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Identification of magnetic dipole lines above 2000 A in several Mo and Zr ions on the PLT tokamak

Description: A number of spectrum lines arising from magnetic dipole transitions in the n = 3 shell ground configurations of molybdenum and zirconium ions have been identified. These lines are particularly suitable for spectroscopic diagnostics in tokamak-type plasmas in the 500 to 1500 eV temperature range.
Date: May 1, 1982
Creator: Suckewer, S.; Hinnov, E.; Cohen, S.; Finkenthal, M. & Sato, K.
Partner: UNT Libraries Government Documents Department

Up-Dating Of Atomic Data Needed For Ionisation Balance Evaluations of Krypton and Molybdenum

Description: Atomic data for both ionization and recombination of Kr and Mo ions are reviewed, since the rate coefficients for these processes need to be regularly up-dated following the publication of new theoretical calculations and new experimental data. Kr is used in magnetic-confinement-fusion devices to produce a peripheral radiating mantle meant to spread the heat confinement-load on the plasma-facing components. In a few tokamaks Mo tiles cover the plasma-facing surfaces, acting in most cases as a plasma-column limiter. The collected atomic data represent the state of the art on the ionization and recombination data for the two considered elements. Samples of rates are proposed for both ionization and recombination along with tables of the fractional abundances at ionization equilibrium. The proposed rates should be included in codes that simulate the impurity behavior in magnetic-confinement-fusion devices, i.e., when radial transport is added to ionization and recombination to predict spatially resolved charge-state distributions that are to be compared with experimental results.
Date: June 1, 2006
Creator: Mattioli, M; Mazzitelli, G; Fournier, K B; Finkenthal, M & Carraro, L
Partner: UNT Libraries Government Documents Department

Laboratory Astrophysics on ASDEX Upgrade: Measurements and Analysis of K-Shell O, F, and Ne Spectra in the 9-20 A region

Description: High-resolution measurements of K-shell emission from O, F, and Ne have been performed at the ASDEX Upgrade tokamak in Garching, Germany. Independently measured temperature and density profiles of the plasma provide a unique test bed for model validation. We present comparisons of measured spectra with calculations based on transport and collisional-radiative models and discuss the reliability of commonly used diagnostic line ratios.
Date: May 1, 2006
Creator: Hansen, S B; Fournier, K B; Finkenthal, M; Smith, R; Puetterich, T & Neu, R
Partner: UNT Libraries Government Documents Department

Emission in the 50-80 A region from highly ionized silver in PLT tokamak plasmas

Description: The spectrum of silver emitted by Princeton Large Torus (PLT) tokamak plasmas has been recorded in the 25 to 150 A region by a multichannel time-resolving grazing-incidence spectrometer. Silver atoms have been introduced in the tokamak plasma using the laser blow-off technique. For the first time, lines emitted within the 3p-3d transitions of Ag XXIX, Ag XXX, and Ag XXXI ions, between 50 and 80 A, have been identified.
Date: September 1, 1985
Creator: Schwob, J.L.; Wouters, A.; Suckewer, S.; Cohen, S.A. & Finkenthal, M.
Partner: UNT Libraries Government Documents Department

Isoelectronic behavior of resonant and intercombination lines in MgI-like ions

Description: Radiative transitions with very different characteristic rates can serve as important diagnostics of local conditions in a plasma. Here, the observed intensity ratio of the 3s{sup 2} {sup 1}S{sub 0} - 3s3p {sup 1}P{sub 1} to the 3s{sup 2} {sup 1}S{sub 0} - 3s3p {sup 3}P{sub 1} transitions in MgI-like ions has always presented those who model plasma spectra with a challenge; the observed intensity of the intercombination line is always several times greater than what simple models predict. The authors offer a model that takes into account the contribution to the MgI-like emission features from autoionizing levels of the adjacent AlI-like charge state. Models in the present work, which include the effects of configuration interaction on ionic wavefunctions, and the contribution of both direct, impact ionization and autoionization channels from the AlI-like ion to the MgI-like ion, give good agreement with the observed resonant/intercombination (R/I) emission ratio only when a departure from ionization equilibrium is assumed. The authors also identify, for the first time, intercombination lines of the form 3s3p {sup 1}P{sub 1} - 3p{sup 2} {sup 3}P{sub 2} in several elements relevant to both astrophysical and magnetically-confined fusion plasmas.
Date: August 1, 1995
Creator: Fournier, K.B.; Goldstein, W.H.; Finkenthal, M.; Bell, R.E. & Terry, J.L.
Partner: UNT Libraries Government Documents Department

