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Doppler backscattering system can measure the perpendicular velocity and fluctuation amplitude of the density turbulence with intermediate wavenumber. An eight-channel Doppler backscattering system has been installed in the Experimental Advanced Superconducting Tokamak (EAST), which can probe eight different radial locations simultaneously by launching eight fixed frequencies (55, 57.5, 60, 62.5, 67.5, 70, 72.5, 75 GHz) into plasma. The quasi-optical system consists of circular corrugated waveguide transmission, a fixed parabolic mirror, and a rotatable parabolic mirror which are integrated with quasi-optics front-end of the profile reflectometer inside the vacuum vessel. The incidence angle can be chosen from 5° to 12°, and the wavenumber range is 2–15/cm with the wavenumber resolution Δk/k ≤ 0.21. Ray tracing simulations are used to calculate the scattering locations and the perpendicular wavenumber. The dynamic range of this new eight-channel Doppler backscattering system can be as large as 40 dB in the EAST. In this article, the hardware design, the ray tracing, and the preliminary experimental results in the EAST will be presented. [ABSTRACT FROM AUTHOR]

A new model-based technique for fast estimation of the pedestal electron density gradient has been developed. The technique uses ordinary mode polarization profile reflectometer time delay data and does not require direct profile inversion. Because of its simple data processing, the technique can be readily implemented via a Field-Programmable Gate Array, so as to provide a real-time density gradient estimate, suitable for use in plasma control systems such as envisioned for ITER, and possibly for DIII-D and Experimental Advanced Superconducting Tokamak. The method is based on a simple edge plasma model with a linear pedestal density gradient and low scrape-off-layer density. By measuring reflectometer time delays for three adjacent frequencies, the pedestal density gradient can be estimated analytically via the new approach. Using existing DIII-D profile reflectometer data, the estimated density gradients obtained from the new technique are found to be in good agreement with the actual density gradients for a number of dynamic DIII-D plasma conditions. [ABSTRACT FROM AUTHOR]

The geodesic acoustic mode (GAM), a coherent form of the zonal flow, plays a critical role in turbulence regulation and cross-magnetic-field transport. In the DIII-D tokamak, unique information on multi-field characteristics and radial structure of eigenmode GAMs has been measured. Two simultaneous and distinct, radially overlapping eigenmode GAMs (i.e., constant frequency vs. radius) have been observed in the poloidal E×B flow in L-mode plasmas. As the plasma transitions from an L-mode to an Ohmic regime, one of these eigenmode GAMs becomes a continuum GAM (frequency responds to local parameters), while the second decays below the noise level. The eigenmode GAMs occupy a radial range of ρ = 0.6-0.8 and 0.75-0.95, respectively. In addition, oscillations at the GAM frequency are observed for the first time in multiple plasma parameters, including ne, Te, and Bθ. The magnitude of Te/Te at the GAM frequency (the magnitude is similar to that of ñe/ne) and measured ne-Te cross-phase (∼140° at the GAM frequency) together indicate that the GAM pressure perturbation is not determined solely by ñe. The magnetic GAM behavior, a feature only rarely reported, is significantly stronger (×18) on the high-field side of the tokamak, suggesting an anti-ballooning nature. Finally, the GAM is also observed to directly modify intermediate-wavenumber ñe levels (kρs ∼ 1.1). The simultaneous temperature, density, flow fluctuations, density-temperature cross-phase, and magnetic behavior present a new perspective on the underlying physics of the GAM. [ABSTRACT FROM AUTHOR]

