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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

Measurement of Molecular Deuterium Fluxes in the DIII-D Edge

Description: In hydrogen-fueled tokamak discharges, the distribution of molecular hydrogen (or deuterium) in the plasma edge region plays a central role in edge fueling, affecting pedestal shape and core density control [1]. In addition to its role in edge fueling, molecular hydrogen is important for plasma edge atomic physics. An example of this is the enhancement of plasma volume recombination known to occur in the presence of vibrationally-excited hydrogen molecules via conversion of H{sup +} ions into molecular ions such as H{sub 2}{sup +} and H{sub 3}{sup +} [2]. Here, measurements of the D{sub 2} molecule flux into the far edge/scrape-off layer (SOL) of the DIII-D tokamak are made using passive visible spectroscopy of the D{sub 2} diagonal Fulcher band (3p-2s triplet Q-branch) line emission over the range {lambda} = 600.640 nm [3]. L-mode, lower-single-null discharges are studied. A multi-chord visible spectrometer with views of both lower divertor legs and the main chamber is used [4]. A schematic of the spectrometer view chords used here, as well as typical magnetic flux surfaces, midplane probe location, and Thomson scattering view locations, are shown in Fig. 1. As a convenient variable to describe the location of each view chord, the poloidal angle {theta} of the corresponding emission volume is used (Fig. 1). Each view chord crosses the SOL twice; in the case of the upper view chords and lower view chords, the emission from the SOL closer to the lower divertor is expected to dominate the measured signal. In the case of the midplane view chord, lineshape (Zeeman splitting) analysis of the D{sub {alpha}} line indicates that the received emission is typically dominated by the inner wall SOL (over the outer wall SOL by {approx} 2-8x).
Date: June 24, 2005
Creator: Hollmann, E; Brezinsek, S; Brooks, N; Groth, M; Lisgo, S; McLean, A et al.
Partner: UNT Libraries Government Documents Department

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

Toroidally Asymmetric Distributions of Hydrocarbon (CD) Emission and Chemical Sputtering Sources in DIII-D

Description: Measurements in DIII-D show that the carbon chemical sputtering sources along the inner divertor and center post are toroidally periodic and highest at the upstream tile edge. Imaging with a tangentially viewing camera and visible spectroscopy were used to monitor the emission from molecular hydrocarbons (CH/CD) at 430.8 nm and deuterium neutrals in attached and partially detached divertors of low-confinement mode plasmas. In contrast to the toroidally periodic CD distribution, emission from deuterium neutrals was observed to be toroidally symmetric along the inner strike zone. The toroidal distribution of the measured tile surface temperature in the inner divertor correlates with that of the CD emission, suggesting larger parallel particle and heat fluxes to the upstream tile edge, either due to toroidal tile gaps or height steps between adjacent tiles.
Date: May 16, 2006
Creator: Groth, M; Brooks, N H; Fenstermacher, M E; Lasnier, C J; McLean, A G & Watkins, J G
Partner: UNT Libraries Government Documents Department

OEDGE Modeling of {sup 13}C Deposition in the Inner Divertor of DIII-D

Description: Use of carbon in tokamaks leads to a major tritium retention issue due to co-deposition. To investigate this process a low power (no beams) L-mode experiment was performed on DIII-D in which {sup 13}CH{sub 4} was puffed into the main vessel through the toroidally-symmetric pumping plenum at the top of lower single-null discharges. Subsequently, the {sup 13}C content of tiles taken from the vessel wall was measured. The interpretive OEDGE code was used to model the results. It was found that the {sup 13}C deposition pattern is controlled by: (a) source strength of {sup 13}C{sup +}, (b) radial location of the {sup 13}C{sup +} source, (c) D{sub {perpendicular}}, (d) M{sub {parallel}}, the scrape-off layer parallel Mach number. Best agreement was found for (a) {approx}50% conversion efficiency {sup 13}CH{sub 4} {yields} {sup 13}C{sup +}, (b) {sup 13}C{sup +} source {approx}3.5 cm outboard of separatrix near {sup 13}CH{sub 4} injection location, (c)D{sub {perpendicular}} {approx} 0.3 m{sup 2}s{sup -1}, (d) M{sub {parallel}} {approx} 0.4 toward inside.
Date: December 1, 2004
Creator: Elder, J; Stangeby, P; Whyte, D; Allen, S; McLean, A; Boedo, J et al.
Partner: UNT Libraries Government Documents Department

Large Area Divertor Temperature Measurements Using A High-speed Camera With Near-infrared FiIters in NSTX

