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Measurement of emitted power in the divertor region in PBX

Description: In strongly indented PBX plasmas, radiated power profiles are calculated by combining data obtained from two bolometer arrays in order to study poloidal asymmetries arising from plasma indentation and characterize emission from the divertor region. A compact, 15-channel bolometer array that views the plasma tangentially along the midplane complements a 19-channel array that scans the plasma vertically in a poloidal plane. Assuming that radiated power density is constant along a magnetic flux surface, the contributions to the irradiance viewed by the poloidal array from the region inside the separatrix can be calculated from the midplane measurements. The difference between this contribution and the measured poloidal distribution is assumed to originate in the expanded boundary divertor. In general, the total radiated power loss constitutes 40% of the total input power, and is independent of beam geometry. However, the radiation profiles in the main plasma and divertor region depend on operating conditions such as beam geometry and gas puffing rates. Radiation from the main plasma accounts for 20% of the input power and radiation from the divertor region accounts for 20%. Accumulation of impurities during neutral-beam-heated discharges can cause peak radiation levels to exceed 1 W/cm/sup 3/, leading to a thermal collapse of the plasma.
Date: September 1, 1986
Creator: Paul, S.F.; Fonck, R.J. & Schmidt, G.L.
Partner: UNT Libraries Government Documents Department

Operation of a tangential bolometer on the PBX tokamak

Description: A compact 15-channel bolometer array that views plasma emission tangentially across the midplane has been installed on the PBX tokamak to supplement a 19-channel poloidal array which views the plasma perpendicular to the toroidal direction. By comparing measurements from these arrays, poloidal asymmetries in the emission profile can be assessed. The detector array consists of 15 discrete 2-mm x 2-mm Thinistors, a mixed semiconductor material whose temperature coefficient of resistance is relatively high. The accumulated heat incident on a detector gives rise to a change in the resistance in each active element. Operated in tandem with an identical blind detector, the resistance in each pair is compared in a Wheatstone bridge circuit. The variation in voltage resulting from the change in resistance is amplified, stored on a CAMAC transient recorder during the plasma discharge, and transferred to a VAX data acquisition computer. The instantaneous power is obtained by digitally smoothing and differentiating the signals in time, with suitable compensation for the cooling of the detector over the course of a plasma discharge. The detectors are ''free standing,'' i.e., they are supported only by their electrical leads. Having no substrate in contact with the detector reduces the response time and increases the time it takes for the detector to dissipate its accumulated heat, reducing the compensation for cooling required in the data analysis. The detectors were absolutely calibrated with a tungsten-halogen filament lamp and were found to vary by +-3%. The irradiance profiles are inverted to reveal the radially resolved emitted power density from the plasma, which is typically in the 0.1 to 0.5 W/cm/sup 3/ range.
Date: April 1, 1987
Creator: Paul, S.F.; Fonck, R.J. & Schmidt, G.L.
Partner: UNT Libraries Government Documents Department

Direct numerical solution of Poisson`s equation in cylindrical (r, z) coordinates

Description: A direct solver method is developed for solving Poisson`s equation numerically for the electrostatic potential {phi}(r,z) in a cylindrical region (r < R{sub wall}, 0 < z < L). The method assumes the charge density {rho}(r,z) and wall potential {phi}(r = R{sub wall}, z) are specified, and {partial_derivative}{phi}/{partial_derivative}z = 0 at the axial boundaries (z = 0, L).
Date: July 22, 1997
Creator: Chao, E.H.; Paul, S.F.; Davidson, R.C. & Fine, K.S.
Partner: UNT Libraries Government Documents Department