A series of carefully designed validation experiments conducted on DIII-D to rigorously test gyrofluid and gyrokinetic predictions of transport and turbulence stiffness in both the ion and electron channels have provided an improved assessment of the experimental fidelity of those models over a range of plasma parameters. The first set of experiments conducted was designed to test predictions of H-mode core transport stiffness at fixed pedestal density and temperature. In low triangularity lower single null plasmas, a factor of 3 variation in neutral beam injection (NBI) heating was obtained, with modest changes to pedestal conditions that slowly increased with applied heating. The measurements and trends with increased NBI heating at both low and high injected torque are generally well-reproduced by the quasilinear trapped gyro-Landau fluid (TGLF) transport model at the lowest heating levels, but with decreasing fidelity (particularly in the electron profiles) as the heating power is increased. Complementing these global stiffness studies, a second set of experiments was performed to quantify the relationship between the local electron energy flux Qe and electron temperature gradient by varying the deposition profile of electron cyclotron heating about a specified reference radius in low density, low current L-mode plasmas. Modelling of these experiments using both the TGLF model and the nonlinear gyrokinetic GYRO code yields systematic underpredictions of the measured fluxes and fluctuation levels. [ABSTRACT FROM AUTHOR]

A critical gradient for long wavelength (kθρs<=0.4) electron temperature fluctuations has been observed in an experiment in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)], where below a threshold value of LTe-1=|∇Te|/Te electron temperature fluctuations are constant and above they steadily increase. Above the critical gradient, the electron heat flux inferred by power balance also increases rapidly. Critical gradients are a predicted attribute of turbulence arising from linear instabilities and are thought to be related to transport stiffness. The presented results are the first direct, systematic demonstration of critical gradient behavior in turbulence measurements in a tokamak. The experiment was performed by changing the deposition location of electron cyclotron heating shot-to-shot to locally scan LTe-1 at r/a = 0.6 in L-mode plasmas; rotation was also varied by changing the momentum input from neutral beam injection. Temperature fluctuations were measured with a correlation electron cyclotron emission (CECE) radiometry system. In addition to the CECE measurements, an array of turbulence measurements were acquired to characterize fluctuations in multiple fields and at multiple scales as LTe-1 and rotation were modified: long wavelength (kθρs<=0.5) density fluctuations were acquired with beam emission spectroscopy, the phase angle between electron temperature and density fluctuations was measured by coupling the CECE system and a reflectometer, intermediate scale (kθρs∼0.8) density fluctuations were measured with a Doppler backscattering (DBS) system, and low frequency flows were also measured with DBS. The accumulated measurements and trends constrain identification of the instability responsible for the observed critical gradient to the ∇Te-driven trapped electron mode. [ABSTRACT FROM AUTHOR]

An extensive set of tests comparing gyrokinetic predictions of temperature-gradient driven electron turbulence to power balance transport analyses and fluctuation measurements are presented. These tests use data from an L-mode validation study on the DIII-D tokamak (Luxon 2002 Nucl. Fusion 42 614) in which the local value of a/LTe = -(a/Te)(dTe/dr) is varied by modulated electron cyclotron heating; the GYRO code (Candy and Waltz 2003 J. Comput. Phys. 186 545) is used to make the gyrokinetic predictions. Using a variety of novel measures, both local and global nonlinear simulations are shown to predict key characteristics of the electron energy flux Qe and longwavelength (low-k) Te fluctuations, but systematically underpredict (by roughly a factor of two) the ion energy flux Qi. A new synthetic diagnostic for comparison to intermediate wavelength Doppler backscattering measurements is presented, and used to compare simulation predictions against experiment. In contrast to the agreement observed in the low-k Te fluctuation comparisons, little agreement is found between the predicted and measured intermediate-k density fluctuation responses. The results presented in this paper significantly expand upon those previously reported in DeBoo et al (2010 Phys. Plasmas 17 056105), comparing transport and multiple turbulence predictions from numerically converged local and global simulations for all four experimental heating configurations (instead of only fluxes and low-k Te fluctuations for one condition) to measurements and power balance analyses. [ABSTRACT FROM AUTHOR]