Description: Fast cameras already installed on the National Spherical Torus Experiment (NSTX) have be equipped with near-infrared (NIR) filters in order to measure the surface temperature in the lower divertor region. Such a system provides a unique combination of high speed (> 50 kHz) and wide fi eld-of-view (> 50% of the divertor). Benchtop calibrations demonstrated the system's ability to measure thermal emission down to 330 oC. There is also, however, signi cant plasma light background in NSTX. Without improvements in background reduction, the current system is incapable of measuring signals below the background equivalent temperature (600 - 700 oC). Thermal signatures have been detected in cases of extreme divertor heating. It is observed that the divertor can reach temperatures around 800 oC when high harmonic fast wave (HHFW) heating is used. These temperature profiles were fi t using a simple heat diffusion code, providing a measurement of the heat flux to the divertor. Comparisons to other infrared thermography systems on NSTX are made.
Date: April 5, 2011
Creator: Lyons, B. C.; Zweben, S. J.; Gray, T. K.; Hosea, J.; Kaita, R.; Kugel, H. W. et al.
Partner: UNT Libraries Government Documents Department

Modification Of The Electron Energy Distribution Function During Lithium Experiments On The National Spherical Torus Experiment

Description: The National Spherical Torus Experiment (NSTX) has recently studied the use of a liquid lithium divertor (LLD). Divertor Langmuir probes have also been installed for making measurements of the local plasma conditions. A non-local probe interpretation method is used to supplement the classical probe interpretation and obtain measurements of the electron energy distribution function (EEDF) which show the occurrence of a hot-electron component. Analysis is made of two discharges within a sequence that exhibited changes in plasma fueling efficiency. It is found that the local electron temperature increases and that this increase is most strongly correlated with the energy contained within the hot-electron population. Preliminary interpretative modeling indicates that kinetic effects are likely in the NSTX.
Date: June 3, 2011
Creator: Jaworski, M. A.; Gray, T. K.; Kaita, R.; Kallman, J.; Kugel, H.; LeBlanc, B. et al.
Partner: UNT Libraries Government Documents Department

First Wall and Operational Diagnostics

Description: In this chapter we review numerous diagnostics capable of measurements at or near the first wall, many of which contribute information useful for safe operation of a tokamak. There are sections discussing infrared cameras, visible and VUV cameras, pressure gauges and RGAs, Langmuir probes, thermocouples, and erosion and deposition measurements by insertable probes and quartz microbalance. Also discussed are dust measurements by electrostatic detectors, laser scattering, visible and IR cameras, and manual collection of samples after machine opening. In each case the diagnostic is discussed with a view toward application to a burning plasma machine such as ITER.
Date: June 19, 2006
Creator: Lasnier, C; Allen, S; Boedo, J; Groth, M; Brooks, N; McLean, A et al.
Partner: UNT Libraries Government Documents Department

Transport and Deposition of 13c From Methane Injection into Detached H-Mode Plasmas in DIII-D

Description: Experiments are described which examine the transport and deposition of carbon entering the main plasma scrape-off layer in DIII-D. {sup 13}CH{sub 4} was injected from a toroidally symmetric source into the crown of lower single-null detached ELMy H-mode plasmas. {sup 13}C deposition, mapped by nuclear reaction analysis of tiles, was high at the inner divertor but absent at the outer divertor, as found previously for low density L-mode plasmas. This asymmetry indicates that ionized carbon is swept towards the inner divertor by a fast flow in the scrape-off layer. In the private flux region between inner and outer strike points, carbon deposition was low for L-mode but high for the H-mode plasmas. OEDGE modeling reproduces observed deposition patterns and indicates that neutral carbon dominates deposition in the divertor from detached H-mode plasmas.
Date: June 1, 2006
Creator: Wampler, W R; McLean, A G; Allen, S L; Brooks, N H; Elder, J D; Fenstermacher, M E et al.
Partner: UNT Libraries Government Documents Department

Scrape-Off Layer Transport and Deposition Studies in DIII-D

Description: Trace {sup 13}CH{sub 4} injection experiments into the main scrape-off layer of low density L-mode and high-density H-mode plasmas have been performed in the DIII-D tokamak [Luxon{_}NF02] to mimic the transport and deposition of carbon arising from a main chamber sputtering source. These experiments indicated entrainment of the injected carbon in plasma flow in the main SOL, and transport toward the inner divertor. Ex-situ surface analysis showed enhanced {sup 13}C surface concentration at the corner formed by the divertor floor and the angled target plate of the inner divertor in L-mode; in H-mode, both at the corner and along the surface bounding the private flux region inboard of the outer strike point. Interpretative modeling was made consistent with these experimental results by imposing a parallel carbon ion flow in the main SOL toward the inner target, and a radial pinch toward the separatrix. Predictive modeling carried out to better understand the underlying plasma transport processes suggests that the deuterium flow in the main SOL is related to the degree of detachment of the inner divertor leg. These simulations show that carbon ions are entrained with the deuteron flow in the main SOL via frictional coupling, but higher charge state carbon ions may be suspended upstream of the inner divertor X-point region due to balance of the friction force and the ion temperature gradient.
Date: October 27, 2006
Creator: Groth, M; Allen, S; Boedo, J; Brooks, N; Elder, J; Fenstermacher, M et al.
Partner: UNT Libraries Government Documents Department

