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Use of a large-area liquid lithium surface as a limiter has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter.

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (?), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500?kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

The paradigm of shear suppression of turbulence as the mechanism for the low to high confinement mode (L to H) transition is examined by quantitative comparison of the predictions of the paradigm with experimental results from the DIII‐D tokamak [Plasma Physics and Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1986), p. 159]. The L to H transition trigger is V×B rotation, not the main ion pressure gradient. The radial electric field Er shear increases before the fluctuation suppression, consistent with increasing Er shear as the cause of the turbulence suppression. The spatial dependence of the turbulence reduction is consistent with shear suppression for negative Er shear. For positive Er shear, the turbulence suppression is consistent with the effect of Er curvature for modes for which an Er well is destabilizing. Finally, the transport barrier depends on the phase angle between the density and potential fluctuations inside the Er well, an effect not included in existing L to H...

Detailed information on the changes in the edge plasma at the L-H transition has been obtained by using charge exchange recombination spectroscopy with high spatial resolution (6 mm) and high time resolution (0.5 ms). Measurements of the radial electric field, Er, show that Er decreases abruptly at the transition resulting in the formation of a negative well like radial profile just within the separatrix. There is a well developed region of gradient in Er which extends inwards 1 to 2 cm into the plasma from the magnetic separatrix. This region of Er' exhibits the greatest increase in impurity ion poloidal rotation at the L-H transition and the greatest reduction in plasma fluctuations. A transport barrier is formed in this same region as is signified by large increases in the observed gradients of ion temperature, carbon density, electron temperature and electron density. The development of the region of significant Er', the increase in impurity ion poloidal rotation, the reduction of plasma turbulence and the formation of the transport barrier all occur simultaneously with the drop in the Dalpha signal which marks the transition. The well like structure of Er is located on and inside the separatrix as is clearly defined by the variation of the signal amplitudes of the spatial chords spanning the separatrix. The well like structure of Er moves with the separatrix as the separatrix location changes and is maintained throughout the H mode

Poloidal and toroidal rotation of the fuel ions (He2+) and the impurity ions (C6+ and B5+) in H-mode helium plasmas have been investigated in the DIII-D tokamak by means of charge exchange recombination spectroscopy, resulting in the discovery that the fuel ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction. The radial electric field obtained from radial force balance analysis of the measured pressure gradients and rotation velocities is shown to be the same regardless of which ion species is used and therefore is a more fundamental parameter than the rotation flows in studying H-mode phenomena. It is shown that the three contributions to the radial electric field (diamagnetic, poloidal rotation, and toroidal rotation terms) are comparable and consequently the poloidal flow does not solely represent the E*B flow. In the high-shear edge region, the density scale length is comparable to the ion poloidal gyroradius, and thus neoclassical theory is not valid there. In view of this new discovery that the fuel and impurity ions rotate in opposite sense, L-H transition theories based on the poloidal rotation may require improvement.

Poloidal and toroidal rotation of the main ions (${\mathrm{He}}^{2+}$) and the impurity ions (${\mathrm{C}}^{6+}$ and ${\mathrm{B}}^{5+}$) in H-mode helium plasmas have been measured via charge exchange recombination spectroscopy in the DIII-D tokamak. It was discovered that the main ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction, in qualitative agreement with the neoclassical theory. The deduced radial electric field in the edge is of the same negative-well shape regardless of which ion species is used, validating the fundamental nature of the electric field in L-H transition phenomenology.

