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Exploration of the Equilibrium Operating Space For NSTX-Upgrade

Description: This paper explores a range of high-performance equilibrium scenarios available in the NSTX-Upgrade device [J.E. Menard, submitted for publication to Nuclear Fusion]. NSTX-Upgrade is a substantial upgrade to the existing NSTX device [M. Ono, et al., Nuclear Fusion 40, 557 (2000)], with significantly higher toroidal field and solenoid capabilities, and three additional neutral beam sources with significantly larger current drive efficiency. Equilibria are computed with freeboundary TRANSP, allowing a self consistent calculation of the non-inductive current drive sources, the plasma equilibrium, and poloidal field coil current, using the realistic device geometry. The thermal profiles are taken from a variety of existing NSTX discharges, and different assumptions for the thermal confinement scalings are utilized. The no-wall and idealwall n=1 stability limits are computed with the DCON code. The central and minimum safety factors are quite sensitive to many parameters: they generally increases with large outer plasmawall gaps and higher density, but can have either trend with the confinement enhancement factor. In scenarios with strong central beam current drive, the inclusion of non-classical fast ion diffusion raises qmin, decreases the pressure peaking, and generally improves the global stability, at the expense of a reduction in the non-inductive current drive fraction; cases with less beam current drive are largely insensitive to additional fast ion diffusion. The non-inductive current level is quite sensitive to the underlying confinement and profile assumptions. For instance, for BT=1.0 T and Pinj=12.6 MW, the non-inductive current level varies from 875 kA with ITER-98y,2 thermal confinement scaling and narrow thermal profiles to 1325 kA for an ST specific scaling expression and broad profiles. This sensitivity should facilitate the determination of the correct scaling of transport with current and field to use for future fully non-inductive ST devices. Scenarios are presented which can be sustained for 8-10 seconds, or (20-30)τCR, at ...
Date: April 25, 2012
Creator: Gerhardt, S. P.; Andre, R. & Menard, J. E.
Partner: UNT Libraries Government Documents Department

Characterization of the plasma current quench during disruptions in the National Spherical Torus Experiment

Description: A detailed analysis of the plasma current quench in the National Spherical Torus Experiment [M.Ono, et al Nuclear Fusion 40, 557 (2000)] is presented. The fastest current quenches are fit better by a linear waveform than an exponential one. Area-normalized current quench times down to .4 msec/m2 have been observed, compared to the minimum of 1.7 msec/m2 recommendation based on conventional aspect ratio tokamaks; as noted in previous ITPA studies, the difference can be explained by the reduced self-inductance at low aspect ratio and high-elongation. The maximum instantaneous dIp/dt is often many times larger than the mean quench rate, and the plasma current before the disruption is often substantially less than the flat-top value. The poloidal field time-derivative during the disruption, which is directly responsible for driving eddy currents, has been recorded at various locations around the vessel. The Ip quench rate, plasma motion, and magnetic geometry all play important roles in determining the rate of poloidal field change.
Date: December 17, 2008
Creator: Gerhardt, S.P., Menard, J.E., and the NSTX Research Team
Partner: UNT Libraries Government Documents Department

Non-ambipolar Transport by Trapped Particles in Tokamaks

Description: Small non-axisymmetric perturbations of the magnetic field can greatly change the performance of tokamaks through non-ambipolar transport. A number of theories have been developed, but the predictions were not consistent with experimental observations in tokamaks. This Letter provides a resolution, with a generalized analytic treatment of the non-ambipolar transport. It is shown that the discrepancy between theory and experiment can be greatly reduced by two effects: (1) The small fraction of trapped particles for which the bounce and precession rates resonate. (2) The non- axisymmetric variation in the field strength along the perturbed magnetic field lines rather than along the unperturbed magnetic field lines. The expected sensitivity of ITER to non-axisymmetries is also discussed.
Date: January 27, 2009
Creator: Park, J.K. . Boozer, A.H and . Menard, J.E
Partner: UNT Libraries Government Documents Department

