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Sample records for high-beta nstx plasmas

  1. Internal Kink Mode Dynamics in HighNSTX Plasmas

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, R.E.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Medley, S.S.; Park, W.; Sabbagh, S.A.; Sontag, A.; Stutman, D.; Tritz, K.; Zhu, W.

    2004-01-01

    Saturated internal kink modes have been observed in many of the highest toroidal beta discharges of the National Spherical Torus Experiment (NSTX). These modes often cause rotation flattening in the plasma core, can degrade energy confinement, and in some cases contribute to the complete loss of plasma angular momentum and stored energy. Characteristics of the modes are measured using soft X-ray, kinetic profile, and magnetic diagnostics. Toroidal flows approaching Alfvenic speeds, island pressure peaking, and enhanced viscous and diamagnetic effects associated with high-beta may contribute to mode nonlinear stabilization. These saturation mechanisms are investigated for NSTX parameters and compared to experimental data

  2. Internal kink mode dynamics in highNSTX plasmas

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, R.E.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Medley, S.S.; Park, W.; Sabbagh, S.A.; Sontag, A.; Zhu, W.; Stutman, D.; Tritz, K.

    2005-01-01

    Saturated internal kink modes have been observed in many of the highest toroidal beta discharges of the National Spherical Torus Experiment (NSTX). These modes often cause rotation flattening in the plasma core, can degrade energy confinement, and in some cases contribute to the complete loss of plasma angular momentum and stored energy. Characteristics of the modes are measured using soft X-ray, kinetic profile, and magnetic diagnostics. Toroidal flows approaching Alfvenic speeds, island pressure peaking, and enhanced viscous and diamagnetic effects associated with high-beta may contribute to mode non-linear stabilization. These saturation mechanisms are investigated for NSTX parameters and compared to experiment. (author)

  3. Predictions and observations of global beta-induced Alfven-acoustic modes in JET and NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Gorelenkov, N N [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Berk, H L [Institute for Fusion Studies, University of Texas, Austin, TX 78712 (United States); Crocker, N A [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Fredrickson, E D [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Kaye, S [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Kubota, S [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Park, H [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Peebles, W [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Sabbagh, S A [Department of Applied Physics, Columbia University, New York, NY 10027-6902 (United States); Sharapov, S E [Euroatom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Stutmat, D [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Tritz, K [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Levinton, F M [Nova Photonics, One Oak Place, Princeton, NJ 08540 (United States); Yuh, H [Nova Photonics, One Oak Place, Princeton, NJ 08540 (United States)

    2007-12-15

    In this paper we report on observations and interpretations of a new class of global MHD eigenmode solutions arising in gaps in the low frequency Alfven-acoustic continuum below the geodesic acoustic mode frequency. These modes have been just reported (Gorelenkov et al 2007 Phys. Lett. 370 70-7) where preliminary comparisons indicate qualitative agreement between theory and experiment. Here we show a more quantitative comparison emphasizing recent NSTX experiments on the observations of the global eigenmodes, referred to as beta-induced Alfven-acoustic eigenmodes (BAAEs), which exist near the extrema of the Alfven-acoustic continuum. In accordance to the linear dispersion relations, the frequency of these modes may shift as the safety factor, q, profile relaxes. We show that BAAEs can be responsible for observations in JET plasmas at relatively low beta <2% as well as in NSTX plasmas at relatively high beta >20%. In NSTX plasma observed magnetic activity has the same properties as predicted by theory for the mode structure and the frequency. Found numerically in NOVA simulations BAAEs are used to explain the observed properties of relatively low frequency experimental signals seen in NSTX and JET tokamaks.

  4. High beta, Long Pulse, Bootstrap Sustained Scenarios on the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.

    2003-01-01

    Long-pulse, high-beta scenarios have been established on the National Spherical Torus Experiment (NSTX). Beta(sub)t(always equal to 2μ(sub)0· /B 2 (sub)t0) ∼ 35% has been achieved during transient discharges. The machine improvements that lead to these results, including error field reduction and high-temperature bakeout of plasma-facing components are described. The highest Beta(sub)t plasmas have high triangularity (delta = 0.8) and elongation (k = 2.0) at low-aspect ratio A always equal to R/a = 1.4. The strong shaping permits large values of normalized current, I(sub)N(always equal to I(sub)p /(aB(sub)t0)) approximately equal to 6 while maintaining moderate values of q(sub)95 = 4. Long-pulse discharges up to 1 sec in duration have been achieved with substantial bootstrap current. The total noninductive current drive can be as high as 60%, comprised of 50% bootstrap current and ∼10% neutral-beam current drive. The confinement enhancement factor H89P is in excess of 2.7. Beta(sub)N * H(sub)89P approximately or greater than 15 has been maintained for 8 * tau(sub)E ∼ 1.6 * tau(sub)CR, where tau(sub)CR is the relaxation time of the first radial moment of the toroidal current density. The ion temperature for these plasmas is significantly higher than that predicted by neoclassical theory

  5. High harmonic fast wave heating experiments on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bell, R.; Bitter, M.

    2001-01-01

    A radio frequency (rf) system has been installed on the National Spherical Torus Experiment (NSTX) with the aim of heating the plasma and driving plasma current. The system consists of six rf transmitters, a twelve element antenna and associated transmission line components to distribute and couple the power from the transmitters to the antenna elements in a fashion to allow control of the antenna toroidal wavenumber spectrum. To date, power levels up to 3.85 MW have been applied to the NSTX plasmas. The frequency and spectrum of the rf waves has been selected to heat electrons via Landau damping and transit time magnetic pumping. The electron temperature has been observed to increase from 400 to 900 eV with little change in plasma density resulting in a plasma stored energy of 59 kJ , a toroidal beta, β T =10% and a normalized beta, β n =2.7. (author)

  6. Advanced ST plasma scenario simulations for NSTX

    International Nuclear Information System (INIS)

    Kessel, C.E.; Synakowski, E.J.; Gates, D.A.; Kaye, S.M.; Menard, J.; Phillips, C.K.; Taylor, G.; Wilson, R.; Harvey, R.W.; Mau, T.K.

    2005-01-01

    Integrated scenario simulations are done for NSTX that address four primary milestones for developing advanced ST configurations: high β and high β N inductive discharges to study all aspects of ST physics in the high beta regime; non-inductively sustained discharges for flattop times greater than the skin time to study the various current drive techniques; non-inductively sustained discharges at high βfor flattop times much greater than a skin time which provides the integrated advanced ST target for NSTX; and non-solenoidal startup and plasma current rampup. The simulations done here use the Tokamak Simulation Code (TSC) and are based on a discharge 109070. TRANSP analysis of the discharge provided the thermal diffusivities for electrons and ions, the neutral beam (NB) deposition profile and other characteristics. CURRAY is used to calculate the High Harmonic Fast Wave (HHFW) heating depositions and current drive. GENRAY/CQL3D is used to establish the heating and CD deposition profiles for electron Bernstein waves (EBW). Analysis of the ideal MHD stability is done with JSOLVER, BALMSC, and PEST2. The simulations indicate that the integrated advanced ST plasma is reachable, obtaining stable plasmas with β ∼ 40% at β N 's of 7.7-9, I P = 1.0 MA and B T = 0.35 T. The plasma is 100% non-inductive and has a flattop of 4 skin times. The resulting global energy confinement corresponds to a multiplier of H 98(y,2 ) = 1.5. The simulations have demonstrated the importance of HHFW heating and CD, EBW off-axis CD, strong plasma shaping, density control, and early heating/H-mode transition for producing and optimizing these plasma configurations (author)

  7. Beta-limiting MHD instabilities in improved performance NSTX spherical torus plasmas

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, M.G.; Bell, R.E.

    2003-01-01

    Global magnetohydrodynamic stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized beta and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during nor- mal operation and increased the maximum achievable beta. As a result of these improvements, peak beta values have reached (not simultaneously) β t = 35%, β N 6.5, N > = 4.5, β / l i =10, and β= 1.4. High β P operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 second with sustained periods of β N ∼ 6. Details of the β limit scalings and β-limiting instabilities in various operating regimes are described. (author)

  8. Predications and Observations of Global Beta-induced Alfven-acoustic Modes in JET and NSTX

    International Nuclear Information System (INIS)

    Gorelenkov, N.N.