DIVIMP Modeling of the Toroidally-Symmetrical Injection of 13CH4 into the Upper SOL of DIII-D

Description: As part of a study of carbon-tritium co-deposition, we carried out an experiment on DIII-D involving a toroidally symmetric injection of {sup 13}CH{sub 4} at the top of a LSN discharge. A Monte Carlo code, DIVIMP-HC, which includes molecular breakup of hydrocarbons, was used to model the region near the puff. The interpretive analysis indicates a parallel flow in the SOL of M{sub l} {approx} 0.4 directed toward the inner divertor. The CH{sub 4} is ionized in the periphery of the SOL and so the particle confinement time, {tau}{sub c}, is not high, only {approx}5 ms, and about 4X lower than if the CH{sub 4} were ionized at the separatrix. For such a wall injection location, however, most of the CH{sub 4} gets ionized to C{sup +}, C{sup ++}, etc., and is efficiently transported along the SOL to the inner divertor, trapping hydrogen by co-deposition there.
Date: December 3, 2004
Creator: McLean, A G; Elder, J D; Stangeby, P C; Allen, S L; Brooks, N H; Fenstermacher, M E et al.
Partner: UNT Libraries Government Documents Department

Far Scrape-Off Layer and Near Wall Plasma Studies in DIII-D

Description: Far scrape-off layer (SOL) plasma parameters in DIII-D depend strongly on the discharge density and confinement regime. In L-mode, cross-field transport increases with the average discharge density and elevates the far SOL density, thus increasing plasma-wall contact. Far SOL density near the low field side (LFS) of the main chamber wall also increases with decreasing plasma current and with decreasing outer wall gap. In H-mode, between edge localized modes (ELMs), plasma-wall contact is weaker than in L-mode. During ELMs plasma fluxes to the LFS wall increase to, or above the L-mode levels. A large fraction of the net cross-field fluxes is convected through the SOL by large amplitude intermittent transport events. In high density L-mode and during ELMs in H-mode, intermittent events propagate all the way to the LFS wall and may cause sputtering.
Date: December 3, 2004
Creator: Rudakov, D; Boedo, J; Moyer, R; Brooks, N; Doerner, R; Evans, T et al.
Partner: UNT Libraries Government Documents Department

HHFW Power Flow Along Magnetic Field Lines In The Scrape-off Layer of NSTX

Description: A significant fraction of high-harmonic fast-wave (HHFW) power applied to NSTX can be lost to the scrape-off layer (SOL) and deposited in bright and hot spirals on the divertor rather than in the core plasma. We show that the HHFW power flows to these spirals along magnetic field lines passing through the SOL in front of the antenna, implying that the HHFW power couples across the entire width of the SOL rather than mostly at the antenna face. This result will help guide future efforts to understand and minimize these edge losses in order to maximize fast wave heating and current drive.
Date: February 27, 2012
Creator: Perkins, Rory; Bell, R. E.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A. et al.
Partner: UNT Libraries Government Documents Department

Particle Control and Plasma Performance in the Lithium Tokamak Experiment (LTX)

Description: The Lithium Tokamak eXperiment (LTX) is a small, low aspect ratio tokamak, which is fitted with a stainless steel-clad copper liner, conformal to the last closed flux surface. The liner can be heated to 350{degree}C. Several gas fueling systems, including supersonic gas injection, and molecular cluster injection have been studied, and produce fueling efficiencies up to 35%. Discharges are strongly affected by wall conditioning. Discharges without lithium wall coatings are limited to plasma currents of order 10 kA, and discharge durations of order 5 msec. With solid lithium coatings discharge currents exceed 70 kA, and discharge durations exceed 30 msec. Heating the lithium wall coating, however, results in a prompt degradation of the discharge, at the melting point of lithium. These results suggest that the simplest approach to implementing liquid lithium walls in a tokamak - thin, evaporated, liquefied coatings of lithium - does not produce an adequately clean surface.
Date: February 21, 2013
Creator: Majeski, Richard Majeski; Abrams, T.; Boyle, D.; Granstedt, E.; Hare, J.; Jacobson, C. M. et al.
Partner: UNT Libraries Government Documents Department

