37 Matching Results

Search Results

Advanced search parameters have been applied.

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

Comparison of Gas Puff Imaging Data in NSTX with the DEGAS 2 Simulation

Description: Gas-Pu -Imaging (GPI) is a two dimensional diagnostic which measures the edge Dα light emission from a neutral Dα gas puff near the outer mid- plane of the National Spherical Torus Experiment (NSTX). DEGAS 2 is a 3-D Monte Carlo code used to model neutral transport and plasma-neutral interactions in fusion plasmas. In this paper, we compare the measured and modeled Dα light emission for speci c NSTX experiments. Both the simulated spatial distribution and radiance of the Dα light emission agree well with the experimental data obtained between Edge Localized Modes (ELMs) in ELMy H-modes.
Date: April 10, 2013
Creator: Cao, B.; Stotler, D. P.; Zweben, S. J.; Bell, M.; Diallo, A. & LeBlanc, B.
Partner: UNT Libraries Government Documents Department

High Speed Imaging of Edge Turbulence in NSTX

Description: The two-dimensional radial versus poloidal structure and motion of edge turbulence in NSTX (National Spherical Torus Experiment) were measured by using high-speed imaging of the visible light emission from a localized neutral gas puff. Edge turbulence images are shown and analyzed for Ohmic, L-mode (low-confinement mode) and H-mode (high-confinement mode) plasma conditions. Typical edge turbulence poloidal correlation lengths as measured using this technique are = 4 {+-} 1 cm and autocorrelation times are 40 {+-} 20 {micro}sec in all three regimes. The relative fluctuation level is typically smaller in H-mode than in L-mode, and transitions from H- to L-mode and can occur remarkably quickly (=30 {micro}sec). The two-dimensional images often show localized regions of strong light emission which move both poloidally and radially through the observed region at a typical speed of =10{sup 5} cm/sec, and sometimes show spatially coherent modes.
Date: March 2003
Creator: Zweben, S. J.; Maqueda, R.; Stotler, D. P.; Keesee, A.; Boedo, J.; Bush, C. et al.
Partner: UNT Libraries Government Documents Department

Final Technical Report: Global Field Aligned Mesh and Gyrokinetic Field Solver in a Tokamak Edge Geometry

Description: This project was a collaboration between researchers at the California Institute of Technology and the University of California, Irvine to investigate the utility of a global field-aligned mesh and gyrokinetic field solver for simulations of the tokamak plasma edge region. Mesh generation software from UC Irvine was tested with specific tokamak edge magnetic geometry scenarios and the quality of the meshes and the solutions to the gyrokinetic Poisson equation were evaluated.
Date: May 15, 2013
Creator: Cummings, Julian C.
Partner: UNT Libraries Government Documents Department

Plot/SurfW: Plotting Utility for EDGE2D Output

Description: This report describes a utility that was developed to display EDGE2D results. The utility is focused on results that relate to impurity density, velocity, and particle fluxes in the SOL and divertor. Due to the complicated nature of 2D impurity sources, the concentration of the thermal force near the separatrix and near the divertor entrance, the impurity flow pattern and impurity densities are not necessarily easy to visualize. Thus, we wanted a utility that allowed simple and quick visualization of the impurity behavior. In order to achieve this we overlaid the divertor hardware for plots inside the divertor and we expanded the appearance of the main chamber SOL by plotting distance along the field lines vs. SOL depth with the density (or velocity or flux or other quantity) the false colour. Also, we allowed for the plotted variable to be a function of the other EDGE2D result variables. __________________________________________________
Date: June 22, 2012
Creator: Strachan, W.M. Davis and J.D.
Partner: UNT Libraries Government Documents Department

Plasma-based Accelerator with Magnetic Compression

Description: Electron dephasing is a major gain-inhibiting effect in plasma-based accelerators. A novel method is proposed to overcome dephasing, in which the modulation of a modest (#24; O(10 kG)), axial, uniform magnetic field in the acceleration channel leads to densification of the plasma through magnetic compression, enabling direct, time-resolved control of the plasma wave properties. The methodology is broadly applicable and can be optimized to improve the leading acceleration approaches, including plasma beat-wave, plasma wakefield, and laser wakefield acceleration. The advantages of magnetic compression compared to other proposed schemes to overcome dephasing are identified.
Date: June 28, 2012
Creator: Schmit, Paul F. & Fisch, Nathaniel J.
Partner: UNT Libraries Government Documents Department

Observation of a High Performance Operating Regime with Small Edge-Localized Modes in the National Spherical Torus Experiment