Confinement of Pure Ion Plasma in a Cylindrical Current Sheet

Description: A novel method for containing a pure ion plasma at thermonuclear densities and temperatures has been modeled. The method combines the confinement properties of a Penning-Malmberg trap and some aspects of the magnetic field geometry of a pulsed theta-pinch. A conventional Penning trap can confine a uniform-density plasma of about 5x1011 cm-3 with a 30-Tesla magnetic field. However, if the axial field is ramped, a much higher local ion density can be obtained. Starting with a 107 cm-3 trapped deuterium plasma in a conventional Penning-Malmberg trap at the Brillouin limit (B = 0.6 Tesla), the field is ramped to 30 Tesla. Because the plasma is comprised of particles of only one sign of charge, transport losses are very low, i.e., the conductivity is high. As a result, the ramped field does not penetrate the plasma and a diamagnetic surface current is generated, with the ions being accelerated to relativistic velocities. To counteract the inward j x B forces from this induced current, additional ions are injected into the plasma along the axis to increase the density (and mutual electrostatic repulsion) of the target plasma. In the absence of the higher magnetic field in the center, the injected ions drift outward until a balance is established between the outward driving forces (centrifugal, electrostatic, pressure gradient) and the inward j x B force. An equilibrium calculation using a relativistic, 1-D, cold-fluid model shows that a plasma can be trapped in a hollow, 49-cm diameter, 0.2-cm thick cylinder with a density exceeding 4 x 1014 cm-3.
Date: December 10, 1999
Creator: Phillips, C.K.; Chao, E.H.; Davidson, R.C. & Paul, S.F.
Partner: UNT Libraries Government Documents Department

A High-Speed Optical Diagnostic that uses Interference Filters to Measure Doppler Shifts

Description: A high-speed, non-invasive velocity diagnostic has been developed for measuring plasma rotation. The Doppler shift is determined by employing two detectors that view line emission from the identical volume of plasma. Each detector views through an interference filter having a passband that varies linearly with wavelength. One detector views the plasma through a filter whose passband has a negative slope and the second detector views through one with a positive slope. Because each channel views the same volume of plasma, the ratio of the amplitudes is not sensitive to variations in plasma emission. With suitable knowledge of the filter characteristics and the relative gain, the Doppler shift is readily obtained in real time from the ratio of two channels without needing a low throughput spectrometer. The systematic errors--arising from temperature drifts, stability, and frequency response of the detectors and amplifiers, interference filter linearity, and ability to thoroughly homogenize the light from the fiber bundle--can be characterized well enough to obtain velocity data with + or - 1 km/sec with a time resolution of 0.3 msec.
Date: August 9, 2004
Creator: Paul, S.F.; Cates, C.; Mauel, M.; Maurer, D.; Navratil, G. & Shilov, M.
Partner: UNT Libraries Government Documents Department

Local carbon diffusion coefficient measurement in the S-1 spheromak

Description: The local carbon diffusion coefficient was measured in the S - 1 spheromak by detecting the radial spread of injected carbon impurity. The radial impurity density profile is determined by the balance of ionization and diffusion. Using measured local electron temperature T/sub e/ and density n/sub e/, the ionization rate is determined from which the particle diffusion coefficient is inferred. The results found in this work are consistent with Bohm diffusion. The absolute magnitude of D/sub /perpendicular// was determined to be (4/approximately/6) /times/ D/sub Bohm/. 25 refs., 13 figs., 2 tabs.
Date: October 1, 1988
Creator: Mayo, R.M.; Levinton, F.M.; Meyerhofer, D.D.; Chu, T.K.; Paul, S.F. & Yamada, M.
Partner: UNT Libraries Government Documents Department

Accounting of the Power Balance for Neutral-beam-heated H-Mode Plasmas in NSTX

Description: A survey of the dependence of power balance on input power, shape, and plasma current was conducted for neutral-beam-heated plasmas in the National Spherical Torus Experiment (NSTX). Measurements of heat to the divertor strike plates and divertor and core radiation were taken over a wide range of plasma conditions. The different conditions were obtained by inducing a L-mode to H-mode transition, changing the divertor configuration [lower single null (LSN) vs. double-null (DND)] and conducting a NBI power scan in H-mode. 60-70% of the net input power is accounted for in the LSN discharges with 20% of power lost as fast ions, 30-45% incident on the divertor plates, up to 10% radiated in the core, and about 12% radiated in the divertor. In contrast, the power accountability in DND is 85-90%. A comparison of DND and LSN data show that the remaining power in the LSN is likely to be directed to the upper divertor
Date: August 9, 2004
Creator: Paul, S.F.; Maingi, R.; Soukhanovskii, V.; Kaye, S.M.; Kugel, H. & Team, the NSTX Research
Partner: UNT Libraries Government Documents Department