Spectra of germanium and selenium in the 50-350 A region from the PLT tokamak plasma

Description: Spectra of germanium and selenium injected into the PLT tokamak plasma were observed in the 50 to 350 A region for GeXIV-XXV (KI to OI-like) and SeXVI-XXIV (KI to NaI-like). A number of 3p/sup k/-3p/sup k-1/3d transitions predicted by isoelectronic sequence extrapolation have been identified. Also, previously identified lines from ions in the AlI to OI-like and KI-like isoelectronic sequences have been observed in the tokamak plasma.
Date: March 1, 1983
Creator: Stratton, B.C.; Hodge, W.L.; Moos, H.W.; Schwob, J.L.; Suckewer, S.; Finkenthal, M. et al.
Partner: UNT Libraries Government Documents Department

Spatially resolved measurements of fully ionized low-Z impurities in the PDX tokamak

Description: Radial distributions of fully ionized oxygen and carbon in the PDX tokamak plasma are reported. These ions were detected via radiation emitted in charge-exchange recombination reactions between the impurities and hydrogen atoms from a non-perturbing neutron beam. The C/sup 6 +/ and O/sup 8 +/ ions are observed out to radii beyond the limiter, which is in contrast to expectations based on coronal equilibrium but consistent with a simple diffusive transport model. Central values of Z/sub eff/ obtained with these measurements agree with values obtained from plasma resistivity and visible bremsstrahlung measurements.
Date: July 1, 1982
Creator: Fonck, R.J.; Finkenthal, M.; Goldston, R.J.; Herndon, D.L.; Hulse, R.A.; Kaita, R. et al.
Partner: UNT Libraries Government Documents Department

Soft X-ray Tangential Imaging of the NSTX Core Plasma by Means of a MPGD Pin-hole Camera

Description: A fast X-ray system based on a Micro Pattern Gas Detector has been used, for the first time, to investigate emission from the plasma core of the National Spherical Tokamak eXperiment (NSTX) at the Princeton Plasma Physics Laboratory. The results presented in this work demonstrate the capability of such a device to measure with a time resolution of the order of 1 ms the curvature and the elongation of the X-ray iso-emissivity contours, under various plasma conditions. Also, comparisons with the magnetic surface structure calculated by the EFIT code show good agreement between reconstructed flux surface and the soft X-ray emissions (SXR) for poloidal beta values up to 0.6. For greater values of beta, X-ray iso-emissivity contours become circular, while magnetic flux surface reconstructions yield elongation 1.5 < k < 2.2. The X-ray images have been acquired with a (statistical) signal to noise ratio (SNR) per pixel of about 30. Thanks to the direct and efficient X-ray conversion and its operation in a photon counting mode, this new diagnostic tool allows the routine investigation of the plasma core with a sampling rate of 1 kHz and extremely high SNR under all experimental conditions in NSTX.
Date: July 24, 2003
Creator: Pacella, D.; Leigheb, M.; Bellazzini, R.; Brez, A.; Finkenthal, M.; Stutman, D. et al.
Partner: UNT Libraries Government Documents Department

2l-nl{prime} x-ray transitions from neonlike charge states of the row 5 metals with 39 {le} Z {le} 46

Description: X-ray spectra of 2l-2l{prime} transitions with 3 {le} n {le} 12 in the row five transition metals zirconium (Z = 40), niobium (Z = 41), molybdenum (Z = 42) and palladium (Z = 46) from charge states around neonlike have been observed from Alcator C-Mod plasmas. Accurate wavelengths ({+-} .2 m{angstrom}) have been determined by comparison with neighboring argon, chlorine and sulfur lines with well known wavelengths. Line identifications have been made by comparison to ab initio atomic structure calculations, using a fully relativistic, parametric potential code. For neonlike ions, calculated wavelengths and oscillator strengths are tabulated for 2p-nd transitions in Y (Z = 39), Tc (Z = 43), Ru (Z = 44) and Rh (Z = 45) with n = 6 and 7. The magnitude of the configuration interaction between the (2p{sup 5}){sub 1/2}6d{sub 3/2} J = 1 level and the (2p{sup 5}){sub 3/2}7D{sub 5/2} J = 1 levels is demonstrated as a function of atomic number for successive neonlike ions. Measured spectra of selected transitions in the aluminum-, magnesium-, sodium- and fluorine like isosequences are also shown.
Date: March 18, 1996
Creator: Rice, J. E.; Terry, J. L.; Marmar, E. S.; Fournier, K. B.; Goldstein, W. H.; Finkenthal, M. et al.
Partner: UNT Libraries Government Documents Department