Comparisons of helium (He) transport and exhaust in L-mode and in an enhanced confinement regime (H-mode), which is induced by a polarizing electrode, have been made for the TEXTOR tokamak. The results show an increased tendency for He accumulation when bulk plasma energy and particle confinement are improved during the polarization induced H-mode. Since these results imply that a high He pumping efficiency may be necessary for H-mode burning plasmas, we have begun exploring He transport in a divertor H-mode, similar to that proposed for International Thermonuclear Experimental Reactor (ITER). A collaborative program has been initiated to measure He transport and scaling on DIII-D during L-mode, H-mode, and ELMing H-mode plasma conditions. To simulate the presence of He ash in DIII-D, a 25 ms He puff is injected into a DIII-D plasma resulting in a He concentration of {approx}5%. The time dependence of the He{sup 2+} density profiles in the plasma core is measured by charge-exchange recombination spectroscopy at 11 radial locations.

Knowledge of the heat deposition profile is crucial to all transport analysis of beam heated discharges. The heat deposition profile can be inferred from the fast ion birth profile which, in turn, is directly related to the loss of neutral atoms from the beam. This loss can be measured spectroscopically by the decrease in amplitude of spectral emissions from the beam as it penetrates the plasma. The spectra are complicated by the motional Stark effect which produces a manifold of nine bright peaks for each of the three beam energy components. A code has been written to analyze these kinds of data. In the first phase of this work, spectra from tokamak shots are fit with a Stark splitting and Doppler shift model that ties together the geometry of several spatial positions when they are fit simultaneously. In the second phase, a relative position‐to‐position intensity calibration will be applied to these results to obtain the spectral amplitudes from which beam atom loss can be estimated. This paper reports ...

Doppler‐shift spectroscopy has shown that finite values of poloidal rotation velocity vθ and of radial electric field Er exist at the edge of a tokamak plasma and that dramatic increases occur in these parameters at an L‐H transition. Er is negative in the L mode and becomes more negative in the H mode; vθ increases in magnitude at the transition. In addition, the radial gradients (shear) of vθ and Er are large and these gradients also increase at the L‐H transition. These results are based on measurements of Doppler shifts of light emitted by He II ions, located in a region about 1–3 cm inside the separatrix. These observations have been made with horizontally viewing and vertically viewing spectrometer systems on the DIII‐D tokamak. The nearly orthogonal views of these systems are used to determine the plasma’s flow velocity in terms of the orthogonal sets vθ and vφ or of v⊥ and v∥. Knowledge of v⊥ is used to calculate Er from the force balance equation for a single ion species. The existing results imp...

Use of a large-area liquid lithium surface as a first wall has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter.

The charge exchange recombination, (CER) diagnostic system on the DIII-D tokamak is used to make spatially and temporally resolved measurements of the ion temperature and toroidal and poloidal rotation velocities. This is performed through visible spectroscopic measurements of the Doppler-broadened and Doppler-shifted He II 468.6-nm, CVI 529.1-nm, and BV 494.5-nm spectral lines which have been excited by charge exchange recombination interactions between the fully stripped ions and the neutral atoms from the heating beams. The plasma viewing optics comprises 32 viewing chords spanning a typical plasma minor radius of 63 cm across the midplane, of which 15 spatial chords span 4.2 cm at the plasma edge just within the separatrix and provide a chord-to-chord spatial resolution of 0.3 cm. Fast camera readout electronics can provide a temporal resolution of 260 mu s per time slice, but the effective minimum integration time, at present, is 1 ms, which is limited by the detected photon flux from the plasma and the decay times of the phosphors used on the multichannel plate image intensifiers. Significant changes in the edge plasma radial electric field at the L-H transition have been observed, as determined from the CER measurements, and these results are being extensively compared to theories which consider the effects of sheared electric fields on plasma turbulence. >

In DIII‐D, visible spectroscopic measurements of the He ii 468.6 nm and C vi 529.2 nm Doppler broadened spectral lines, resulting from charge exchange recombination interactions between beam neutral atoms and plasma ions, are performed to determine ion temperatures, and toroidal and poloidal rotation velocities. The diagnostics system comprises 32 viewing chords spanning a typical minor radius of 63 cm across the midplane, of which 16 spatial chords span 11 cm of the plasma edge just within the separatrix. A temporal resolution of 260 μs per time slice can be obtained as a result of using MCP phosphors with short decay times and fast camera readout electronics. Results from this system will be used in radial electric field comparisons with theory at the L–H transition and ion transport analysis.