Neutral beam emission spectroscopy diagnostic for measurement of density fluctuations on the TFTR tokamak

Description: A multi-channel diagnostic for measuring low amplitude, long wavelength (k{sub {perpendicular}{rho}i} < 0.5) density fluctuations along the outer half of the plasma has been installed on TFTR. It is based on observing fluctuations in the H{sub {alpha}} fluorescence of a neutral heating beam due to collisional excitation from the plasma and impurity ions. Both radial and poloidal correlation lengths as short as 2--3 cm can be determined, with the spatial resolution limited primarily by the width and geometry of the three neutral beam sources. Optical fibers transmit the light from a 20-cm diameter vacuum window, re-entrant mirror, and lens assembly to sixteen interference filter/photomultiplier combinations located outside the radiation area. Initially, the fibers comprise a fixed 55-channel radial array and readily movable 10-channel vertical arrays which can be positioned at 27 radial locations. The filters are designed to accept the Doppler-shifted H{sub {alpha}} emission from primary energy component of the neutral beam, and reject background lines and unshifted edge H{sub {alpha}}. The measurable fluctuation amplitude ( S/N = 1) is limited to 0.5% over a 100 kHz bandwidth by the photon noise associated with the DC level of the beam emission. The contribution of impurities to the total beam fluorescence will be determined directly by measuring impurity density fluctuations using charge exchange recombination emission from the n = 8 {minus} 7 CVI line at 5292 {angstrom}. 6 refs., 2 figs., 2 tabs.
Date: June 1, 1990
Creator: Paul, S.F. (Princeton Univ., NJ (USA). Plasma Physics Lab.) & Fonck, R.J. (Wisconsin Univ., Madison, WI (USA). Dept. of Nuclear Engineering)
Partner: UNT Libraries Government Documents Department

Impurity behavior during ion-Bernstein wave heating in PBX-M

Description: Ion-Bernstein-wave heating (IBWH) has been tested in several tokamaks. In some cases the results have been quite positive, producing temperature increases and also improving both energy and particle confinement times, whereas in others, no distinctive changes were observed. Most recently, IBWH has been utilized in the Princeton Beta Experiment-Modified (PBX-M) where the long-range goal is the achievement of operation in the second stable region by current and pressure profile control. Investigations have been performed in this machine using IBWH as the sole source of auxiliary power or using IBWH in conjunction with neutral-beam injection (NBI) or with lower-hybrid current drive (LHCD). Impurity studies seem particularly important for IBWH since not only have influxes often been observed to increase, but the global impurity confinement time has also been shown to lengthen as the confinement of the working gas improved. The authors present here a set of characteristic experimental results regarding the impurity behavior in PBX-M; in general, these are consonant with previous observations in other tokamaks.
Date: September 1, 1994
Creator: Isler, R. C.; Post-Zwicker, A. P.; Paul, S. F.; Tighe, W.; Ono, M.; LeBlanc, B. P. et al.
Partner: UNT Libraries Government Documents Department

The relationship between turbulence measurements and transport in different heating regimes in TFTR

Description: The scaling of broad band density fluctuations in the confinement zone of TFTR measured by microwave scattering, beam emission spectroscopy (BES), and reflectometry show a relationship between these fluctuations and energy transport measured from power balance calculations. In L-mode plasmas scattering and BES indicates that the density fluctuation level, {delta}n{sup 2}, in the confinement zone for 0.2 < k{perpendicular}ps < 1.0 depends qualitatively on P{sub aux} and I{sub p} in a way that is consistent with variations in energy transport. Fluctuation levels measured with all systems increase strongly toward the edge in all heating regimes following increases in energy transport coefficients. Measurements using BES have shown that poloidal and radial correlation lengths in the confinement zone of L-mode and supershot plasmas fall in the range of 1 to 2 cm. with a wave structure which has k{sub max} {approx} 1 cm{sup {minus}1} (k{perpendicular}ps {approx} 0.2) in the poloidal direction and k{sub max} approaching zero in the radial direction. A simple estimate of the diffusion coefficient based on a measured radial correlation length and correlation time indicates good agreement with power balance calculations. Similar estimates using reflectometry give radial coherence lengths at 10 to 20 kHz in low density ohmic and supershot plasmas of between I and 2 cm.
Date: October 1, 1992
Creator: Bretz, N. L.; Mazzucato, E.; Nazikian, R.; Paul, S. F.; Hammett, G.; Rewoldt, G. et al.
Partner: UNT Libraries Government Documents Department

Plasma-Material Interface Development for Future Spherical Tokamak-based Devices in NSTX.