    2008-01-01

    In this paper we report on observations and interpretations of a new class of global MHD eigenmode solutions arising in gaps in the low frequency Alfven-acoustic continuum below the geodesic acoustic mode frequency. These modes have been just reported (Gorelenkov et al 2007 Phys. Lett. 370 70-7) where preliminary comparisons indicate qualitative agreement between theory and experiment. Here we show a more quantitative comparison emphasizing recent NSTX experiments on the observations of the global eigenmodes, referred to as beta-induced Alfven-acoustic eigenmodes (BAAEs), which exist near the extrema of the Alfven-acoustic continuum. In accordance to the linear dispersion relations, the frequency of these modes may shift as the safety factor, q, profile relaxes. We show that BAAEs can be responsible for observations in JET plasmas at relatively low beta 20%. In NSTX plasma observed magnetic activity has the same properties as predicted by theory for the mode structure and the frequency. Found numerically in NOVA simulations BAAEs are used to explain the observed properties of relatively low frequency experimental signals seen in NSTX and JET tokamaks

  9. Advanced ST Plasma Scenario Simulations for NSTX

    International Nuclear Information System (INIS)

    Kessel, C.E.; Synakowski, E.J.; Gates, D.A.; Harvey, R.W.; Kaye, S.M.; Mau, T.K.; Menard, J.; Phillips, C.K.; Taylor, G.; Wilson, R.

    2004-01-01

    Integrated scenario simulations are done for NSTX [National Spherical Torus Experiment] that address four primary milestones for developing advanced ST configurations: high β and high β N inductive discharges to study all aspects of ST physics in the high-beta regime; non-inductively sustained discharges for flattop times greater than the skin time to study the various current-drive techniques; non-inductively sustained discharges at high β for flattop times much greater than a skin time which provides the integrated advanced ST target for NSTX; and non-solenoidal start-up and plasma current ramp-up. The simulations done here use the Tokamak Simulation Code (TSC) and are based on a discharge 109070. TRANSP analysis of the discharge provided the thermal diffusivities for electrons and ions, the neutral-beam (NB) deposition profile, and other characteristics. CURRAY is used to calculate the High Harmonic Fast Wave (HHFW) heating depositions and current drive. GENRAY/CQL3D is used to establish the heating and CD [current drive] deposition profiles for electron Bernstein waves (EBW). Analysis of the ideal-MHD stability is done with JSOLVER, BALMSC, and PEST2. The simulations indicate that the integrated advanced ST plasma is reachable, obtaining stable plasmas with β ∼ 40% at β N 's of 7.7-9, I P = 1.0 MA, and B T = 0.35 T. The plasma is 100% non-inductive and has a flattop of 4 skin times. The resulting global energy confinement corresponds to a multiplier of H 98(y,2) 1.5. The simulations have demonstrated the importance of HHFW heating and CD, EBW off-axis CD, strong plasma shaping, density control, and early heating/H-mode transition for producing and optimizing these plasma configurations

  10. Beta-limiting MHD Instabilities in Improved-performance NSTX Spherical Torus Plasmas

    International Nuclear Information System (INIS)

    J.E. Menard; M.G. Bell; R.E. Bell; E.D. Fredrickson D.A. Gates: S.M. Kaye; B.P. LeBlanc; R. Maingi; D. Mueller; S.A. Sabbagh; D. Stutman; C.E. Bush; D.W. Johnson; R. Kaita; H.W. Kugel; R.J. Maqueda; F. Paoletti; S.F Paul; M. Ono; Y.-K.M. Peng; C.H. Skinner; E.J. Synakowski; the NSTX Research Team

    2003-01-01

    Global magnetohydrodynamic stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized beta and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable beta. As a result of these improvements, peak beta values have reached (not simultaneously) β t = 35%, β N = 6.4, N > = 4.5, β N /l i = 10, and β P = 1.4. High β P operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 second with sustained periods of β N ∼ 6 above the ideal no-wall limit and near the with-wall limit. Details of the β limit scalings and β-limiting instabilities in various operating regimes are described

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

    International Nuclear Information System (INIS)

    Solomon, W.M.; Kaye, S.M.; Bell, S.M.; LeBlanc, B.P.; Menard, B.P.; Rewoldt, B.P.; Wang, W.; Levinton, F.M.; Yuh, H.; Sabbagh, S.A.

    2008-01-01

    Experiments have been conducted on NSTX to study both steady state and perturbative momentum 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

  12. Diagnostic Development for ST Plasmas on NSTX

    International Nuclear Information System (INIS)

    Johnson, D.

    2003-01-01

    Spherical tokamaks (STs) have much lower aspect ratio (a/R) and lower toroidal magnetic field, relative to tokamaks and stellarators. This paper will highlight some of the challenges and opportunities these features pose in the diagnosis of ST plasmas on the National Spherical Torus Experiment (NSTX), and discuss some of the corresponding diagnostic development that is underway. The low aspect ratio necessitates a small center stack, with tight space constraints and large thermal excursions, complicating the design of magnetic sensors in this region. The toroidal magnetic field on NSTX is less than or equal to 0.6 T, making it impossible to use ECE as a good monitor of electron temperature. A promising new development for diagnosing electron temperature is electron Bernstein wave (EBW) radiometry, which is currently being pursued on NSTX. A new high-resolution charge exchange recombination spectroscopy system is being installed. Since non-inductive current initiation and sustainment ar e top-level NSTX research goals, measurements of the current profile J(R) are essential to many planned experiments. On NSTX several modifications are planned to adapt the MSE technique to lower field, and two novel MSE systems are being prototyped. Several high speed 2-D imaging techniques are being developed, for viewing both visible and x-ray emission. The toroidal field is comparable to the poloidal field at the outside plasma edge, producing a large field pitch (>50 o ) at the outer mid-plane. The large shear in pitch angle makes some fluctuation diagnostics like beam emission spectroscopy very difficult, while providing a means of achieving spatial localization for microwave scattering investigations of high-k turbulence, which are predicted to be virulent for NSTX plasmas. A brief description of several of these techniques will be given in the context of the current NSTX diagnostic set

  13. High Harmonic Fast Wave Heating Experiments on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bell, R.; Bitter, M.; Bonoli, P.

    2000-01-01

    A radio frequency (rf) system has been installed on the National Spherical Torus Experiment (NSTX) with the aim of heating the plasma and driving plasma current. The system consists of six rf transmitters, a twelve element antenna and associated transmission line components to distribute and couple the power from the transmitters to the antenna elements in a fashion to allow control of the antenna toroidal wavenumber spectrum. To date, power levels up to 3.85 MW have been applied to the NSTX plasmas. The frequency and spectrum of the rf waves has been selected to heat electrons via Landau damping and transit time magnetic pumping. The electron temperature has been observed to increase from 400 to 900 eV with little change in plasma density resulting in a plasma stored energy of 59 kJ and a toroidal beta, bT , =10% and bn = 2.7

  14. Physics of integrated high-performance NSTX plasmas

    International Nuclear Information System (INIS)

    Menard, J. E.; Bell, M. G.; Bell, R. E.; Fredrickson, E. D.; Gates, D. A.; Heidbrink, W.; Kaita, R.; Kaye, S. M.; Kessel, C. E.; Kugel, H.; LeBlanc, B. P.; Lee, K. C.; Levinton, F. M.; Maingi, R.; Medley, S. S.; Mikkelsen, D. R.; Mueller, D.; Nishino, N.; Ono, M.; Park, H.; Park, W.; Paul, S. F.; Peebles, T.; Peng, M.; Raman, R.; Redi, M.; Roquemore, L.; Sabbagh, S. A.; Skiner, C. H.; Sontag, A.; Soukhanovskii, V.; Stratton, B.; Stutman, D.; Synakowski, E.; Takase, Y.; Taylor, G.; Tritz, K.; Wade, M.; Wilson, J. R.; Zhu, W.

    2005-01-01

    An overarching goal of magnetic fusion research is the integration of steady state operation with high fusion power density, high plasma β, good thermal and fast particle confinement, and manageable heat and particle fluxes to reactor internal components. NSTX has made significant progress in integrating and understanding the interplay between these competing elements. Sustained high elongation up to 2.5 and H-mode transitions during the I p ramp-up have increased β p and reduced l i at high current resulting in I p flat-top durations exceeding 0.8s for I p >0.8MA. These shape and profile changes delay the onset of deleterious global MHD activity yielding β N values >4.5 and β T ∼20% maintained for several current diffusion times. Higher ∫ N discharges operating above the non-wall limit are sustained via rotational stabilization of the RWM. H-mode confinement scaling factors relative to H98(y,2) span the range 1±0.4 for B T >4kG and show a stron (Nearly linear) residual scaling with B T . Power balance analysis indicates the electron thermal transport dominates the loss power in beam-heated H m ode discharges, but the core χ e can be significantly reduced through current profile modification consistent with reversed magnetic shear. Small ELM regimes have been obtained in high performance plasmas on NSTX, but the ELM type and associated pedestal energy loss are found to depend sensitively on the boundary elongation, magnetic balance, and edge collisionality. NPA data and TRANSP analysis suggest resonant interactions with mid-radius tearing modes may lead to large fast-ion transport. The associated fast-ion diffusion and/or loss likely impact(s) both the driven current and power deposition profiles from NBI heating. Results from experiments to initiate the plasma without the ohmic solenoid and integrated scenario with the TSC code will also be described. (Author)

