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Dynamics of the Disruption Halo Current Toroidal Asymmetry in NSTX

Description: This paper describes the dynamics of disruption halo current non-axisymmetries in the lower divertor of the National Spherical Torus Experiment [M. Ono, et al. Nuclear Fusion 40, 557 (2000)]. While. The halo currents typically have a strongly asymmetric structure where they enter the divertor floor, and this asymmetry has been observed to complete up to 7 toroidal revolutions over the duration of the halo current pulse. However, the rotation speed and toroidal extend of the asymmetry can vary significantly during the pulse. The rotation speed, halo current pulse duration, and total number of revolutions tend to be smaller in cases with large halo currents. The halo current pattern is observed to become toroidally symmetric at the end of the halo current pulse. It is proposed that this symmeterization is due to the loss of most or all of the closed field line geometry in the final phase of the vertical displacement event.
Date: September 27, 2012
Creator: Gerhardt, S.P.
Partner: UNT Libraries Government Documents Department

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 Disruption Halo Currents in the National Spherical Torus Experiment

Description: This paper describes the general characteristics of disruptions halo currents in the National Spherical Torus Experiment [M. Ono, et al. Nuclear Fusion 40, 557 (2000)]. The commonly observed types of vertical motion and resulting halo current patterns are described, and it is shown that plasma discharges developing between components can facilitate halo current flow. The halo current fractions and toroidal peaking factors at various locations in the device are presented. The maximum product of these two metrics for localized halo current measurements is always significantly less than the worst-case expectations from conventional aspect ratio tokamaks (which are typically written in terms of the total halo current). The halo current fraction and impulse is often largest in cases with the fastest plasma current quenches and highest quench rates. The effective duration of the halo current pulse is comparable to or shorter than the plasma current quench time. The largest halo currents have tended to occur in lower β and lower elongation plasmas. The sign of the poloidal halo current is reversed when the toroidal field direction is reversed.
Date: April 25, 2012
Creator: Gerhardt, S. P.; Menard, J.; Sabbagh, S. & Scotti, F.
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

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

Strike Point Control for the National Spherical Torus Experiment (NSTX)

Description: This paper presents the first control algorithm for the inner and outer strike point position for a Spherical Torus (ST) fusion experiment and the performance analysis of the controller. A liquid lithium divertor (LLD) will be installed on NSTX which is believed to provide better pumping than lithium coatings on carbon PFCs. The shape of the plasma dictates the pumping rate of the lithium by channeling the plasma to LLD, where strike point location is the most important shape parameter. Simulations show that the density reduction depends on the proximity of strike point to LLD. Experiments were performed to study the dynamics of the strike point, design a new controller to change the location of the strike point to desired location and stabilize it. The most effective PF coils in changing inner and outer strike points were identified using equilibrium code. The PF coil inputs were changed in a step fashion between various set points and the step response of the strike point position was obtained. From the analysis of the step responses, PID controllers for the strike points were obtained and the controller was tuned experimentally for better performance. The strike controller was extended to include the outer-strike point on the inner plate to accommodate the desired low outer-strike points for the experiment with the aim of achieving "snowflake" divertor configuration in NSTX.
Date: July 9, 2010
Creator: Kolemen, E.; Gates, D. A.; Rowley, C. W.; Kasdin, N. J.; Kallman, J.; Gerhardt, S. et al.
Partner: UNT Libraries Government Documents Department

Physics Design of a 28 GHz Electron Heating System for the National Spherical Torus Experiment Upgrade

Description: A megawatt-level, 28 GHz electron heating system is being designed to support non-inductive (NI) plasma current (I{sub p}) start-up and local heating and current drive (CD) in H-mode discharges in the National Spherical Torus Experiment Upgrade (NSTX-U). The development of fully NI I{sub p} start-up and ramp-up is an important goal of the NSTX-U research program. 28 GHz electron cyclotron (EC) heating is predicted to rapidly increase the central electron temperature (T{sub e}(0)) of low density NI plasmas generated by Coaxial Helicity Injection (CHI). The increased T{sub e}(0) will significantly reduce the Ip decay rate of CHI plasmas, allowing the coupling of fast wave heating and neutral beam injection. Also 28 GHz electron Bernstein wave (EBW) heating and CD can be used during the I{sub p} flat top in NSTX-U discharges when the plasma is overdense. Ray tracing and Fokker-Planck numerical simulation codes have been used to model EC and EBW heating and CD in NSTX-U. This paper presents a pre-conceptual design for the 28 GHz heating system and some of the results from the numerical simulations.
Date: July 9, 2013
Creator: Taylor, G.; Bertelli, N.; Ellis, R. A.; Gerhardt, S. P.; Harvey, R. W.; Hosea, J. C. et al.
Partner: UNT Libraries Government Documents Department

