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DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (~30 kPa) and stored energy (3.2 MJ) with H 98y2 ≈ 1.6–2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric independent of current between q 95 = 2.8–3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses ≈0 injected torque and the operating space is more ITER-relevant. Finally, the high- (⩽3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving ~40% divertor heat flux reduction using either argon or neon with P tot up to 15 MW.

A full-scale ITER toroidal interferometer and polarimeter (TIP) prototype, including an active feedback alignment system, has been installed and tested on the DIII-D tokamak. In the TIP prototype, a two-color interferometry measurement of line-integrated density is carried out at 10.59 μm and 5.22 μm using a CO2 and quantum cascade laser, respectively, while a separate polarimetry measurement of the plasma-induced Faraday effect is made at 10.59 μm. The TIP prototype is equipped with a piezo tip/tilt stage active feedback alignment system that minimizes noise in the measurement and keeps the diagnostic aligned throughout DIII-D discharges. The measured phase resolution for the polarimeter and interferometer is 0.05° (100 Hz bandwidth) and 1.9° (1 kHz bandwidth), respectively. The corresponding line-integrated density resolution for the vibration-compensated interferometer is δnL = 1.5 × 1018 m-2, and the magnetic field-weighted line-integrated density from the polarimeter is δnBL = 1.5 × 1019 Tm-2. Both interferometer and polarimeter measurements during DIII-D discharges compare well with the expectations based on calculations using Thomson scattering measured density profiles and magnetic equilibrium reconstructions. Additionally, larger bandwidth interferometer measurements show that the diagnostic is a sensitive monitor of core density fluctuations with demonstrated measurements of Alfven eigenmodes and tearing modes.

Details of the low-high (L-H) confinement transition are studied by a new technique which allows for an arbitrarily slow transition between the L- and H-modes on DIII-D. During the transition, the plasma is in an intermediate state (IM-mode) of temperature, density, confinement and edge shear flow. The IM-mode is characterized by periodic bursts of an edge instability, governed by relaxation oscillations, which evolve into type III edge localized modes (ELMs) as the neutral beam heating power is raised. An ELM-free H-mode is achieved when the edge pressure gradient is large enough to support shear flow sufficient to quell the IM-mode edge instability.

Recent measurements from the DIII-D tokamak indicate that shear in the group velocity of the edge density fluctuations is at least partly responsible for the factor 2 change in the H-mode power threshold that is observed when the direction of the ion ∇B drift relative to the X-point location is reversed. Spatially resolved edge density fluctuation measurements show a change in the poloidal group velocity of the fluctuations when the ∇B drift direction was changed, even though the edge profiles of density and temperature remained nearly the same. High (low) shear in the poloidal velocity is associated with a low (high) power threshold.

Detailed measurements in two dimensions by probes and Thomson scattering reveal unexpected local electric potential and electron pressure (pe) maxima near the divertor X point in L-mode plasmas in the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)]. The potential drives E×B circulation about the X point, thereby exchanging plasma between closed and open magnetic surfaces at rates that can be comparable to the total cross-separatrix transport. The potential is consistent with the classical parallel Ohm’s law. A simple model is proposed to explain the pressure and potential hills in low power, nearly detached plasmas. Recent two-dimensional edge transport modeling with plasma drifts also shows X-point pressure and potential hills but by a different mechanism. These experimental and theoretical results demonstrate that low power tokamak plasmas can be far from poloidal uniformity in a boundary layer just inside the separatrix. Additional data, although preliminary and incomplete, ...

Electric potential, electron temperature, and electron density were measured in two dimensions (R,Z) throughout the divertor and X-point in DIII-D tokamak plasmas. An electric potential hill (∼100 eV) and an associated electron pressure hill were discovered at the divertor X-point in L-mode plasmas. The potential hill extends previously reported divertor E x B circulation, convecting particles, energy and toroidal momentum into and out of closed magnetic surfaces, and contributes significantly to transport across the boundary. The potential is explained by classical parallel (to B) plasma physics, when the X-point T i is clamped lower than upstream T i . The low X-point T i state might be self-sustaining at low heating power due to the same E x B circulation. We speculate that if the circulation transport is incompatible with H-mode, then the spontaneous L-H transition might require as a precondition that the X-point T i become equalized on the near-separatrix magnetic surfaces.