Investigation of Ion Absorption of the High Harmonic Fast Wave in NSTX using HPRT

Description: Understanding high harmonic fast wave (HHFW) power absorption by ions in a spherical torus (ST) is of critical importance to assessing the wave's viability as a means of heating and especially driving current. In this work, the HPRT code is used to calculate absorption for helium and deuterium, with and without minority hydrogen in National Spherical Torus Experiment (NSTX) plasmas using experimental EFIT code equilibria and kinetic profiles. HPRT is a two-dimensional ray-tracing code which uses the full hot plasma dielectric to compute the perpendicular wave number along the hot electron and cold ion plasma ray path. Ion and electron absorption dependence on antenna phasing, ion temperature, beta (subscript t), and minority temperature and concentration is analyzed. These results form the basis for comparisons with other codes, such as CURRAY, METS, TORIC, and AORSA.
Date: May 18, 2001
Creator: Rosenberg, A.; Menard, J.E. & and LeBlanc, B.P.
Partner: UNT Libraries Government Documents Department

Bootstrap Current for the Edge Pedestal Plasma in a Diverted Tokamak Geometry

Description: The edge bootstrap current plays a critical role in the equilibrium and stability of the steep edge pedestal plasma. The pedestal plasma has an unconventional and difficult neoclassical property, as compared with the core plasma. It has a narrow passing particle region in velocity space that can be easily modified or destroyed by Coulomb collisions. At the same time, the edge pedestal plasma has steep pressure and electrostatic potential gradients whose scale-lengths are comparable with the ion banana width, and includes a magnetic separatrix surface, across which the topological properties of the magnetic field and particle orbits change abruptly. A driftkinetic particle code XGC0, equipped with a mass-momentum-energy conserving collision operator, is used to study the edge bootstrap current in a realistic diverted magnetic field geometry with a self-consistent radial electric field. When the edge electrons are in the weakly collisional banana regime, surprisingly, the present kinetic simulation confirms that the existing analytic expressions [represented by O. Sauter et al. , Phys. Plasmas 6 , 2834 (1999)] are still valid in this unconventional region, except in a thin radial layer in contact with the magnetic separatrix. The agreement arises from the dominance of the electron contribution to the bootstrap current compared with ion contribution and from a reasonable separation of the trapped-passing dynamics without a strong collisional mixing. However, when the pedestal electrons are in plateau-collisional regime, there is significant deviation of numerical results from the existing analytic formulas, mainly due to large effective collisionality of the passing and the boundary layer trapped particles in edge region. In a conventional aspect ratio tokamak, the edge bootstrap current from kinetic simulation can be significantly less than that from the Sauter formula if the electron collisionality is high. On the other hand, when the aspect ratio is close to unity, the collisional ...
Date: August 10, 2012
Creator: Koh, S.; Chang, C. S.; Ku, S.; Menard, J. E.; Weitzner, H. & Choe, W.
Partner: UNT Libraries Government Documents Department

Real-time Equilibrium Reconstruction and Isoflux Control of Plasma Shape and Position in the National Spherical Torus Experiment (NSTX)

Description: The implementation of the rtEFIT-isoflux algorithm in the digital control system for NSTX has led to improved ability to control the plasma shape. In particular, it has been essential for good gap control for radio-frequency experiments, for control of drsep in H-mode studies, and for X-point height control and {kappa} control in a variety of experiments.
Date: August 11, 2004
Creator: Mueller, D.; Gates, D.A.; Menard, J.E.; Ferron, J.R. & Sabbagh, S.A.
Partner: UNT Libraries Government Documents Department