Estimates if population inversion for deep-UV transitions in Kr-like Y,Zr,Nb and Mo in a high-current reflex discharge

Description: Kr-like ions are good candidates for FUV lasing since they can be produced in plasmas quite easily. We present results from a spectroscopic investigation of Y IV emission from a high current density, cold cathode reflex discharge. The Y II to Y V emission is recorded in the 200-3000 {angstrom} range using photometrically calibrated spectrometers, while the emission of trace aluminum ions serves for plasma diagnostics. The intensities of the Y IV 4d - 5p and 5s - 5p transitions strongly increase relative to lines from Y II and Y III with increasing plasma current. The spectra studied here are obtained at a current density of 1.75 A/cm{sup 2}. Experimental Y IV intensity ratios spanning several excited configurations are compared with collisional radiative predictions of the HULLAC atomic physics package. Good agreement is found for the measured and predicted ratios of 4p{sup 5}5p to 4p{sup 5}5s level populations per statistical weight. Finally, the response of the Kr-like system to a fast, transient excitation pulse is examined using the RADEX code. Large transient gains are predicted for several 5s - 5p transitions in Y IV, Zr V, Nb VI and Mo VII.
Date: July 6, 1999
Creator: Finkenthal, M.; May, M. J.; Fournier, K.; Goldstein, W. H.; Shlyaptsev, V. N.; Soukhanovskii, V. et al.
Partner: UNT Libraries Government Documents Department

O-shell emission of uranium in a high temperature, low density plasma

Description: We present models for the soft X ray emission spectra of UXXIV, UXXV, UXXX, UXXXI and UXXXII under tokamak conditions. The spectra are calculated from collisional-radiative models and cover the 60 to 200{Angstrom} range. A fully relativistic parametric potential code has been used for the ab initio atomic structure calculations, and electron impact excitation rates have been computed in the distorted wave approximation. The ions considered here display a range of spectroscopic phenomena to be found in the O-shell of heavy elements, that can be compared with experimental observations. These include the configuration interaction between 5s{sup 2}5p{sup 6}5d{sup k{minus}1}5f and 5s{sup 2}5p{sup 5}5d{sup k+1} in ions with 5s{sup 2}5p{sup 6}5d{sup k} ground configurations, represented by UXXIV (k=l); configuration interaction between 5s{sup 2}5p{sup k{minus}1}5d and 5s{sup 2}5p{sup k+l} in ions with 5s{sup 2}5p{sup k} ground configurations, such as UXXX (k=l); the unique system of satellite emission lines to the resonant 5s-5p transitions in UXXXII arising from highly metastable 4f{sup 13}5s{sup 2} levels; and the bright, isolated resonant lines associated with the closed shell UXXXI and systems.
Date: December 1, 1993
Creator: Fournier, K. B.; Osterheld, A. L.; Goldstein, W. H.; Finkenthal, M.; Holmes, C. P. & Moos, H. W.
Partner: UNT Libraries Government Documents Department

Empirical evaluation of the radiative cooling coefficient for krypton gas in the FTU plasma

Description: For future fusion reactors, a careful balance must be achieved between the cooling of the outer plasma via impurity radiation and the deleterious effects of inevitable core penetration by impurity ions. We have injected krypton gas into the Frascati Tokamak Upgrade (FTU) plasma. The measured visible bremsstrahlung and bolometric signals from krypton have been inverted and the resulting radial impurity density profile and power loss profile for krypton gas are extracted. Using the measured electron density and temperature profiles, the radiative cooling coefficient for krypton is derived. The level of intrinsic impurities (Mo, Cr, Mn and Fe) in the plasma during the krypton puffing is monitored with a VUV SPRED spectrometer. Models for krypton emissivity from the literature are compared to our measured results. 7 figs.
Date: November 18, 1997
Creator: Fournier, K.B.; Pacella, D.; Gregory, B.C.; May, M.J.; Mazzitelli, G.; Gabellieri, L. et al.
Partner: UNT Libraries Government Documents Department