Analysis of DIII-D discharge data with fluid plasma and Monte Carlo neutrals transport codes reveals that core particle fueling stays relatively constant between the L-mode and the ELM-free H-mode phase immediately following the L–H transition. This indicates that in the ELM-free phase nearly all of the increase in plasma electron density comes from a decrease in the cross-field transport rate and an increase in the impurity influx. This result differs from conclusions of previous work in that the effects of the thinner H-mode scrape-off-layer do not appear to be as important in a plasma that is fueled primarily from divertor recycling as would be expected if the fueling from limiter recycling were dominant. In both L-mode and H-mode the calculated core particle confinement times are less than, but within 50% of, the corresponding energy confinement times.

The specific size and structure of the edge current profile has important effects on the magnetohydrodynamic stability and ultimate performance of many advanced tokamak (AT) operating modes. This is true for both bootstrap and externally driven currents that may be used to tailor the edge shear. Absent a direct local measurement of j(r), the best alternative is a determination of the poloidal field. Measurements of the precision (0.1°–0.01° in magnetic pitch angle and 1–10 ms) necessary to address issues of stability and control and provide constraints for EFIT are difficult to do in the region of interest (ρ=0.9–1.1). Using Zeeman polarization spectroscopy of the 2S–2P lithium resonance line emission from the DIII-D LIBEAM [D. M. Thomas, Rev. Sci. Instrum. 66, 806 (1995); D. M. Thomas, A. W. Hyatt, and M. P. Thomas, Rev. Sci. Instrum. 61, 340 (1990)] measurements of the various field components may be made to the necessary precision in exactly the region of interest to these studies. Because of the negli...

Theories predict that neutrals play a role in the low to high (L-H) confinement mode transition in tokamak plasmas via charge exchange damping and other effects. Previous estimates of neutral damping have been based on calculations of the edge neutral density. This work introduces a new method of measuring the neutral density near the X point, where simulations predict it to be a maximum. The technique employed uses Dα light from a TV camera reconstructed onto a poloidal plane, along with Thomson scattering measurements of the electron temperature and density. Measured neutral densities span the range 109-1013cm-3. Good agreement, considering the neutral density error bars, is found between the measurements and the 2-D simulations. This work represents the first step in verifying previous 2-D simulations and in corroborating previous conclusions that the neutral damping is large enough to play a role in the L-H transition process.

In order to study possible effects of neutrals on the power threshold for transition from L-mode to H-mode confinement in DIII-D, the problem of reconstructing neutral particle distributions inside the separatrix from available edge, scrape-off layer, and divertor plasma diagnostic data is addressed. Neutral particle profiles in the shear layer are reconstructed primarily with the DEGAS code using plasma parameters obtained by fitting discharge data with the B2.5 code. A series of discharges in which the ion grad B drift direction is toward the X-point is analyzed. For reversed Bt, the L to H transition power threshold is 2–4 times higher than normal Bt discharges at the same density: one discharge of this type is analyzed. The variation of the L–H transition power threshold divided by the density is seen to be correlated with the neutral density scale length in the shear layer for both directions of the magnetic field. Radial and poloidal neutral particle profile results for three low density discharges in which the L–H transition power threshold varies from 1.3 to 5.3 MW are presented.

To study the effect of edge neutrals on the low-to-high confinement transition threshold, a broad range of plasma discharges has been analyzed. These discharges vary by gas puffing and pumping rates, position of the X point, and line-averaged density. It is shown that the determination of the neutral density (or neutral pressure) in the scrape-off layer (SOL) can give a misleading indication of the neutral population inside the separatrix. An increase of neutral density in the SOL creates an increase of plasma density that, in turn, increases the opacity to the neutrals and results in reduced neutral penetration. At a constant magnetic field, the transition power divided by the density appears to be a function of a single parameter measuring the neutrals effect. From this analysis, this parameter cannot be uniquely identified. For instance, it may be the radial decay length of the neutral profile or the charge-exchange damping rate at about r/a≈0.95. A similar correlation exists between these neutral para...