Description: The divertor plasma-material interface (PMI) must be able to withstand steady-state heat fluxes up to 10 MW/m{sup 2} (a limit imposed by the present day divertor material and engineering constraints) with minimal material erosion, as well as to provide impurity control and ion density pumping capabilities. In spherical tokamaks (STs), the compact divertor geometry and the requirement of low core electron collisionality n*{sub e} at n{sub e} &lt; 0.5-0.7 n{sub G} (where n{sub G} is the Greenwald density) for increased neutral beam current drive efficiency impose much greater demands on divertor and first-wall particle and heat flux mitigation solutions. In NSTX, divertor heat flux mitigation and impurity control with an innovative 'snowflake' divertor configuration and ion density pumping by evaporated lithium wall and divertor coatings are studied. Lithium coatings have enabled ion density reduction up to 50% in NSTX through the reduction of wall and divertor recycling rates. The 'snowflake' divertor configuration was obtained in NSTX in 0.8-1 MA 4-6 MW NBI-heated H-mode lithium-assisted discharges using three divertor coils. The snowflake divertor formation was always accompanied by a partial detachment of the outer strike point with an up to 50% increase in divertor radiation from intrinsic carbon, the peak divertor heat flux reduction from 3-6 MW/m{sup 2} to 0.5-1 MW/m{sup 2}, and a significant increase in divertor volume recombination. High core confinement was maintained with the snowflake divertor, evidenced by the t{sub E}, W{sub MHD} and the H98(y,2) factors similar to those of the standard divertor discharges. Core carbon concentration and radiated power were reduced by 30-70%, apparently as a result of reduced divertor physical and chemical sputtering in the snowflake divertor and ELMs. In the SFD discharges, the MHD stability of the H-mode pedestal region was altered leading to the re-appearance of medium size (DW/W = 5-10%), Type I, ...
Date: September 24, 2011
Creator: Soukhanovskii, V. A.; Battaglia, D.; Bell, M G.; Bell, R. E.; Diallo, A.; Gerhardt, S. et al.
Partner: UNT Libraries Government Documents Department

Transport in Auxiliary Heated NSTX Discharges

Description: The NSTX spherical torus (ST) provides a unique platform to investigate magnetic confinement in auxiliary-heated plasmas at low aspect ratio. Auxiliary power is routinely coupled to ohmically heated plasmas by deuterium neutral-beam injection (NBI) and by high-harmonic fast waves (HHFW) launch. While theory predicts both techniques to preferentially heat electrons, experiment reveals the electron temperature is greater than the ion temperature during HHFW, but the electron temperature is less than the ion temperature during NBI. In the following we present the experimental data and the results of transport analyses.
Date: July 10, 2003
Creator: LeBlanc, B.P.; Bell, M.G.; Bell, R.E.; Bitte, M.L.; Bourdelle, C.; Gates, D.A. et al.
Partner: UNT Libraries Government Documents Department

Confinement Studies of Auxiliary Heated NSTX Plasmas

Description: The confinement of auxiliary heated NSTX discharges is discussed. The energy confinement time in plasmas with either L-mode or H-mode edges is enhanced over the values given by the ITER97L and ITER98Pby(2) scalings, being up to 2-3 times L-mode and 1.5 times H-mode. TRANSP calculations based on the kinetic profile measurements reproduce the magnetics-based determination of stored energy and the measured neutron production rate. Power balance calculations reveal that, in a high power neutral beam heated H-mode discharge, the ion thermal transport is near neoclassical levels, and well below the electron thermal transport, which is the main loss channel. Perturbative impurity injection techniques indicate the particle diffusivity to be slightly above the neoclassical level in discharges with L-mode edge. High-harmonic fast-wave (HHFW) bulk electron heating is described and thermal transport is discussed. Thermal ion transport is found to be above neoclassical level, but thermal electron transport remains the main loss mechanism. Evidences of an electron thermal internal transport barrier obtained with HHFW heating are presented. A description of H-mode discharges obtained during HHFW heating is presented.
Date: May 6, 2003
Creator: LeBlanc, B.P.; Bell, M.G.; Bell, R.E.; Bitter, M.L.; Bourdelle, C.; Gates, D.A. et al.
Partner: UNT Libraries Government Documents Department

