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D A Gates,
C Kessel,
J Menard,
G Taylor,
J R Wilson,
M G Bell,
R E Bell,
S Bernabei,
J Bialek,
T Biewer, [......],
M Wade,
R White,
J Wilgen,
M Williams,
W Zhu,
S J Zweben,
R Akers,
P Beiersdorfer,
R Betti,
T Bigelow
The Review of scientific instruments 12/2009; 80(12):129901. · 1.52 Impact Factor
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[show abstract]
[hide abstract]
ABSTRACT: A variety of magnetohydrodynamic (MHD) phenomena have been observed on NSTX. Many of these affect fast particle losses, which are of major concern for future burning plasma experiments. Usual diagnostics for studying these phenomena are arrays of Mirnov coils for magnetic oscillations and p-i-n diode arrays for soft x-ray emission from the plasma core. Data reported here are from a unique fast soft x-ray imaging camera (FSXIC) with a wide-angle (pinhole) tangential view of the entire plasma minor cross section. The camera provides a 64x64 pixel image, on a charge coupled device chip, of light resulting from conversion of soft x rays incident on a phosphor to the visible. We have acquired plasma images at frame rates of 1-500 kHz (300 frames/shot) and have observed a variety of MHD phenomena: disruptions, sawteeth, fishbones, tearing modes, and edge localized modes (ELMs). New data including modes with frequency >90 kHz are also presented. Data analysis and modeling techniques used to interpret the FSXIC data are described and compared, and FSXIC results are compared to Mirnov and p-i-n diode array results.
The Review of scientific instruments 11/2008; 79(10):10E928. · 1.52 Impact Factor
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D. A. Gates,
R. Maingi,
J. Menard,
S. Kaye,
S. A. Sabbagh,
G. Taylor,
J. R. Wilson,
M. G. Bell,
R. E. Bell,
S. Bernabei, [......],
M. Wade,
R. White,
J. Wilgen,
M. Williams,
W. Zhu,
S. J. Zweben,
R. Akers,
P. Beiersdorfer,
R. Betti,
T. Bigelow
[show abstract]
[hide abstract]
ABSTRACT: The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including the stability of global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Improved shaping capability has been crucial to achieving βt ∼ 40%. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved elongation κ ∼ 2.8 and triangularity δ ∼ 0.8. Ideal MHD theory predicts increased stability at high values of shaping factor S ≡ q95Ip/(aBt), which has been observed at large values of the S ∼ 37[MA/(m∙T)] on NSTX. The behavior of ELMs is observed to depend on plasma shape. A description of the ELM regimes attained as shape is varied will be presented. Increased shaping is predicted to increase the bootstrap fraction at fixed Ip. The achievement of strong shaping has enabled operation with 1 s pulses with Ip = 1 MA, and for 1.6 s for Ip = 700 kA. Analysis of the noninductive current fraction as well as empirical analysis of the achievable plasma pulse length as elongation is varied will be presented. Data are presented showing a reduction in peak divertor heat load due to increasing in flux expansion.
Physics of Plasmas 05/2006; 13(5):056122-056122-7. · 2.15 Impact Factor
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D.A. Gates,
C. Kessel,
J. Menard,
G. Taylor,
J.R. Wilson,
M.G. Bell,
R.E. Bell,
S. Bernabei,
J. Bialek,
T. Biewer, [......],
M. Wade,
R. White,
J. Wilgen,
M. Williams,
W. Zhu,
S.J. Zweben,
R. Akers,
P. Beiersdorfer,
R. Betti,
T. Bigelow
[show abstract]
[hide abstract]
ABSTRACT: In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ~ 2.1 to κ ~ 2.6—approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βt with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ~ 0.8) at elevated elongation (κ ~ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented.
Nuclear Fusion 01/2006; 46(3):S22. · 4.09 Impact Factor
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S.M. Kaye,
M.G. Bell,
R.E. Bell,
S. Bernabei,
J. Bialek,
T. Biewer,
W. Blanchard,
J. Boedo,
C. Bush,
M.D. Carter, [......],
M. Schaffer,
I. Semenov,
K.C. Shaing,
M.A. Shapiro,
K. Shinohara,
P. Sichta,
X. Tang,
R. Vero,
D. Walker,
W. Wampler
[show abstract]
[hide abstract]
ABSTRACT: The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfvén eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.