High spatially resolved measurements of the radial electric field, Er, have been made across the transition from L mode to H mode plasmas for many different plasma parameters and conditions. The evolution of the well-like structure of the Er profile formed at the L-H transition has been investigated. No distinct variation in the shape or width of the Er well at the L-H transition is observed as a function of the plasma parameters investigated, such as the plasma current, the toroidal magnetic field and the plasma density. The value of Er is negative just inside the last closed flux surface (LCFS) for all the plasmas studied. There is a variation in the depth of the Er well for different conditions. The experimental results have been compared with theoretical predictions for suppression of plasma turbulence by sheared E × B plasma flow.

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma separates the first wall from the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma - wall interaction. In this paper we will describe a set of experiments on DIII-D designed to provide detailed two-dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. We will also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modelling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process.

Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne⩽1021 m−3, 0.5 eV⩽Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion–neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3–5, consistent with the measurements. Absolutely calibrated vacuum ultravi...

In this paper we describe the newly installed divertor Thomson scattering system for the DIII-D tokamak and present initial results from plasma discharges. Measured plasma densities have ranged from 5 × 1018 to 5 × 1020 m−3 and divertor plasma temperatures from 1 to 500 eV. These data are compared with earlier Langmuir probe data and qualitatively compared with UEDGE computer simulations. The divertor Thomson system uses one of the eight existing core Thomson scattering lasers (1 J, 20 Hz) which has been re-directed to probe the divertor region of the DIII-D vessel. Scattered light from this multipulse Nd:Yag laser is viewed with an f/6.8 collection optics system which provides eight spatial channels from 1–21 cm above the vessel floor (divertor target), each with 1.5 cm vertical resolution. Translating the plasma across the vessel floor using position controls provides a full scan of the divertor plasma.

Thomson scattering measurements of ne and Te in the divertor region of a Tokamak are reported. These data are used as input to boundary physics codes such as UEDGE and DEGAS and to benchmark the predictive capabilities of these codes. These measurements have also contributed to the characterization of tokamak disruptions. A Nd:YAG laser (20 Hz, 1 J, 15 ns, 1064 nm) is directed vertically through the lower divertor region of the DIII–D Tokamak. A custom, aspherical collection lens (f/6.8) images the laser beam from 1 to 21 cm above the target plates into eight spatial channels with 1.5 cm vertical and 0.3 cm radial resolution. Two-dimensional mapping of the divertor region is achieved by sweeping the divertor X-point location radially through the fixed laser beam location. Fiber optics carry the light to polychromators whose interference filters have been optimized for low Te measurements. Silicon avalanche photodiodes measure both the scattered and plasma background light. Temperatures and densities are typically in the range of 5–200 eV and 1–10×1019 m−3, respectively. Low temperatures, Te 8×1020 m−3 have been observed in detached plasmas. Background light levels have not been a significant problem. Reduction of the laser stray light permits Rayleigh calibration. Because of access difficulties, no in-vessel vacuum alignment target could be used. Instead, an in situ laser alignment monitor provides alignment information for each laser pulse. Results are compared with Langmuir probe measurements where good agreement is found except for regions of high ne and low Te as measured by Thomson scattering.