Excitation of toroidal Alfven eigenmodes in TFTR

Description: Deuterium neutral beams with energies up to 110 keV were injected into TFTR (Tokamak Fusion Test Reactor) plasmas at low magnetic field such that the beam injection velocities were comparable to the Alfven velocity. Excitation of toroidal Alfven eigenmodes was observed by Mirnov coils and beam emission spectroscopy. 10 refs., 4 figs.
Date: March 1, 1991
Creator: Wong, K.L.; Fonck, R.J.; Paul, S.F.; Roberts, D.R.; Fredrickson, E.D.; Nazikian, R. et al.
Partner: UNT Libraries Government Documents Department

The relationship between turbulence measurements and transport in different heating regimes in TFTR

Description: The scaling of broad band density fluctuations in the confinement zone of TFTR measured by microwave scattering, beam emission spectroscopy (BES), and reflectometry show a relationship between these fluctuations and energy transport measured from power balance calculations. In L-mode plasmas scattering and BES indicates that the density fluctuation level, [delta]n[sup 2], in the confinement zone for 0.2 < k[perpendicular]ps < 1.0 depends qualitatively on P[sub aux] and I[sub p] in a way that is consistent with variations in energy transport. Fluctuation levels measured with all systems increase strongly toward the edge in all heating regimes following increases in energy transport coefficients. Measurements using BES have shown that poloidal and radial correlation lengths in the confinement zone of L-mode and supershot plasmas fall in the range of 1 to 2 cm. with a wave structure which has k[sub max] [approx] 1 cm[sup [minus]1] (k[perpendicular]ps [approx] 0.2) in the poloidal direction and k[sub max] approaching zero in the radial direction. A simple estimate of the diffusion coefficient based on a measured radial correlation length and correlation time indicates good agreement with power balance calculations. Similar estimates using reflectometry give radial coherence lengths at 10 to 20 kHz in low density ohmic and supershot plasmas of between I and 2 cm.
Date: January 1, 1992
Creator: Bretz, N.L.; Mazzucato, E.; Nazikian, R.; Paul, S.F.; Hammett, G.; Rewoldt, G. et al.
Partner: UNT Libraries Government Documents Department

Investigation of global Alfven instabilities in TFTR

Description: Toroidal Alfven Eigenmodes (TAE) were excited by the energetic neutral beam ions tangentially injected into TFTR plasmas at low magnetic field such that the injection velocities were comparable to the Alfven speed. The modes were identified by measurements from Mirnov coils and beam emission spectroscopy (BES). TAE modes appear in bursts whose repetition rate increases with beam power. The neutron emission rate exhibits sawtooth-like behavior and the crashes always coincide with TAE bursts. This indicates ejection of fast ions from the plasma until these modes are stabilized. The dynamics of growth and stabilization was investigated at various plasma current and magnetic field. The results indicate that the instability can effectively clamp the number of energetic ions in the plasma. The observed instability threshold is discussed in the light of recent theories. In addition to these TAE modes, intermittent oscillations at three times the fundamental TAE frequency were observed by Mirnov coils, but no corresponding signal was found in BES. It appears that these high frequency oscillations do not have direct effect on the plasma neutron source strength.
Date: January 1, 1992
Creator: Wong, K. L.; Paul, S. F.; Fredrickson, E. D.; Nazikian, R.; Park, H. K.; Bell, M. et al.
Partner: UNT Libraries Government Documents Department

Beta-limiting MHD Instabilities in Improved-performance NSTX Spherical Torus Plasmas

