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Response of NSTX Liquid Lithium divertor to High Heat Loads

Description: Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ~1.5 MW/m2 for 1-3 seconds. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the "bare" sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface. __________________________________________________
Date: July 18, 2012
Creator: Abrams, Tyler; Kallman, J; Kaitaa, R; Foley, E L; Grayd, T K; Kugel, H et al.
Partner: UNT Libraries Government Documents Department

Upward-facing Lithium Flash Evaporator for NSTX-U

Description: NSTX plasma performance has been significantly enhanced by lithium conditioning [1]. To date, the lower divertor and passive plates have been conditioned by downward facing lithium evaporators (LITER) as appropriate for lower null plasmas. The higher power operation expected from NSTX-U requires double null plasma operation in order to distribute the heat flux between the upper and lower divertors making it desirable to coat the upper divertor region with Li as well. An upward aiming LITER (U-LITER) is presently under development and will be inserted into NSTX-U using a horizontal probe drive located in a 6" upper midplane port. In the retracted position the evaporator will be loaded with up to 300 mg of Li granules utilizing one of the calibrated NSTX Li powder droppers[2]. The evaporator will then be inserted into the vessel in a location within the shadow of the RF limiters and will remain in the vessel during the discharge. About 10 seconds before a discharge, it will be rapidly heated and the lithium completely evaporated onto the upper divertor, thus avoiding the complication of a shutter that prevents evaporation during the shot when the diagnostic shutters are open. The minimal time interval between the evaporation and the start of the discharge will avoid the passivation of the lithium by residual gases and enable the study of the conditioning effects of un-passivated Li surfaces [3]. Two methods are being investigated to accomplish the rapid (few second) heating of the lithium. A resistive method relies on passing a large current through a Li filled crucible. A second method requires using a 3 kW e-beam gun to heat the Li. In this paper the evaporator systems will be described and the pros and cons of each heating method will be discussed.
Date: July 9, 2013
Creator: Roquemore, A. L.
Partner: UNT Libraries Government Documents Department

Recent Progress in the NSTX/NSTX-U Lithium Program and Prospects for Reactor-Relevant Liquid-Lithium Based Divertor Development

Description: Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. While tungsten has been identified as the most attractive solid divertor material, the NSTX/NSTX-U lithium (Li) program is investigating the viability of liquid lithium (LL) as a potential reactor compatible divertor plasma facing component (PFC) . In the near term, operation in NSTX-U is projected to provide reactor-like divertor heat loads < 40 MW/m^2 for 5 s. During the most recent NSTX campaign, ~ 0.85 kg of Li was evaporated onto the NSTX PFCs where a ~50% reduction in heat load on the Liquid Lithium Divertor (LLD) was observed, attributable to enhanced divertor bolometric radiation. This reduced divertor heat flux through radiation observed in the NSTX LLD experiment is consistent with the results from other lithium experiments and calculations. These results motivate an LL-based closed radiative divertor concept proposed here for NSTX-U and fusion reactors. With an LL coating, the Li is evaporated from the divertor strike point surface due to the intense heat. The evaporated Li is readily ionized by the plasma due to its low ionization energies, and the ionized Li ions can radiate strongly, resulting in a significant reduction in the divertor heat flux. Due to the rapid plasma transport in divertor plasma, the radiation values can be significantly enhanced up to ~ 11 MJ/cc of LL. This radiative process has the desired function of spreading the focused divertor heat load to the entire divertor chamber facilitating the divertor heat removal. The LL divertor surface can also provide a "sacrificial" surface to protect the substrate solid material from transient high heat flux such as the ones caused by the ELMs. The closed radiative LLD concept has the advantages of providing some degree of partition in terms of plasma disruption ...
Date: October 27, 2012
Creator: M. Ono, et al.
Partner: UNT Libraries Government Documents Department