  15. Progress toward commissioning and plasma operation in NSTX-U

    Science.gov (United States)

    Ono, M.; Chrzanowski, J.; Dudek, L.; Gerhardt, S.; Heitzenroeder, P.; Kaita, R.; Menard, J. E.; Perry, E.; Stevenson, T.; Strykowsky, R.; Titus, P.; von Halle, A.; Williams, M.; Atnafu, N. D.; Blanchard, W.; Cropper, M.; Diallo, A.; Gates, D. A.; Ellis, R.; Erickson, K.; Hosea, J.; Hatcher, R.; Jurczynski, S. Z.; Kaye, S.; Labik, G.; Lawson, J.; LeBlanc, B.; Maingi, R.; Neumeyer, C.; Raman, R.; Raftopoulos, S.; Ramakrishnan, R.; Roquemore, A. L.; Sabbagh, S. A.; Sichta, P.; Schneider, H.; Smith, M.; Stratton, B.; Soukhanovskii, V.; Taylor, G.; Tresemer, K.; Zolfaghari, A.; The NSTX-U Team

    2015-07-01

    The National Spherical Torus Experiment-Upgrade (NSTX-U) is the most powerful spherical torus facility at PPPL, Princeton USA. The major mission of NSTX-U is to develop the physics basis for an ST-based Fusion Nuclear Science Facility (FNSF). The ST-based FNSF has the promise of achieving the high neutron fluence needed for reactor component testing with relatively modest tritium consumption. At the same time, the unique operating regimes of NSTX-U can contribute to several important issues in the physics of burning plasmas to optimize the performance of ITER. NSTX-U further aims to determine the attractiveness of the compact ST for addressing key research needs on the path toward a fusion demonstration power plant (DEMO). The upgrade will nearly double the toroidal magnetic field BT to 1 T at a major radius of R0 = 0.93 m, plasma current Ip to 2 MA and neutral beam injection (NBI) heating power to 14 MW. The anticipated plasma performance enhancement is a quadrupling of the plasma stored energy and near doubling of the plasma confinement time, which would result in a 5-10 fold increase in the fusion performance parameter nτ T. A much more tangential 2nd NBI system, with 2-3 times higher current drive efficiency compared to the 1st NBI system, is installed to attain the 100% non-inductive operation needed for a compact FNSF design. With higher fields and heating powers, the NSTX-U plasma collisionality will be reduced by a factor of 3-6 to help explore the favourable trend in transport towards the low collisionality FNSF regime. The NSTX-U first plasma is planned for the Summer of 2015, at which time the transition to plasma operations will occur.

  16. Electron Bernstein Wave Research on NSTX and CDX-U

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C.; Jones, B.; Bell, G.L.; Bers, A.; Bigelow, T.S.; Carter, M.D.; Harvey, R.W.; Ram, A.K.; Rasmussen, D.A.; Smirnov, A.P.; Wilgen, J.B.; Wilson, J.R.

    2003-01-01

    Studies of thermally emitted electron Bernstein waves (EBWs) on CDX-U and NSTX, via mode conversion (MC) to electromagnetic radiation, support the use of EBWs to measure the Te profile and provide local electron heating and current drive (CD) in overdense spherical torus plasmas. An X-mode antenna with radially adjustable limiters successfully controlled EBW MC on CDX-U and enhanced MC efficiency to ∼ 100%. So far the X-mode MC efficiency on NSTX has been increased by a similar technique to 40-50% and future experiments are focused on achieving * 80% MC. MC efficiencies on both machines agree well with theoretical predictions. Ray tracing and Fokker-Planck modeling for NSTX equilibria are being conducted to support the design of a 3 MW, 15 GHz EBW heating and CD system for NSTX to assist non-inductive plasma startup, current ramp up, and to provide local electron heating and CD in high beta NSTX plasmas

  17. Plasma control system upgrade and increased plasma stability in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Mastrovito, D., E-mail: dmastrovito@pppl.go [Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543 (United States); Gates, D.; Gerhard, S.; Lawson, J.; Ludescher-Furth, C.; Marsala, R. [Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543 (United States)

    2010-07-15

    Plasma control on the National Spherical Torus Experiment (NSTX) was previously accomplished using eight 333 MHz G4 processors built by Sky computers. Several planned improvements and additional control algorithms required significant upgrades to our real-time control computers and real-time data acquisition infrastructure. Several in-house modules have been designed and implemented including: the digital time stamp module (DITS) and for digital/analog front panel data port (FPDP) output, the FPDP output module digital/analog (FOMD/A). Standard Linux based Intel computers perform the real-time control tasks and InfiniBand as been employed for communication between a user-accessible 'host' server and the real-time computer. In addition to several independent real-time processes the General Atomics developed PCS (Bell (2006) ) system infrastructure continues to be used on NSTX. While maintaining previous functionality, improvements in the control system software include: an RWM feedback algorithm, beta feedback NBI control, more comprehensive error logging and trapping, more user-friendly interface, more complete archiving and restoring functionality, and better status reporting and diagnostic tools. Once completed, we succeeded in increasing overall plasma stability and decreasing control system latency by several times.

  18. Kinetic Profiles in NSTX Plasmas

    International Nuclear Information System (INIS)

    Bell, R.E.; LeBlanc, B.P.; Bourdelle, C.; Ernst, D.R.; Fredrickson, E.D.; Gates, D.A.; Hosea, J.C.; Johnson, D.W.; Kaye, S.M.; Maingi, R.; Medley, S.; Menard, J.E.; Mueller, D.; Ono, M.; Paoletti, F.; Peng, M.; Sabbagh, S.A.; Stutman, D.; Swain, D.W.; Synakowski, E.J.; Wilson, J.R.

    2001-01-01

    The National Spherical Torus Experiment (NSTX) is a low aspect ratio (R/a approximately 1.3) device with auxiliary heating from neutral-beam injection (NBI) and high-harmonic fast-wave heating (HHFW). Typical NSTX parameters are R(subscript ''0'') = 85 cm, a = 67 cm, I(subscript ''p'') = 0.7-1.4 MA, B(subscript ''phi'') = 0.25-0.45 T. Three co-directed deuterium neutral-beam sources have injected P(subscript ''NB'') less than or equal to 4.7 MW. HHFW plasmas typically have delivered P(subscript ''RF'') less than or equal to 3 MW. Important to the understanding of NSTX confinement are the new kinetic profile diagnostics: a multi-pulse Thomson scattering system (MPTS) and a charge-exchange recombination spectroscopy (CHERS) system. The MPTS diagnostic currently measures electron density and temperature profiles at 30 Hz at ten spatial locations. The CHERS system has recently become available to measure carbon ion temperature and toroidal flow at 17 radial positions spanning the outer half of the minor radius with 20 msec time resolution during NBI. Experiments conducted during the last year have produced a wide range of kinetic profiles in NSTX. Some interesting examples are presented below

  19. Overview of impurity control and wall conditioning in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    KUGEL,H.W.; MAINGI,R.; BELL,M.; BLANCHARD,W.; GATES,D.; JOHNSON,D.; KAITA,R.; KAYE,S.; MARQUEDA,R.; MENARD,J.; MUELLER,D.; ONO,M.; PENG,Y-K.M.; RAMAN,R.; RAMSEY,A.; ROQUEMORE,A.; SKINNER,C.; SABBAGH,S.; STUTMAN,D.; WAMPLER,WILLIAM R.; WILSON,J.R.; ZWEBEN,S.

    2000-05-25

    The National Spherical Torus Experiment (NSTX) started plasma operations in February 1999, and promptly achieved high current, inner wall limited, double null, and single null plasma discharges, initial Coaxial Helicity Injection, and High Harmonic Fast Wave results. NSTX is designed to study the physics of Spherical Tori (ST) in a device that can produce non-inductively sustained high-{beta} discharges in the 1 MA regime and to explore approaches toward a small, economical high power density ST reactor core. As expected, discharge reproducibility and performance were strongly affected by wall conditions. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results.

  20. Solenoid-free plasma startup in NSTX using transient CHI

    International Nuclear Information System (INIS)

    Raman, R.; Jarboe, T.R.; Nelson, B.A.; Mueller, D.; Bell, M.G.; Bell, R.; Gates, D.; Gerhardt, S.; Hosea, J.; Kaita, R.; Kugel, H.; LeBlanc, B.; Menard, J.; Ono, M.; Paul, S.; Roquemore, L.; Maingi, R.; Maqueda, R.; Nagata, M.; Sabbagh, S.