The multi-field/multi-scale core (ρ ∼ 0.5-0.8) turbulence response to electron cyclotron heating (ECH) of DIII-D Ohmic plasmas is reported for the first time. Long wavelength (low-k) electron temperature (Te/Te) and high-k density turbulence levels (ñe/ne) are observed to strongly increase during ECH. In contrast, low-k and intermediate-k ñe/ne showed little change, whereas the cross-phase between local low-k electron temperature and density fluctuations (αneTe) was significantly modified. The increase in the electron thermal diffusivity determined from power balance is consistent with the increased turbulent transport correlated with the measured increases in low-k Te/Te and high-k ñe/ne. Linear stability analysis using the trapped gyro-Landau fluid (TGLF) model indicates an enhanced growth rate for electron modes [e.g., trapped electron mode (TEM)] at low-k consistent with the observed modifications in Te/Te and αneTe. TGLF also predicts an increase in high-k electron mode growth rates for normalized wavenumbers kθρs > 7, where electron temperature gradient (ETG) modes exist, which is consistent with the observed increase in high-kñe/ne turbulence [ABSTRACT FROM AUTHOR]

Recent studies on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] have elucidated key aspects of the dependence of stability, confinement, and density control on the plasma magnetic configuration, leading to the demonstration of nearly noninductive operation for >1 s with pressure 30% above the ideal no-wall stability limit. Achieving fully noninductive tokamak operation requires high pressure, good confinement, and density control through divertor pumping. Plasma geometry affects all of these. Ideal magnetohydrodynamics modeling of external kink stability suggests that it may be optimized by adjusting the shape parameter known as squareness (ζ). Optimizing kink stability leads to an increase in the maximum stable pressure. Experiments confirm that stability varies strongly with ζ, in agreement with the modeling. Optimization of kink stability via ζ is concurrent with an increase in the H-mode edge pressure pedestal stability. Global energy confinement is optimized at the lowest ζ tested, with increased pedestal pressure and lower core transport. Adjusting the magnetic divertor balance about a double-null configuration optimizes density control for improved noninductive auxiliary current drive. The best density control is obtained with a slight imbalance toward the divertor opposite the ion grad(B) drift direction, consistent with modeling of these effects. These optimizations have been combined to achieve noninductive current fractions near unity for over 1 s with normalized pressure of 3.5<βN<3.9, bootstrap current fraction of >65%, and a normalized confinement factor of H98(y,2)≈1.5. [ABSTRACT FROM AUTHOR]

Small-scale density turbulence (k⊥ρi∼4–10) and electron thermal flux are both observed to increase during electron cyclotron heating (ECH) of a high-temperature tokamak plasma (k⊥ is the turbulent wavenumber and ρi the ion gyroradius). In contrast, large- (k⊥ρi≤1) and intermediate-scale (k⊥ρi∼1–3) turbulence ñ/n levels and ion thermal transport remain effectively constant. This implies that the small-scale turbulence is not a remnant or tail of the ubiquitous, large-scale or intermediate-scale turbulence, and also indicates a potentially important role in determining anomalous electron thermal transport. Radial scans of small-scale turbulence during ECH indicate decreased fluctuations in the deep core compared with increased levels towards the edge. This trend is consistent with linear gyrokinetic growth rate predictions for electron temperature gradient driven instabilities. [ABSTRACT FROM AUTHOR]