Comparison of H-Mode Plasmas Diverted to Solid and Liquid Lithium Surfaces

Description: Experiments were conducted with a Liquid Lithium Divertor (LLD) in NSTX. Among the goals was to use lithium recoating to sustain deuterium (D) retention by a static liquid lithium surface, approximating the ability of flowing liquid lithium to maintain chemical reactivity. Lithium evaporators were used to deposit lithium on the LLD surface. Improvements in plasma edge conditions were similar to those with lithiated graphite plasma-facing components (PFCs), including an increase in confinement over discharges without lithiumcoated PFCs and ELM reduction during H-modes. With the outer strike point on the LLD, the D retention in the LLD was about the same as that for solid lithium coatings on graphite, or about two times that achieved without lithium PFC coatings. There were also indications of contamination of the LLD surface, possibly due erosion and redeposition of carbon from PFCs. Flowing lithium may thus be needed for chemically active PFCs during long-pulse operation.
Date: July 20, 2012
Creator: R. Kaita, et. al.
Partner: UNT Libraries Government Documents Department

Rapidly Moving Divertor Plates In A Tokamak

Description: It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.
Date: May 16, 2011
Creator: Zweben, S.
Partner: UNT Libraries Government Documents Department

Plasma Facing Surface Composition During NSTX Li Experiments

Description: Lithium conditioned plasma facing surfaces have lowered recycling and enhanced plasma performance on many fusion devices. However, the nature of the plasma-lithium surface interaction has been obscured by the difficulty of in-tokamak surface analysis. We report laboratory studies of the chemical composition of lithium surfaces exposed to typical residual gases found in tokamaks. Solid lithium and a molybdenum alloy (TZM) coated with lithium has been examined using x-ray photoelectron spectroscopy, temperature programmed desorption, and Auger electron spectroscopy both in ultrahigh vacuum conditions and after exposure to trace gases. Lithium surfaces near room temperature were oxidized after exposure to 1-2 Langmuirs of oxygen or water vapor. The oxidation rate by carbon monoxide was four times less. Lithiated PFC surfaces in tokamaks will be oxidized in about 100 s depending on the tokamak vacuum conditions.
Date: July 20, 2012
Creator: Skinner, C. H.; Sullenberger, R.; Koel, B. E.; Jaworski, M. A. & Kugel, H. W.
Partner: UNT Libraries Government Documents Department

Mini-Conference on the First Microns of the First Wall

Description: Interactions between plasmas and their surrounding materials (plasma facing components) are of great interest to present and future magnetic fusion experiments, and ITER [ITER Physics Basis Editors, ITER Physics Exper Group Chairs, ITER Joint Central Team, and Physics Inte gration Unit, Nucl. Fusion 39, 2137 (1999)] in particular. This interest is the result of concerns with the survivability of these materials, as well as the impact of these interactions back on the plasma. These interactions begin on the surface, but can have consequences a few microns into the material.This mini-conference on these "first microns" was designed to bring to the Division of Plasma Physics Meeting experts on these topics who would otherwise not attend. At the same time, the mini-conference was intended to expose the broader fusion community to these issues. The mini-conference covered in three, half-day sessions the topics of lithium coatings and surfaces, mixed materials characteristics, and issues associated with graphite.
Date: March 20, 2008
Creator: D.P. Stotler, T.D. Rognlien and S.I. Krasheninnikov
Partner: UNT Libraries Government Documents Department

Tritium Removal from Carbon Plasma Facing Components

Description: Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating.
Date: November 24, 2003
Creator: Skinner, C.H.; Coad, J.P. & Federici, G.
Partner: UNT Libraries Government Documents Department

Tritiated Dust Levitation by Beta Induced Static Charge

Description: Tritiated particles have been observed to spontaneously levitate under the influence of a static electric field. Tritium containing co-deposits were mechanically scraped from tiles that had been used in the Tokamak Fusion Test Reactor (TFTR) inner limiter during the deuterium-tritium campaign and were placed in a glass vial. On rubbing the plastic cap of the vial a remarkable ''fountain'' of particles was seen inside the vial. Particles from an unused tile or from a TFTR co-deposit formed during deuterium discharges did not exhibit this phenomenon. It appears that tritiated particles are more mobile than other particles and this should be considered in assessing tokamak accident scenarios and in occupational safety.
Date: June 4, 2003
Creator: Skinner, C.H.; Gentile, C.A.; Ciebiera, L. & Langish, S.
Partner: UNT Libraries Government Documents Department

Is Carbon a Realistic Choice for ITER's Divertor?