Implementation of BN Control in the National Spherical Torus Experiment

Description: We have designed and constructed a system for control of the normalized B in the National Spherical Torus Experiment [M. Ono, et al., Nuclear Fusion 40, 557 (2000)]. A PID operator is applied to the difference between the present value of B N (from realtime equilibrium reconstruction) and a time-dependent request, in order to calculate the required injected power. This injected power request is then turned into modulations of the neutral beams. The details of this algorithm are described, including the techniques used to develop the appropriate control gains. Example uses of the system are shown
Date: September 15, 2012
Creator: Gerhardt, S.; Bell, M. G.; Cropper, M.; Gates, D. A.; Koleman, E.; Lawson, J. et al.
Partner: UNT Libraries Government Documents Department

Comparison of Options for a Pilot Plant Fusion Nuclear Mission

Description: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The pilot plant mission encompassed component test and fusion nuclear science missions plus the requirement to produce net electricity with high availability in a device designed to be prototypical of the commercial device. Three magnetic configuration options were developed around this mission: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS). With the completion of the study and separate documentation of each design option a question can now be posed; how do the different designs compare with each other as candidates for meeting the pilot plant mission? In a pro/con format this paper will examine the key arguments for and against the AT, ST and CS magnetic configurations. Key topics addressed include: plasma parameters, device configurations, size and weight comparisons, diagnostic issues, maintenance schemes, availability influences and possible test cell arrangement schemes.
Date: August 27, 2012
Creator: Brown, T; Goldston, R J; El-Guebaly, L; Kessel, C; Neilson, G H; Malang, S et al.
Partner: UNT Libraries Government Documents Department

Detection of Disruptions in the High-β Spherical Torus NSTX

Description: This paper describes the prediction of disruptions based on diagnostic data in the high-β spherical torus NSTX [M. Ono, et al., Nuclear Fusion 40 , 557 (2000)]. The disruptive threshold values on many signals are examined. In some cases, raw diagnostic data can be used as a signal for disruption prediction. In others, the deviations of the plasma data from simple models provides the signal used to determine the proximity to disruption. However, no single signal and threshold value can form the basis for disruption prediction in NSTX; thresholds that produce an acceptable false positive rate have too large a missed or late warning rate, while combinations that produce an acceptable rate of missed or late warnings have an unacceptable false positive rate. To solve this problem, a novel means of combining multiple threshold tests has been developed. After being properly tuned, this algorithm can produce a false positive rate of 2.8%, with a late warning rate of 3.7% when applied to a database of ~2000 disruptions collected from three run campaigns. Furthermore, many of these false positives are triggered by near-disruptive MHD events that might indeed be disruptive in larger plasmas with more stored energy. However, the algorithm is less efficient at detecting the MHD event that prompts the disruption process.
Date: January 16, 2013
Creator: Gerhardt, S P; Bell, R E; LeBlanc, B P; Menard, J E; Mueller, D; Roquemore, A L et al.
Partner: UNT Libraries Government Documents Department

Disruptions, Disruptivity, and Safer Operating Windows in the High-β Spherical Torus NSTX

Description: This paper discusses disruption rates, disruption causes, and disruptivity statistics in the high- βN National Spherical Torus Experiment (NSTX) [M. Ono, et al. Nuclear Fusion 40, 557 (2000)]. While the overall disruption rate is rather high, configurations with high βN , moderate q*, strong boundary shaping, sufficient rotation, and broad pressure and current profiles are found to have the lowest disruptivity; active n=1 control further reduces the disruptivity. The disruptivity increases rapidly for q*<2.7, which is substantially above the ideal MHD current limit. In quiescent conditions, qmin >1.25 is generally acceptable for avoiding the onset of core rotating n=1 kink/tearing modes; when EPM and ELM disturbances are present, the required qmin for avoiding those modes is raised to ~1.5. The current ramp and early flat-top phase of the discharges are prone to n=1 core rotating modes locking to the wall, leading to a disruption. Small changes to the discharge fueling during this phase can often mitigate the rotation damping associated with these modes and eliminate the disruption. The largest stored energy disruptions are those that occur at high current when a plasma current rampdown is initiated incorrectly.
Date: September 27, 2012
Creator: Gerhardt, S P; Diallo, A; Gates, D; LeBlanc, B P; Menard, J E; Mueller, D et al.
Partner: UNT Libraries Government Documents Department

