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Fast wave heating and current drive in DIII-D discharges with negative central shear

Description: The noninductive current driven by fast Alfven waves (FWCD) has been applied to discharges in DIII-D with negative central shear. Driven currents as high as 275 kA have been achieved with up to 3 MW of fast wave power with the efficiency and profile as predicted by theory-based modeling. When counter-current FWCD was applied to discharges with negative central shear, the negative shear was strengthened and prolonged, showing that FWCD can help to control the current profile in advanced tokamak discharges. Under some conditions in negative central shear, the plasma spontaneously makes a transition into a regime of improved performance, with a reduction in both the ion and the electron heat diffusivities. Up to 3 MW of fast wave power has been successfully coupled into H-mode discharges with large edge localized modes through use of an innovative decoupler/hybrid power splitter combination.
Date: October 1, 1996
Creator: Prater, R.; Austin, M.E. & Baity, F.W.
Partner: UNT Libraries Government Documents Department

Observation and control of resistive wall modes

Description: Two approaches to achieving long-time scale stabilization of the ideal kink mode with a real, finite conductivity wall are considered: plasma rotation and active feedback control, DIII-D experiments have demonstrated stabilization of the resistive wall mode (RWM) by sustaining beta greater than the no-wall limit for up to 200 ms, much longer than the wall penetration time of a few ms. These plasmas are typically terminated by an m = 3, n = 1 mode as the plasma rotation slows below a few kHz. Recent temperature profile data shows an ideal MHD mode structure, as expected for the resistive wall mode at beta above the no-wall limit. The critical rotation rate for stabilization is in qualitative agreement with recent theories for dissipative stabilization in the absence of magnetic islands. However, drag by small-amplitude RWMs or damping of stable RWMs may contribute to an observed slowing of rotation at high beta, rendering rotational stabilization more difficult. An initial open-loop active control experiment, using non-axisymmetric external coils and a new array of saddle loop detectors, has yielded encouraging results, delaying the onset of the RWM.
Date: December 1998
Creator: Strait, E. J.; Garofalo, A. M. & Austin, M. E.
Partner: UNT Libraries Government Documents Department

Determination of wall reflectivity for ECE frequencies in DIII-D

Description: The significance of cyclotron radiation losses in next-generation tokamaks depends on the reflectivity of first wall materials. An experimental study of the effective reflectivity for electron cyclotron frequencies in the graphite-walled DIII-D tokamak is reported. Measurements of optically-thin harmonics ({omega} = n{omega}{sub ce}, n > 4) are made for two polarizations from thermal plasma discharges using an absolutely calibrated Michelson interferometer. The reflectivity r and polarization transfer fraction p are obtained by matching measured spectra to simulations from an ECE radiation transport code with adjustable wall parameters. For the frequency range 150-400 GHz average values of r = 0.76 and p = 0.19 are found.
Date: May 1, 1997
Creator: Austin, M.E.; Ellis, R.F. & Luce, T.C.
Partner: UNT Libraries Government Documents Department

Electron cyclotron heating experiments on the DIII-D tokamak

Description: Initial experiments on heating and current drive using second harmonic electron cyclotron heating (ECH) are being performed on the DIII-D tokamak using the new 110 GHz ECH system. Modulation of the ECH power in the frequency range 50 to 300 Hz and detection of the temperature perturbation by ECE diagnostics is used to validate the location of the heating. This technique also determines an upper bound on the width of the deposition profile. Analysis of electron cyclotron current drive indicates that up to 0.17 MA of central current is driven, resulting in a negative loop voltage near the axis.
Date: January 1998
Creator: Prater, R.; Austin, M. E. & Bernabei, S.
Partner: UNT Libraries Government Documents Department

Evidence for modified transport due to sheared E x B flows in high-temperature plasmas

Description: 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 x 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, a rapid improvement in confinement in the boundary layer 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 confinement in the plasma core leading to a regime called the VH-mode. In both examples, the region of improved confinement is characterized by an increase of shear in the radial electric field E{sub r}, reduced levels of turbulence and increases in gradients of temperatures and densities. These observations are consistent with the hypothesis that the improved confinement is caused by an increase in shear of the E x B velocity which leads to a reduction of turbulence. For the VH-mode, the dominant term controlling E{sub r} is the toroidal rotation v{sub {phi}}, indicating that the E{sub r} profile is controlled by the source and transport of toroidal momentum.
Date: November 1, 1994
Creator: Groebner, R. J.; Burrell, K. H. & Austin, M. E.
Partner: UNT Libraries Government Documents Department