Description: Global magnetohydrodynamic stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized beta and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable beta. As a result of these improvements, peak beta values have reached (not simultaneously) {beta}{sub t} = 35%, {beta}{sub N} = 6.4, &lt;{beta}{sub N}&gt; = 4.5, {beta}{sub N}/l{sub i} = 10, and {beta}{sub P} = 1.4. High {beta}{sub P} operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 second with sustained periods of {beta}{sub N} {approx} 6 above the ideal no-wall limit and near the with-wall limit. Details of the {beta} limit scalings and {beta}-limiting instabilities in various operating regimes are described.
Date: May 29, 2003
Creator: Menard, J.E.; Bell, M.G.; Bell, R.E.; Kaye, E.D. Fredrickson D.A. Gates: S.M.; LeBlanc, B.P.; Maingi, R. et al.
Partner: UNT Libraries Government Documents Department

Investigation of global Alfven instabilities in TFTR

Description: Toroidal Alfven Eigenmodes (TAE) were excited by the energetic neutral beam ions tangentially injected into TFTR plasmas at low magnetic field such that the injection velocities were comparable to the Alfven speed. The modes were identified by measurements from Mirnov coils and beam emission spectroscopy (BES). TAE modes appear in bursts whose repetition rate increases with beam power. The neutron emission rate exhibits sawtooth-like behavior and the crashes always coincide with TAE bursts. This indicates ejection of fast ions from the plasma until these modes are stabilized. The dynamics of growth and stabilization was investigated at various plasma current and magnetic field. The results indicate that the instability can effectively clamp the number of energetic ions in the plasma. The observed instability threshold is discussed in the light of recent theories. In addition to these TAE modes, intermittent oscillations at three times the fundamental TAE frequency were observed by Mirnov coils, but no corresponding signal was found in BES. It appears that these high frequency oscillations do not have direct effect on the plasma neutron source strength.
Date: January 1, 1992
Creator: Wong, K.L.; Paul, S.F.; Fredrickson, E.D.; Nazikian, R.; Park, H.K.; Bell, M. et al.
Partner: UNT Libraries Government Documents Department

Excitation of high frequency pressure driven modes by non-axisymmetric equilibrium at high {beta}{sub pol} in PBX-M

Description: High-frequency pressure-driven modes have been observed in high-poloidal-{beta} discharges in the Princeton Beta Experiment-Modification (PBX-M). These modes are excited in a non-axisymmetric equilibrium characterized by a large, low frequency m{sub 1}=1/n{sub 1}=1 island, and they are capable of expelling fast ions. The modes reside on or very close to the q=1 surface, and have mode numbers with either m{sub h}=n{sub h} or (less probably) m{sub h}/n{sub h}=m{sub h}/(m{sub h}-1), with m{sub h} varying between 3 and 10. Occasionally, these modes are, simultaneously localized in the vicinity of the m{sub 1}=2/n{sub 1}=1 island. The high frequency modes near the q=1 surface also exhibit a ballooning character, being significantly stronger on the large major radius side of the plasma. When a large m{sub 1}=1/n{sub 1}=1 island is present the mode is poloidally localized in the immediate vicinity of the x-point of the island. The modes, which occur exclusively in high-{beta} discharges, appear to be driven by the plasma pressure or pressure gradient. They can thus be a manifestation of either a toroidicity-induced shear Alfven eigenmode (TAE) at q=(2m{sub h}+ 1)/2n{sub h}, a kinetic ballooning mode (KBM), or some other type of pressure-driven mode. Theory predicts that the TAE mode is a gap mode, but the high frequency modes in PBX-M are found exclusively on or in the immediate neighborhood of magnetic surfaces with low rational numbers.
Date: June 1, 1993
Creator: Sesnic, S.; Kaita, R.; Kaye, S.; Okabayashi, M.; Takahashi, H.; Bell, R. E. et al.
Partner: UNT Libraries Government Documents Department

NSTX Plasma Response to Lithium Coated Divertor

Description: NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, <0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.
Date: January 21, 2011
Creator: Kugel, H. W.; Bell, M. G.; Allain, J. P.; Bell, R. E.; Ding, S.; Gerhardt, S. P. et al.
Partner: UNT Libraries Government Documents Department