Impurity transport studies of intrinsic Mo and injected Ge in high temperature ECRH heated FTU tokamak plasmas

Description: FTU plasmas reached a peak electron temperature up to 11 keV with ECRH heating during the current ramp up phase. For these plasmas X-ray emission of highly ionized molybdenum, the dominant intrinsic impurity, are presented in section II, and VUV spectra of injected germanium are presented in section III. In section IV the conclusions are discussed.
Date: June 1, 1999
Creator: Bracco, F; Crisanti, M; Finkenthal, M; Fournier, K; Gabellieri, G; Granucci, G 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

CDX-U Operation with a Large Area Liquid Lithium Limiter

Description: The Current Drive experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has begun experiments with a fully toroidal liquid lithium limiter. CDX-U is a compact [R = 34 cm, a = 22 cm, B(subscript)toroidal = 2 kG, I(subscript)P = 100 kA, T(subscript)e(0) {approx} 100 eV, n(subscript)e(0) {approx} 5 x 10{sup 19} m{sup -3}] short-pulse (<25 msec) spherical torus with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, can be filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium prior to the experiment. Surface coatings are evident on part of the lithium. Despite the surface coatings, tokamak discharges operated in contact with the lithium-filled tray show evidence of reduced impurities and recycling. The reduction in recycling is largest when the lithium is liquefied by heating to 250 degrees Celsius.
Date: July 12, 2002
Creator: Majeski, R.; Boaz, M.; Hoffman, D.; Jones, B.; Kaita, R.; Kugel, H. et al.
Partner: UNT Libraries Government Documents Department

Liquid Lithium Limiter Effects on Tokamak Plasmas and Plasma-Liquid Surface Interactions

Description: We present results from the first experiments with a large area liquid lithium limiter in a magnetic fusion device, and its effect on improving plasma performance by reducing particle recycling. Using large area liquid metal surfaces in any major fusion device is unlikely before a test on a smaller scale. This has motivated its demonstration in the CDX-U spherical torus with a unique, fully toroidal lithium limiter. The highest current discharges were obtained with a liquid lithium limiter. There was a reduction in recycling, as indicated by a significant decrease in the deuterium-alpha emission and oxygen radiation. How these results might extrapolate to reactors is suggested in recycling/retention experiments with liquid lithium surfaces under high-flux deuterium and helium plasma bombardment in PISCES-B. Data on deuterium atoms retained in liquid lithium indicate retention of all incident ions until full volumetric conversion to lithium deuteride. The PISCES-B results also show a material loss mechanism that lowers the maximum operating temperature compared to that for the liquid surface equilibrium vapor pressure. This may restrict the lithium temperature in reactors.
Date: October 15, 2002
Creator: Kaita, R.; Majeski, R.; Doerner, R.; Antar, G.; Baldwin, M.; Conn, R. et al.
Partner: UNT Libraries Government Documents Department

Surface Treatment of a Lithium Limiter for Spherical Torus Plasma Experiments

Description: The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. As part of investigations to determine the feasibility of this approach, plasma interaction questions in a toroidal plasma geometry are being addressed in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The first experiments involved a toroidally local lithium limiter (L3). Measurements of pumpout rates indicated that deuterium pumping was greater for the L3 compared to conventional boron carbide limiters. The difference in the pumpout rates between the two limiter types decreased with plasma exposure, but argon glow discharge cleaning was able to restore the pumping effectiveness of the L3. At no point, however, was the extremely low recycling regime reported in previous lithium experiments achieved. This may be due to the much larger lithium surfaces that were exposed to the plasma in the earlier work. The possibility will be studied in the next set of CDX-U experiments, which are to be conducted with a large area, fully toroidal lithium limiter.
Date: March 20, 2001
Creator: Kaita, R.; Majeski, R.; Doerner, R.; Antar, G.; Timberlake, J.; Spaleta, J. et al.
Partner: UNT Libraries Government Documents Department