This paper describes divertor density and temperature measurements using both a new reciprocating Langmuir probe (XPT-RCP) which plunges vertically above the divertor floor up to the X-point height and swept, single, Langmuir probes fixed horizontally across the divertor floor. These types of measurements are important for testing models of the SOL and divertor which then are used to determine engineering design criteria for plasma facing components in reactor size tokamaks. The 6 mm diameter fixed single probes (19 domed and 2 flat at radially equivalent locations) are incorporated into the lower divertor floor at 19 radial locations and swept at 250 Hz. These probes are critical for determining plasma detachment from the floor during operation with high density, highly radiating divertors. By sweeping the divertor strike point across the fixed probes, different regions of the target plate incident flux profile can be sampled and a high resolution spatial profile can be obtained from each probe tip as the strike point moves past. The X-point reciprocating probe (XPT-RCP) provides ne and Te profiles with high spatial (2 mm) and temporal (0.5 ms) resolution from the target plate to the X-point along a single vertical chord at the same radial location as a fixed probe tip at a different azimuthal location. The probe ne and Te are compared to the divertor Thomson scattering (DTS) ne and Te (eight vertical points at 20 Hz, RThomson = RX-point-rcp). Recent observations have also shown divertor densities from 3 × 1019 to 4 × 1020 m−3 near the target plate with the highest densities observed with D2 gas puffing. Electron temperature is typically of the order of 15–25 eV at the target rising to about 70 eV near the X-point. Lower temperature, higher density plasmas are observed along the inner leg. Generally good agreement among the XPT-RCP, the fixed floor probes, and the DTS is observed. Differences between these diagnostic measurements will also be discussed with respect to different operating regimes and conditions.

The next generation of tokamaks present unique challenges to plasma diagnostic design due to the physical size of the devices and the radiation environment. The need for a density measurement for density feedback control for a prototype reactor such as ITER is well established and several proposals for line average measurements have been put forward. In this article, a design for a line average density diagnostic for ITER using collinear interferometry and Faraday rotation measurements will be presented. Plasma effects on both types of measurements and density resolution will be discussed along with the possibility of combining the information from the two collinear measurements to improve the reliability and quality of the density profile. Survivability of the plasma facing mirrors, in particular the surface flatness and surface roughness, are critical issues and preliminary analysis suggests these may limit the wavelength of probing beams. Thermal and stress analysis of the plasma facing mirrors will be...

Measurements of local electron density n{sub e}, electron temperature T{sub e}, and ion temperature T{sub i} have been made at the very edge of the plasma just prior to the transition into H-mode for four different single parameter scans in the DIII-D tokamak. The means and standard derivations of n{sub e}, T{sub e}, and T{sub i} under these conditions for a value of the normalized toroidal flux of 0.98 are respectively, 1.5 {+-} 0.7 x 10{sup 19} m{sup -3}, 0.051 {+-} 0.016 keV, and 0.14 {+-} 0.03 keV. The threshold condition for the transition is more sensitive to temperature than to density. The data indicate that the dependence is not as simple as a requirement for a fixed value of the ion collisionality.

DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D{sub 2} and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D{sub 2} gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design.

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...

A regime of improved H mode energy confinement, the VH mode, was obtained following boronization of the DIII-D Tokamak vacuum vessel. The gradual confinement improvement in VH mode is associated with expansion of the H mode pedestal and reduction of thermal diffusivity in a region just inside the pedestal. Disappearance of density fluctuation bursts is associated with a confinement improvement in the latter region. The VH mode is terminated in a rapid global energy loss which is initiated by a mode with toroidal mode number, n approximately 5, consistent with calculated instability to an edge localized ideal kink. The improved confinement in VH mode is consistent with the extension of the region of the E*B velocity shear turbulence suppression zone further in from the plasma boundary. Expansion of the ideal ballooning second stability region and of the region with drift reversal stabilization of trapped particle modes are also considered as possible explanations for the confinement improvement

A regime of improved H-mode energy confinement, VH-mode, is obtained in the DIII-D tokamak with adequate vessel conditioning. The improved confinement in VH-mode is consistent with the extension of the region of high E*B velocity shear turbulence suppression zone further in from the plasma boundary. The energy confinement enhancement in VH-mode can be limited by ELMs, localized momentum transfer events, or operation at high heating power or low q. Energy confinement enhancement improves with increasing triangularity of the plasma cross section and is independent of elongation. The termination of VH-mode is associated with an edge localized kink mode which is destabilized by both the large edge pressure gradient and edge current density.