Limiter H-mode experiments on TFTR

Description: Limiter H-modes with centrally peaked density profiles have been obtained in TFTR using a highly conditioned graphite limiter. The transition to these centrally peaked H-modes takes place from the supershot to the H-mode rather than the usual L- to H-mode transition observed on other tokamaks. Bidirectional beam heating is required to induce the transition. Density peaking factors, n{sub e}(0)/<n{sub e}>, greater than 2.3 are obtained and at the same time the H-mode characteristics are similar to those of limiter H-modes on other tokamaks, while the global confinement, {tau}{sub E}, can be >2.5 times L-mode scaling. The transport analysis of the data shows that transport in these H-modes is similar to that of supershots within the inner 0.6 m core of the plasma, but the stored electron energy (calculated using measured values of T{sub e} and n{sub e}) is higher for the H-mode at the plasma edge. Microwave scattering data for the edge plasma shows broad spectra at k = 5.5 cm{sup {minus}1} which begin at the drop in D{sub {alpha}} radiation and are strongly shifted in the electron diamagnetic drift direction. At the same time, beam emission spectroscopy (BES) shows a coherent mode near the boundary which propagates in the ion direction with m = 15--20 at 20--30 kHz. During the ELM event these apparent rotations cease and Mirnov fluctuations in the frequency range of 50--500 kHz increase in intensity. 16 refs., 8 figs.
Date: May 1, 1991
Creator: Bush, C.E. (Oak Ridge National Lab., TN (USA)); Bretz, N.L.; Fredrickson, E.D.; McGuire, K.M.; Nazikian, R.; Park, H.K. et al.
Partner: UNT Libraries Government Documents Department

Results of NSTX Heating Experiments

Description: The National Spherical Torus Experiment (NSTX) at Princeton is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 5 MW of neutral beam injection, NBI, at an injection energy of 90 keV and up to 6 MW of high harmonic fast wave, HHFW, at 30 MHz. NSTX has achieved beta T of 32%. A variety of MHD phenomena have been observed to limit eta. NSTX has now begun addressing E scaling, eta limits and current drive issues. During the NBI heating experiments, a broad Ti profile with Ti up to 2 keV, Ti &gt; Te and a large toroidal rotation. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, E is up to two times the ITER-89P L-Mode scaling and exceeds the ITER-98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of tE. after the transition in some conditions. During HH FW heating, Te &gt; Ti and Te up to 3.5 keV were observed. Current drive has been studied using coaxial helicity injection (CHI), which has produced 390 kA of toroidal current and HHFW, which has produced H-modes with significant bootstrap current fraction at low Ip, high q and high{sub etap}.
Date: June 18, 2002
Creator: Mueller, D.; Ono, M.; Bell, M.G.; Bell, R.E.; Bitter, M.; Bourdelle, C. et al.
Partner: UNT Libraries Government Documents Department

The National Spherical Torus Experiment (NSTX) Research Program and Progress Towards High Beta, Long Pulse Operating Scenarios

Description: A major research goal of the National Spherical Torus Experiment is establishing long-pulse, high-beta, high-confinement operation and its physics basis. This research has been enabled by facility capabilities developed over the last two years, including neutral-beam (up to 7 MW) and high-harmonic fast-wave heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with &lt;beta {sub T}&gt; up to 35%. Normalized beta values often exceed the no wall limit, and studies suggest that passive wall mode stabilization is enabling this for broad pressure profiles characteristic of H-mode plasmas. The viability of long, high bootstrap-current fraction operations has been established for ELMing H-mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fueling are likely contributing to a reduction in H-mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary-heated plasmas examined thus far. High-harmonic fast-wave (HHFW) effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is by comparing of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. A peak heat flux of 10 MW/m superscript ''2'' has been measured in the H-mode, with large asymmetries in the power deposition being observed between the inner and outer strike points. Noninductive plasma start-up studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current-drive techniques have begun.
Date: October 15, 2002
Creator: Synakowski, E. J.; Bell, M. G.; Bell, R. E.; Bigelow, T.; Bitter, M.; Blanchard, W. 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