Diagnostics for liquid lithium experiments in CDX-U

Description: A flowing liquid lithium first wall or diverter target could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls in fusion reactors. To investigate the interaction of a spherical torus plasma with liquid lithium limiters, large area diverter targets, and walls, discharges will be established in the Current Drive Experiment-Upgrade (CDX-U) where the plasma-wall interactions are dominated by liquid lithium surfaces. Among the unique CDX-U lithium diagnostics is a multi-layer mirror (MLM) array, which will monitor the 135 {angstrom} LiIII line for core lithium concentrations. Additional spectroscopic diagnostics include a grazing incidence XUV spectrometer (STRS) and a filterscope system to monitor D{sub {alpha}} and various impurity lines local to the lithium limiter. Profile data will be obtained with a multichannel tangential bolometer and a multipoint Thomson scattering system configured to give enhanced edge resolution. Coupons on th e inner wall of the CDX-U vacuum vessel will be used for surface analysis. A 10,000 frame per second fast visible camera and an IR camera will also be available.
Date: June 21, 2000
Creator: Kaita, R.; Efthimion, P.; Hoffman, D.; Jones, B.; Kugel, H.; Majeski, R. 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

Plasma Performance Improvements with Liquid Lithium Limiters in CDX-U

Description: The use of flowing liquid lithium as a first wall for a reactor has potentially attractive physics and engineering features. The Current Drive experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has begun experiments with a fully toroidal liquid lithium limiter. CDX-U is a compact [R = 34 cm, a = 22 cm, Btoroidal = 2 kG, IP =100 kA, T(subscript)e(0) {approx} 100 eV, n(subscript)e(0) {approx} 5 x 10{sup 19} m-3] short-pulse (<25 msec) spherical tokamak with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, can be filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium prior to the experiment. The total area of the tray is approximately 2000 cm{sup 2}. The tokamak edge plasma, when operated in contact with the lithium-filled tray, shows evidence of reduced impurities and recycling. The reduction in re cycling and impurities is largest when the lithium is liquefied by heating to 250 degrees Celsius. Discharges which are limited by the liquid lithium tray show evidence of performance enhancement. Radiated power is reduced and there is spectroscopic evidence for increases in the core electron temperature. Furthermore, the use of a liquid lithium limiter reduces the need for conditioning discharges prior to high current operation. The future development path for liquid lithium limiter systems in CDX-U is also discussed.
Date: July 12, 2002
Creator: Majeski, R.; Boaz, M.; Hoffman, D.; Jones, B.; Kaita, R.; Kugel, H. et al.
Partner: UNT Libraries Government Documents Department

Spherical Torus Plasma Interactions with Large-area Liquid Lithium Surfaces in CDX-U

Description: The Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego. Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance.
Date: January 18, 2002
Creator: Kaita, R.; Majeski, R.; Boaz, M.; Efthimion, P.; Jones, B.; Hoffman, D. 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

High Performance Plasmas on the National Spherical Torus Experiment

Description: The National Spherical Torus Experiment (NSTX) has produced toroidal plasmas at low aspect ratio (A = R/a = 0.86 m/0.68 m approximately equal to 1.3, where R is the major radius and a is the minor radius of the torus) with plasma currents of 1.4 MA. The rapid development of the machine has led to very exciting physics results during the first full year of physics operation. Pulse lengths in excess of 0.5 sec have been obtained with inductive current drive. Up to 4 MW of High Harmonic Fast Wave (HHFW) heating power has been applied with 6 MW planned. Using only 2 MW of HHFW heating power clear evidence of electron heating is seen with HHFW, as observed by the multi-point Thomson scattering diagnostic. A noninductive current drive concept known as Coaxial Helicity Injection (CHI) has driven 260 kA of toroidal current. Neutral-beam heating power of 5 MW has been injected. Plasmas with beta toroidal (= 2 mu(subscript ''0'')<p>/B(superscript ''2'') = a measure of magnetic confinement efficiency ) of 22% have been achieved, as calculated using the EFIT equilibrium reconstruction code. Beta-limiting phenomena have been observed, and the maximum beta toroidal scales with I(subscript ''p'')/aB(subscript ''t''). High frequency (>MHz) magnetic fluctuations have been observed. High-confinement mode plasmas are observed with confinement times of >100 msec. Beam-heated plasmas show energy confinement times in excess of those predicted by empirical scaling expressions. Ion temperatures in excess of 2.0 keV have been measured, and power balance suggests that the power loss from the ions to the electrons may exceed the calculated classical input power to the ions.
Date: July 10, 2001
Creator: Gates, D.A.; Bell, M.G.; Bell, R.E.; Bialek, J.; Bigelow, T.; Bitter, M. et al.
Partner: UNT Libraries Government Documents Department