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Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

Description: At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700oC. However, at a sufficiently high operating temperature (700 - 1000 oC), tungsten is selfannealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment. The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium. Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands1 and fusion experiments2,3 to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 - 400 oC), it has been now been used as a PFC in several confinement experiments (TFTR, T11- M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.
Date: January 15, 2010
Creator: Majeski, R.
Partner: UNT Libraries Government Documents Department

Direct Electron Heating at Moderate Harmonic Number for Compact Ignition Devices

Description: Direct electron heating of compact ignition devices by radio-frequency power in the 300-400 MHz,range is discussed. The possible advantage of this approach to heating an ignition device, as opposed to resonant heating of an ion population, is the insensitivity to the exact value of the magnitude field. Heating with central power deposition during a toroidal field ramp is therefore possible.
Date: July 1, 1999
Creator: Majeski, R.
Partner: UNT Libraries Government Documents Department

The Thomson Scattering System on the Lithium Tokamak eXperiment (LTX)

Description: The Lithium Tokamak eXperiment (LTX) is a spherical tokamak with R0 = 0.4m, a = 0.26m, BTF ~ 3.4kG, IP ~ 400kA, and pulse length ~ 0.25s. The goal of LTX is to investigate tokamak plasmas that are almost entirely surrounded by a lithium-coated plasma-facing shell conformal to the last closed magnetic flux surface. Based on previous experimental results and simulation, it is expected that the low-recycling liquid lithium surfaces will result in higher temperatures at the plasma edge, flatter overall temperature profiles, centrally-peaked density profiles, and an increased confinement time. To test these predictions, the electron temperature and density profiles in LTX will be measured by a multi-point Thomson scattering system (TVTS). Initially, TS measurements will be made at up to 12 simultaneous points between the plasma center and plasma edge. Later, high resolution edge measurements will be deployed to study the lithium edge physics in greater detail. Technical challenges to implementing the TS system included limited "line of sight" access to the plasma due to the plasma-facing shell and problems associated with the presence of liquid lithium.
Date: July 31, 2008
Creator: T. Strickler, R. Majeski, R. Kaita, B. LeBlanc
Partner: UNT Libraries Government Documents Department

The Impact Of Lithium Wall Coatings On NSTX Discharges And The Engineering Of The Lithium Tokamak eXperiment (LTX)

Description: Recent experiments on the National Spherical Torus eXperiment (NSTX) have shown the benefits of solid lithium coatings on carbon PFC's to diverted plasma performance, in both Land H- mode confinement regimes. Better particle control, with decreased inductive flux consumption, and increased electron temperature, ion temperature, energy confinement time, and DD neutron rate were observed. Successive increases in lithium coverage resulted in the complete suppression of ELM activity in H-mode discharges. A liquid lithium divertor (LLD), which will employ the porous molybdenum surface developed for the LTX shell, is being installed on NSTX for the 2010 run period, and will provide comparisons between liquid walls in the Lithium Tokamak eXperiment (LTX) and liquid divertor targets in NSTX. LTX, which recently began operations at the Princeton Plasma Physics Laboratory, is the world's first confinement experiment with full liquid metal plasma-facing components (PFCs). All materials and construction techniques in LTX are compatible with liquid lithium. LTX employs an inner, heated, stainless steel-faced liner or shell, which will be lithium-coated. In order to ensure that lithium adheres to the shell, it is designed to operate at up to 500 - 600 oC to promote wetting of the stainless by the lithium, providing the first hot wall in a tokamak to operate at reactor-relevant temperatures. The engineering of LTX will be discussed.
Date: March 18, 2010
Creator: Majeski, R.; Kugel, H. & Kaita, R.
Partner: UNT Libraries Government Documents Department