High values of normalized β (βN∼4) and safety factor (qmin∼2) have been sustained simultaneously for ∼2 s in DIII-D [J.L. Luxon, Nucl. Fusion 42, 64 (2002)], suggesting a possible path to high fusion performance, steady-state tokamak scenarios with a large fraction of bootstrap current. The combination of internal transport barrier and negative central magnetic shear at high β results in high confinement (H89P>2.5) and large bootstrap current fraction (fBS>60%) with good alignment. Previously, stability limits in plasmas with core transport barriers have been observed at moderate values of βN (<3) because of the pressure peaking which normally develops from improved core confinement. In recent DIII-D experiments, the internal transport barrier is clearly observed in the electron density and in the ion temperature and rotation profiles at ρ∼0.5 but not in the electron temperature profile, which is very broad. The misalignment of Ti and Te gradients may help to avoid a large local pressure gradient. Furthermore, at low internal inductance ∼0.6, the current density gradients are close to the vessel and the ideal kink modes are strongly wall-coupled. Simultaneous feedback control of both external and internal sets of n=1 magnetic coils was used to maintain optimal error field correction and resistive wall mode stabilization, allowing operation above the free-boundary β limit. Large particle orbits at high safety factor in the core help to broaden both the pressure and the beam-driven current profiles, favorable for steady-state operation. At plasma current flat top and β∼5%, a noninductive current fraction of ∼100% has been observed. Stability modeling shows the possibility for operation up to the ideal-wall limit at β∼6%. [ABSTRACT FROM AUTHOR]

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, βT=3.6%, normalized beta, βN=3.5, and confinement factor, H89=2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1 s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2 s duration (1τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario. [ABSTRACT FROM AUTHOR]

Results are presented from comparisons of modeling and experiment in studies to assess the best choices of safety factor q profile, pressure profile, and discharge shape for high β, steady-state, noninductive advanced tokamak operation in the DIII-D device [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. These studies are motivated by the need for high qminβN to maximize the self-driven bootstrap current while maintaining high toroidal β to increase fusion gain. Modeling shows that increases in the normalized beta βN stable to ideal, low toroidal mode number (n=1,2), instabilities can be obtained through broadening of the pressure profile and use of a symmetric double-null divertor shape. Experimental results are in agreement with this prediction. The general trend is for qminβN to increase with the minimum q value (qmin) although βN decreases as qmin increases. By broadening the pressure profile, βN≈4 is obtained with qmin≈2. Modeling of equilibria with near 100% bootstrap current indicates that operation with βN≈5 should be possible with a sufficiently broad pressure profile. [ABSTRACT FROM AUTHOR]

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. [ABSTRACT FROM AUTHOR]

High-confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (Intemational Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation that is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 s or >25 energy confinement times τ[sub E]), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved β[sub N]H[sub 89]:7 for up to 5 times the τ[sub E] of 150 ms. β[sub N]H[sub 89] values of 7 substantially exceed the value of 4 routinely achieved in the standard ELMing H mode. The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density and impurity control in the quiescent H mode is possible because of the presence of an edge magnetohydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected. [ABSTRACT FROM AUTHOR]

Significant progress has been made in obtaining high-performance discharges for many energy confinement times in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. Normalized performance (measured by the product of β[sub N]H[sub 89] and indicative of the proximity to both conventional β limits and energy confinement quality, respectively) ∼10 has been sustained for >5 τ[sub E] with q[sub min] > 1.5. These edge localized modes (ELMing) H-mode discharges have &beta∼5%, which is limited by the onset of resistive wall modes slightly above the ideal no-wall n = 1 limit, with approximately 75% of the current driven noninductively. The remaining Ohmic current is localized near the half-radius. The DIII-D electron cyclotron heating system is being upgraded to replace this inductively driven current with localized electron cyclotron current drive (ECCD). Density control, which is required for effective ECCD, has been successfully demonstrated in long-pulse high-performance ELMing H-mode discharges with &beta[sub N]H[sub 89]∼ 7 for up to 6.3 s. In plasma shapes compatible with good density control in the present divertor configuration, the achieved β[sub N] is somewhat less than that in the high β[sub N]H[sub 89] = 10 discharges. © 2001 American Institute of Physics. [DOI: 10.1063/1. [ABSTRACT FROM AUTHOR]