Disruptions, Disruptivity, and Safer Operating Windows in the High-β Spherical Torus NSTX

Description: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant. The mission of the pilot plant was set to encompass component test and fusion nuclear science missions yet produce net electricity with high availability in a device designed to be prototypical of the commercial device. The objective of the study was to evaluate three different magnetic configuration options, the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS) in an effort to establish component characteristics, maintenance features and the general arrangement of each candidate device. With the move to look beyond ITER the fusion community is now beginning to embark on DEMO reactor studies with an emphasis on defining configuration arrangements that can meet a high availability goal. This paper reviews the AT pilot plant design, detailing the selected maintenance approach, the device arrangement and sizing of the in-vessel components. Details of interfacing auxiliary systems and services that impact the ability to achieve high availability operations will also be discussed.
Date: September 26, 2012
Creator: Brown, T; Goldston, R J; El-Guebaly, L; Kessel, C; Neilson, G H; Malang, S et al.
Partner: UNT Libraries Government Documents Department

Physics Design of the National High-power Advanced Torus Experiment

Description: Moving beyond ITER toward a demonstration power reactor (Demo) will require the integration of stable high fusion gain in steady-state, advanced methods for dissipating very high divertor heat-fluxes, and adherence to strict limits on in-vessel tritium retention. While ITER will clearly address the issue of high fusion gain, and new and planned long-pulse experiments (EAST, JT60-SA, KSTAR, SST-1) will collectively address stable steady-state highperformance operation, none of these devices will adequately address the integrated heat-flux, tritium retention, and plasma performance requirements needed for extrapolation to Demo. Expressing power exhaust requirements in terms of Pheat/R, future ARIES reactors are projected to operate with 60-200MW/m, a Component Test Facility (CTF) or Fusion Development Facility (FDF) for nuclear component testing (NCT) with 40-50MW/m, and ITER 20-25MW/m. However, new and planned long-pulse experiments are currently projected to operate at values of Pheat/R no more than 16MW/m. Furthermore, none of the existing or planned experiments are capable of operating with very high temperature first-wall (Twall = 600-1000C) which may be critical for understanding and ultimately minimizing tritium retention with a reactor-relevant metallic first-wall. The considerable gap between present and near-term experiments and the performance needed for NCT and Demo motivates the development of the concept for a new experiment — the National High-power advanced-Torus eXperiment (NHTX) — whose mission is to study the integration of a fusion-relevant plasma-material interface with stable steady-state high-performance plasma operation.
Date: July 18, 2007
Creator: Menard, J E; Fu, G -Y; Gorelenkov, N; Kaye, S M; Kramer, G; Maingi, R et al.
Partner: UNT Libraries Government Documents Department

Momentum Transport Studies in High E x B Shear Plasmas in NSTX

Description: Experiments have been conducted on NSTX to study both steady state and perturbative mo mentum transport. These studies are unique in their parameter space under investigation, where the low aspect ratio of NSTX results in rapid plasma rotation with E x B shearing rates high enough to suppress low-k turbulence. In some cases, the ratio of momentum to energy confinement time is found to exceed five. Momentum pinch velocities of order 10-40 m/s are inferred from the measured angular momentum flux evolution after non-resonant magnetic perturbations are applied to brake the plasma.
Date: June 26, 2008
Creator: Solomon, W M; Bell, R E; LeBlanc, B P; Menard, J E; Rewoldt, G; Wang, W et al.
Partner: UNT Libraries Government Documents Department