Recent results from the DIII-D tokamak have provided significant contributions to the understanding of many of the elements of tokamak physics and the application of this understanding to the design of next generation devices including ITER and CIT. The limitations of magnetohydrodynamic stability on the values of plasma beta (the ratio of kinetic pressure to the containing pressure of the magnetic field) that can be attained has been experimentally demonstrated and found to be described by existing theory. Values of beta (10.7%) well in excess of those required for proposed devices (ITER and CIT) have been demonstrated. Regimes of confinement (H-mode) have been established that scale favorably to proposed next generation devices, and experiments demonstrating the dependence of the energy confinement on plasma size have been completed. Understanding of confinement is rapidly developing especially in the areas of bulk transport and the role of turbulence in the plasma edge. Key experimental results in areas of plasma transport and edge plasma phenomena are in agreement with theories based on short wavelength turbulence. Control of the divertor heat loads and impurity influx has been demonstrated, and new progress has been made in the understanding of plasma edge phenomena. Experiments with ion Bernstein wave heating have not found regimes in which these waves can produce effective central ion heating. Electron cyclotron current drive experiments have demonstrated 70 kA of driven current in 400 kA discharges.

Marked changes in the edge radial electric field ${\mathit{E}}_{\mathit{r}}$ and in the edge poloidal rotation velocity ${\mathit{v}}_{\mathrm{\ensuremath{\theta}}}$ are important signatures of the L-H transition in the DIII-D tokamak. Shear exists in ${\mathit{E}}_{\mathit{r}}$ and ${\mathit{v}}_{\mathrm{\ensuremath{\theta}}}$ and increases from the L mode to the H mode. A comparison of experiment with theory shows that shear in ${\mathit{E}}_{\mathit{r}}$ and ${\mathit{v}}_{\mathrm{\ensuremath{\theta}}}$ is sufficient to suppress edge fluctuations, that ion-orbit loss is large enough to be the source of ${\mathit{v}}_{\mathrm{\ensuremath{\theta}}}$, and that ${\mathit{E}}_{\mathit{r}}$ and ${\mathit{v}}_{\mathrm{\ensuremath{\theta}}}$ may play a causal role in the L-H transition.

Charge exchange recombination (CER) spectroscopy has become a standard diagnostic for Tokamaks. CER measurements have been used to determine spatially and temporally resolved ion temperature, toroidal and poloidal ion rotation speed, impurity density, and radial electric field. Knowledge of the spatial profile and temporal evolution of the electric field shear in the plasma edge is crucial to understanding the physics of the L to H transition. High speed CER measurements are also valuable for edge localized mode studies. Since the 0.52 ms minimum time resolution of our present system is barely adequate to study the time evolution of these phenomena, we have developed a new charge coupled device (CCD) detector system with about a factor of 2 better time resolution. In addition, our existing system detects sufficient photons to utilize the shortest time resolution only under exceptional conditions. The new CCD detector has a quantum efficiency of about 0.65, which is a factor of 7 better than our previous i...

Recent studies have shown that fairly weak dilution of the D‐T fuel mixture by the thermal alpha particle population may quench ignition in ITER. The alpha particle exhaust issue is not well understood experimentally, especially in H mode. On DIII‐D, a large set of diagnostics for studying alpha particle thermalization and exhaust are being developed. Helium density profiles in the core plasma are inferred from data obtained by the existing DIII‐D charge‐exchange recombination spectroscopy system which has 32 chords that span the plasma cross section and has been absolutely calibrated so that absolute emission intensities can be inferred from the measured data. Relative chord‐to‐chord intensity calibrations have been made by injecting neutral beams into neutral gas with no magnetic fields. A multifiber system viewing the divertor floor has been absolutely calibrated to measure the hydrogen and helium recycling profile across the divertor floor. Other diagnostics in development include a Penning discharge gauge to measure helium partial pressure under the divertor baffle in DIII‐D. Preliminary measurements from these diagnostics will be presented. This work supported by U.S. Department of Energy (DOE) Contract DE‐AC03‐89ER51114 and by appointment to the DOE Fusion Energy Postdoctoral Research Program administered by Oak Ridge Associated Universities.