Evolution of toroidal Alfven eigenmode instability in TFTR

Description: The nonlinear behavior of the Toroidal Alfven Eigenmode (TAE) driven unstable by energetic ions in TFTR is studied. The evolution of instabilities can take on several scenarios: a single mode or several modes can be driven unstable at the same time, the spectrum can be steady or pulsating and there can be negligible or anomalous loss associated with the instability. This paper presents a comparison between experimental results and recently developed nonlinear theory. The authors find many features observed in experiment are compatible with the consequences of the nonlinear theory. Examples include the structure of the saturated pulse that emerges from the onset of instability of a single mode and the decrease but persistence of TAE signals when the applied rf power is reduced or shut off.
Date: July 1, 1996
Creator: Wong, K. L.; Majeski, R. & Petrov, M.
Partner: UNT Libraries Government Documents Department

Mode conversion studies in TFTR

Description: Mode converted Ion Bernstein Waves (IBW) have important potential applications in tokamak reactors. These applications include on or off axis electron heating and current drive and the channeling of alpha particle power for both current drive and increased reactivity. Efficient mode conversion electron heating with a low field side antenna, with both on and off axis power deposition, has been demonstrated for the first time in TFTR in D{sup 3}He-{sup 4}He plasmas. Up to 80% of the Ion Cyclotron Range of Frequency (ICRF) power is coupled to electrons at the mode conversion surface. Experiments during deuterium and tritium neutral beam injection (NBI) indicate that good mode conversion efficiency can be maintained during NBI if sufficient {sup 3}He is present. No evidence of strong alpha particle heating by the IBW is seen. Recent modeling indicates that if the mode converted IBW is preferentially excited off the horizontal midplane then the resultant high poloidal mode number wave may channel alpha particle power to either electrons or ions. In TFTR both the propagation of the IBW and its effect on the alpha particle population is being investigated. Experiments with 2 MW of ICRF power launched with {+-} 90{degree} antenna phasing for current drive show that electron heating and sawtooth activity depend strongly on the direction of the launched wave. The noninductively driven current could not be experimentally determined in these relatively high plasma current, short pulse discharges. Experiments at higher RF power and lower plasma current are planned to determine on and off axis current drive efficiency.
Date: March 1, 1995
Creator: Majeski, R.; Fisch, N.J. & Adler, H.
Partner: UNT Libraries Government Documents Department

Recent results from the TFTR ICRF DT Program

Description: The first experiments to be performed with ICRF heating of DT plasmas are reported. ICRF heating of minority ions, tritium (second harmonic resonance), as well as direct electron heating are being performed during the DT phase of TFTR. RF power modulation and Fourier transform techniques are used to attempt to elucidate the competition between tritium second harmonic, direct electron, and {sup 3}He fundamental heating in DT plasmas. A significant fraction of the RF power has been found to couple to the tritium ions via second harmonic heating. Relevant RF coupling physics is investigated using {sup 3}He minority heating (43 MHz), H minority heating (64 MHz), and mode conversion (43 MHz, comparable densities of {sup 3}He and {sup 4}He) at a toroidal field of 4.5T.
Date: March 1, 1995
Creator: Rogers, J.H.; Darrow, D. & Majeski, R.
Partner: UNT Libraries Government Documents Department

Mode Conversion Heating Scenarios for the National Compact Stellarator Experiment

Description: Radio-frequency heating scenarios for the National Compact Stellarator eXperiment (NCSX) are considered. The focus here is on mode conversion from the fast to the slow ion Bernstein wave as either an electron or ''bulk'' ion heating technique, using a high-field side launch to directly access the ion-ion hybrid layer. Modeling for the planned parameters of NCSX [R(subscript ave) {approximately} 1.4 m, a(subscript ave) {approximately} 0.4 m, B(subscript T)(0) {approximately} 1.2-2 T, n(subscript e)(0) {approximately} 2-5 x 10(superscript19) m(superscript -3), T(subscript e)(0) {approximately} T(subscript i)(0) {approximately} 1-2 keV] for mode conversion in D-H and D-3He plasmas is presented. Possible types of high-field side antennas are also briefly discussed.
Date: May 18, 2001
Creator: Majeski, R.; Wilson, J.R. & and Zarnstorff, M.
Partner: UNT Libraries Government Documents Department