FWCD and ECCD experiments on DIII-D

Description: Fast wave current drive and electron cyclotron current drive experiments have been performed on the DIII-D tokamak as part of the advanced tokamak program. The goal of this program is to develop techniques for controlling the profile of the current density in order to access regimes of improved confinement and stability. The experiments on fast wave current drive used a four strap antenna with 900 phasing between straps. A decoupler was used to help maintain the phasing, and feedback control of the plasma position was used to keep the resistive loading constant. RF pickup loops demonstrate that the directivity of the antenna is as expected. Plasma currents up to 0.18 MA were driven by 1.5 MW of fast wave power. Electron cyclotron current drive experiments at 60 GHz have shown 0.1 MA of plasma current driven by 1 MW of power. New fast wave and electron cyclotron heating systems are in development for DIII-D, so that the goals of the advanced tokamak program can be carried out.
Date: February 1, 1994
Creator: Prater, R.; Austin, M. E. & Baity, F. W.
Partner: UNT Libraries Government Documents Department

Noninductive current drive experiments on DIII-D, and future plans

Description: Experiments on DIII-D (and other tokamaks) have shown that improved performance can follow from optimization of the current density profile. Increased confinement of energy and a higher limit on beta have both been found in discharges in which the current density profile is modified through transient means, such as ramping of current or elongation. Peaking of the current distribution to obtain discharges with high internal inductance {ell}{sub i} has been found to be beneficial. Alternatively, discharges with broader profiles, as in the VH-mode or with high beta poloidal, have shown improved performance. Noninductive current drive is a means to access these modes of improved confinement on a steady state basis. Accordingly, experiments on noninductive current drive are underway on the DIII-D tokamak using fast waves, electron cyclotron waves. Recent experiments on fast wave current drive have demonstrated the ability to drive up to 180 kA of noninductive current using 1.5 MW of power at 60 MHz, including the contribution from 1 MW of ECCD and the bootstrap current. Higher power rf current drive systems are needed to strongly affect the current profile on DIII-D. An upgrade to the FWCD system is underway to increase the total power to 6 MW, using two additional antennas and two new 30 to 120 MHz transmitters. Additionally, a 1 MW prototype ECH system at 110 GHz is being developed (with eventual upgrade to 10 MW). With these systems, noninductive current drive at the 1 MA level will be available for experiments on profile control in DIII-D.
Date: February 1, 1994
Creator: Prater, R.; Austin, M. E. & Baity, F. W.
Partner: UNT Libraries Government Documents Department

Improved operation of the Michelson interferometer ECE diagnostic on DIII-D

Description: The measurement of accurate temperature profiles is critical for transport analysis and equilibrium reconstruction in the DIII-D tokamak. Recent refinements in the Michelson interferometer diagnostic have produced more precise electron temperature measurements from electron cyclotron emission and made them available for a wider range of discharge conditions. Replacement of a lens-relay with a low-loss corrugated waveguide transmission system resulted in an increase in throughput of 6 dB and reduction of calibration error to around 5%. The waveguide exhibits a small polarization scrambling fraction of 0.05 at the quarter wavelength frequency and very stable transmission characteristics over time. Further reduction in error has been realized through special signal processing of the calibration and plasma interferograms.
Date: May 1, 1996
Creator: Austin, M.E.; Ellis, R.F.; Doane, J.L. & James, R.A.
Partner: UNT Libraries Government Documents Department

Heatpulse Propagation Studies on DIII-D and TFTR

Description: Sawtooth phenomena have been studied on DIII-D and TFTR. In the experiments, with high power neutral beam injection the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. A strong ballistic electron heat pulse is found on DIII-D, stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the sawtooth precursor. Fast, 2 msec interval, measurements on DIII-D were made of the ion temperature evolution following the sawtooth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, ''ballistic'' behavior seen for the electrons. Further, both the electron and ion heat pulses from partial sawtooth crashes and similar events are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities. These results suggest that the fast sawtooth induced heat pulse propagation is not a feature of non-linear transport models, but that MHD events can have a strong effect on thermal transport.
Date: April 19, 2000
Creator: Fredrickson, E.; Austin, M.E.; Groebner, R.; Manicham, J. & al, et
Partner: UNT Libraries Government Documents Department

Polarization Measurements During Electron Cyclotron Heating Experiments in the DIII-D Tokamak

Description: The polarization of the launched electron cyclotron wave has been optimized for coupling to the X-mode by adjusting the inclination of grooved mirrors located in two consecutive mitre bends of the waveguide. The unwanted O-mode component of the launched beam can be positively identified by the difference in the power deposition profiles between X-mode and O-mode. The optimal polarization for X-mode launch is in good agreement with theoretical expectations.
Date: July 1, 1999
Creator: Petty, C.C.; Luce, T.C.; Austin, M.E.; Ikezi, H.; Lohr, J. & Prater, R.
Partner: UNT Libraries Government Documents Department