In recent experiments in DIII-D, comparison of specific turbulence characteristics with linear and nonlinear modeling has identified common features associated with the ion temperature gradient (ITG) mode. A low-frequency turbulence feature is observed in high-density saturated Ohmic confinement discharges, which is absent in low-density linear Ohmic confinement discharges. The feature is in a range of wavelength (k[sub ⊥]p[sub s] ≈ 0.2 − 0.5) and the frequency expected for the ITG mode and onset of the feature is coincident with onset of confinement saturation. The density profile is significantly broader in the high-density discharge, a known destabilizing effect on the ITG mode. Gyrokinetic stability calculations of the growth rate of the most unstable drift ballooning mode show the ITG mode to be more unstable in the high-density discharges. [ABSTRACT FROM AUTHOR]

The profile reflectometer system on the DIII-D tokamak has been significantly upgraded in order to improve time coverage, data quality, and profile availability. The performance of the reflectometer system, which utilizes continuous frequency modulated (FMCW) radar techniques, has been improved as follows: First, a new PC-based data acquisition system has been installed, providing higher data sampling rates and larger memory depth. The higher sampling rate enables use of faster frequency sweeps of the FMCW microwave source, improving time resolution, and increasing profile accuracy. The larger memory depth enables longer data records, so that profiles can now be obtained throughout 5 s discharges at 100 Hz profile measurement rates, while continuous sampling at 10 MHz is available for 1 s for high time resolution physics studies. Second, an initial automated between-shots profile analysis capability is now available. Third, availability of the profiles to end users has been significantly improved. © 2001 American Institute of Physics. [ABSTRACT FROM AUTHOR]

Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection taking place when the heating power exceeds a threshold value of about 9 MW, contrasting to co-NBI injection, where P[sub threshold]<2.5 MW. Expansion of the ITB is enhanced compared to similar co-injected discharges. Both differences are believed to arise from modification of the ExB shear dynamics when the sign of the rotation contribution is reversed. Sustainment of an AT regime with β[sub N]H[sub 89]=9 for 16 confinement times has been accomplished in a discharge combining an ELMing H-mode (edge localized, high confinement mode) edge and an ITB, and exhibiting ion thermal transport down to 2-3 times neoclassical. The microinstabilities usually associated with ion thermal transport are predicted stable, implying that another mechanism limits performance. High frequency magnetohydrodynamic (MHD) activity is identified as the probable cause. © 2000 American Institute of Physics. [ABSTRACT FROM AUTHOR]

In DIII-D [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] tokamak plasmas with an internal transport barrier (ITB), the comparison of gyrokinetic linear stability (GKS) predictions with experiments in both low and strong negative magnetic shear plasmas provide improvedunderstanding for electron thermal transport within the plasma. Within a limited region just inside the ITB, the electron temperature gradient (ETG) modes appear to control the electron temperature gradient and, consequently, the electron thermal transport. The increase in the electron temperature gradient with more strongly negative magnetic shear is consistent with the increase in the ETG mode marginal gradient. Closer to the magnetic axis the T3 profile flattens and the ETG modes are predicted to be stable. With additional core electron heating, FIR scattering measurements near the axis show the presence of high k fluctuations (12 cm-1), rotating in the electron diamagnetic drift direction. This turbulence could impact electron transport and possibly also ion transport. Thermal diffusivities for electrons, and to a lesser degree ions, increase. The ETG mode can exist at this wave number, but it is computed to be robustly stable near the axis. Consequently, in the plasmas we have examined, calculations of drift wave linear stability do not explain the observed transport near the axis in plasmas with or without additional electron heating, and there are probably other processes controling transport in this region.. [ABSTRACT FROM AUTHOR]