Observation of Dust in DIII-D Divertor and SOL by Visible Imaging

Description: Dust is commonly found in fusion devices. Though generally of no concern in the present day machines, dust may pose serious safety and operational concerns for ITER. Micron-size dust usually dominates the samples collected from tokamaks. During a plasma discharge micron-size dust particles can become highly mobile and travel over distances of a few meters. Once inside the plasma, dust particles heat up to over 3000 K and emit thermal radiation that can be detected by visible imaging techniques. Observations of naturally occurring and artificially introduced dusts have been performed in DIII-D divertor and scrape-off layer (SOL) using standard frame rate CMOS cameras, a gated-intensified CID camera, and a fast-framing CMOS camera. In the first 2-3 plasma discharges after a vent with personnel entry inside the vacuum vessel ('dirty vent') dust levels were quite high with thousands of particles observed in each discharge. Individual particles moving at velocities of up to a few hundred m/s and breakup of larger particles into pieces were observed. After about 15 discharges dust was virtually gone during the stationary portion of a discharge, and appeared at much reduced levels during the plasma initiation and termination phases. After a few days of plasma operations (about 70 discharges) dust levels were further reduced to just a few observed events per discharge except in discharges with current disruptions that produced significant amounts of dust. An injection of a few milligram of micron-size (6 micron median diameter) carbon dust into a high-power lower single-null ELMing H-mode discharge with strike points swept across the lower divertor floor was performed. A significant increase of the core carbon radiation was observed for about 250 ms after the injection, as the total radiated power increased twofold. Dust particles from the injection were observed by the fast framing camera in the outboard SOL ...
Date: April 2, 2007
Creator: Rudakov, D L; West, W P; Groth, M; Yu, J H; Wong, C C; Boedo, J A et al.
Partner: UNT Libraries Government Documents Department

13C-Tracer Experiments in DIII-D Preliminary to Thermal Oxidation Experiments to Understand Tritium Recovery in DIII-D, JET, C-Mod, and MAST

Description: Retention of tritium in carbon co-deposits is a serious concern for ITER. Developing a reliable in-situ removal method of the co-deposited tritium would allow the use of carbon plasma-facing components which have proven reliable in high heat flux conditions and compatible with high performance plasmas. Thermal oxidation is a potential solution, capable of reaching even hidden locations. It is necessary to establish the least severe conditions to achieve adequate tritium recovery, minimizing damage and reconditioning time. The first step in this multi-machine project is {sup 13}C-tracer experiments in DIII-D, JET, C-Mod and MAST. In DIII-D and JET, {sup 13}CH{sub 4} has been (and in C-Mod and MAST, will be) injected toroidally symmetrically, facilitating quantification and interpretation of the results. Tiles have been removed, analyzed for {sup 13}C content and will next be evaluated in a thermal oxidation test facility in Toronto with regard to the ability of different severities of oxidation exposure to remove the different types of (known and measured) {sup 13}C co-deposit. Removal of D/T from B on Mo tiles from C-Mod will also be tested. OEDGE interpretive code analysis of the {sup 13}C deposition patterns is used to generate the understanding needed to apply findings to ITER. First results are reported here for the {sup 13}C injection experiments IN DIII-D.
Date: June 19, 2006
Creator: Stangeby, P; Allen, S; Bekris, N; Brooks, N; Christie, K; Chrobak, C 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} < 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

DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

Description: A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited virtually no carbon deposition.
Date: October 2, 2006
Creator: Rudakov, D; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W et al.
Partner: UNT Libraries Government Documents Department

Dust Measurements in Tokamaks

Description: Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.
Date: April 23, 2008
Creator: Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R et al.
Partner: UNT Libraries Government Documents Department

Spectroscopic Characterization and Simulation of Chemical Sputtering Using the DiMES Porous Plug Injector in DIII-D

Description: A self-contained gas injection system for the Divertor Material Evaluation System (DiMES) on DIII-D has been employed for in-situ study of chemical erosion in the tokamak divertor environment. The Porous Plug Injector (PPI) releases methane, a major component of molecular influx due to chemical sputtering of graphite, from the tile surface into the plasma at a controlled rate through a porous graphite surface. Perturbation to local plasma is minimized, while also simulating the immediate environment of methane molecules released from a solid graphite surface. The release rate was chosen to be of the same order of magnitude as natural sputtering. Photon efficiencies of CH{sub 4} for measured local plasma conditions are reported. The contribution of chemical versus physical sputtering to the source of C{sup +} at the target is assessed through measurement of CII and CD/CH band emissions during release of CH{sub 4} from the PPI, and due to intrinsic emission.
Date: May 15, 2006
Creator: McLean, A G; Davis, J W; Stangeby, P C; Brooks, N H; Whyte, D G; Allen, S L et al.
Partner: UNT Libraries Government Documents Department