Description: We report observation of a high performance scenario in the National Spherical Torus Experiment with very small edge-localized modes (ELMs). These ELMs have no measurable impact on stored energy and are consistent with high bootstrap current operation with line average density approaching Greenwald scaling. The ELM perturbation is observed to typically originate near the lower divertor region, as opposed to the outer midplane for ELMs described in the literature. If extrapolable, this scenario would provide an attractive operating regime for next step fusion experiments
Date: May 13, 2004
Creator: Maingi, R.; Tritz, K.; Fredrickson, E.D.; Menard, J.E.; Sabbagh, S.A.; Stutman, D. et al.
Partner: UNT Libraries Government Documents Department

Images of Edge Turbulence in NSTX

Description: The 2-D structure of edge plasma turbulence has been measured in the National Spherical Torus Experiment (NSTX) by viewing the emission of the Da spectral line of deuterium. Images have been made at framing rates of up to 250,000 frames/sec using an ultra-high speed CCD camera developed by Princeton Scientific Instruments. A sequence of images showing the transition between L-mode and H-mode states is shown.
Date: July 16, 2004
Creator: Zweben, S.J.; Bush, C.E.; Maqueda, R.; Munsat, T.; Stotler, D.; Lowrance, J. et al.
Partner: UNT Libraries Government Documents Department

Dynamical Evolution of Pedestal Parameters in ELMy H-mode in the National Spherical Torus Experiment

Description: Characterizations of the pedestal parameter dynamics throughout the edge localized modes(ELM) cycles are performed on the National Spherical Torus Experiment (NSTX, [M. Ono et al., Nucl. Fusion 40, 557 (2000)]). A clear buildup of the pedestal height is observed between ELMs for three di erent plasma currents, which tends to saturate prior to the onset of ELM at low and medium plasma current. Similarly, the pedestal width increases with no clear evidence of saturation during an ELM cycle. The maximum pedestal gradient increases as a function of plasma current, reaches a nominal value after the ELM crash, and remains constant until the end of the ELM cycle. The pedestal height just prior to the onset of ELM is shown to increase quadratically with plasma current. The pedestal width Δ is proportional to the square-root of the poloidal Β at the top of the pedestal. Coherent density uctuations strongly increasing at the plasma edge are observed to be maximum after the ELM crash and to decay during the rest of the ELM cycle. Finally, the pedestal parameters evolution during the ELM cycle as well as the scaling with Ip of the pedestal pressure prior to the onset ELM are found to be qualitatively consistent with the peeling ballooning theory.
Date: July 27, 2011
Creator: Diallo, A; Kubota, S; Sontag, A; Osborne, T; Podesta, M; Bell, R E et al.
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

ELMs and the H-mode Pedestal in NSTX

Description: We report on the behavior of ELMs in NBI-heated H-mode plasmas in NSTX. It is observed that the size of Type I ELMs, characterized by the change in plasma energy, decreases with increasing density, as observed at conventional aspect ratio. It is also observed that the Type I ELM size decreases as the plasma equilibrium is shifted from a symmetric double-null toward a lower single-null configuration. Type III ELMs have also been observed in NSTX, as well as a high-performance regime with small ELMs which we designate Type V. These Type V ELMs are consistent with high bootstrap current operation and density approaching Greenwald scaling. The Type V ELMs are characterized by an intermittent n=1 MHD mode rotating counter to the plasma current. Without active pumping, the density rises continuously through the Type V phase. However, efficient in-vessel pumping should allow density control, based on particle containment time estimates.
Date: July 16, 2004
Creator: Maingi, R.; Sabbagh, S.A.; Bush, C.E.; Fredrickson, E.D.; Menard, J.E.; Stutman, D. 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

Recent Liquid Lithium Limiter Experiments in CDX-U

Description: Recent experiments in the Current Drive eXperiment-Upgrade (CDX-U) 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) {approx} 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 pool 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 pool 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: May 3, 2005
Creator: Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J. et al.
Partner: UNT Libraries Government Documents Department

Edge Minority Heating Experiment in Alcator C-Mod

Description: An attempt was made to control global plasma confinement in the Alcator C-Mod tokamak by applying ion cyclotron resonance heating (ICRH) power to the plasma edge in order to deliberately create a minority ion tail loss. In theory, an edge fast ion loss could modify the edge electric field and so stabilize the edge turbulence, which might then reduce the H-mode power threshold or improve the H-mode barrier. However, the experimental result was that edge minority heating resulted in no improvement in the edge plasma parameters or global stored energy, at least at power levels of radio-frequency power is less than or equal to 5.5 MW. A preliminary analysis of these results is presented and some ideas for improvement are discussed.
Date: March 25, 2005
Creator: Zweben, S.J.; Terry, J.L.; Bonoli, P.; Budny, R.; Chang, C.S.; Fiore, C. et al.
Partner: UNT Libraries Government Documents Department