β-Suppression of Alfvén Cascade Modes in the National Spherical Torus Experiment

Description: The coupling of Alfvén Cascade (AC) modes or reversed-shear Alfvén eigenmodes (rsAE) to Geodesic Acoustic Modes (GAM) implies that the range of the AC frequency sweep is reduced as the electron β is increased. This model provides an explanation for the otherwise surprising absence of AC modes in reverse shear NSTX plasmas, given the rich spectrum of beam-driven instabilities typically seen in NSTX. In experiments done at very low β to investigate this prediction, AC modes were seen, and as the βe was increased from shot to shot, the range of the AC frequency sweep was reduced, in agreement with this theoretical prediction.
Date: June 29, 2007
Creator: Fredrickson, E D; Gorelenkov, N N; Heidbrink, W W; Kubota, S; Levinton, F M; Yuh, H et al.
Partner: UNT Libraries Government Documents Department

Ideal MHD Stability Characteristics of Advanced Operating Regimes in Spherical Torus Plasmas and the Role of High Harmonic Fast Waves

Description: The ARIES reactor study group has found an economically attractive ST-based reactor configuration with: A = 1.6, {kappa} = 3.4, {delta} = 0.65, {beta} = 50%, {beta}{sub N} = 7.3, f{sub BS} = 0.95, R{sub 0} = 3.2 meters, B{sub t0} = 2.08 Tesla, and I{sub P} = 28.5 MA which yields a cost of electricity of approximately 80mils/kWh. MHD stability analysis finds that a broad pressure profile is optimal for wall-stabilizing the pressure driven kink modes typical of such configurations, and that wall stabilization is crucial to achieving the high {beta} needed for an economical power plant. The 6MW high-harmonic fast wave system presently being installed on NSTX should allow real-time control of the plasma {beta}, and in combination with NBI may permit experimental investigations of the effect of pressure profile peaking on MHD stability in the near-term. In the longer term, ejection of ions through resonant interaction with HHFW might be used to induce a controllable edge radial electric field with potentially interesting effects on edge MHD and confinement.
Date: June 1, 1999
Creator: Kessel, C.E.; Manickam, J.; Menard, J.E.; Jardin, S.C. & others], S.M. Kaye
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

Shape Optimization for DIII-D Advanced Tokamak Plasmas

Description: The advanced tokamak program on DIII-D is targeting the full integration of high-beta and high-bootstrap/noninductive current fraction for long-pulse lengths and the high confinement consistent with these features. Central to achieving these simultaneously is access to the highest ideal beta limits possible to maximize the headroom for experimental operation with RWM control. A study of the ideal-MHD stability is done for plasmas modeled after DIII-D advanced tokamak plasmas, varying the plasma elongation, triangularity, and outboard squareness. The highest beta(sub)N limits reach 6-7 for the n=1 kink mode for all elongation, outer squareness values, and plasma triangularity equals 0.8.
Date: July 30, 2003
Creator: Kesse, C.E.; Ferron, J.R.; Greenfield, C.M.; Menard, J.E. & Taylor, T.S.
Partner: UNT Libraries Government Documents Department

Modeling of Neoclassical Tearing Mode Stability for Generalized Toroidal Geometry

Description: Neoclassical tearing modes (NTMs) can lead to disruption and loss of confinement. Previous analysis of these modes used large aspect ratio, low beta (plasma pressure/magnetic pressure) approximations to determine the effect of NTMs on tokamak plasmas. A more accurate tool is needed to predict the onset of these instabilities. As a follow-up to recent theoretical work, a code has been written which computes the tearing mode island growth rate for arbitrary tokamak geometry. It calls PEST-3 [A. Pletzer et al., J. Comput. Phys. 115, 530 (1994)] to compute delta prime, the resistive magnetohydrodynamic (MHD) matching parameter. The code also calls the FLUXGRID routines in NIMROD [A.H. Glasser et al., Plasma Phys. Controlled Fusion 41, A747 (1999)] for Dnc, DI and DR [C.C. Hegna, Phys. Plasmas 6, 3980 (1999); A.H. Glasser et al., Phys. Fluids 18, 875 (1975)], which are the bootstrap current driven term and the ideal and resistive interchange mode criterion, respectively. In addition to these components, the NIMROD routines calculate alphas-H, a new correction to the Pfirsch-Schlter term. Finite parallel transport effects were added and a National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)] equilibrium was analyzed. Another program takes the output of PEST-3 and allows the user to specify the rational surface, island width, and amount of detail near the perturbed surface to visualize the total helical flux. The results of this work will determine the stability of NTMs in an spherical torus (ST) [Y.-K.M. Peng et al., Nucl. Fusion 26, 769 (1986)] plasma with greater accuracy than previously achieved.
Date: August 21, 2002
Creator: Rosenberg, A.L.; Gates, D.A.; Pletzer, A.; Menard, J.E.; Kruger, S.E.; Hegna, C.C. et al.
Partner: UNT Libraries Government Documents Department