We present the global analysis of a recent survey of the H-mode power threshold in DIII-D using D degrees to D+ NBI after boronization of the vacuum vessel. Single parameter scans of BT, Ip, density, and plasma shape have been carried out on the DIII-D tokamak for neutral beam heated single-null and double-null diverted plasmas. In single-null discharges, the power threshold is found to increase approximately linearly with BT and ne but remains independent of Ip. In double-null discharges, the power threshold is found to be approximately independent of both BT and ne. Various shape parameters such as plasma-wall gaps had only a weak effect on the power threshold. Imbalancing the double null configuration resulted in a large increase in the threshold power.

Combined theoretical and experimental work has resulted in the creation of a paradigm which has allowed semi-quantitative understanding of the edge confinement improvement that occurs in the H-mode. Shear in the E*B flow of the fluctuations in the plasma edge can lead to decorrelation of the fluctuations, decreased radial correlation lengths and reduced turbulent transport. Changes in the radial electric field, the density fluctuations and the edge transport consistent with shear stabilization of turbulence have been seen in several tokamaks. The purpose of this paper is to discuss the most recent data in the light of the basic paradigm of electric field shear stabilization and to critically compare the experimental results with various theories.

Accurate equilibrium reconstruction and detailed stability analysis of a strongly shaped, double‐null, βT=11% discharge shows that the plasma core is in the second stable regime to ideal ballooning modes. The equilibrium reconstruction using all the available data (coil currents, poloidal magnetic loops, motional Stark effect data, the kinetic pressure profile, the magnetic axis location, and the location of the two q=1 surfaces) shows a region of negative magnetic shear near the magnetic axis, an outer positive shear region, and a low shear region connecting the two. The inner negative shear region allows a large positive shear region near the boundary, even at low q (q95=2.6), permitting a large outer region pressure gradient to be first regime stable. The inner region is in the second stable regime, consistent with the observed axial beta [βT(0)=44%]. In the low shear region p’ vanishes, consistent with Mercier stability. This is one way to extend the ballooning limit in shaped plasmas while maintainin...

There is a need for a real time, reliable density measurement compatible with the restricted access and radiation environment on International Thermonuclear Experimental Reactor (ITER). Due to the large plasma path length, high density and field, refraction, and Faraday rotation effects makes the use of contemporary long wavelength (≳50 μm) interferometers impractical. In this paper we consider the design of a short wavelength vibration compensated interferometer which allows operation without a prohibitively large vibration isolated structure and permits the optics to be conveniently mounted directly in or on the tokamak. A density interferometer design for ITER incorporating a 10.6 μm CO2 interferometer with vibration compensation provided by a 3.39 μm HeNe laser is discussed. The proposed interferometer design requires only a small intrusion into the ITER tokamak without a large support structure, refraction and Faraday rotation problems are avoided, and it provides a density resolution of at least 0.5%. Results are presented from an interferometer installed on the DIII‐D tokamak incorporating essential elements of the proposed ITER design including 10.6 and 3.39 μm lasers, a retroreflector mounted on the vacuum wall of the DIII‐D tokamak and real‐time density feedback control. In this paper we consider a short wavelength interferometer design that incorporates vibration compensation for use on ITER. Our primary concern is to develop an interferometer design that will produce a reliable real time density monitor. We use the ITER conceptual design activity report as the basis of the design.

This paper describes the design and operation of a 40 spatial channel Thomson scattering system that uses multiple 20‐Hz Nd:YAG lasers to measure the electron temperature and density profiles periodically throughout an entire plasma discharge. As many as eight lasers may be fired alternately for an average measurement frequency of 160 Hz, or they may be fired in rapid succession (

A regime of very high confinement (VH mode) has recently been observed in DIII‐D with global energy confinement times up to a factor of 3.5 above the ITER89‐P L‐mode scaling [Nucl. Fusion 30, 1999 (1990)] and 1.5 to 2 times greater than the DIII‐D/JET edge‐localized‐mode‐free H‐mode scaling relation [Nucl. Fusion 31, 73 (1991)]. These discharges were obtained after boronization in DIII‐D and are characterized by low radiated power and Zeff, increasing confinement time during the VH phase of the discharge and low Ohmic target density. The low radiated power and Zeff are a consequence of the boronization. During the VH phase these discharges exhibit an inward shift in the region of highest electric field shear and a large calculated edge bootstrap current. The outer region (ρ≳0.85) is calculated to be in the second stable regime to ideal ballooning modes.