Methane Screening in JET Reverse Field Experiments

Description: JET plasmas with reverse magnetic field feature a different SOL flow than those with normal field. The observed carbon fueling efficiency from injecting methane gas was similar in reverse and normal field. EDGE2D modeling used an externally applied force to create the SOL flows, without specifying the origin of the force. The resulting flow agreed reasonably with the experimental values between the separatrix and 4 cm mid-plane depth in the SOL. The effect of the flow on the calculated carbon screening was 5 to 15% higher carbon fueling efficiency for the low flow velocity with reverse field.
Date: May 17, 2004
Creator: Strachan, J.D.; Alper, B.; Corrigan, G.; Erents, S.K.; Giroud, C.; Korotkov, A. et al.
Partner: UNT Libraries Government Documents Department

Observations of Anisotropic Ion Temperature in the NSTX Edge during RF Heating

Description: A new spectroscopic diagnostic on the National Spherical Torus Experiment (NSTX) measures the velocity distribution of ions in the plasma edge with both poloidal and toroidal views. An anisotropic ion temperature is measured during the presence of high-power high-harmonic fast-wave (HHFW) radio-frequency (RF) heating in helium plasmas, with the poloidal ion temperature roughly twice the toroidal ion temperature. Moreover, the measured spectral distribution suggests that two populations are present and have temperatures of 500 eV and 50 eV with rotation velocities of -50 km/s and -10 km/s, respectively. This bi-modal distribution is observed in both the toroidal and poloidal views (in both He{sup +} and C{sup 2+} ions), and is well correlated with the period of RF power application to the plasma. The temperature of the hot edge ions is observed to increase with the applied RF power, which was scanned between 0 and 4.3 MW. The ion heating mechanism is likely to be ion-Bernstein waves (IBW) from nonlinear decay of the launched HHFW.
Date: October 21, 2004
Creator: Biewer, T.M.; Bell, R.E.; Wilson, J.R. & Ryan, P.M.
Partner: UNT Libraries Government Documents Department

The H-mode Pedestal and Edge Localized Modes in NSTX

Description: The research program of the National Spherical Torus Experiment (NSTX) routinely utilizes the H-mode confinement regime to test and extend beta and pulse length limits. As in conventional aspect ratio tokamaks, NSTX observes a variety of edge localized modes (ELMs) in H-mode. Hence a significant part of the research program is dedicated to ELMs studies.
Date: July 16, 2004
Creator: Maingi, R.; Fredrickson, E.D.; Menard, J.E.; Nishino, N.; Roquemore, A.L.; Sabbagh, S.A. et al.
Partner: UNT Libraries Government Documents Department

Effect of Gas Fueling Location on H-mode Access in NSTX

Description: The dependence of H-mode access on the poloidal location of the gas injection source has been investigated in the National Spherical Torus Experiment (NSTX). We find that gas fueling from the center stack midplane area produces the most reproducible H-mode access with generally the lowest L-H threshold power in lower single-null configuration. The edge toroidal rotation velocity is largest (in direction of the plasma current) just before the L-H transition with center stack midplane fueling, and then reverses direction after the L-H transition. Simulation of these results with a 2-D guiding-center Monte Carlo neoclassical transport code is qualitatively consistent with the trends in the measured velocities. Double-null discharges exhibit H-mode access with gas fueling from either the center stack midplane or center stack top locations, indicating a reduced sensitivity of H-mode access on fueling location in that shape.
Date: October 9, 2003
Creator: Maingi, R.; Bell, M.; Bell, R.; Biewer, T.; Bush, C.; Chang, C.S. et al.
Partner: UNT Libraries Government Documents Department

L-H Mode Transitions in the National Spherical Torus Experiment

Description: Edge data from plasmas in the National Spherical Torus Experiment (NSTX) [S. Kaye et al., Fusion Technology 36 (1999) 16] have been compared to theories of transport suppression that have been used to develop a physics framework for low-confinement (L) to high-confinement (H) mode transitions. The NSTX data were obtained in low aspect ratio (R/a approximately equal to 1.3) discharges taken from a variety of discharge phases, including L-modes, L-H transitions, and H-modes with and without edge localized modes (ELMs). The comparisons show that the group of points taken just before the L-H mode transition are well mixed with the purely L-mode group to within the measurement uncertainties, indicating that changes in these parameters leading up to the transition are subtle. One of the theory parameters, alpha{sub MHD} = -R{sub q}{sup 2}dbeta{sub t/dr}, does show a clear threshold alpha{sub MHD} = 1 to 2 between the H-mode grouping of points and those remaining in the L-mode or taken just prior to the transition. Additionally, there is no evidence for an edge temperature threshold necessary for transitioning into the H-mode. NSTX data indicate further a possible connection between L-H transitions and non-ambipolar beam ion losses.
Date: July 24, 2003
Creator: Kaye, S.M.; Bush, C.E.; Fredrickson, E.; LeBlanc, B.; Maingi, R. & Sabbagh, S.A.
Partner: UNT Libraries Government Documents Department

