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A global fitting code for multichordal neutral beam spectroscopic data

Description: Knowledge of the heat deposition profile is crucial to all transport analysis of beam heated discharges. The heat deposition profile can be inferred from the fast ion birth profile which, in turn, is directly related to the loss of neutral atoms from the beam. This loss can be measured spectroscopically be the decrease in amplitude of spectral emissions from the beam as it penetrates the plasma. The spectra are complicated by the motional Stark effect which produces a manifold of nine bright peaks for each of the three beam energy components. A code has been written to analyze this kind of data. In the first phase of this work, spectra from tokamak shots are fit with a Stark splitting and Doppler shift model that ties together the geometry of several spatial positions when they are fit simultaneously. In the second phase, a relative position-to-position intensity calibration will be applied to these results to obtain the spectral amplitudes from which beam atom loss can be estimated. This paper reports on the computer code for the first phase. Sample fits to real tokamak spectral data are shown.
Date: May 1, 1992
Creator: Seraydarian, R.P.; Burrell, K.H. & Groebner, R.J.
Partner: UNT Libraries Government Documents Department

High spatial and temporal resolution visible spectroscopy of the plasma edge in DIII-D

Description: In DIII-D, visible spectroscopic measurements of the He II 468.6 nm and C VI 529.2 nm Doppler broadened spectral lines, resulting from charge exchange recombination interactions between beam neutral atoms and plasma ions, are performed to determine ion temperatures, and toroidal and poloidal rotation velocities. The diagnostics system comprises 32 viewing chords spanning a typical minor radius of 63 cm across the midplane, of which 16 spatial chords span 11 cm of the plasma edge just within the separatrix. A temporal resolution of 260 {mu}s per time slice can be obtained as a result of using MCP phosphors with short decay times and fast camera readout electronics. Results from this system will be used in radial electric field comparisons with theory at the L-H transition and ion transport analysis. 6 refs., 3 figs.
Date: October 1, 1990
Creator: Gohil, P.; Burrell, K.H.; Groebner, R.J. & Seraydarian, R.P.
Partner: UNT Libraries Government Documents Department

Spectroscopic study of edge poloidal rotation and radial electric fields in the DIII-D tokamak

Description: Doppler-shift spectroscopy has shown that finite values of poloidal rotation velocity {upsilon}{sub {theta}} and of radial electric field E{sub r} exist at the edge of a tokamak plasma and that dramatic increases occur in these parameters at an L-H transition. E{sub r} is negative in the L-mode and becomes more negative in the H-mode; {upsilon}{sub {theta}} increases in magnitude at the transition. In addition, the radial gradients (shear) of {upsilon}{sub {theta}} and E{sub r} are large and these gradients also increase at the L-H transition. These results are based on measurements of Doppler shifts of light emitted by He II ions, located in a region about 1--3 cm inside the separatrix. These observations have been made with horizontally-viewing and vertically-viewing spectrometer systems on the DIII-D tokamak. The nearly orthogonal views of these systems are used to determine the plasma's flow velocity in terms of the orthogonal sets {upsilon}{sub {theta}} and {upsilon}{sub {phi}} or of {upsilon}{sub {perpendicular}} and {upsilon}{sub {parallel}}. Knowledge of {upsilon}{sub {perpendicular}} is used to calculate E{sub r} from the force balance equation for a single ion species. The existing results impose constraints on theories of the L-H transition. More detailed studies of the spatial profiles and temporal evolution of {upsilon}{sub {theta}} and E{sub r} will be made with upgraded instrumentation, which is now coming on-line. 28 refs.
Date: October 1, 1990
Creator: Groebner, R.J.; Burrell, K.H.; Gohil, P. & Seraydarian, R.P.
Partner: UNT Libraries Government Documents Department

The charge exchange recombination diagnostic system on the DIII-D tokamak

Description: The charge exchange recombination (CER) diagnostic system on the DIII-D tokamak is used to make spatially and temporally resolved measurements of the ion temperature and toroidal and poloidal rotation velocities. This is performed through visible spectroscopic measurements of the Doppler broadened and Doppler shifted HE II 468.6 nm, the CVI 529.1 nm, and the BV 494.5 nm spectral lines which have been excited by charge exchange recombination interactions between the fully stripped ions and the neutral atoms from the heating beams. The plasma viewing optics comprises 32 viewing chords spanning a typical plasma minor radius of 63 cm across the midplane, of which 15 spatial chords span 4.2 cm at the plasma edge just within the separatrix and provide a chord-to-chord spatial resolution of 0.3 cm. Fast camera readout electronics can provide a temporal resolution of 260 {mu}s per time slice, but the effective minimum integration time, at present, is 1 ms which is limited by the detected photon flux from the plasma and the decay times of the phosphors used on the multichannel plate image intensifiers. Significant changes in the edge plasma radial electric field at the L-H transition have been observed, as determined from the CER measurements, and these results are being extensively compared to theories which consider the effects of sheared electric fields on plasma turbulence. 13 refs., 10 figs.
Date: November 1, 1991
Creator: Gohil, P.; Burrell, K.H.; Groebner, R.J.; Kim, J.; Martin, W.C.; McKee, E.L. et al.
Partner: UNT Libraries Government Documents Department

Liquid Lithium Limiter Experiments in CDX-U

Description: Recent experiments in the Current Drive Experiment-Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, B{sub toroidal} = 2 kG, I{sub P} = 100 kA, T{sub e}(0) = 100 eV, n{sub e}(0) {approx} 5 x 10{sup 19} m{sup -3}) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm{sup 2} (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium-limited discharges are consistent with Z{sub effective} < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced.
Date: October 28, 2004
Creator: Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J. et al.
Partner: UNT Libraries Government Documents Department

Recent Liquid Lithium Limiter Experiments in CDX-U

Description: Recent experiments in the Current Drive eXperiment-Upgrade (CDX-U) provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R=34 cm, a=22 cm, B{sub toroidal} = 2 kG, I{sub P} =100 kA, T{sub e}(0) {approx} 100 eV, n{sub e}(0) {approx} 5 x 10{sup 19} m{sup -3}) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium pool limiter with an area of 2000 cm{sup 2} (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium pool limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium limited discharges are consistent with Z{sub effective} < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced.
Date: May 3, 2005
Creator: Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J. et al.
Partner: UNT Libraries Government Documents Department

Testing of Liquid Lithium Limiters in CDX-U

Description: Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described.
Date: July 30, 2004
Creator: Majeski, R.; Kaita, R.; Boaz, M.; Efthimion, P.; Gray, T.; Jones, B. et al.
Partner: UNT Libraries Government Documents Department

Effects of Large Area Liquid Lithium Limiters on Spherical Torus Plasmas

Description: Use of a large-area liquid lithium surface as a first wall has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter.
Date: June 7, 2004
Creator: Kaita, R.; Majeski, R.; Boaz, M.; Efthimion, P.; Gettelfinger, G.; Gray, T. et al.
Partner: UNT Libraries Government Documents Department

Progress Towards High Performance, Steady-state Spherical Torus

Description: Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} ...
Date: October 2, 2003
Creator: Ono, M.; Bell, M.G.; Bell, R.E.; Bigelow, T.; Bitter, M.; Blanchard, W. et al.
Partner: UNT Libraries Government Documents Department