An Inexpensive Ohmic Transformer Firing Circuit for the CDX-U Spherical Torus

Description: We have designed and modeled a simple, efficient circuit for delivering power to the CDX-U ohmic transformer solenoid. Inexpensive electrolytic capacitors are used to provide the bulk of the stored energy. One small high-voltage oil-filled capacitor bank is used in the ignitron-based circuit. Several design objectives are met, including the production of a solenoid current waveform well suited to the breakdown and ohmic current-drive of a tokamak plasma, making efficient use of the available loop volt-seconds. The electrolytic capacitors are protected from reverse-bias conditions, and the ohmic solenoid is protected from voltages above 1 kV, well within the voltage rating, under normal operation and any forseeable fault conditions.
Date: October 1, 1999
Creator: Majeski, R. & Munsat, T.
Partner: UNT Libraries Government Documents Department

Alfvenic behavior of alpha particle driven ion cyclotron emission in TFTR

Description: Ion cyclotron emission (ICE) has been observed during D-T discharges in the Tokamak Fusion Test Reactor (TFTR), using rf probes located near the top and bottom of the vacuum vessel. Harmonics of the alpha cyclotron frequency ({Omega}{sub {alpha}}) evaluated at the outer midplane plasma edge are observed at the onset of the beam injection phase of TFTR supershots, and persist for approximately 100-250 ms. These results are in contrast with observations of ICE in JET, in which harmonics of {Omega}{sub {alpha}} evolve with the alpha population in the plasma edge. Such differences are believed to be due to the fact that newly-born fusion alpha particles are super-Alfvenic near the edge of JET plasmas, while they are sub-Alfvenic near the edge of TFTR supershot plasmas. In TFTR discharges with edge densities such that newly-born alpha particles are super-Alfvenic, alpha cyclotron harmonics are observed to persist. These results are in qualitative agreement with numerical calculations of growth rates due to the magnetoacoustic cyclotron instability.
Date: July 1, 1995
Creator: Cauffman, S.; Majeski, R. & McClements, K.G.
Partner: UNT Libraries Government Documents Department

NCSX Plasma Heating Methods

Description: The NCSX (National Compact Stellarator Experiment) has been designed to accommodate a variety of heating systems, including ohmic heating, neutral-beam injection, and radio-frequency. Neutral beams will provide one of the primary heating methods for NCSX. In addition to plasma heating, beams are also expected to provide a means for external control over the level of toroidal plasma rotation velocity and its profile. The plan is to provide 3 MW of 50 keV balanced neutral-beam tangential injection with pulse lengths of 500 msec for initial experiments, and to be upgradeable to pulse lengths of 1.5 sec. Subsequent upgrades will add 3 MW of neutral-beam injection. This Chapter discusses the NCSX neutral-beam injection requirements and design issues, and shows how these are provided by the candidate PBX-M (Princeton Beta Experiment-Modification) neutral-beam injection system. In addition, estimations are given for beam-heating efficiencies, scaling of heating efficiency with machine size an d magnetic field level, parameter studies of the optimum beam-injection tangency radius and toroidal injection location, and loss patterns of beam ions on the vacuum chamber wall to assist placement of wall armor and for minimizing the generation of impurities by the energetic beam ions. Finally, subsequent upgrades could add an additional 6 MW of radio-frequency heating by mode-conversion ion-Bernstein wave (MCIBW) heating, and if desired as possible future upgrades, the design also will accommodate high-harmonic fast-wave and electron-cyclotron heating. The initial MCIBW heating technique and the design of the radio-frequency system lend themselves to current drive, so that if current drive became desirable for any reason only minor modifications to the heating system described here would be needed. The radio-frequency system will also be capable of localized ion heating (bulk or tail), and possibly ion-Bernstein-wave-generated sheared flows.
Date: February 28, 2003
Creator: Kugel, H.W.; Spong, D.; Majeski, R. & Zarnstorff, M.
Partner: UNT Libraries Government Documents Department

Magnetic Diagnostics for Equilibrium Reconstructions in the Presence of Nonaxisymmetric Eddy Current Distributions in Tokamaks

Description: The lithium tokamak experiment #2;LTX#3; is a modest-sized spherical tokamak #2;R0=0.4 m and a =0.26 m#3; designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 oC. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.
Date: December 10, 2010
Creator: Kaita, R.; Kozub, T.; Logan, N.; Majeski, R.; Menard, J. & Zakharov, L.
Partner: UNT Libraries Government Documents Department