The L to H transition in the DIII‐D tokamak [Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. I, p. 159] is associated with a decrease in the edge density and magnetic fluctuations. In addition, in single‐null divertor plasmas, a reduction in the heat flux asymmetry between the inner and outer divertor hit spots occurs. These observations indicate that the L to H transition is associated with the reduction in anomalous, fluctuation‐connected transport across the outer midplane of the plasma. Magnetic fluctuations are measured with a poloidally distributed set of Mirnov loops while density fluctuations are detected with multiple fixed‐frequency microwave reflectometers. Spectroscopic observations of edge poloidal and toroidal rotation have allowed the inference that the radial electric field just inside the separatrix is negative in the L mode and becomes more negative at the L to H transition. These changes in fluctuations and in the edge electric field occur in plasma...

Local details of the edge current profile can have profound effects on the stability and ultimate performance of many advanced tokamak (AT) modes1-3. This is true for both bootstrap and externally driven currents that may be used to tailor the edge shear. Absent a direct local measurement of j(r), the best alternative is a determination of the poloidal field. On the DIII-D tokamak4 we are using precision polarimetry of an injected lithium beam (LIBEAM) to make measurements of the necessary precision in the region of interest (ρ = 0.7–1.1)5. Combined polarimetry/spectroscopy of the various Zeeman- split 2S-2P lithium resonance line components gives direct information on the local magnetic field components. Because of the negligible Stark mixing of the relevant atomic levels, this method of determining j(r) is insensitive to the large local electric fields typically found in enhanced confinement (H-mode) edges, and thus avoids an ambiguity present in motional Stark effect (MSE) measurements of B6,7. Beam intensities and energy are suitable for measurements in the edge region of even high density discharges. Key issues for utilizing this technique include good beam quality, an optimum viewing geometry, and a suitable optical pre-filter to isolate the polarized emission line. Details of the existing DIII-D diagnostic are presented. We also discuss details of the analysis necessary to determine the field component from these measurements, and our plans for future improvements to the system.

The transition from low-confinement (L-mode) to high-confinement (H-mode) plasmas has been directly produced by injecting frozen deuterium pellets in the DIII-D tokamak. H-mode transitions were produced at edge electron and ion temperatures below the L-mode values. This implies that a critical edge temperature is not necessary for H-mode transitions. The experimentally determined edge plasma parameters were well below those predicted by several theories of the H-mode transition to trigger the H-mode, indicating a need for revision of these theories.

A novel imaging diagnostic has recently become operational on the DIII‐D tokamak for the study of density fluctuations at the outer edge of the plasma. The phase contrast imaging approach overcomes the limitations of conventional scattering techniques in the spectral range of interest for transport‐related phenomena, by allowing detection of long‐wavelength modes (up to 7.6 cm) with excellent spatial resolution (5 mm) in the radial direction. Additional motivation for the diagnostic is provided by wave‐plasma interactions during heating and current drive experiments in the ion cyclotron range of frequencies. Density perturbations of 4×107 cm−3 with a 1‐MHz bandwidth can be resolved. The diagnostic employs a 7.6‐cm‐diam CO2 laser beam launched vertically across the plasma edge. An image of the plasma is then created on a 16‐element detector array: the detector signals are directly proportional to the density fluctuations integrated along each chord. Wavelengths and correlation lengths can be inferred from the spatial mapping. The phase contrast method and its application to DIII‐D are described and tests and first plasma data are presented.