An instrument has been developed to spectroscopically measure ion temperature and rotation speed spatial profiles in neutral‐beam‐heated tokamak plasmas. Up to eight spatial chords of data can be obtained for 32 time slices during a single discharge with selectable resolution times as short as 2 ms. A special arrangement of fiber optics allows simultaneous data acquisition from all chords without the use of scanning mirrors or other moving parts. Techniques used to analyze the data from the system are discussed.

Two kinds of plasma-related damage, spotty damage and wide area damage, have been photographically recognized on the TiC-coated surface of the new graphite primary limiter in Doublet III. TV and infrared-camera observations revealed the relationship of these damages to the plasma operations. The spotty damage was generated by a single runaway discharge. The runaway electron energy was deposited not only on the surface but also penetrated into the bulk graphite. Thermal analysis showed that runaway electron of 5–6 MeV penetrates into the limiter to approximately 1.5–2.0 cm. Wide area damage was caused by one or more successive discharges which exhibited rapid termination of the plasma current. In this case, the electron energy was deposited only on the limiter surface.

The authors' data indicate that the L-mode to H-mode transition in the DIII-D tokamak is associated with the sudden reduction in anomalous, fluctuation-connected transport across the outer midplane of the plasma. In addition to the reduction in edge density and magnetic fluctuations observed at the transition, the edge radial electric field becomes more negative after the transition. They have determined the scaling of the H-mode power threshold with various plasma parameters; the roughly linear increase with plasma density and toroidal field are particularly significant. Control of the ELM frequency and duration by adjusting neutral beam input power has allowed us to produce H-mode plasmas with constant impurity levels and durations up to 5 s. Energy confinement time in ohmic H-mode plasmas and in deuterium H-mode plasmas with deuterium beam injection can exceed saturated ohmic confinement times by at least a factor of two. Energy confinement times above 0.3 s have been achieved in these beam-heated plasmas with plasma currents in the range of 2.0 to 2.5 MA. Local transport studies have shown that electron and ion thermal diffusivities and angular momentum diffusivity are comparable in magnitude and all decrease with increasing plasma current.

Analysis of the expanded boundary divertor in Doublet III reveals a medium density operating regime ( n e ≲ 4 × 10 13 cm -3 ), in which (1) the divertor density ne(Div) rises with the cube of the main plasma line average density, n e , and (2) the neutral pressure at the exhaust end of the divertor has a quadratic dependence on n e (with divertor plasma temperatures ≲ 15 eV). In this regime, injection of argon into the discharge results in a steady state concentration of argon, which is two orders of magnitude higher in the divertor plasma than in the main discharge. This trapping effect suggests that the argon concentration in the divertor region may be useful as an externally controllable parameter; that is, injected argon effectively cools the plasma boundary, yet is not deleterious to the plasma as a whole. A model which successfully describes the behavior of the edge and divertor plasmas in ohmically heated discharges suggests that these desirable trapping and radiative features may be maintained in the high powered neutral beam heating phase.

Tokamak discharges using the expanded boundary divertor in the DIII-D device exhibit H-mode confinement. With neutral-beam power up to 6 MW, energy confinement remains comparable to the Ohmic value at a plasma current of 1 MA. Confinement is also independent of plasma density and toroidal field. Confinement increases with plasma current, but the exact functional dependence is, as yet, uncertain. These results show that the H mode can be achieved in a reactor-compatible open divertor configuration.