Nuclear Fusion 10/2005; 45(10):S168. · 4.09 Impact Factor
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E.J. Synakowski,
M.G. Bell,
R.E. Bell,
T. Bigelow,
M. Bitter,
W. Blanchard,
J. Boedo,
C. Bourdelle,
C. Bush,
D.S. Darrow, [......],
B.C. Stratton,
Y. Takase,
X. Tang,
R. Vero,
W.R. Wampler,
G.A. Wurden,
X.Q. Xu,
J.G. Yang,
L. Zeng,
W. Zhu
[show abstract]
[hide abstract]
ABSTRACT: A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7 MW) and high harmonic fast wave (HHFW) heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10 MW m−2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.
Nuclear Fusion 12/2003; 43(12):1653. · 4.09 Impact Factor
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Y.-K.M. Peng,
W. Reiersen,
S.M. Kaye,
S.C. Jardin,
J. Menard,
D. Gates, J. Robinson,
F. Dahlgren,
L.R. Grisham,
R. Majeski, [......],
J.A. Schmidt,
J.R. Wilson,
R. Woolley,
E.T. Cheng,
R.J. Cerbone,
D.J. Strickler,
J.D. Galambos,
I.N. Sviatoslavsky,
K.C. Shaing,
X. Wang
[show abstract]
[hide abstract]
ABSTRACT: The scientific parameters and the technology issues for a modest size spherical torus (ST) at 10 MA plasma current are discussed. This class of devices includes a DT-capable ST experiment (DTST, R0 = 1.2 m) for extended plasma performance tests for limited pulse lengths and neutron fluences, and a volume neutron source (VNS, R0 = 1.1 m) for steady state energy technology testing to high neutron fluences. The scientific issues of interest for DTST include non-inductive ramp-up of plasma current on a limited timescale (~30 s), the confinement needed for high Q burn, the behaviour of energetic particles, the physics and techniques to handle intense plasma exhaust, and the possibility of high performance plasma regimes free of disruptions or large disruption impact. Of further interest for the VNS would be steady state operation using large external current drive, possibly at a modest Q (~1-2), achieving significant neutron wall loading (~1 MW/m2) and a configuration relatively amenable to remote maintenance. A much longer timescale would be permitted in a VNS for non-inductive current ramp-up. The centre leg of the toroidal field coils, possibly multiturn for DTST and necessarily single turn for a VNS without significant nuclear shielding, presents technical and material issues of unique importance to the ST. Positive ion neutral beam injection and high harmonic fast wave (~80 MHz) heating and current drive systems already available are likely to be adequate for DTST following pulse length extension to ~50 s. Given an adequate physics database, the remaining enabling technologies needed for the VNS appear largely similar in nature to those of the ITER EDA design.
Nuclear Fusion 05/2002; 40(3Y):583. · 4.09 Impact Factor
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M. Ono,
S.M. Kaye,
Y.-K.M. Peng,
G. Barnes,
W. Blanchard,
M.D. Carter,
J. Chrzanowski,
L. Dudek,
R. Ewig,
D. Gates, [......], J. Robinson,
A.L. Roquemore,
P. Ryan,
S. Sabbagh,
D. Swain,
E.J. Synakowski,
M. Viola,
M. Williams,
J.R. Wilson,
NSTX Team
[show abstract]
[hide abstract]
ABSTRACT: The National Spherical Torus Experiment (NSTX) is being built at Princeton Plasma Physics Laboratory to test the fusion physics principles for the spherical torus concept at the MA level. The NSTX nominal plasma parameters are R0 = 85 cm, a = 67 cm, R/a ≥ 1.26, Bt = 3 kG, Ip = 1 MA, q95 = 14, elongation κ ≤ 2.2, triangularity δ ≤ 0.5 and a plasma pulse length of up to 5 s. The plasma heating/current drive tools are high harmonic fast wave (6 MW, 5 s), neutral beam injection (5 MW, 80 keV, 5 s) and coaxial helicity injection. Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes, including very high plasma β, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well and high pressure driven sheared flow. In addition, the NSTX programme plans to explore fully non-inductive plasma startup as well as a dispersive scrape-off layer for heat and particle flux handling.