Description: OAK-B135 For over twenty years, data acquisition hardware at DIII-D has been based on the CAMAC platform. These rugged and reliable systems, however, are gradually becoming obsolete due to end-of-life issues, ever-decreasing industry support of older hardware, and the availability of modern alternative hardware with superior performance. Efforts are underway at DIII-D to adopt new data acquisition solutions which exploit modern technologies and surpass the limitations of the CAMAC standard. These efforts have involved the procurement and development of data acquisition systems based on the PCI and Compact-PCI platform standards. These systems are comprised of rack-mount computers containing data acquisition boards (digitizers), Ethernet connectivity, and the drivers and software necessary for control. Each digitizer contains analog-to-digital converters, control circuitry, firmware and memory to collect, store, and transfer waveform data acquired using internal or external triggers and clocks. Software has been developed which allows DIII-D computers to program the operational parameters of the digitizers, as well as to upload acquired data into the DIII-D acquisition database. All communication between host computers and the new acquisition systems occurs via standard Ethernet connections, a vast improvement over the slower, serial loop highways used for control and data transfer with CAMAC systems. In addition, the capabilities available in modern integrated and printed circuit manufacture result in digitizers with high channel count and memory density. Cost savings are also realized by utilizing a platform based on standards of the personal computer industry. Details of the new systems at DIII-D are presented, along with initial experience with their use, and plans for future expansion and improvement.
Date: October 1, 2003
Partner: UNT Libraries Government Documents Department

Initial results from the multi-megawatt 110 GHz ECH system for the DIII-D tokamak

Description: The first of three MW-level 110 GHz gyrotrons was operated into the DIII-D tokamak in late 1996. Two additional units will be commissioned during 1997. Each gyrotron is connected to the tokamak by a low loss, windowless, evacuated transmission line using circular corrugated waveguide carrying the HE{sub 11} mode. The microwave beam spot is well focused with a spot size of approximately 6 cm and can be steered poloidally from the center to the outer edge of the plasma. The initial operation with about 0.5 MW delivered to a low density plasma for 0.5 s showed good central electron heating, with peak temperature in excess of 10 keV. The injection was 19{degree} off perpendicular for current drive.
Date: April 1, 1997
Creator: Callis, R.W.; Lohr, J.; O`Neill, R.C.; Ponce, D.; Luce, T.C.; Prater, R. et al.
Partner: UNT Libraries Government Documents Department

Polarization, propagation, and deposition measurements during ECCD experiments on the DIII-D tokamak

Description: The power deposition profiles for different poloidal and toroidal launch angles have been determined by modulating the ECH power and measuring the electron temperature response. The peak of the measured power density follows the poloidal steering of the ECH launcher, and perpendicular launch gives a narrower deposition profile than does oblique (current drive) launch. The difference in wave refraction between X-mode and O-mode allows positive identification of an unwanted O-mode component of the launched beam.
Date: March 1, 1999
Creator: Petty, C.C.; Luce, T.C.; Lin-Liu, Y.R.; Lohr, J.; Prater, R. & Austin, M.E.
Partner: UNT Libraries Government Documents Department

Observation of SOL Current Correlated with MHD Activity in NBI-heated DIII-D Tokamak Discharges

Description: This work investigates the potential roles played by the scrape-off-layer current (SOLC) in MHD activity of tokamak plasmas, including effects on stability. SOLCs are found during MHD activity that are: (1) slowly growing after a mode-locking-like event, (2) oscillating in the several kHz range and phase-locked with magnetic and electron temperature oscillations, (3) rapidly growing with a sub-ms time scale during a thermal collapse and a current quench, and (4) spiky in temporal behavior and correlated with spiky features in Da signals commonly identified with the edge localized mode (ELM). These SOLCs are found to be an integral part of the MHD activity, with a propensity to flow in a toroidally non-axisymmetric pattern and with magnitude potentially large enough to play a role in the MHD stability. Candidate mechanisms that can drive these SOLCs are identified: (a) toroidally non-axisymmetric thermoelectric potential, (b) electromotive force (EMF) from MHD activity, and (c) flux swing, both toroidal and poloidal, of the plasma column. An effect is found, stemming from the shear in the field line pitch angle, that mitigates the efficacy of a toroidally non-axisymmetric SOLC to generate a toroidally non-axisymmetric error field. Other potential magnetic consequences of the SOLC are identified: (i) its error field can introduce complications in feedback control schemes for stabilizing MHD activity and (ii) its toroidally non-axisymmetric field can be falsely identified as an axisymmetric field by the tokamak control logic and in equilibrium reconstruction. The radial profile of a SOLC observed during a quiescent discharge period is determined, and found to possess polarity reversals as a function of radial distance.
Date: March 26, 2004
Creator: Takahashi, H.; Fredrickson, E.D.; Schaffer, M.J.; Austin, M.E.; Evans, T.E.; Lao, L.L. et al.
Partner: UNT Libraries Government Documents Department