Measurements of ion temperature profiles, obtained on a single-shot basis, have been coupled with transport analysis to obtain spatial profiles of the ion thermal diffusivity ?i for a series of 18 neutral-beam-heated discharges in the Doublet III tokamak. The expanded boundary divertor discharges examined in this study had the following range of parameters: Bt = 2.53 T, Ip = 0.35?0.89 MA, vertical elongation = 1.4?1.9, n?e = (3.7?7.9) ? 1013 cm?3, PT = 3.5?6.6 MW, and ?i* = 0.03?0.22 , for r/a = 0.2?0.7. The experimentally inferred values of ?i have been compared with the values of the ion thermal diffusivity predicted by neoclassical theory. On the basis of a detailed error analysis, the results for all discharges examined are the same: ?i is approximately equal to near the plasma centre, but for r/a between about 0.2 and 0.8, ?i is both larger than and almost certainly has a spatial variation different from that of . The differences between ?i and are larger than can be explained by errors in the experimental measurements and indicate that there are processes operative for the loss of ion heat which are not included in the usual transport models for ion confinement.

Angular momentum confinement has been studied in neutral beam heated limiter and divertor discharges in Doublet III. Scaling of the angular momentum confinement time with plasma current and neutral beam power has been determined. For divertor discharges, the angular momentum confinement time and the energy confinement time are equal and scale in the same way. This suggests that a single mechanism governs the transport of both energy and angular momentum in divertor discharges. For limiter discharges, the angular momentum confinement time is less than the energy confinement time and the scalings are dissimilar. In both limiter and divertor discharges the measured angular rotation speed profiles are consistent with the theoretical prediction that angular rotation speed is a function of the flux surface co-ordinate only and is independent of poloidal angle.

A pair of visible/near‐UV spectrometers with eight viewing chords apiece have been installed on the DIII‐D tokamak. Each system views a neutral heating beam and can acquire up to 250 complete spectra from each chord with 5–20‐ms time resolution. Each viewing chord covers 60 A with 0.27‐A spectral resolution, and the chords span about (2)/(3) of the plasma’s full width. By viewing Doppler‐broadened spectral lines from charge exchange recombination (CER) reactions between beam neutrals and plasma ions, ion temperatures up to 4 keV have been measured, and the bulk Doppler shift of these same lines has yielded plasma rotation velocities up to 200 km/s. The constancy of temperature on a magnetic flux surface and the rigid rotor model of a flux surface have been confirmed. These instruments have also been used to measure the neutral beam deposition profile, and preliminary experimental results agree with theoretical calculations of the beam deposition profile.

A major portion of the DIII-D program includes studies of the L-H transition, of the VH-mode, of particle transport and control and of the power-handling capability of a divertor. Significant progress has been made in all of these areas and the purpose of this paper is to summarize the major results obtained during the last two years. An increased understanding of the origin of improved confinement in H-mode and in VH-mode discharges has been obtained, good impurity control has been achieved in several operating scenarios, studies of helium transport provide encouraging results from the point of view of reactor design, an actively pumped divertor chamber has controlled the density in H-mode discharges and a radiative divertor is a promising technique for controlling the heat flux from the main plasma.

Single shot, multipoint ion temperature and plasma rotation profiles have been routinely obtained on the Doublet III tokamak for 32 consecutive time slices with 20‐ms resolution. A six‐chord tangentially viewing spectroscopic diagnostic has been built to look at radiation emitted by fully stripped low‐Z impurity ions (He, C, O) that have undergone charge exchange recombination with hydrogen atoms from a 3‐MW heating beam. The main components of the instrument are a single monochromator for wavelength dispersion, a single image intensifier tube for photon gain, and a pair of 1024‐element linear photodiode arrays for detection. A special arrangement of fiber optics allows simultaneous data acquisition from all chords without the use of scanning mirrors or other moving parts. Ion temperature profiles taken under a variety of plasma conditions will be presented.