Measuring the Density of a Molecular Cluster Injector via Visible Emission from an Electron Beam

Description: A method to measure the density distribution of a dense hydrogen gas jet is pre- sented. A Mach 5.5 nozzle is cooled to 80K to form a flow capable of molecular cluster formation. A 250V, 10mA electron beam collides with the jet and produces Hα emission that is viewed by a fast camera. The high density of the jet, several 10<sup>16</sup>cm<sup>-3</sup>, results in substantial electron depletion, which attenuates the H<sub>α</sub> emission. The attenuated emission measurement, combined with a simplified electron-molecule collision model, allows us to determine the molecular density profile via a simple iterative calculation.
Date: June 28, 2010
Creator: Lundberg, D. P.; Kaita, R.; Majeski, R. M. & Stotler, D. P.
Partner: UNT Libraries Government Documents Department

Liquid Lithium Wall Experiments in CDX-U

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. Sputtering and erosion tests are currently underway in the PISCES device at the University of California at San Diego (UCSD). To complement this effort, plasma interaction questions in a toroidal plasma geometry will be addressed by a proposed new groundbreaking experiment in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The CDX-U plasma is intensely heated and well diagnosed, and an extensive liquid lithium plasma-facing surface will be used for the first time with a toroidal plasma. Since CDX-U is a small ST, only approximately1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment.
Date: October 1, 1999
Creator: Doerner, R.; Kaita, R.; Majeski, R.; Luckhardt, S. & al, et
Partner: UNT Libraries Government Documents Department

High-harmonic fast wave heating experiments in CDX-U

Description: One of the primary objectives of the proposed National Spherical Tokamak Experiment (NSTX) is the investigation of very high {beta} regimes. Consequently, finding efficient methods of non-inductive heating and current drive required to heat and sustain such plasmas is of considerable importance. High-frequency fast waves are a promising candidate in this regard. However, in NSTX, the field-line pitch at the outer midplane will range from 0 up to 60 degrees from plasma start-up to current flattop. Thus, antenna strap orientation with respect to the edge magnetic field may have a serious impact on power coupling and absorption. To address this issue, the vacuum vessel of the Current Drive Experiment -- Upgrade (CDX-U) spherical tokamak has been upgraded to accommodate a rotatable two-strap antenna capable of handling several hundred kilowatts in short pulses. Details of the antenna design and results from loading measurements made as a function of power, strap angle, and strap phasing will be presented. Results from microwave scattering experiments will also be discussed.
Date: December 1, 1997
Creator: Menard, J.; Majeski, R.; Ono, M.; Wilson, J.R.; Munsat, T. & Seki, T.
Partner: UNT Libraries Government Documents Department

New Electron Cyclotron Emission Diagnostic Based Upon the Electron Bernstein Wave

Description: Most magnetically confined plasma devices cannot take advantage of standard Electron Cyclotron Emission (ECE) diagnostics to measure temperature. They either operate at high density relative to their magnetic field or they do not have sufficient density and temperature to reach the blackbody condition. The standard ECE technique measures the electromagnetic waves emanating from the plasma. Here we propose to measure electron Bernstein waves (EBW) to ascertain the local electron temperature in these plasmas. The optical thickness of EBW is extremely high because it is an electrostatic wave with a large k(subscript i). One can reach the blackbody condition with a plasma density approximately equal to 10(superscript 11) cm(superscript -3) and electron temperature approximately equal to 1 eV. This makes it attractive to most plasma devices. One serious issue with using EBW is the wave accessibility. EBW may be accessible by either direct coupling or mode conversion through an extremely narrow layer (approximately 1-2 mm) in low field devices.
Date: May 1, 1999
Creator: Efthimion, P.C.; Hosea, J.C.; Kaita, R.; Majeski, R. & Taylor, G.
Partner: UNT Libraries Government Documents Department