    2009-01-01

    Experiments in NSTX have now demonstrated the coupling of toroidal plasmas produced by the technique of coaxial helicity injection (CHI) to inductive sustainment and ramp-up of the toroidal plasma current. In these discharges, the central Ohmic transformer was used to apply an inductive loop voltage to discharges with a toroidal current of about 100 kA created by CHI. The coupled discharges have ramped up to >700 kA and transitioned into an H-mode demonstrating compatibility of this startup method with conventional operation. The electron temperature in the coupled discharges reached over 800 eV and the resulting plasma had low inductance, which is preferred for long-pulse high-performance discharges. These results from NSTX in combination with the previously obtained record 160 kA non-inductively generated startup currents in an ST or tokamak in NSTX demonstrate that CHI is a viable solenoid-free plasma startup method for future STs and tokamaks.

  1. Application of Electron Bernstein Wave heating and current drive to high beta plasmas

    International Nuclear Information System (INIS)

    Efthimion, P.C.

    2002-01-01

    Electron Bernstein Waves (EBW) can potentially heat and drive current in high-beta plasmas. Electromagnetic waves can convert to EBW via two paths. O-mode heating, demonstrated on W-7AS, requires waves be launched within a narrow k-parallel range. Alternately, in high-beta plasmas, the X-mode cutoff and EBW conversion layers are millimeters apart, so the fast X-mode can tunnel to the EBW branch. We are studying the conversion of EBW to the X-mode by measuring the radiation temperature of the cyclotron emission and comparing it to the electron temperature. In addition, mode conversion has been studied with an approximate kinetic full-wave code. We have enhanced EBW mode conversion to ∼ 100% by encircling the antenna with a limiter that shortens the density scale length at the conversion layer in the scrape off of the CDX-U spherical torus (ST) plasma. Consequently, a limiter in front of a launch antenna achieves efficient X-mode coupling to EBW. Ray tracing and Fokker-Planck codes have been used to develop current drive scenarios in NSTX high-beta (∼ 40%) ST plasmas and a relativistic code will examine the potential synergy of EBW current drive with the bootstrap current. (author)

  2. Initial operation of NSTX with plasma control

    International Nuclear Information System (INIS)

    Gates, D.; Bell, M.; Ferron, J.; Kaye, S.; Menard, J.; Mueller, D.; Neumeyer, C.; Sabbagh, S.

    2000-01-01

    First plasma, with a maximum current of 300kA, was achieved on NSTX in February 1999. These results were obtained using preprogrammed coil currents. The first controlled plasmas on NSTX were made starting in August 1999 with the full 1MA plasma current achieved in December 1999. The controlled quantities were plasma position (R, Z) and current (Ip). Variations in the plasma shape are achieved by adding preprogrammed currents to those determined by the control parameters. The control system is fully digital, with plasma position and current control, data acquisition, and power supply control all occurring in the same four-processor real time computer. The system uses the PCS (Plasma Control Software) system designed at General Atomics. Modular control algorithms, specific to NSTX, were written and incorporated into the PCS. The application algorithms do the actual control calculations, with the PCS handling data passing. The control system, including planned upgrades, will be described, along with results of the initial controlled plasma operations. Analysis of the performance of the control system will also be presented

  3. Initial Studies of Core and Edge Transport of NSTX Plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Bell, M.G.; Bell, R.E.; Bush, C.E.; Bourdelle, C.; Darrow, D.; Dorland, W.; Ejiri, A.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; Kubota, S.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.J.; Menard, J.E.; Mueller, D.; Rosenberg, A.; Sabbagh, S.A.; Stutman, D.; Taylor, G.; Johnson, D.W.; Kaita, R.; Ono, M.; Paoletti, F.; Peebles, W.; Peng, Y-K.M.; Roquemore, A.L.; Skinner, C.H.; Soukhanovskii, V.A.

    2001-01-01

    Rapidly developing diagnostic, operational, and analysis capability is enabling the first detailed local physics studies to begin in high-beta plasmas of the National Spherical Torus Experiment (NSTX). These studies are motivated in part by energy confinement times in neutral-beam-heated discharges that are favorable with respect to predictions from the ITER-89P scaling expression. Analysis of heat fluxes based on profile measurements with neutral-beam injection (NBI) suggest that the ion thermal transport may be exceptionally low, and that electron thermal transport is the dominant loss channel. This analysis motivates studies of possible sources of ion heating not presently accounted for by classical collisional processes. Gyrokinetic microstability studies indicate that long wavelength turbulence with k(subscript ''theta'') rho(subscript ''i'') ∼ 0.1-1 may be suppressed in these plasmas, while modes with k(subscript ''theta'') rho(subscript ''i'') ∼ 50 may be robust. High-harmonic fast-wave (HHFW) heating efficiently heats electrons on NSTX, and studies have begun using it to assess transport in the electron channel. Regarding edge transport, H-mode [high-confinement mode] transitions occur with either NBI or HHFW heating. The power required for low-confinement mode (L-mode) to H-mode transitions far exceeds that expected from empirical edge-localized-mode-free H-mode scaling laws derived from moderate aspect ratio devices. Finally, initial fluctuation measurements made with two techniques are permitting the first characterizations of edge turbulence

  4. Soft X-ray Tangential Imaging of the NSTX Core Plasma by Means of a MPGD Pin-hole Camera

    International Nuclear Information System (INIS)

    Pacella, D.; Leigheb, M.; Bellazzini, R.; Brez, A.; Finkenthal, M.; Stutman, D.; Kaita, R.; Sabbagh, S.A.

    2003-01-01

    A fast X-ray system based on a Micro Pattern Gas Detector has been used, for the first time, to investigate emission from the plasma core of the National Spherical Tokamak eXperiment (NSTX) at the Princeton Plasma Physics Laboratory. The results presented in this work demonstrate the capability of such a device to measure with a time resolution of the order of 1 ms the curvature and the elongation of the X-ray iso-emissivity contours, under various plasma conditions. Also, comparisons with the magnetic surface structure calculated by the EFIT code show good agreement between reconstructed flux surface and the soft X-ray emissions (SXR) for poloidal beta values up to 0.6. For greater values of beta, X-ray iso-emissivity contours become circular, while magnetic flux surface reconstructions yield elongation 1.5 < k < 2.2. The X-ray images have been acquired with a (statistical) signal to noise ratio (SNR) per pixel of about 30. Thanks to the direct and efficient X-ray conversion and its operation in a photon counting mode, this new diagnostic tool allows the routine investigation of the plasma core with a sampling rate of 1 kHz and extremely high SNR under all experimental conditions in NSTX

  5. Implications of NSTX lithium results for magnetic fusion research

    Energy Technology Data Exchange (ETDEWEB)

    Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Bell, R.E.; Kaita, R.; Kugel, H.W.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Canik, J.M.; Diem, S. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Gerhardt, S.P.; Hosea, J.; Kaye, S.; Mansfield, D. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Menard, J.; Paul, S.F. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Raman, R. [University of Washington at Seattle, Seattle, WA (United States); Sabbagh, S.A. [Columbia University, New York, NY (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Taylor, G. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2010-11-15

    Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to {approx}100 g of lithium onto the lower divertor plates between lithium re-loadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, edge localized mode (ELM) control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.

  6. Implications of NSTX Lithium Results for Magnetic Fusion Research

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.; Kaita, R.; Kugel, H.W.; LeBlanc, B.P.; Canik, J.M.; Diem, S.; Gerhardt, S.P.; Hosea, J.; Kaye, S.; Mansfield, D.; Maingi, R.; Menard, J.; Paul, S.F.; Raman, R.; Sabbagh, S.A.; Skinner, C.H.; Soukhanovskii, V.; Taylor, G.

    2010-01-01

    Lithium wall coating techniques have been experimentally explored on NSTX for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to ∼ 100 g of lithium onto the lower divertor plates between lithium reloadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, ELM control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.

  7. Implications of NSTX lithium results for magnetic fusion research

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.; Kaita, R.; Kugel, H.W.; LeBlanc, B.P.; Canik, J.M.; Diem, S.; Gerhardt, S.P.; Hosea, J.; Kaye, S.; Mansfield, D.; Maingi, R.; Menard, J.; Paul, S.F.; Raman, R.; Sabbagh, S.A.; Skinner, C.H.; Soukhanovskii, V.; Taylor, G.

    2010-01-01

    Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to ∼100 g of lithium onto the lower divertor plates between lithium re-loadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, edge localized mode (ELM) control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.

  8. NSTX Diagnostics for Fusion Plasma Science Studies

    International Nuclear Information System (INIS)

    Kaita, R.; Johnson, D.; Roquemore, L.; Bitter, M.; Levinton, F.; Paoletti, F.; Stutman, D.