Description: Tritium retention by co-deposition with carbon on the divertor target plate is predicted to limit ITER's DT burning plasma operations (e.g. to about 100 pulses for the worst conditions) before the in-vessel tritium inventory limit, currently set at 350 g, is reached. At this point, ITER will only be able to continue its burning plasma program if technology is available that is capable of rapidly removing large quantities of tritium from the vessel with over 90% efficiency. The removal rate required is four orders of magnitude faster than that demonstrated in current tokamaks. Eighteen years after the observation of co-deposition on JET and TFTR, such technology is nowhere in sight. The inexorable conclusion is that either a major initiative in tritium removal should be funded or that research priorities for ITER should focus on metal alternatives.
Date: May 13, 2005
Creator: Skinner, C.H. & Federici, G.
Partner: UNT Libraries Government Documents Department

Lithium Wall Conditioning And Surface Dust Detection On NSTX

Description: Lithium evaporation onto NSTX plasma facing components (PFC) has resulted in improved energy confinement, and reductions in the number and amplitude of edge-localized modes (ELMs) up to the point of complete ELM suppression. The associated PFC surface chemistry has been investigated with a novel plasma material interface probe connected to an in-vacuo surface analysis station. Analysis has demonstrated that binding of D atoms to the polycrystalline graphite material of the PFCs is fundamentally changed by lithium - in particular deuterium atoms become weakly bonded near lithium atoms themselves bound to either oxygen or the carbon from the underlying material. Surface dust inside NSTX has been detected in real-time using a highly sensitive electrostatic dust detector. In a separate experiment, electrostatic removal of dust via three concentric spiral-shaped electrodes covered by a dielectric and driven by a high voltage 3-phase waveform was evaluated for potential application to fusion reactors
Date: May 23, 2011
Creator: Skinner, C. H.; Bell, M. G.; Friesen, F. Q. L.; Heim, B.; Jaworski, M. A.; Kugel, H. et al.
Partner: UNT Libraries Government Documents Department

NSTX Report on FES Joint Facilities Research Milestone 2010

Description: Annual Target: Conduct experiments on major fusion facilities to improve understanding of the heat transport in the tokamak scrape-off layer (SOL) plasma, strengthening the basis for projecting divertor conditions in ITER. The divertor heat flux profiles and plasma characteristics in the tokamak scrape-off layer will be measured in multiple devices to investigate the underlying thermal transport processes. The unique characteristics of C-Mod, DIII-D, and NSTX will enable collection of data over a broad range of SOL and divertor parameters (e.g., collisionality ν*, beta β, parallel heat flux q||, and divertor geometry). Coordinated experiments using common analysis methods will generate a data set that will be compared with theory and simulation.
Date: March 24, 2011
Creator: Maingi, R.; Ahn, J.-W.; Gray, T. K.; McLean, A. G. & Soukhanovskii, V. A.
Partner: UNT Libraries Government Documents Department

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

He Puff System For Dust Detector Upgrade

Description: Local detection of surface dust is needed for the safe operation of next-step magnetic fusion devices such as ITER. An electrostatic dust detector, based on a 5 cm x 5 cm grid of interlocking circuit traces biased to 50 V, has been developed to detect dust on remote surfaces and was successfully tested for the first time on the National Spherical Torus Experiment (NSTX). We report on a helium puff system that clears residual dust from this detector and any incident debris or fibers that might cause a permanent short circuit. The entire surface of the detector was cleared of carbon particles by two consecutive helium puffs delivered by three nozzles of 0.45 mm inside diameter. The optimal configuration was found to be with the nozzles at an angle of 30o with respect to the surface of the detector and a helium backing pressure of 6 bar. __________________________________________________
Date: October 1, 2010
Creator: Rais, B.; Skinner, C. H. & Roquemore, A. L.
Partner: UNT Libraries Government Documents Department

A Simple Apparatus for the Injection of Lithium Aerosol into the Scrape-Off Layer of Fusion Research Devices