Liquid Lithium Experiments in CDX-U

Description: The initial results of experiments involving the use of liquid lithium as a plasma facing component in the Current Drive Experiment-Upgrade (CDX-U) are reported. Studies of the interaction of a steady-state plasma with liquid lithium in the Plasma Interaction with Surface and Components Experimental Simulator (PISCES-B) are also summarized. In CDX-U a solid or liquid lithium covered rail limiter was introduced as the primary limiting surface for spherical torus discharges. Deuterium recycling was observed to be reduced, but so far not eliminated, for glow discharge-cleaned lithium surfaces. Some lithium influx was observed during tokamak operation. The PISCES-B results indicate that the rates of plasma erosion of lithium can exceed predictions by an order of magnitude at elevated temperatures. Plans to extend the CDX-U experiments to large area liquid lithium toroidal belt limiters are also described.
Date: November 15, 2000
Creator: Majeski, R.; Doerner, R.; Kaita, R.; Antar, G.; Timberlake, J. & al, et
Partner: UNT Libraries Government Documents Department

Microscopic Motion of Liquid Metal Plasma Facing Components In A Diverted Plasma

Description: Liquid metal plasma facing components (PFCs) have been identified as an alternative material for fusion plasma experiments. The use of a liquid conductor where significant magnetic fields are present is considered risky, with the possibility of macroscopic fluid motion and possible ejection into the plasma core. Analysis is carried out on thermoelectric magnetohydrodynamic (TEMHD) forces caused by temperature gradients in the liquid-container system itself in addition to scrape-off-layer currents interacting with the PFC from a diverted plasma. Capillary effects at the liquid-container interface will be examined which govern droplet ejection criteria. Stability of the interface is determined using linear stability methods. In addition to application to liquidmetal PFCs, thin film liquidmetal effects have application to current and future devices where off-normal events may liquefy portions of the first wall and other plasma facing components.
Date: September 22, 2010
Creator: Jaworski, M. A.; Morley, N. B.; Abrams, T; Kaita, R; Kallman, J; Kugel, H et al.
Partner: UNT Libraries Government Documents Department

Majority ion heating near the ion-ion hybrid layer in tokamaks

Description: Efficient direct majority ion heating in a deuterium-tritium (D-T) reactor-grade plasma via absorption of fast magnetosonic waves in the ion cyclotron range of frequencies (ICRF) is discussed. Majority ion heating results from resonance overlap between the cyclotron layers and the D-T ion-ion hybrid layer in hot, dense plasmas for fast waves launched with high parallel wavenumbers. Analytic and numerical models are used to explore the regime in ITER plasmas.
Date: August 1, 1995
Creator: Phillips, C.K.; Hosea, J.C.; Ignat, D.; Majeski, R.; Rogers, J.H.; Schilling, G. et al.
Partner: UNT Libraries Government Documents Department

Radio frequency current drive for small aspect ratio tori

Description: Non-inductive current drive (CD) is required during plasma initiation and for current sustainment in NSTX. The physics of high harmonic fast waves (HHFW) and the design of an antenna system for NSTX are studied. It is found that the theoretical current drive efficiency for HHFW can be high, and a general survey of parameters gives a good target for the antenna design. The primary issue for HHFW during plasma initiation is loading since the CD efficiency is very high for low density plasmas. For high beta operation at full current launching in the usual manner from the equatorial plane may lead to marginal CD performance. However, advanced antenna designs exploiting the theoretical results show some promise for high beta operation. Two methods to optimize the CD efficiency have been explored. The first non-zero poloidal mode excitation, provides enhanced efficiency because of improved penetration and a reduction of detrimental trapped particle effects. A second, placement of the antenna away from the equatorial plane, can also be used to reduce trapped particle effects. These methods can be used separately or together, yielding potential improvements of more than a factor of 2 in CD efficiency for NSTX.
Date: December 31, 1994
Creator: Carter, M.D.; Jaeger, E.F.; Batchelor, D.B.; Strickler, D.J. & Majeski, R.
Partner: UNT Libraries Government Documents Department