    2001-01-01

    This paper will discuss how plasma science issues are addressed by the diagnostics for the National Spherical Torus Experiment (NSTX), the newest large-scale machine in the magnetic confinement fusion (MCF) program. The development of new schemes for plasma confinement involves the interplay of experimental results and theoretical interpretations. A fundamental requirement, for example, is a determination of the equilibria for these configurations. For MCF, this is well established in the solutions of the Grad-Shafranov equation. While it is simple to state its basis in the balance between the kinetic and magnetic pressures, what they are as functions of space and time are often not easy to obtain. Quantities like the plasma pressure and current density are not directly measurable. They are derived from data that are themselves complex products of more basic parameters. The same difficulties apply to the understanding of plasma instabilities. Not only are the needs for spatial and temporal resolution more stringent, but the wave parameters which characterize the instabilities are difficult to resolve. We will show how solutions to the problems of diagnostic design on NSTX, and the physics insight the data analysis provides, benefits both NSTX and the broader scientific community

  9. Lithium Surface Coatings for Improved Plasma Performance in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H W; Ahn, J -W; Allain, J P; Bell, R; Boedo, J; Bush, C; Gates, D; Gray, T; Kaye, S; Kaita, R; LeBlanc, B; Maingi, R; Majeski, R; Mansfield, D; Menard, J; Mueller, D; Ono, M; Paul, S; Raman, R; Roquemore, A L; Ross, P W; Sabbagh, S; Schneider, H; Skinner, C H; Soukhanovskii, V; Stevenson, T; Timberlake, J; Wampler, W R

    2008-02-19

    NSTX high-power divertor plasma experiments have shown, for the first time, significant and frequent benefits from lithium coatings applied to plasma facing components. Lithium pellet injection on NSTX introduced lithium pellets with masses 1 to 5 mg via He discharges. Lithium coatings have also been applied with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium depositions from a few mg to 1 g have been applied between discharges. Benefits from the lithium coating were sometimes, but not always seen. These improvements sometimes included decreases plasma density, inductive flux consumption, and ELM frequency, and increases in electron temperature, ion temperature, energy confinement and periods of MHD quiescence. In addition, reductions in lower divertor D, C, and O luminosity were measured.

  10. Predictions and observations of low-shear beta-induced shear Alfven-acoustic eigenmodes in toroidal plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Gorelenkov, N.N. [Princeton Plasma Physics Laboratory, Princeton University (United States)], E-mail: ngorelen@pppl.gov; Berk, H.L. [IFS, Austin, Texas (United States); Fredrickson, E. [Princeton Plasma Physics Laboratory, Princeton University (United States); Sharapov, S.E. [Euroatom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire (United States)

    2007-10-08

    New global MHD eigenmode solutions arising in gaps in the low frequency Alfven-acoustic continuum below the geodesic acoustic mode (GAM) frequency have been found numerically and have been used to explain relatively low frequency experimental signals seen in NSTX and JET tokamaks. These global eigenmodes, referred to here as Beta-induced Alfven-Acoustic Eigenmodes (BAAE), exist in the low magnetic safety factor region near the extrema of the Alfven-acoustic continuum. In accordance to the linear dispersion relations, the frequency of these modes shifts as the safety factor, q, decreases. We show that BAAEs can be responsible for observations in JET plasmas at relatively low beta <2% as well as in NSTX plasmas at relatively high-beta >20%. In contrast to the mostly electrostatic character of GAMs the new global modes also contain an electromagnetic (magnetic field line bending) component due to the Alfven coupling, leading to wave phase velocities along the field line that are large compared to the sonic speed. Qualitative agreement between theoretical predictions and observations are found.

  11. Turbulence of high-beta plasma

    International Nuclear Information System (INIS)

    Khvesyuk, V.I.; Chirkov, A.Y.

    1999-01-01

    Principals of numerical modelling of turbulence in high-beta plasma (β > 0.1) are discussed. Creation of transport model for axial symmetric nonuniform confining magnetic field is considered. Numerical model of plasma turbulence in FRC is presented. The physical and mathematical models are formulated from nonuniform axial symmetric high-beta plasma. It is shown that influence of waves arise under this plasma conditions lead to chaotic motion of charged particles across magnetic field. (author)

  12. Measured improvement of global magnetohydrodynamic mode stability at high-beta, and in reduced collisionality spherical torus plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Berkery, J. W.; Sabbagh, S. A.; Balbaky, A. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States); Bell, R. E.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B. P.; Manickam, J.; Menard, J. E.; Podestà, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Betti, R. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States)

    2014-05-15

    Global mode stability is studied in high-β National Spherical Torus Experiment (NSTX) plasmas to avoid disruptions. Dedicated experiments in NSTX using low frequency active magnetohydrodynamic spectroscopy of applied rotating n = 1 magnetic fields revealed key dependencies of stability on plasma parameters. Observations from previous NSTX resistive wall mode (RWM) active control experiments and the wider NSTX disruption database indicated that the highest β{sub N} plasmas were not the least stable. Significantly, here, stability was measured to increase at β{sub N}∕l{sub i} higher than the point where disruptions were found. This favorable behavior is shown to correlate with kinetic stability rotational resonances, and an experimentally determined range of measured E × B frequency with improved stability is identified. Stable plasmas appear to benefit further from reduced collisionality, in agreement with expectation from kinetic RWM stabilization theory, but low collisionality plasmas are also susceptible to sudden instability when kinetic profiles change.

  13. Impact of the wall conditioning program on plasma performance in NSTX

    International Nuclear Information System (INIS)

    Kugel, H.W.; Soukhanovskii, V.; Bell, M.; Blanchard, W.; Gates, D.; LeBlanc, B.; Maingi, R.; Mueller, D.; Na, H.K.; Paul, S.; Skinner, C.H.; Stutman, D.; Wampler, W.R.

    2003-01-01

    High performance operating regimes have been achieved on NSTX through impurity control and wall conditioning techniques. These techniques include HeGDC-aided boronization using deuterated trimethylboron, inter-discharge HeGDC, 350 deg. C PFC bake-out followed by D 2 and HeGDC, and experiments to test fueling discharges with either a He-trimethylboron mixture or pure trimethylboron. The impact of this impurity and density control program on recent advances in NSTX plasma performance is discussed

  14. Impact of the Wall Conditioning Program on Plasma Performance in NSTX

    International Nuclear Information System (INIS)

    H.W. Kuge; V. Soukhanovskii; M. Bell; , W. Blanchard; D. Gates; B. LeBlanc; R. Maingi; D. Mueller; H.K. Na; S. Paul; C.H. Skinner; D. Stutman; and W.R. Wampler

    2002-01-01

    High performance operating regimes have been achieved on NSTX (National Spherical Torus Experiment) through impurity control and wall-conditioning techniques. These techniques include HeGDC-aided boronization using deuterated trimethylboron, inter-discharge HeGDC, 350 C PFC bake-out followed by D2 and HeGDC, and experiments to test fueling discharges with either a He-trimethylboron mixture or pure trimethylboron. The impact of this impurity and density control program on recent advances in NSTX plasma performance is discussed

  15. A Neutral Beam Injector Upgrade for NSTX

    International Nuclear Information System (INIS)

    Stevenson, T.; McCormack, B.; Loesser, G.D.; Kalish, M.; Ramakrishnan, S.; Grisham, L.; Edwards, J.; Cropper, M.; Rossi, G.; Halle, A. von; Williams, M.

    2002-01-01

    The National Spherical Torus Experiment (NSTX) capability with a Neutral Beam Injector (NBI) capable of 80 kiloelectronvolt (keV), 5 Megawatt (MW), 5 second operation. This 5.95 million dollar upgrade reused a previous generation injector and equipment for technical, cost, and schedule reasons to obtain these specifications while retaining a legacy capability of 120 keV neutral particle beam delivery for shorter pulse lengths for possible future NSTX experiments. Concerns with NBI injection included power deposition in the plasma, aiming angles from the fixed NBI fan array, density profiles and beam shine through, orbit losses of beam particles, and protection of the vacuum vessel wall against beam impingement. The upgrade made use of the beamline and cryo panels from the Neutral Beam Test Stand facility, existing power supplies and controls, beamline components and equipment not contaminated by tritium during DT [deuterium-tritium] experiments, and a liquid Helium refrigerator plant to power and cryogenically pump a beamline and three ion sources. All of the Tokamak Fusion Test Reactor (TFTR) ion sources had been contaminated with tritium, so a refurbishment effort was undertaken on selected TFTR sources to rid the three sources destined for the NSTX NBI of as much tritium as possible. An interconnecting duct was fabricated using some spare and some new components to attach the beamline to the NSTX vacuum vessel. Internal vacuum vessel armor using carbon tiles was added to protect the stainless steel vacuum vessel from beam impingement in the absence of plasma and interlock failure. To date, the NBI has operated to 80 keV and 5 MW and has injected requested power levels into NSTX plasmas with good initial results, including high beta and strong heating characteristics at full rated plasma current

  16. Predictions and Observations of Low-shear Beta-induced Alfvén-acoustic Eigenmodes in Toroidal Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Gorelenkov, N. N.; Berk, H. L.; Fredrickson, E.; Sharapov, S. E.