Observation and Analysis of Resistive Instabilities in Negative Central Shear in DIII-D Discharges with L-Mode Edge

Description: In DIII-D plasmas with L-mode edge and negative central shear (q{sub axis}-q{sub min} {approx}0.3 to 0.5), an interchange-like instability has been observed [1]. The instability and a subsequent tearing mode cause reduction of the core electron temperature and plasma rotation, and therefore the instability affects discharge evolution and the desired high performance is not achieved. Stability analyses indicate robust ideal stability, while the Resistive Interchange Mode criterion is marginal and the instability appears to be localized initially. Based on this, we believe that the mode is, most likely, a Resistive Interchange Mode. The amplitude of the instability is correlated with the location of the q{sub min} surface and inversely with the fast-ion pressure. There is indication that the interchange-like instability may be ''seeding'' the tearing mode that sometimes follows the interchange-like instability.
Date: July 1, 2002
Creator: Jayakumar, R.J.; Austin, M.E.; Brennan, D.P.; Chu, M.S.; Luce, T.C.; Strait, E.J. et al.
Partner: UNT Libraries Government Documents Department


Description: Experiments have been performed where the T{sub e} profile stiffness was tested at several spatial locations by varying the ECH resonance location. Propagation of the pulses was Fourier analyzed and compared to simulations based on several transport models. The plasma appears to be near the critical T{sub e} gradient for ETG modes and marginally stable to ITG modes. However, the local T{sub e} response to a locally applied heat pulse does not indicate a nonlinear, critical gradient model where T{sub e} is clipped when trying to rise above a critical gradient. The response can be simply understood as the plasma integrating the ECH power, producing an increase in T{sub e} which equilibrates to a new local level with an exponential time constant representing the local confinement time.
Date: July 1, 2002
Creator: DeBOO, J.C.; AUSTIN, M.E.; BRAVENEC, R.V.; KINSEY, J.E; LOHR, J.; LUCE, T.C. et al.
Partner: UNT Libraries Government Documents Department

Effects of ExB Velocity Shear and Magnetic Shear in the Formation of Core Transport Barriers in the DIII-D Tokamak

Description: Core transport barriers can be reliably formed in DIII-D by tailoring the evolution of the current density profile. This paper reports studies of the relative role of magnetic and ExB shear in creating core transport barriers in the DIII-D tokamak and considers the detailed dynamics of the barrier formation. The core barriers seen in DIII-D negative shear discharges form in a stepwise fashion during the initial current ramp. The reasons for the stepwise formation is not known; these steps do not correlate with integer values of q(O) or minimum q. The data from DIII-D is consistent with previous results that negative magnetic shear facilitates the formation of core transport barriers in the ion channel but is not necessary. However, strongly negative magnetic shear does allow formation of transport barriers in particle, electron thermal, ion thermal and angular momentum transport channels. Shots with strong negative magnetic shear have produced the steepest ion temperature and toroidal rotation profiles seen yet in DIII-D. In addition, the ExB shearing rates seen in these shots exceed the previous DIII-D record value by a factor of four.
Date: December 31, 1997
Creator: Burrell, K.H.; Greenfield, C.M.; Lao, L.L.; Staebler, G.M.; Austin, M.E.; Rice, B.W. et al.
Partner: UNT Libraries Government Documents Department

Observation of Abrupt- and Fast-rising SOL Current during Trigger Phase of ELMs in DIII-D Tokamak

Description: Extensive studies to date of edge localized modes (ELMs) have sought their origin inside the separatrix, i.e., MHD instability from steep gradients in the plasma edge, and examined their consequences outside the separatrix, i.e., transport of heat and particles in the scrape-off-layer (SOL) and divertors. Recent measurement by a high-speed scrape-off-layer current (SOLC) diagnostic may indicate that the ELM trigger process lies, in part, in the SOL. Thermoelectrically driven SOLC precedes, or co-evolves with, other parameters of the ELM process, and thus can potentially play a causal role: error field generated by non-axisymmetric SOLC, flowing in the immediate vicinity (approximately 1 cm) of the plasma edge, may contribute toward destabilizing MHD modes. The SOLC, observed concurrently with MHD activity, including ELMs, has been reported elsewhere.
Date: June 27, 2005
Creator: Takahashi, H.; Fredrickson, E. D.; Schaffer, M. J.; Austin, M. E.; Brooks, N. H.; Evans, T. E. et al.
Partner: UNT Libraries Government Documents Department