Mission and Readiness Assessment for Fusion Nuclear Facilities

Description: Magnetic fusion development toward DEMO will most likely require a number of fusion nuclear facilities (FNF), intermediate between ITER and DEMO, to test and validate plasma and nuclear technologies and to advance the level of system integration. The FNF mission space is wide, ranging from basic materials research to net electricity demonstration, so there is correspondingly a choice among machine options, scope, and risk in planning such a step. Readiness requirements to proceed with a DEMO are examined, and two FNF options are assessed in terms of the contributions they would make to closing DEMO readiness gaps, and their readiness to themselves proceed with engineering design about ten years from now. An advanced tokamak (AT) pilot plant with superconducting coils and a mission to demonstrate net electricity generation would go a long way toward DEMO. As a next step, however, a pilot plant would entail greater risk than a copper-coil FNSF-AT with its more focussed mission and technology requirements. The stellarator path to DEMO is briefly discussed. Regardless of the choice of FNF option, an accompanying science and technology development program, also aimed at DEMO readiness, is absolutely essential.
Date: December 12, 2012
Creator: Neilson, G. H.; Brown, T. G.; Gates, D. A.; Kessel, C. E.; Menard, J. E.; Prager, S. C. et al.
Partner: UNT Libraries Government Documents Department

Electron Gyro-scale Fluctuation Measurements in National Spherical Torus Experiment H-mode Plasmas

Description: A collective scattering system has measured electron gyro-scale fluctuations in National Spherical Torus Experiment (NSTX) H-mode plasmas to investigate electron temperature gradient (ETG) turbulence. Observations and results pertaining to fluctuation measurements in ETGstable regimes, the toroidal field scaling of fluctuation amplitudes, the relation between between fluctuation amplitudes and transport quantities, and fluctuation magnitudes and k-spectra are presented. Collectively, the measurements provide insight and guidance for understanding ETG turbulence and anomalous electron thermal transport.
Date: August 10, 2009
Creator: Smith, D R; Lee, W; Mazzucato, E; Park, H K; Bell, R E; Domier, C W et al.
Partner: UNT Libraries Government Documents Department

Observations of Reduced Electron Gyro-scale Fluctuations in National Spherical Torus Experiment H-mode Plasmas with Large E × B Flow Shear

Description: Electron gyro-scale fluctuation measurements in National Spherical Torus Experiment (NSTX) H-mode plasmas with large toroidal rotation reveal fluctuations consistent with electron temper- ature gradient (ETG) turbulence. Large toroidal rotation in NSTX plasmas with neutral beam injection generates E × B flow shear rates comparable to ETG linear growth rates. Enhanced fluctuations occur when the electron temperature gradient is marginally stable with respect to the ETG linear critical gradient. Fluctuation amplitudes decrease when the E × B flow shear rate exceeds ETG linear growth rates. The observations indicate E × B flow shear can be an effective suppression mechanism for ETG turbulence.
Date: February 13, 2009
Creator: Smith, D. R.; Kaye, S. M.; Lee, W.; Mazzucato, E.; Park, H. K.; Bell, R. E. et al.
Partner: UNT Libraries Government Documents Department

