39 Matching Results

Search Results

Advanced search parameters have been applied.
in subjects: "Plasma-Wall Interaction"  
Results: 1 - 24 next
open access

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 develo… more
Date: July 18, 2012
Creator: Abrams, Tyler; Kallman, J; Kaitaa, R; Foley, E L; Grayd, T K; Kugel, H et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

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… more
Date: October 27, 2012
Creator: M. Ono, et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

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-L… more
Date: July 9, 2013
Creator: Roquemore, A. L.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

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 o… more
Date: July 20, 2012
Creator: R. Kaita, et. al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

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 spectrosco… more
Date: July 20, 2012
Creator: Skinner, C. H.; Sullenberger, R.; Koel, B. E.; Jaworski, M. A. & Kugel, H. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 requ… more
Date: May 13, 2005
Creator: Skinner, C. H. & Federici, G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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… more
Date: May 16, 2011
Creator: Zweben, S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

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 … more
Date: March 20, 2008
Creator: D.P. Stotler, T.D. Rognlien and S.I. Krasheninnikov
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 durin… more
Date: June 4, 2003
Creator: Skinner, C. H.; Gentile, C. A.; Ciebiera, L. & Langish, S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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,… more
Date: November 24, 2003
Creator: Skinner, C. H.; Coad, J. P. & Federici, G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 c… more
Date: October 1, 2010
Creator: Rais, B.; Skinner, C. H. & Roquemore, A. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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… more
Date: March 24, 2011
Creator: Maingi, R.; Ahn, J.-W.; Gray, T. K.; McLean, A. G. & Soukhanovskii, V. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 col… more
Date: June 15, 2004
Creator: Skinner, C. H.; Roquemore, A. L.; NSTX team; Bader, A. & Wampler, W. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 … more
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
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 ope… more
Date: January 15, 2010
Creator: Majeski, R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

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… more
Date: May 23, 2011
Creator: Skinner, C. H.; Bell, M. G.; Friesen, F. Q. L.; Heim, B.; Jaworski, M. A.; Kugel, H. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

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 disch… more
Date: March 18, 2010
Creator: Majeski, R.; Kugel, H. & Kaita, R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 discha… more
Date: March 8, 2004
Creator: Dunaevsky, A. & Fisch, N. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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… more
Date: August 9, 2004
Creator: Paul, S. F.; Maingi, R.; Soukhanovskii, V.; Kaye, S. M. & Kugel, H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 … more
Date: August 22, 2002
Creator: Kaganovich, Igor D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Heuristic Drift-based Model of the Power Scrape-off width in H-mode Tokamaks

Description: An heuristic model for the plasma scrape-off width in H-mode plasmas is introduced. Grad B and curv B drifts into the SOL are balanced against sonic parallel flows out of the SOL, to the divertor plates. The overall particle flow pattern posited is a modification for open field lines of Pfirsch-Shlüter flows to include sinks to the divertors. These assumptions result in an estimated SOL width of ~ 2aρp/R. They also result in a first-principles calculation of the particle confinement time of H-m… more
Date: April 29, 2011
Creator: Goldston, Robert J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Development of NSTX Particle Control Techniques

Description: The National Spherical Torus Experiment (NSTX) High Harmonic Fast Wave (HHFW) current-drive discharges will require density control for acceptable efficiency. In NSTX, this involves primarily controlling impurity influxes and recycling. We have compared boronization on hot and cold surfaces, varying helium glow discharge conditioning (HeGDC) durations, helium discharge cleaning, brief daily boronization, and between discharge boronization to reduce and control spontaneous density rises. Access … more
Date: July 30, 2004
Creator: Kugel, H. W.; Maingi, R.; Bell, M.; Gates, D.; Hill, K.; LeBlanc, B. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Towards a Revised Monte Carlo Neutral Particle Surface Interaction Model

Description: The components of the neutral- and plasma-surface interaction model used in the Monte Carlo neutral transport code DEGAS 2 are reviewed. The idealized surfaces and processes handled by that model are inadequate for accurately simulating neutral transport behavior in present day and future fusion devices. We identify some of the physical processes missing from the model, such as mixed materials and implanted hydrogen, and make some suggestions for improving the model.
Date: June 9, 2005
Creator: Stotler, D. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

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 volt… more
Date: February 11, 2005
Creator: Raitses, Y.; Staack, D.; Keidar, M. & Fisch, N.J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
Back to Top of Screen