Description: A simple device has been developed to deposit elemental lithium onto plasma facing components in the National Spherical Torus Experiment. Deposition is accomplished by dropping lithium powder into the plasma column. Once introduced, lithium particles quickly become entrained in scrape-off layer flow as an evaporating aerosol. Particles are delivered through a small central aperture in a computer-controlled resonating piezoelectric disk on which the powder is supported. The device has been used to deposit lithium both during discharges as well as prior to plasma breakdown. Clear improvements to plasma performance have been demonstrated. The use of this apparatus provides flexibility in the amount and timing of lithium deposition and, therefore, may benefit future fusion research devices.
Date: October 11, 2010
Creator: D. K. Mansfield, A.L Roquemore, H. Schneider, J. Timberlake, H. Kugel, M.G. Bell and the NSTX Research Team
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

Deposition Diagnostics for Next-step Devices

Description: The scale-up of deposition in next-step devices such as ITER will pose new diagnostic challenges. Codeposition of hydrogen with carbon needs to be characterized and understood in the initial hydrogen phase in order to mitigate tritium retention and qualify carbon plasma facing components for DT operations. Plasma facing diagnostic mirrors will experience deposition that is expected to rapidly degrade their reflectivity, posing a new challenge to diagnostic design. Some eroded particles will collect as dust on interior surfaces and the quantity of dust will be strictly regulated for safety reasons - however diagnostics of in-vessel dust are lacking. We report results from two diagnostics that relate to these issues. Measurements of deposition on NSTX with 4 Hz time resolution have been made using a quartz microbalance in a configuration that mimics that of a typical diagnostic mirror. Often deposition was observed immediately following the discharge suggesting that diagnostic shutters should be closed as soon as possible after the time period of interest. Material loss was observed following a few discharges. A novel diagnostic to detect surface particles on remote surfaces was commissioned on NSTX.
Date: June 15, 2004
Creator: Skinner, C.H.; Roquemore, A.L.; team, the NSTX; Bader, A. & Wampler, W.R.
Partner: UNT Libraries Government Documents Department

Operation of Ferroelectric Plasma Sources in a Gas Discharge Mode

Description: Ferroelectric plasma sources in vacuum are known as sources of ablative plasma, formed due to surface discharge. In this paper, observations of a gas discharge mode of operation of the ferroelectric plasma sources (FPS) are reported. The gas discharge appears at pressures between approximately 20 and approximately 80 Torr. At pressures of 1-20 Torr, there is a transition from vacuum surface discharge to the gas discharge, when both modes coexist and the surface discharges sustain the gas discharge. At pressures between 20 and 80 Torr, the surface discharges are suppressed, and FPS operate in pure gas discharge mode, with the formation of almost uniform plasma along the entire surface of the ceramics between strips. The density of the expanding plasma is estimated to be about 1013 cm-3 at a distance of 5.5 mm from the surface. The power consumption of the discharge is comparatively low, making it useful for various applications. This paper also presents direct measurements of the yield of secondary electron emission from ferroelectric ceramics, which, at low energies of primary electrons, is high and dependent on the polarization of the ferroelectric material
Date: March 8, 2004
Creator: Dunaevsky, A. & Fisch, N.J.
Partner: UNT Libraries Government Documents Department

How to Patch Active Plasma and Collisionless Sheath: Practical Guide

Description: Most plasmas have a very thin sheath compared with the plasma dimension. This necessitates separate calculations of the plasma and sheath. The Bohm criterion provides the boundary condition for calculation of plasma profiles. To calculate sheath properties, a value of electric field at the plasma-sheath interface has to be specified in addition to the Bohm criterion. The value of the boundary electric field and robust procedure to approximately patch plasma and collisionless sheath with a very good accuracy are reported.
Date: August 22, 2002
Creator: Kaganovich, Igor D.
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

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

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

Electron-wall Interaction in Hall Thrusters

Description: Electron-wall interaction effects in Hall thrusters are studied through measurements of the plasma response to variations of the thruster channel width and the discharge voltage. The discharge voltage threshold is shown to separate two thruster regimes. Below this threshold, the electron energy gain is constant in the acceleration region and therefore, secondary electron emission (SEE) from the channel walls is insufficient to enhance electron energy losses at the channel walls. Above this voltage threshold, the maximum electron temperature saturates.
Date: February 11, 2005
Creator: Raitses, Y.; Staack, D.; Keidar, M. & Fisch, N.J.
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