Status of the Control System on the National Spherical Torus Experiment (NSTX)

Description: In 2003, the NSTX plasma control system was used for plasma shape control using real-time equilibrium reconstruction (using the rtEFIT code - J. Ferron, et al., Nucl. Fusion 38 1055 (1998)). rtEFIT is now in routine use for plasma boundary control [D. A. Gates, et al., submitted to Nuclear Fusion (2005)]. More recently, the system has been upgraded to support feedback control of the resistive wall mode (RWM). This paper describes the hardware and software improvements that were made in support of these physics requirements. The real-time data acquisition system now acquires 352 channels of data at 5kHz for each NSTX plasma discharge. The latency for the data acquisition, which uses the FPDP (Front Panel Data Port) protocol, is measured to be {approx}8 microseconds. A Stand-Alone digitizer (SAD), designed at PPPL, along with an FPDP Input multiplexing module (FIMM) allows for simple modular upgrades. An interface module was built to interface between the FPDP output of the NSTX control system and the legacy Power Conversion link (PCLINK) used for communicating with the PPPL power supplies (first used for TFTR). Additionally a module has been built for communicating with the switching power amplifiers (SPA) recently installed on NSTX. In addition to the hardware developments, the control software [D. Mastrovito, Fusion Eng. And Design 71 65 (2004)] on the NSTX control system has been upgraded. The control computer is an eight processor (8x333MHz G4) built by Sky Computers (Helmsford, MA). The device driver software for the hardware described above will be discussed, as well as the new control algorithms that have been developed to control the switching power supplies for RWM control. An important initial task in RWM feedback is to develop a reliable mode detection algorithm.
Date: August 5, 2005
Creator: Gates, D.A.; Ferron, J.R.; Bell, M.; Gibney, T.; Johnson, R.; Marsala, R.J. et al.
Partner: UNT Libraries Government Documents Department

Aspect Ratio Scaling of Ideal No-wall Stability Limits in High Bootstrap Fraction Tokamak Plasmas

Description: Recent experiments in the low aspect ratio National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40 (2000) 557] have achieved normalized beta values twice the conventional tokamak limit at low internal inductance and with significant bootstrap current. These experimental results have motivated a computational re-examination of the plasma aspect ratio dependence of ideal no-wall magnetohydrodynamic stability limits. These calculations find that the profile-optimized no-wall stability limit in high bootstrap fraction regimes is well described by a nearly aspect ratio invariant normalized beta parameter utilizing the total magnetic field energy density inside the plasma. However, the scaling of normalized beta with internal inductance is found to be strongly aspect ratio dependent at sufficiently low aspect ratio. These calculations and detailed stability analyses of experimental equilibria indicate that the nonrotating plasma no-wall stability limit has been exceeded by as much as 30% in NSTX in a high bootstrap fraction regime.
Date: November 25, 2003
Creator: Menard, J. E.; Bell, M. G.; Bell, R. E.; Gates, D. A.; Kaye, S. M.; LeBlanc, B. P. et al.
Partner: UNT Libraries Government Documents Department

Plasma Shape Control on the National Spherical Torus Experiment (NSTX) using Real-time Equilibrium Reconstruction

Description: Plasma shape control using real-time equilibrium reconstruction has been implemented on the National Spherical Torus Experiment (NSTX). The rtEFIT code originally developed for use on DIII-D was adapted for use on NSTX. The real-time equilibria provide calculations of the flux at points on the plasma boundary, which is used as input to a shape control algorithm known as isoflux control. The flux at the desired boundary location is compared to a reference flux value, and this flux error is used as the basic feedback quantity for the poloidal-field coils on NSTX. The hardware that comprises the control system is described, as well as the software infrastructure. Examples of precise boundary control are also presented.
Date: April 15, 2005
Creator: Gates, D.A.; Ferron, J.R.; Bell, M.; Gibney, T.; Johnson, R.; Marsala, R.J. et al.
Partner: UNT Libraries Government Documents Department