    2007-07-02

    New global MHD eigenmode solutions arising in gaps in the low frequency Alfvén -acoustic continuum below the geodesic acoustic mode (GAM) frequency have been found numerically and have been used to explain relatively low frequency experimental signals seen in NSTX and JET tokamaks. These global eigenmodes, referred to here as Beta-induced Alfvén-Acoustic Eigenmodes (BAAE), exist in the low magnetic safety factor region near the extrema of the Alfvén-acoustic continuum. In accordance to the linear dispersion relations, the frequency of these modes shifts as the safety factor, q, decreases. We show that BAAEs can be responsible for observations in JET plasmas at relatively low beta < 2% as well as in NSTX plasmas at relatively high beta > 20%. In contrast to the mostly electrostatic character of GAMs the new global modes also contain an electromagnetic (magnetic field line bending) component due to the Alfvén coupling, leading to wave phase velocities along the field line that are large compared to the sonic speed. Qualitative agreement between theoretical predictions and observations are found.

  17. Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Brooks, A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Lopes-Cardozo, N. [TU/Eindhoven, Eindhoven (Netherlands); Menard, J.; Ono, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Rindt, P. [TU/Eindhoven, Eindhoven (Netherlands); Tresemer, K. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2016-11-15

    Highlights: • An upgrade path for the NSTX-U tokamak is proposed that maintains scientific productivity while enabling exploration of novel, liquid metal PFC. • Pre-filled liquid metal divertor targets are proposed as an intermediate step that mitigates technical and scientific risks associated with liquid metal PFC. • Analysis of leading edge features show a strong link between engineering design considerations and expected performance as a PFC. • A method for optimizing porous liquid metal targets restrained by capillary forces is provided indicating pore-sizes well within current technical capabilities. - Abstract: Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. Two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

  18. National Spherical Torus Experiment (NSTX) Engineering Overview and Research Results 1999 - 2000

    International Nuclear Information System (INIS)

    Neumeyer, C.

    2000-01-01

    The NSTX is a new US facility for the study of plasma confinement, heating, and current drive in a low aspect ratio, spherical torus (ST) configuration. The ST configuration is an alternate magnetic confinement concept which is characterized by high beta (ratio plasma pressure to magnetic field pressure) and low toroidal field compared to conventional tokamaks, and could provide a pathway to the realization of a practical fusion power source. NSTX achieved first plasma in February 1999, and since that time has completed and commissioned all components and systems within the machine proper. Routine operation with inductively driven plasma current less than or equal to 1MA and flat top less than or equal to 0.3 seconds has been established, and the ohmic characterization phase of the research program is underway. Radio Frequency (RF) and Neutral Beam Injection (NBI) systems have been installed and are presently being commissioned. This paper describes the NSTX mission, gives an overview of the engineering design, and summarizes the research results obtained thus far

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

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Maingi, R.; Sabbagh, S.A.; Soukhanovskii, V.; Stutman, D.

    2003-01-01

    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

  20. Overview of Results from the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.; Ahn, J.; Allain, R.; Andre, R.; Bastasz, R.; Bell, M.; Bell, R.; Belova, E.; Berkery, J.; Betti, R.

    2009-01-01

    The mission of NSTX is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale-length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of High Harmonic Fast-Waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance l i ∼ 0.4 with strong shaping (κ ∼ 2.7, (delta) ∼ 0.8) with β N approaching the with-wall beta limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction f NI ∼ 71%. Instabilities driven by super-Alfvenic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvenic. Linear TAE thresholds and appreciable fast-ion loss during multi-mode bursts are measured and these results are compared to theory. The impact of n > 1 error fields on stability is a important result for ITER. RWM/RFA feedback combined with n=3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are: results of lithium coating

  1. Progress on advanced tokamak and steady-state scenario development on DIII-D and NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, California 90095 (United States); Garofalo, A M [Columbia University, New York, New York 10027 (United States); Greenfield, C M [General Atomics, San Diego, California 92186-5608 (United States); Kaye, S M [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Menard, J E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Murakami, M [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Sabbagh, S A [Columbia University, New York, New York 10027 (United States); Austin, M E [University of Texas-Austin, Austin, Texas 78712 (United States); Bell, R E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Burrell, K H [General Atomics, San Diego, California 92186-5608 (United States); Ferron, J R [General Atomics, San Diego, California 92186-5608 (United States); Gates, D A [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Groebner, R J; Hyatt, A W; Luce, T C; Petty, C C; Wade, M R; Waltz, R E [General Atomics, San Diego, California 92186-5608 (United States); Jayakumar, R J [Lawrence Livermore National Lab., Livermore, California 94550 (United States); Kinsey, J E [Lehigh Univ., Bethlehem, Pennsylvania 18015 (United States); LeBlanc, B P [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); McKee, G R [Univ. of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Okabayashi, M [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Peng, Y-K M [Oak Ridge National Lab., Oak Ridge, Tennessee 37831 (United States); Politzer, P A [General Atomics, San Diego, California 92186-5608 (United States); Rhodes, T L [Dept. of Electrical Engineering and PSTI, Univ. of California, Los Angeles, California 90095 (United States)

    2006-12-15

    Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f{sub BS} {>=} 60%, q{sub 95} {approx} 4-5 and G {identical_to}{beta}{sub N}H{sub scaling}/q{sub 95}{sup 2} {>=}0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E ? B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at {beta}{sub N} {>=} 4 with q{sub min} {>=} 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. {beta}{sub N} = 4, H{sub 89} = 2.5, with f{sub BS} > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q {>=} 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

  2. Plasma boundary shape control and real-time equilibrium reconstruction on NSTX-U

    Science.gov (United States)

    Boyer, M. D.; Battaglia, D. J.; Mueller, D.; Eidietis, N.; Erickson, K.; Ferron, J.; Gates, D. A.; Gerhardt, S.; Johnson, R.; Kolemen, E.; Menard, J.; Myers, C. E.; Sabbagh, S. A.; Scotti, F.; Vail, P.

    2018-03-01

    The upgrade to the National Spherical Torus eXperiment (NSTX-U) included two main improvements: a larger center-stack, enabling higher toroidal field and longer pulse duration, and the addition of three new tangentially aimed neutral beam sources, which increase available heating and current drive, and allow for flexibility in shaping power, torque, current, and particle deposition profiles. To best use these new capabilities and meet the high-performance operational goals of NSTX-U, major upgrades to the NSTX-U control system (NCS) hardware and software have been made. Several control algorithms, including those used for real-time equilibrium reconstruction and shape control, have been upgraded to improve and extend plasma control capabilities. As part of the commissioning phase of first plasma operations, the shape control system was tuned to control the boundary in both inner-wall limited and diverted discharges. It has been used to accurately track the requested evolution of the boundary (including the size of the inner gap between the plasma and central solenoid, which is a challenge for the ST configuration), X-point locations, and strike point locations, enabling repeatable discharge evolutions for scenario development and diagnostic commissioning.

  3. Status of National Spherical Torus Experiment (NSTX)*

    Science.gov (United States)

    Ono, Masayuki

    2001-10-01

    The main aim of National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the innovative spherical torus (ST) concept. The NSTX experimental facility has been operating reliably and its capabilities steadily improving. Due to relatively efficient ohmic current drive and benign halo current behavior, the plasma current was increased to 1.4 MA, which is well above the design value of 1 MA. The plasmas at 1 MA are now routinely heated by NBI to the average toroidal beta value of 20 percent range at 3 kG with electrons and ions in the 1-2 keV range. Even with the “L-mode” edge, the energy confinement time can well exceed the so-called L-mode (and even H-mode) scaling values. As a part of ST tool development, High Harmonic Fast Wave (HHFW) heating has demonstrated efficient electron heating with the central electron temperatures reaching 3.7 keV. HHFW induced H-modes have been also observed. For CHI (Coaxial Helicity Injection) non-inductive start-up, CHI discharges of up to 300 kA of toroidal current and 300 msec duration have been produced from zero current using = 25 kA of injected current. The poster presentation will also include the near term NSTX facility upgrade plan.