Nuclear Fusion 05/2002; 40(3Y):557. · 4.09 Impact Factor
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M. Ono,
S.M. Kaye,
C. Neumeyer,
Y.-K.M. Peng,
M. Williams,
G. Barnes,
M. Bell,
J. Bialek,
T. Bigelow,
W. Blanchard, [......],
P. Ryan,
S. Sabbagh,
P. Sichta,
T. Stevenson,
D. Swain,
M. Viola,
A. Von Halle,
J.R. Wilson,
G. Wurden,
S. Zweben
[show abstract]
[hide abstract]
ABSTRACT: The NSTX (National Spherical Torus Experiment) facility located at
Princeton Plasma Physics Laboratory is the newest national fusion
science experimental facility for the restructured US Fusion Energy
Science Program. The NSTX project was approved in FY 97 as the first
proof-of-principle national fusion facility dedicated to the spherical
torus research. On Feb. 15, 1999, the first plasma was achieved 10 weeks
ahead of schedule. The project was completed on budget and with an
outstanding safety record. This paper gives an overview of the NSTX
facility construction and the initial plasma operations
Fusion Engineering, 1999. 18th Symposium on; 02/1999
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M. Ono,
M. G. Bell,
R. E. Bell,
T. Bigelow,
T. Bitter,
W. Blanchard,
D. S. Darrow,
E. D. Fredrickson,
D. A. Gates,
L. R. Grisham, [......],
P. Sichta,
D. Stotler,
B. C. Stratton,
Y. Takase,
W. R. Wampler,
G. A. Wurden,
X. Q. Xu,
J. G. Yang,
L. Zeng,
W. Zhu
[show abstract]
[hide abstract]
ABSTRACT: The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current I-p was successfully brought up to the design value of I MA on 14 December 1999. The planned plasma shaping parameters, elongation kappa = 1.6-2.2 and, triangularity delta = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons, were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 x 10(13) cm(-3), increasing the plasma energy to 59 kJ and the toroidal beta, beta (T), to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (P-NBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to beta (T) approximate to 18% at a plasma current of 1.1 MA.
Nuclear Fusion. 41(10):1435-1447.
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M. Ono,
M. G. Bell,
R. E. Bell,
T. Bigelow,
M. Bitter,
W. Blanchard,
J. Boedo,
C. Bourdelle,
C. Bush,
W. Choe, [......], J. Robinson,
P. Roney,
K. Shaing,
S. Shiraiwa,
P. Sichta,
D. Stotler,
B. C. Stratton,
R. Vero,
W. R. Wampler,
G. A. Wurden
[show abstract]
[hide abstract]
ABSTRACT: Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values beta(T) of up to 35% with a near unity central beta(T) have been obtained. NSTX will be exploring advanced regimes where beta(T) up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (similar to60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to I-p similar to 500 kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.
Plasma Physics and Controlled Fusion. 45:A335-A350.
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E. J. Synakowski,
M. G. Bell,
R. E. Bell,
T. Bigelow,
M. Bitter,
W. Blanchard,
J. Boedo,
C. Bourdelle,
C. Bush,
D. S. Darrow, [......],
B. C. Stratton,
Y. Takase,
X. Tang,
R. Vero,
W. R. Wampler,
G. A. Wurden,
X. Q. Xu,
J. G. Yang,
L. Zeng,
W. Zhu
[show abstract]
[hide abstract]
ABSTRACT: A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7 MW) and high harmonic fast wave (HHFW) heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with beta(T) equivalent to
/(B-T0(2)/2mu(0)) of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10 MW m(-2) has been measured in the H-mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.
Nuclear Fusion. 43(12):1653-1664.
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M. Ono,
S. M. Kaye,
C. Neumeyer,
Y. K. M. Peng,
M. Williams,
G. Barnes,
M. Bell,
J. Bialek,
T. Bigelow,
W. Blanchard, [......],
S. Sabbagh,
P. Sichta,
T. Stevenson,
D. Swain,
M. Viola,
A. Von Halle,
J. R. Wilson,
G. Wurden,
S. Zweben,
Nstx Team
[show abstract]
[hide abstract]
ABSTRACT: The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, the first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations.
18th Ieee/Npss Symposium on Fusion Engineering.