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 two clear signatures: edge density fluctuations are abruptly suppressed (in ≈ 100 μsec), while the edge poloidal rotation velocity ν[sub θ] increases, implying that the radial electric field E[sub r] becomes more negative. Detailed new spectroscopic profile measurements show that the changes in ν[sub θ] and E[sub r] generate a region of sheared electric field and poloidai flow of width ≈ 3-5 cm. This. region develops simultaneously with, and has the same spatial extent as, the edge fluctuation suppression zone as measured using a refteetometer system. Furthermore, the radial extent of the shear and fluctuation suppression zones encompass the location of the H-mode edge transport barrier. These observations are consistent with recent theoretical models of the L-H transition, and a comparison with these theories is presented. Data are also presented on the evolution of edge parameters and density fluctuations after the transition: the shear and fluctuation suppression layers are maintained for the duration of the quiescent H-mode phase, while relative density fluctuation levels decrease and interior plasma confinement gradually improves. Precursors to several different types of edge localized mode (ELM's) are also discussed. [ABSTRACT FROM AUTHOR]

Describes the implementation of a highly reliable density profile reflectometer system on the DIII-D tokamak via fast-sweep broadband frequency-modulated reflectometry. Density profile measurements; Signal quality variations with changing frequency sweep rates; Planned improvements to the system.

Discusses the use of millimeter wave reflectometry to determine radio frequency wave electric field profiles in the DIII-D Tokamak. Determination of internal fast magnetosonic wave electric field profiles; Factors that determine the normalized reflectometer signal.

Plasma tomographic reconstructions are subject to aliasing ambiguity as a result of the limited angular and radial sampling rates for the line-integrated data. The two-view interferometer installed on the Microtor tokamak yields unambiguous information, specifically, a collection of six low-order alias-free moments of the 2-D electron density distribution. The unspoiled coefficients can be related to physically intuitive quantities and so yield information pertinent to the equilibrium and dynamical behavior of the plasma column. Alternatively, they can be used as constraints for a maximum entropy reconstruction of the source to produce an image free of aliasing artifacts. [ABSTRACT FROM AUTHOR]

Conventional interferometer systems used on fusion devices are typically limited to approximately ten channels or views, thus precluding the use of tomographic reconstruction techniques. Instead, a priori assumptions have to be made in order to obtain the electron density distribution. To overcome these problems, interferometer systems have been developed at the University of California, Los Angeles (UCLA), which use imaging techniques to provide many channels per probe beam. Using a two-dimensional, 40-channel, imaging far infrared (FIR) interferometer measurements have been made of the density distribution in the minor cross section of the UCLA Microtor tokamak. The heterodyne interferometer consists of two probe beams expanded so as to completely fill orthogonal views of the plasma cross section. These views are then imaged onto two 20-channel, linear microbolometer detector arrays. Using data from this system, the spatial electron density distribution in a tokamak plasma has been recovered for the first time using tomographic reconstruction techniques. [ABSTRACT FROM AUTHOR]

Studies the confinement and the stability of properties of the DIII-D tokamak high-performance discharges in terms of rotational and magnetic shear. Experimental results obtained from the negative central magnetic shear (NCS); Core transport barrier that forms inside the NCS region; Performance of NCS discharges with strongly peaked pressure profiles.

A broadband tunable (∼50-75 GHz) correlation reflectometer system is installed and operational on the DIII-D tokamak. The same launch and receive horns are utilized for the two probing frequencies ensuring that they sample the same path within the plasma. Very wide band IF electronics allow the separation frequency to be continuously tuned over the range 100 MHz→10 GHz corresponding to a maximum spatial separation of ∼7 cm. Sweeping the separation frequency during the plasma shot produces multiple spatial points (>10) on a single discharge reducing shot to shot reproducibility problems and increasing the spatial resolution. A description of the microwave system, laboratory tests, and representative data from a tokamak discharge are presented. Difficulties with the method and some possible resolutions are discussed. [ABSTRACT FROM AUTHOR]

Anomalous transport in fusion plasmas remains an enigma requiring explanation. A predictive capability is highly desirable if confinement enhancement regimes such as H mode or super shots are to be extrapolated to the next phase in development of the International Fusion program, epitomized, for example in ITER. Therefore, identification of the role that electrostatic turbulence plays in confinement is a critical issue requiring detailed experimental data capable of testing and challenging existing theoretical models. This article presents microturbulence measurements obtained on the DIII-D and TEXT tokamaks utilizing heterodyne, far-infrared collective scattering, and reflectometry techniques. The experimental systems are described on both machines and emphasis placed on results obtained during the L-H transition, ELM activity, and saturated ohmic operation where ion temperature gradient driven (ITGD) turbulence is theoretically predicted to exist. [ABSTRACT FROM AUTHOR]