  4. Exploration of spherical torus physics in the NSTX device

    Science.gov (United States)

    Ono, M.; Kaye, S. M.; Peng, Y.-K. M.; Barnes, G.; Blanchard, W.; Carter, M. D.; Chrzanowski, J.; Dudek, L.; Ewig, R.; Gates, D.; Hatcher, R. E.; Jarboe, T.; Jardin, S. C.; Johnson, D.; Kaita, R.; Kalish, M.; Kessel, C. E.; Kugel, H. W.; Maingi, R.; Majeski, R.; Manickam, J.; McCormack, B.; Menard, J.; Mueller, D.; Nelson, B. A.; Nelson, B. E.; Neumeyer, C.; Oliaro, G.; Paoletti, F.; Parsells, R.; Perry, E.; Pomphrey, N.; Ramakrishnan, S.; Raman, R.; Rewoldt, G.; Robinson, J.; Roquemore, A. L.; Ryan, P.; Sabbagh, S.; Swain, D.; Synakowski, E. J.; Viola, M.; Williams, M.; Wilson, J. R.; NSTX Team

    2000-03-01

    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 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.

  5. Heating and current drive on NSTX

    Science.gov (United States)

    Wilson, J. R.; Batchelor, D.; Carter, M.; Hosea, J.; Ignat, D.; LeBlanc, B.; Majeski, R.; Ono, M.; Phillips, C. K.; Rogers, J. H.; Schilling, G.

    1997-04-01

    Low aspect ratio tokamaks pose interesting new challenges for heating and current drive. The NSTX (National Spherical Tokamak Experiment) device to be built at Princeton is a low aspect ratio toroidal device that has the achievement of high toroidal beta (˜45%) and non-inductive operation as two of its main research goals. To achieve these goals significant auxiliary heating and current drive systems are required. Present plans include ECH (Electron cyclotron heating) for pre-ionization and start-up assist, HHFW (high harmonic fast wave) for heating and current drive and eventually NBI (neutral beam injection) for heating, current drive and plasma rotation.

  6. Improvement in Plasma Performance with Lithium Coatings in NSTX

    International Nuclear Information System (INIS)

    Kaita, R.

    2009-01-01

    Lithium as a plasma-facing material has attractive features, including a reduction in the recycling of hydrogenic species and the potential for withstanding high heat and neutron fluxes in fusion reactors. Dramatic effects on plasma performance with lithium-coated plasma-facing components (PFC's) have been demonstrated on many fusion devices, including TFTR, T-11M, and FT-U. Using a liquid-lithium-filled tray as a limiter, the CDX-U device achieved very significant enhancement in the confinement time of ohmically heated plasmas. The recent NSTX experiments reported here have demonstrated, for the first time, significant and recurring benefits of lithium PFC coatings on divertor plasma performance in both L- and H- mode regimes heated by neutral beams.

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

    International Nuclear Information System (INIS)

    Rosenberg, A.; Menard, J.E.; LeBlanc, B.P.

    2001-01-01

    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

  8. NSTX Plasma Response to Lithium Coated Divertor

    Energy Technology Data Exchange (ETDEWEB)

    H.W. Kugel, M.G. Bell, J.P. Allain, R.E. Bell, S. Ding, S.P. Gerhardt, M.A. Jaworski, R. Kaita, J. Kallman, S.M. Kaye, B.P. LeBlanc, R. Maingi, R. Majeski, R. Maqueda, D.K. Mansfield, D. Mueller, R. Nygren, S.F. Paul, R. Raman, A.L. Roquemore, S.A. Sabbagh, H. Schneider, C.H. Skinner, V.A. Soukhanovskii, C.N. Taylor, J.R. Timberlak, W.R. Wampler, L.E. Zakharov, S.J. Zweben, and the NSTX Research Team

    2011-01-21

    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.

  9. NSTX plasma response to lithium coated divertor

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Allain, J.P.; Bell, R.E.; Ding, S.; Gerhardt, S.P.; Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.M.; LeBlanc, B.P.; Maingi, Rajesh; Majeski, R.; Maqueda, R.J.; Mansfield, D.K.; Mueller, D.; Nygren, R.E.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Sabbagh, S.A.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.A.; Taylor, C.N.; Timberlake, J.; Wampler, W.R.; Zakharov, L.E.; Zweben, S.J.

    2011-01-01

    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 Z(eff) 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.

  10. Synthetic Aperture Microwave Imaging (SAMI) of the plasma edge on NSTX-U

    Science.gov (United States)

    Vann, Roddy; Taylor, Gary; Brunner, Jakob; Ellis, Bob; Thomas, David

    2016-10-01

    The Synthetic Aperture Microwave Imaging (SAMI) system is a unique phased-array microwave camera with a +/-40° field of view in both directions. It can image cut-off surfaces corresponding to frequencies in the range 10-34.5GHz; these surfaces are typically in the plasma edge. SAMI operates in two modes: either imaging thermal emission from the plasma (often modified by its interaction with the plasma edge e.g. via BXO mode conversion) or ``active probing'' i.e. injecting a broad beam at the plasma surface and imaging the reflected/back-scattered signal. SAMI was successfully pioneered on the Mega-Amp Spherical Tokamak (MAST) at Culham Centre for Fusion Energy. SAMI has now been installed and commissioned on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton Plasma Physics Laboratory. The firmware has been upgraded to include real-time digital filtering, which enables continuous acquisition of the Doppler back-scattered active probing data. In this poster we shall present SAMI's analysis of the plasma edge on NSTX-U including measurements of the edge pitch angle on NSTX-U using SAMI's unique 2-D Doppler-backscattering capability.

  11. Overview of the initial NSTX experimental results

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.; Bell, R.

    2001-01-01

    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 1 million amperes on December 14, 1999. The planned plasma shaping parameters, κ=1.6-2.2 and δ=0.2-0.4, were achieved in inner limited, single null and double null configurations. The CHI (Coaxial Helicity Injection) and HHFW (High Harmonic Fast Wave) experiments were also initiated. A CHI injected current of 27 kA produced up to 260 kA of toroidal current without using an ohmic solenoid. With an injection of 2.3 MW of HHFW power, using twelve antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5x10 13 cm -3 increasing the plasma energy to 59 kJ and the toroidal beta, β T to 10 %. Finally, the NBI system commenced operation in Sept. 2000. The initial results with two ion sources (P NBI =2.8MW) shows good heating, producing a total plasma stored energy of 90 kJ corresponding to β T ∼18% at a plasma current of 1.1 MA. (author)

  12. Overview of the Initial NSTX Experimental Results

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.; Bell, R. E.; Bigelow, T.; Bitter, M.

    2000-01-01

    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 Ip was successfully brought up to the design value of 1 million amperes on December 14, 1999. The planned plasma shaping parameters, k = 1.6 ± 2.2 and d = 0.2 ± 0.4, were achieved in inner limited, single null and double null configurations. The CHI (Coaxial Helicity Injection) and HHFW (High Harmonic Fast Wave) experiments were also initiated. A CHI injected current of 27 kA produced up to 260 kA of toroidal current without using an ohmic solenoid. With an injection of 2.3 MW of HHFW power, using twelve 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 1013 cm-3 increasing the plasma energy to 59 kJ and the toroidal beta, bT to 10 %. Finally, the NBI system commenced operation in Sept. 2000. The initial results with two ion sources (PNBI = 2.8 MW) shows good heating, producing a total plasma stored energy of 90 kJ corresponding to bT = 18 % at a plasma current of 1.1 MA

  13. ECH on NSTX

    International Nuclear Information System (INIS)

    Bigelow, T.S.; Batchelor, D.B.; Carter, M.D.; Peng, M.; Wilson, J.R.

    1997-01-01

    Electron Cyclotron Heating has been proposed for plasma initiation, startup assistance and non-inductive startup on NSTX. One physics goal of NSTX will be to establish entirely non-inductive plasma operation by utilizing ECH to provide a sufficient start-up plasma to support further current drive from other heating systems. Scaling of previous ECH-only startup experiments on CDX-U and DIII-D indicate that 400 kW of ECH should be capable of driving 42 kA of pressure driven current on NSTX and possibly higher levels after optimizing the process. Due to the low NSTX magnetic field, over-dense plasmas exist during most of the discharge so conventional ECH operation is limited to the low density startup phase. To extend the useful operating range for ECH, a scheme involving mode conversion to the electron Bernstein Wave (EBW) from either O r X mode launch is being investigated for bulk heating and current drive applications at higher density. Microwave equipment, including 18 GHz klystrons and 28 GHz gyrotrons are available at ORNL and appear ideal for use on NSTX. Preliminary pre-ionization and start-up system configurations are presented here along with discussions on various operation modes

  14. Surface chemistry analysis of lithium conditioned NSTX graphite tiles correlated to plasma performance

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, C.N., E-mail: chase.taylor@inl.gov [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Birck Nanotechnology Center, Discovery Park, West Lafayette, IN 47907 (United States); Luitjohan, K.E. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Heim, B. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Birck Nanotechnology Center, Discovery Park, West Lafayette, IN 47907 (United States); Kollar, L. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Allain, J.P. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47906 (United States); Birck Nanotechnology Center, Discovery Park, West Lafayette, IN 47907 (United States); Skinner, C.H.; Kugel, H.W.; Kaita, R.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2013-12-15