Description: The onset of tearing modes and the resulting negative effects on plasma performance set significant limits on the operational domain of tokamaks. Modes with toroidal mode number (n) larger than two cause only a minor reduction in energy confinement (<10%). Modes which have a dominant poloidal mode number (m) of three and n=2 lead to a significant reduction in confinement (<30%) at fixed power. The plasma pressure {beta} (normalized to the magnetic field pressure) can be raised further, albeit with very small incremental confinement. Pushing to higher {beta} often destabilizes the m=2/n=1 tearing mode which can lock to the wall and lead to a complete and rapid disruption of the plasma with potentially serious consequences for the tokamak. The {beta} values at which these modes usually appear in conventional tokamak discharges are well below the limits calculated using ideal MHD theory. Therefore, the tearing modes can set effective upper limits on energy confinement and pressure. Significant progress has been made in stabilizing these modes by local current generation using electron cyclotron waves. The tearing mode is essentially a deficit in current flowing helically, resonant with the spatial structure of the local magnetic field. This forms an ''island'' where the magnetic flux is no longer monotonic. It was predicted theoretically [1,2] that replacement of this ''missing'' current would return the plasma to the state prior to the instability. Experiments on the ASDEX-Upgrade [3], JT-60U [4], and DIII-D [5] tokamaks have demonstrated stabilization of m=3/n=2 modes using electron cyclotron current drive (ECCD) to replace the current in the island. Following these initial experiments, recent work on the DIII-D tokamak has demonstrated two significant advances in application of this technique--extending the operational domain stable to m=3/n=2 modes to higher {beta} and the first suppression of the more dangerous m=2/n=1 mode.
Date: July 1, 2002
Creator: LUCE, T.C.; LaHAYE, R.J.; D.A.HUMPHREYS; PETTY, C.C.; PRATER, R.; AUSTIN, M.E. et al.
Partner: UNT Libraries Government Documents Department

Stability in High Gain Plasmas in DIII-D

Description: Fusion power gain has been increased by a factor of 3 in DIII-D plasmas through the use of strong discharge shaping and tailoring of the pressure and current density profiles. H-mode plasmas with weak or negative central magnetic shear are found to have neoclassical ion confinement throughout most of the plasma volume. Improved MHD stability is achieved by controlling the plasma pressure profile width. The highest fusion power gain Q (ratio of fusion power to input power) in deuterium plasmas was 0.0015. which extrapolates to an equivalent Q of 0.32 in a deuterium-tritium plasma and is similar to values achieved in tokamaks of larger size and magnetic fields.
Date: January 1, 1997
Creator: Lazarus, E. A.; Hong, R. M.; Navratil, G. A.; Sabbagh, S.; Strait, E. J.; Rice, B. W. et al.
Partner: UNT Libraries Government Documents Department


Description: 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 Phys. and Contr. Nucl. Fusion Research 1986 (International 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 which 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 seconds or >25 energy confinement times {tau}{sub E}), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved {beta}{sub N} H{sub 89} = 7 for up to 5 times the {tau}{sub E} of 150 ms. The {beta}{sub N} H{sub 89} values of 7 substantially exceed the value of 4 routinely achieved in 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 magnetic hydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.
Date: November 1, 2000
Creator: BURRELL, K.H.; AUSTIN, M.E.; BRENNAN, D.P.; DeBOO, J.C.; DOYLE, E.J.; FENZI, C. et al.
Partner: UNT Libraries Government Documents Department


Description: Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance {approx}10 has been sustained for >5 {tau}{sub E} with q{sub min} >1.5. (The normalized performance is measured by the product {beta}{sub N} H{sub 89} indicating the proximity to the conventional {beta} limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and {beta} {approx}{le} 5%. The limit to increasing {beta} is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate {approx}75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and {beta} control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with {beta}{sub N}H{sub 89} {approx} 7 for up to 6.3 s or {approx} 34 {tau}{sub E}. These discharges appear to be in resistive equilibrium with q{sub min} {approx} 1.05, in agreement with the current profile relaxation time of 1.8 s.
Date: October 1, 2000
Creator: LUCE, T.C.; WADE, M.R.; POLITZER, P.A.; ALLEN, S.L.; AUSTIN, M E.; BAKER, D.R. et al.
Partner: UNT Libraries Government Documents Department