Transient Transport Experiments in the CDX-U Spherical Torus

Description: Electron transport has been measured in the Current Drive Experiment-Upgrade (CDX-U) using two separate perturbative techniques. Gas modulation at the plasma edge was used to introduce cold-pulses which propagate towards the plasma center, providing time-of-flight information leading to a determination of chi(subscript e) as a function of radius. Sawteeth at the q=1 radius (r/a {approx} 0.15) induced heat-pulses which propagated outward towards the plasma edge, providing a complementary time-of-flight based chi(subscript e) profile measurement. This work represents the first localized measurement of chi(subscript e) in a spherical torus. It is found that chi(subscript e) = 1-2 meters squared per second in the plasma core (r/a &lt; 1/3), increasing by an order of magnitude or more outside of this region. Furthermore, the chi(subscript e) profile exhibits a sharp transition near r/a = 1/3. Spectral and profile analyses of the soft X-rays, scanning interferometer, and edge probe data show no evidence of a significant magnetic island causing the high chi(subscript e) region.
Date: June 12, 2001
Creator: Munsat, T.; Efthimion, P.C.; Jones, B.; Kaita, R.; Majeski, R.; Stutman, D. et al.
Partner: UNT Libraries Government Documents Department

The Development of RF Heating of Magnetically Confined Deuterium-Tritium Plasmas

Description: The experimental and theoretical development of ion cyclotron radiofrequency heating (ICRF) in toroidal magnetically-confined plasmas recently culminated with the demonstration of ICRF heating of D-T plasmas, first in the Tokamak Fusion Test Reactor (TFTR) and then in the Joint European Torus (JET). Various heating schemes based on the cyclotron resonances between the plasma ions and the applied ICRF waves have been used, including second harmonic tritium, minority deuterium, minority helium-3, mode conversion at the D-T ion-ion hybrid layer, and ion Bernstein wave heating. Second harmonic tritium heating was first shown to be effective in a reactor-grade plasma in TFTR. D-minority heating on JET has led to the achievement of Q = 0.22, the ratio of fusion power produced to RF power input, sustained over a few energy confinement times. In this paper, some of the key building blocks in the development of rf heating of plasmas are reviewed and prospects for the development of advanced methods of plasma control based on the application of rf waves are discussed.
Date: June 1, 1999
Creator: LeBlanc, B.P.; Phillips, C.K.; Hosea, J.C.; Majeski, R. & others], S. Bernabei
Partner: UNT Libraries Government Documents Department

Electron Bernstein wave electron temperature profile diagnostic

Description: Electron cyclotron emission (ECE) has been employed as a standard electron temperature profile diagnostic on many tokamaks and stellarators, but most magnetically confined plasma devices cannot take advantage of standard ECE diagnostics to measure temperature. They are either overdense, operating at high density relative to the magnetic field (e.g. where the plasma frequency is much greater than the electron cyclotron frequency, as in a spherical torus) or they have insufficient density and temperature to reach the blackbody condition. Electron Bernstein waves (EBWs) are electrostatic waves that can propagate in overdense plasmas and have a high optical thickness at the electron cyclotron resonance layers, as a result of their large perpendicular wavenumber. This paper reports on measurements of EBW emission on the CDX-U spherical torus, where B{sub o} {approximately} 2 kG, &lt;n{sub e}&gt; {approximately}10{sup 13} cm{sup {minus}3} and T{sub e} {approx} to 10 -- 200 eV. Results are presented for electromagnetic measurements of EBW emission, mode-converted near the plasma edge. The EBW emission was absolutely calibrated and compared to the electron temperature profile measured by a multi-point Thomson scattering diagnostic. Depending on the plasma conditions, the mode converted EBW radiation temperature was found to be less than or equal to T{sub e} and the emission source was determined to be radially localized at the electron cyclotron resonance layer. A Langmuir triple probe and a 140 GHz interferometer were employed to measure changes in edge density profile in the vicinity of the upper hybrid resonance, where the mode conversion of the EBWs is expected to occur. Initial results suggest EBW emission and EBW heating are viable concepts for overdense plasmas.
Date: July 20, 2000
Creator: Taylor, G.; Efthimion, P.; Jones, B.; Munsat, T.; Spaleta, J.; Hosea, J. et al.
Partner: UNT Libraries Government Documents Department