Edge Turbulence Imaging on NSTX and Alcator C-Mod

Description: Edge turbulence images have been made using an ultra-high speed CCD camera on both NSTX and Alcator C-Mod. In both cases, the D-alpha or HeI (587.6 nm) line emission from localized deuterium or helium gas puffs was viewed along a local magnetic field line near the outer midplane. Fluctuations in this line emission reflect fluctuations in electron density and/or electron temperature through the atomic excitation rates, which can be modeled using the DEGAS-2 code. The 2-D structure of the measured turbulence can be compared with theoretical simulations based on 3-D fluid models.
Date: July 10, 2002
Creator: Zweben, S.J.; Maqueda, R.A.; Terry, J.L.; Bai, B.; Boswell, C.J.; Bush, C.E. et al.
Partner: UNT Libraries Government Documents Department

EDGE2D Simulations of JET{sup 13}C Migration Experiments

Description: Material migration has received renewed interest due to tritium retention associated with carbon transport to remote vessel locations. Those results influence the desirability of carbon usage on ITER. Subsequently, additional experiments have been performed, including tracer experiments attempting to identify material migration from specific locations. In this paper, EDGE2D models a well-diagnosed JET{sup 13}C tracer migration experiment. The role of SOL flows upon the migration patterns is identified.
Date: June 16, 2004
Creator: Strachan, J.D.; Coad, J.P.; Corrigan, G.; Matthews, G.F. & Spence, J.
Partner: UNT Libraries Government Documents Department

Intermittency in the Scrape-off Layer of the National Spherical Torus Experiment During H-mode Confinement

Description: A gas puff imaging diagnostic is used in the National Spherical Tokamak Experiment [M. Ono, et al., Nucl. Fusion 40, 557 (2000)] to study the edge turbulence and intermittency present during H-mode discharges. In the case of low power Ohmic H-modes the suppression of turbulence/blobs is maintained through the duration of the (short lived) H-modes. Similar quiescent edges are seen during the early stages of H-modes created with the use of neutral beam injection. Nevertheless, as time progresses following the L-H transition, turbulence and blobs reappear although at a lower level than that typically seen during L-mode confinement. It is also seen that the time-averaged SOL emission profile broadens, as the power loss across the separatrix increases. These broad profiles are characterized by a large level of fluctuations and intermittent events.
Date: November 22, 2010
Creator: Maqueda, R. J.; Stotler, D. P. & Zweben, S. J.
Partner: UNT Libraries Government Documents Department

Edge Plasma Boundary Layer Generated By Kink Modes in Tokamaks

Description: This paper describes the structure of the electric current generated by external kink modes at the plasma edge using the ideally conducting plasma model. It is found that the edge current layer is created by both wall touching and free boundary kink modes. Near marginal stability, the total edge current has a universal expression as a result of partial compensation of the δ-functional surface current by the bulk current at the edge. The resolution of an apparent paradox with the pressure balance across the plasma boundary in the presence of the surface currents is provided.
Date: November 22, 2010
Creator: Zakharov, L.E.
Partner: UNT Libraries Government Documents Department

Basics of Fusion-Fissison Research Facility (FFRF) as a Fusion Neutron Source

Description: FFRF, standing for the Fusion-Fission Research Facility represents an option for the next step project of ASIPP (Hefei, China) aiming to a first fusion-fission multifunctional device [1]. FFRF strongly relies on new, Lithium Wall Fusion plasma regimes, the development of which has already started in the US and China. With R/a=4/1m/m, Ipl=5 MA, Btor=4-6 T, PDT=50- 100 MW, Pfission=80-4000MW, 1 m thick blanket, FFRF has a unique fusion mission of a stationary fusion neutron source. Its pioneering mission of merging fusion and fission consists in accumulation of design, experimental, and operational data for future hybrid applications.
Date: June 3, 2011
Creator: Zakharov, Leonid E.
Partner: UNT Libraries Government Documents Department