Although it has long been suggested that microturbulence is responsible for anomalous transport, relatively little is known about turbulence drive and suppression mechanisms which determine the observed fluctuation levels. In the DIII-D (DIII-D) tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] during H-mode discharges, microturbulence has been observed to change in two stages. Collective far-infrared scattering has confirmed and quantified a rapid (... 100 µsec) suppression of edge microturbulence at the L→H transition, as well as locally increased edge electric field shear, consistent with theoretical models which predict microturbulence suppression due to E×B sheared flow. Additionally, during the subsequent H-mode phase, a slower (10-100 msec) reduction of interior microturbulence is observed, coincident with increased electric field shear and reduced Δn[sub e] in the same region. This reduction is much larger in boronized discharges which also generally have larger rotation velocity. Experiments have been performed whereby the interior radial electric field is reduced while maintaining similar pressure profiles through the use of "magnetic braking" of toroidal rotation. As expected, the E×B Doppler frequency shift of the scattered fluctuation spectrum decreased as the electric field was reduced. The microturbulence levels were observed to increase when the interior electric field and the associated shear were reduced. [ABSTRACT FROM AUTHOR]

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 plasmas where the H mode is created by neutral beam heating, electron cyclotron heating, and even by Ohmic heating alone. Accordingly, these changes are characteristic of the H mode and are not artifacts of the heating method. Several published theories of the L to H transition involve the effects of the electric field and edge plasma rotation. The observations in this paper have been critically compared with these theories. [ABSTRACT FROM AUTHOR]

Describes a divertor reflectometer system that was designed and installed on the DIII-D tokamak. System description; Radiation propagation; Attachment of the divertor reflectometer to a large table which is separate from the DIII-D vacuum vessel; Density profiles.

A far-infrared collective scattering system has been installed and is operational on the DIII-D tokamak. The system consists of a 245 GHz (λ = 1.22 mm) twin frequency laser and uses heterodyne detection techniques which enables us to determine the propagation direction of fluctuations in the laboratory frame of reference. At present, three receiver channels are available to study poloidal propagating fluctuations along the entire midplane of DIII-D in a wave number range of 2 ≤ k ≤ 16 cm[sup -1]. The beam geometry of the system dictates a wave number resolution of Δk ≈ 0,7 cm[sup -1] and a radial spatial resolution of ± 20 cm at k = 10 cm[sup -1]. Fluctuation spectra and propagation direction measured during Ohmic operation were found to depend on plasma parameters. Large plasma toroidal rotation was measured which correlated with modifications in the fluctuation spectra. During H-mode operation, a dramatic drop of low-frequency fluctuations (< 250 kHz) is observed coincident with the drop in D-alpha light. This is followed by the appearance of a highfrequency component possibly connected with profile modification and rotation effects. [ABSTRACT FROM AUTHOR]

An O-mode, homodyne reflectometer system utilizing seven discrete channels spanning 15-75 GHz is in operation on DIII-D. A frequency-tunable X-mode system is also used to take fluctuation data from reflecting layers tying between those of the fixed frequency, O-mode system. In order to demonstrate the power and flexibility of this diagnostic, results are presented from several different plasma regimes, including L to H transitions, ELMS, and high beta operation. One specific example is that at all L-H transitions, high-frequency (≥ 50 kHz) fluctuations are dramatically suppressed in a localized region near the edge of the plasma on a timescale of ≈ 100 µs. [ABSTRACT FROM AUTHOR]