Characteristics of Short Wavelength Compressional Alfven Eigenmodes

Description: Most Alfvenic activity in the frequency range between Toroidal Alfven Eigenmodes and roughly one half of the ion cyclotron frequency on NSTX [M. Ono, et al., Nucl. Fusion 40 (2000) 557], that is, approximately 0.3 MHz up to ≈ 1.2 MHz, are modes propagating counter to the neutral beam ions. These have been modeled as Compressional and Global Alfven Eigenmodes (CAE and GAE) and are excited through a Doppler-shifted cyclotron resonance with the beam ions. There is also a class of co-propagating modes at higher frequency than the counter-propagating CAE and GAE. These modes have been identified as CAE, and are seen mostly in the company of a low frequency, n=1 kink-like mode. In this paper we present measurements of the spectrum of these high frequency CAE (hfCAE), and their mode structure. We compare those measurements to a simple model of CAE and present evidence of a curious non-linear coupling of the hfCAE and the low frequency kink-like mode.
Date: December 19, 2012
Creator: Fredrickson, E. D.; Podesta, M.; Bortolon, A.; Crocker, N. A.; Gerhardt, S. P.; Bell, R. E. et al.
Partner: UNT Libraries Government Documents Department

Relationship Between Onset Thresholds, Trigger Types, and Rotation Shear for the m/n=2/1 Neoclassical Tearing Mode in a High-β Spherical Torus

Description: The onset conditions for the m/n=2/1 neoclassical tearing mode (NTM) are studied in terms of neoclassical drive, triggering instabilities, and toroidal rotation or rotation shear, in the spherical torus NSTX [M. Ono, et al., Nuclear Fusion 40, 557 (2000)]. There are three typical onset conditions for these modes, given in order of increasing neoclassical drive required for mode onset: triggering by energetic particle modes, triggering by edge localized modes, and cases where the modes appear to grow without a trigger. In all cases, the required drive increases with toroidal rotation shear, implying a stabilizing effect from the shear.
Date: February 24, 2009
Creator: Gerhardt, S. P.; Brennan, D. P.; Buttery, R.; La Haye, R. J.; Sabbagh, S.; Strait, E. et al.
Partner: UNT Libraries Government Documents Department

First Observation Of ELM Pacing With Vertical Jogs In A Spherical Torus

Description: Experiments in a number of conventional aspect ratio tokamaks have been successful in pacing edge localized modes (ELMs) by rapid vertical jogging of the plasma. This paper demonstrates the first pacing of ELMs in a spherical torus plasma. Applied 30 Hz vertical jogs synchronized the ELMs with the upward motion of the plasma. 45 Hz jogs also lead to an increase in the ELM frequency, though the synchronization of the ELMs and jogs was unclear. A reduction in the ELM energy was observed at the higher driven ELM frequencies. __________________________________________________
Date: July 15, 2010
Creator: Gerhardt, S. P.; Canik, J. M.; Maingi, R.; Bell, R.; Gates, D.; Goldston, R. et al.
Partner: UNT Libraries Government Documents Department

Transient Enhancement ('Spike-on-Tail') Observed on Neutral-Beam-Injected Energetic Ion Spectra Using the E||B Neutral Particle Analyzer in the National Spherical Torus Experiment