There is a need for real time, reliable density measurement for tokamak plasma density control, compatible with the restricted access and radiation environment on ITER. Line average density measurements using microwave or laser interferometry techniques have proven to be robust and reliable for density control on contemporary tokamaks. In ITER, the large path length, high density and high density gradients, limit the wavelength of a probing beam to shorter then about 50 µm due to refraction effects. In this paper we consider the design of short wavelength vibration compensated interferometers and Faraday rotation techniques for density measurements on ITER. These techniques allow operation of the diagnostics without a prohibitively large vibration isolated structure and permit the optics to be mounted directly on the radial port plugs on ITER. A beam path designed for 10.6 µm (CO2 laser) with a tangential path through the plasma allows both an interferometer and a Faraday rotation measurement of the line average density with good density resolution while avoiding refraction problems. Plasma effects on the probing beams and design tradeoffs will be discussed along with radiation and long pulse issues. A proposed layout of the diagnostic for ITER will be presented.

Sheared mass flows are generated in many fluids and are often important for the dynamics of instabilities in these fluids. Similarly, large values of the E×B velocity have been observed in magnetic confinement machines and there is theoretical and experimental evidence that sufficiently large shear in this velocity may stabilize important instabilities. Two examples of this phenomenon have been observed in the DIII–D tokamak. In the first example, sufficient heating power can lead to the L‐H transition (transition from low‐mode to high‐mode confinement), a rapid improvement in confinement in the boundary layer (narrow region just inside the last closed flux surface) of the plasma. For discharges with heating power close to the threshold required to get the transition, changes in the edge radial electric field are observed to occur prior to the transition itself. In the second example, certain classes of discharges with toroidal momentum input from neutral beam injection exhibit a further improvement of co...

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.

The divertor magnetic geometry has a significant effect on the poloidal velocity and resulting velocity shear of turbulent density fluctuations in the outer region of L-mode tokamak plasmas, as determined via two-dimensional measurements of density fluctuations with beam emission spectroscopy on DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. Plasmas with similar parameters, except that in one case the ion ∇B drift points towards the divertor X point (lower-single-null, LSN), and in the other case, the ion ∇B drift points away from the divertor X point (upper-single-null, USN), are compared. Inside r∕a=0.9, the turbulence characteristics (density fluctuation amplitude, flow direction, correlation lengths) are similar in both cases, while near r∕a=0.92, a dramatic reversal of the poloidal velocity of turbulent eddies relative to the core flow direction is observed in plasmas with the ion ∇B drift pointing towards the divertor X point. No such velocity reversal is observed in plasmas with the ion ∇B drif...

The first Thomson scattering measurements of ne and Te in the divertor region of a tokamak are reported. These data are used as input to boundary physics codes such as UEDGE and DEGAS and to benchmark the predictive capabilities of these codes. These measurements have also contributed to the characterization of tokamak disruptions. A Nd:YAG laser (20 Hz, 1 J, 15 ns, 1064 nm) is directed vertically through the lower divertor region of the DIII–D tokamak. A custom, aspherical collection lens (f/6.8) images the laser beam from 1 to 21 cm above the target plates into eight spatial channels with 1.5 cm vertical and 0.3 cm radial resolution. Two-dimensional mapping of the divertor region is achieved by sweeping the divertor X-point location radically through the fixed laser beam location. Fiber optics carry the light to polychromators whose interference filters have been optimized for low-Te measurements. Silicon avalanche photodiodes measure both the scattered and plasma background light. Temperatures and dens...

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.

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.

Both external and internal magnetic probes have been used in low‐current discharges and external probes in high‐current discharges to study the magnetic field configuration and its fluctuations in the OHTE toroidal pinch with toroidal field reversal. The equilibrium magnetic field configuration is close to that of the Taylor state in the central half of the plasma (μ≡μ0bJ/B is constant) but differs in the outer half (μ gradually goes to zero at the plasma edge). Measurements of the magnetic fluctuations indicate that the dominant fluctuation mode observed is the resistive internal kink with m=1, n≂18. The measured relative level of magnetic field fluctuations scales as B/B∼S−1/2, where S is the magnetic Reynolds number.