The experimental observation of spatially asymmetric scattering on the TOSCA tokamak has led to the identification of two distinct mechanisms responsible for such effects. The first is the influence of the plasma interaction length L on the scattered field (volume effect). When L is small, the plasma behaves like a thin phase grating and equal-amplitude positive and negative diffraction orders are generated (Raman-Nath scattering). As L increases, however, the scattering takes on three-dimensional properties similar to x-ray diffraction from a crystal; only one diffraction order can occur, and then only if the Bragg condition is satisfied (Bragg scattering). The second effect occurs whenever the scattering volume is large enough to contain counterpropagating waves. In this case, constructive/destructive interference can affect the measured intensity of the scattered light. [ABSTRACT FROM AUTHOR]

Development efforts are directed at advanced reflectometry systems for next generation tokamaks such as ITER. Laboratory tests have been successfully completed for a pulsed radar refiectometer (τ...200 ps) to be tested on DIII-D. In addition, there exists a need for sources and components beyond those commercially available. Monolithic quasi-optical spatial power combined arrays of solid state devices will provide continuous, watt level swept source coverage up to 200 GHz as well as electronically controlled beam steering, focusing, and modulation. Studies are also underway to address the feasibility of instantaneous profile coverage using monolithic impulse generators (τ...1-3 ps) comprised of spatially combined nonlinear transmission lines, employing novel quantum barrier varactors. [ABSTRACT FROM AUTHOR]

Presents an erratum on an article on understanding and control of transport in advanced tokamak regimes in DIII-D. Revision of the diagram depicting transport barrier in ion thermal channel; Correct equations used in the calculations; Misprint of diagrams.

During a machine vent in December 1991, two new reflectometer systems were successfully installed and tested on the DIII-D tokamak. The first is an X-mode broadband system primarily intended for density profile measurements, utilizing BWO sources and covering Q and V frequency bands (33-50 and 50-75 GHz). The second system is an adaptation of a pre-existing inside launch (high field side) ECRH waveguide to provide an inside launch reflectometer capability at the same frequencies and polarization as an outside launch fixed frequency O-mode system. The new systems will have a dual role in both directly supporting the DIII-D physics program, and also acting as flexible and adaptable test beds for the development of reactor relevant refiectometer systems, such as required for ITER. Specific examples of planned measurements include investigation of possible in/out plasma asymmetries at the L-H transition and ELMs, and demonstration of routine and reliable density profile measurements. It is expected that preliminary data from the inside launch system will be available by the time of the conference. This work is supported by the U. S. Department of Energy under Grant No. DE-FG03-86-ER53225 and General Atomics subcontract SC120536 under DOE Contract No. DE-AC03-89ER51114. [ABSTRACT FROM AUTHOR]

A critical gradient threshold has been observed for the first time in a systematic, controlled experiment for a locally measured turbulent quantity in the core of a confined high-temperature plasma. In an experiment in the DIII-D tokamak where LTe-1 = ∣∇Te∣/Te and toroidal rotation were varied, long wavelength (kθps ≲ 0.4) electron temperature fluctuations exhibit a threshold in LTe-1: below, they change little; above, they steadily increase. The increase in δTe/Te is concurrent with increased electron heat flux and transport stiffness. Observations were insensitive to rotation. Accumulated evidence strongly enforces the identification of the experimentally observed threshold with ∇Te-driven trapped electron mode turbulence. [ABSTRACT FROM AUTHOR]

Direct evidence of zonal flow (ZF) predator-prey oscillations and the synergistic roles of ZF- and equilibrium EX B flow shear in triggering the low- to high-confinement (L- to H-mode) transition in the DIII-D tokamak is presented. Periodic turbulence suppression is first observed in a narrow layer at and just inside the separatrix when the shearing rate transiently exceeds the turbulence decorrelation rate. The final transition to H mode with sustained turbulence and transport reduction is controlled by equilibrium E X B shear due to the increasing ion pressure gradient. [ABSTRACT FROM AUTHOR]