity transport is observed in the central core of PBX H-mode discharges by measuring radial profiles for both low- and high-Z intrinsic impurities in the presence of strong sawtooth activity. Transport coefficients are derived by simulating the data with a diffusive/convective transport model which includes an internal disruption model. The time evolving Z/sub eff/ profile and central metallic densities are fitted during the quiescent phase between internal disruptions to yield a diffusion coefficient of D = 1 x 10/sup 3/ cm/sup 2//s for all species and an inward convective speed of v = 2.2 x 10/sup 2/ (r/a) cm/s for oxygen and carbon and v = 1.1 x 10/sup 3/ (r/a) cm/s for metals. These transport coefficients are in quantitative agreement with simple cylindrical neoclassical estimates for the region within q less than or equal to 1. 17 refs., 4 figs.

During tangential neutral-beam injection into the PBX tokamak, bursts of two types of instabilities are observed. One instability occurs in the frequency range 120-210 kHz and the other oscillates predominantly near the frequency of bulk plasma rotation (20-30 kHz). Both instabilities correlate with drops in neutron emission and bursts in charge-exchange neutral flux, indicating that beam ions are removed from the center of the plasma by the instabilities. The central losses are comparable to the losses induced by the fishbone instability during perpendicular injection. © 1986 The American Physical Society.

Ion-Bernstein-wave heating is investigated in the JIPPT-II-U tokamak plasma, ${\overline{n}}_{e}\ensuremath{\simeq}1.5\ifmmode\times\else\texttimes\fi{}{10}^{13}$ ${\mathrm{cm}}^{\ensuremath{-}3}$, ${T}_{e0}\ensuremath{\simeq}700$ eV, and ${T}_{i0}\ensuremath{\simeq}300$ eV for ${P}_{\mathrm{rf}}\ensuremath{\lesssim}100$ kW. When the $\frac{3}{2}{\ensuremath{\Omega}}_{\mathrm{H}}$ layer is placed near the plasma minor axis, the bulk-hydrogen-ion temperature shows a significant rise, $\ensuremath{\Delta}{T}_{i\ensuremath{\perp}}l~700$ eV and $\ensuremath{\Delta}{T}_{i\ensuremath{\parallel}}l~300$ eV. The ion-heating dependence on the magnetic field and rf power suggests a presence of a direct bulk-hydrogen heating mechanism at $\ensuremath{\omega}\ensuremath{\simeq}\frac{3}{2}{\ensuremath{\Omega}}_{\mathrm{H}}$.

The improvement of core electron heat confinement has been realized in a wide range of helical devices such as CHS, LHD, TJ-II and W7-AS. Strongly peaked electron temperature profiles and large positive radial electric field,Er, in the core region are common features for this improved confinement. Such observations are consistent with a transition to the ‘electron-root’ solution of the ambipolarity condition forErin the context of neoclassical transport, which is unique to non-axisymmetric configurations. Based on this background, this improved confinement has been collectively dubbed ‘core electron-root confinement’ (CERC). The thresholds for CERC establishment are found for the collisionality and electron cyclotron heating power. The magnetic configuration properties (e.g. effective ripple and magnetic islands/rational surfaces) play important roles for CERC establishment.

Li pellet injection has provided a complex plasma with a large fraction of Li ions, which is characterized by intense emissions from Li I and III. The spatial profiles of the fully ionized Li 3+ ions are measured by charge exchange recombination spectroscopy with a resolution of 13 mm and the local decay time of the injected Li ion has been estimated. The spectral profile of the charge exchange recombination line of Li III from n = 5 to n = 4 shows a complicated structure, which depends on Li 3+ density. The effects on other intrinsic impurities and recycled Li are also discussed.

In the Large Helical Device (LHD), upgrading of the key hardware system since the last EPS conference has led to (1) higher Te (Te(0) = 4.4 keV at ne = 5.3×1018 m-3 and Pabs = 1.8 MW); (2) higher confinement (τE = 0.3 s, Te(0) = 1.1 keV at ne = 6.5×1019 m-3 and Pabs = 2.0 MW); (3) higher stored energy Wpdia = 880 kJ; and (4) the highest β value in helical devices (2.4% at B = 1.3 T). The energy confinement was systematically higher than that predicted by the International Stellarator Scaling 95 up to a factor of 1.6. Ion cyclotron range of frequencies (ICRF) power up to 1.3 MW was reliably injected into the plasma without significant impurity contamination and a plasma with stored energy of 200 kJ was sustained for 5 s by ICRF alone. Long pulse discharges greater than 1 min in duration have been successfully achieved with ICRF power alone and with neutral beam injection (NBI) power alone. With NBI heating an 80 s discharge was achieved with a heating power of 0.5 MW at 2.75 T. The electron density was maintained at around 1.6×1019 m-3 by controlled gas puffing. The central electron and ion temperatures were kept around 1.5 keV. With ICRF heating, a similar long pulse discharge was achieved for 68 s with a heating power of 0.85 MW. The sustained plasma parameters are: Wp~110 kJ, Te(0)~Ti(0) = 2.0 keV and ne = 1.0×1019 m-3. During these discharges, no increase in radiation power has been observed.

Evidence for a reduction of heat transport inside the magnetic island O point is observed from the propagation of a cold pulse produced by a tracer encapsulated solid pellet in the Large Helical Device. A small peak and slow propagation of the cold pulse are observed inside the island. A significant result is that electron heat diffusivity inside the island is estimated to be $0.2\text{ }\text{ }{\mathrm{m}}^{2}/\mathrm{s}$ which is smaller than that outside the island by an order of magnitude.