Description: An increase of up to four-fold in the E||B Neutral Particle Analyzer (NPA) charge exchange neutral flux localized at the Neutral Beam (NB) injection full energy is observed in the National Spherical Torus Experiment (NSTX). Termed the High-Energy Feature (HEF), it appears on the NB-injected energetic ion spectrum only in discharges where tearing or kink-type modes (f < 10 kHz) are absent, TAE activity (f ~ 10-150 kHz) is weak (δBrms < 75 mGauss) and CAE/GAE activity (f ~ 400 – 1200 kHz) is robust. The feature exhibits a growth time of ~ 20 - 80 ms and occasionally develops a slowing down distribution that continues to evolve over periods of 100's of milliseconds, a time scale long compared with the typical ~ 10's ms equilibration time of the NB injected particles. The HEF is observed only in H-mode (not L-mode) discharges with injected NB power of 4 MW or greater and in the field pitch range v||/v ~ 0.7 – 0.9; i.e. only for passing (never trapped) energetic ions. The HEF is suppressed by vessel conditioning using lithium deposition at rates ~ 100 mg/shot, a level sufficient to suppress ELM activity. Increases of ~ 10 - 30 % in the measured neutron yield and total stored energy are observed to coincide with the feature along with broadening of measured Te(r), Ti(r) and ne(r) profiles. However, TRANSP analysis shows that such increases are driven by plasma profile changes and not the HEF phenomenon itself. Though a definitive mechanism has yet to be developed, the HEF appears to be caused by a form of TAE/CAE wave-particle interaction that distorts of the NB fast ion distribution in phase space.
Date: June 1, 2010
Creator: Medley, S. S.; Gorelenkov, N. N.; Bell, R. E.; Fredrickson, E. D.; Gerhardt, S. P.; LeBlanc, B. P. 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

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} &lt; 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

Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

Description: Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW~1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with longpulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.
Date: September 26, 2008
Creator: Kugel, W.; Bell, M.; Berzak,L.; Brooks, A.; Ellis, R.; Gerhardt, S. et al.
Partner: UNT Libraries Government Documents Department

Evaporated Lithium Surface Coatings in NSTX

Description: Two lithium evaporators were used to evaporate more than 100 g of lithium on to the NSTX lower divertor region. Prior to each discharge, the evaporators were withdrawn behind shutters, where they also remained during the subsequent HeGDC applied for periods up to 9.5 min. After the HeGDC, the shutters were opened and the LITERs were reinserted to deposit lithium on the lower divertor target for 10 min, at rates of 10-70 mg/min, prior to the next discharge. The major improvements in plasma performance from these lithium depositions include: 1) plasma density reduction as a result of lithium deposition; 2) suppression of ELMs; 3) improvement of energy confinement in a low-triangularity shape; 4) improvement in plasma performance for standard, high-triangularity discharges; 5) reduction of the required HeGDC time between discharges; 6) increased pedestal electron and ion temperature; 7) reduced SOL plasma density; and 8) reduced edge neutral density.
Date: April 9, 2009
Creator: Kugel, H. W.; Mansfield, D.; Maingi, R.; Bel, M. G.; Bell, R. E.; Allain, J. P. et al.
Partner: UNT Libraries Government Documents Department

Demonstration of Tokamak Ohmic Flux Saving by Transient Coaxial Helicity Injection on NSTX

Description: Transient Coaxial Helicity Injection (CHI) started discharges in NSTX have attained peak currents up to 300 kA and when these discharges are coupled to induction, it has produced up to 200 kA additional current over inductive-only operation. CHI in NSTX has shown to be energetically quite efficient, producing a plasma current of about 10 A/Joule of capacitor bank energy. In addition, for the first time, the CHI produced toroidal current that couples to induction continues to increase with the energy supplied by the CHI power supply at otherwise similar values of the injector flux, indicating the potential for substantial current generation capability by CHI in NSTX and in future toroidal devices. __________________________________________________
Date: April 23, 2010
Creator: Raman, R.; Mueller, D.; Nelson, B. A.; Jarboe, T. R.; Gerhardt, S.; Kugel, H. W. et al.
Partner: UNT Libraries Government Documents Department

NSTX Plasma Response to Lithium Coated Divertor

Description: NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, <0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.
Date: January 21, 2011
Creator: Kugel, H. W.; Bell, M. G.; Allain, J. P.; Bell, R. E.; Ding, S.; Gerhardt, S. P. et al.
Partner: UNT Libraries Government Documents Department