    Lithium wall conditioning in NSTX has resulted in reduced divertor recycling, improved energy confinement, and reduced frequency of edge-localized modes (ELMs), up to the point of complete ELM suppression. NSTX tiles were removed from the vessel following the 2008 campaign and subsequently analyzed using X-ray photoelectron spectroscopy as well as nuclear reaction ion beam analysis. In this paper we relate surface chemistry to deuterium retention/recycling, develop methods for cleaning of passivated NSTX tiles, and explore a method to effectively extract bound deuterium from lithiated graphite. Li–O–D and Li–C–D complexes characteristic of deuterium retention that form during NSTX operations are revealed by sputter cleaning and heating. Heating to ∼850 °C desorbed all deuterium complexes observed in the O 1s and C 1s photoelectron energy ranges. Tile locations within approximately ±2.5 cm of the lower vertical/horizontal divertor corner appear to have unused Li-O bonds that are not saturated with deuterium, whereas locations immediately outboard of this region indicate high deuterium recycling. X-ray photo electron spectra of a specific NSTX tile with wide ranging lithium coverage indicate that a minimum lithium dose, 100–500 nm equivalent thickness, is required for effective deuterium retention. This threshold is suspected to be highly sensitive to surface morphology. The present analysis may explain why plasma discharges in NSTX continue to benefit from lithium coating thickness beyond the divertor deuterium ion implantation depth, which is nominally <10 nm.

  15. Dependence of recycling and edge profiles on lithium evaporation in high triangularity, high performance NSTX H-mode discharges

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R., E-mail: rmaingi@pppl.gov [Princeton Plasma Physics Laboratory, Receiving 3, Route 1 North, Princeton, NJ 08543 (United States); Osborne, T.H. [General Atomics, 3550 General Atomics Ct., San Diego, CA 92121 (United States); Bell, M.G.; Bell, R.E.; Boyle, D.P. [Princeton Plasma Physics Laboratory, Receiving 3, Route 1 North, Princeton, NJ 08543 (United States); Canik, J.M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Diallo, A.; Kaita, R.; Kaye, S.M.; Kugel, H.W.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Receiving 3, Route 1 North, Princeton, NJ 08543 (United States); Sabbagh, S.A. [Applied Physics and Applied Math Dept., Columbia University, New York, NY 10027 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Receiving 3, Route 1 North, Princeton, NJ 08543 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 7000 East Ave, PO Box 808, Livermore, CA 94551 (United States)

    2015-08-15

    In this paper, the effects of a pre-discharge lithium evaporation variation on highly shaped discharges in the National Spherical Torus Experiment (NSTX) are documented. Lithium wall conditioning (‘dose’) was routinely applied onto graphite plasma facing components between discharges in NSTX, partly to reduce recycling. Reduced D{sub α} emission from the lower and upper divertor and center stack was observed, as well as reduced midplane neutral pressure; the magnitude of reduction increased with the pre-discharge lithium dose. Improved energy confinement, both raw τ{sub E} and H-factor normalized to scalings, with increasing lithium dose was also observed. At the highest doses, we also observed elimination of edge-localized modes. The midplane edge plasma profiles were dramatically altered, comparable to lithium dose scans at lower shaping, where the strike point was farther from the lithium deposition centroid. This indicates that the benefits of lithium conditioning should apply to the highly shaped plasmas planned in NSTX-U.

  16. Heating and current drive on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Batchelor, D.; Carter, M.; Hosea, J.; Ignat, D.; LeBlanc, B.; Majeski, R.; Ono, M.; Phillips, C.K.; Rogers, J.H.; Schilling, G.

    1997-01-01

    Low aspect ratio tokamaks pose interesting new challenges for heating and current drive. The NSTX (National Spherical Tokamak Experiment) device to be built at Princeton is a low aspect ratio toroidal device that has the achievement of high toroidal beta (∼45%) and non-inductive operation as two of its main research goals. To achieve these goals significant auxiliary heating and current drive systems are required. Present plans include ECH (Electron cyclotron heating) for pre-ionization and start-up assist, HHFW (high harmonic fast wave) for heating and current drive and eventually NBI (neutral beam injection) for heating, current drive and plasma rotation. copyright 1997 American Institute of Physics

  17. High beta plasma operation in a toroidal plasma producing device

    International Nuclear Information System (INIS)

    Clarke, J.F.

    1978-01-01

    A high beta plasma is produced in a plasma producing device of toroidal configuration by ohmic heating and auxiliary heating. The plasma pressure is continuously monitored and used in a control system to program the current in the poloidal field windings. Throughout the heating process, magnetic flux is conserved inside the plasma and the distortion of the flux surfaces drives a current in the plasma. As a consequence, the total current increases and the poloidal field windings are driven with an equal and opposing increasing current. The spatial distribution of the current in the poloidal field windings is determined by the plasma pressure. Plasma equilibrium is maintained thereby, and high temperature, high beta operation results

  18. ECH on NSTX

    International Nuclear Information System (INIS)

    Bigelow, T.S.; Batchelor, D.B.; Carter, M.D.; Peng, M.; Wilson, J.R.

    1997-01-01

    Electron Cyclotron Heating has been proposed for plasma initiation, startup assistance and non-inductive startup on NSTX. One physics goal of NSTX will be to establish entirely non-inductive plasma operation by utilizing ECH to provide a sufficient start-up plasma to support further current drive from other heating systems. Scaling of previous ECH-only startup experiments on CDX-U and DIII-D indicate that 400 kW of ECH should be capable of driving 42 kA of pressure driven current on NSTX and possibly higher levels after optimizing the process. Due to the low NSTX magnetic field, over-dense plasmas exist during most of the discharge so conventional ECH operation is limited to the low density startup phase. To extend the useful operating range for ECH, a scheme involving mode conversion to the electron Bernstein Wave (EBW) from either O or X mode launch is being investigated for bulk heating and current drive applications at higher density. Microwave equipment, including 18 GHz klystrons and 28 GHz gyrotrons are available at ORNL and appear ideal for use on NSTX. Preliminary pre-ionization and start-up system configurations are presented here along with discussions on various operation modes. copyright 1997 American Institute of Physics

  19. Solenoid-free Plasma Startup in NSTX using Coaxial Helicity Injection

    International Nuclear Information System (INIS)

    Roger Raman; Jarboe, Thomas R.; Bell, Michael G.; Dennis Mueller; Nelson, Brian A.; Benoit LeBlanc; Charles Bush; Masayoshi Nagata; Ted Biewer

    2005-01-01

    The favorable properties of the Spherical Torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. Coaxial Helicity Injection (CHI) is a promising candidate for solenoid-free plasma startup in a ST. Recent experiments on the HIT-II ST at the University of Washington, have demonstrated the capability of a new method, referred to as transient CHI, to produce a high quality, closed-flux equilibrium that has then been coupled to induction, with a reduced requirement for transformer flux [R. Raman, T.R. Jarboe, B.A. Nelson, et al., Phys. Rev. Lett. 90 (February 2003) 075005-1]. An initial test of this method on the National Spherical Torus Experiment (NSTX) has produced about 140 kA of toroidal current. Modifications are now underway to improve capability for transient CHI in NSTX

  20. Ramp-up of CHI Initiated Plasmas on NSTX

    International Nuclear Information System (INIS)

    Mueller, D.; Bell, M.G.; Bell, R.E.; LeBlanc, B.; Roquemore, A.L.; Raman, R.; Jarboe, T.R.; Nelson, B.A.; Soukhanovskii, V.

    2009-01-01

    Experiments on the National Spherical Torus (NSTX) have now demonstrated flux savings using transient coaxial helicity injection (CHI). In these discharges, the discharges initiated by CHI are ramped up with an inductive transformer and exhibit higher plasma current than discharges without the benefit of CHI initiation.

  1. High spatial sampling global mode structure measurements via multichannel reflectometry in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Crocker, N A; Peebles, W A; Kubota, S; Zhang, J [Department of Physics and Astronomy, University of California-Los Angeles, Los Angeles, CA 90095-7099 (United States); Bell, R E; Fredrickson, E D; Gorelenkov, N N; LeBlanc, B P; Menard, J E; Podesta, M [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Sabbagh, S A [Department of Applied Physics and Applied Mathematics, Columbia University, New York, NY 10027 (United States); Tritz, K [Johns Hopkins University, Baltimore, MD 21218 (United States); Yuh, H [Nova Photonics, Princeton, NJ 08540 (United States)

    2011-10-15

    Global modes-including kinks and tearing modes (f <{approx} 50 kHz), toroidicity-induced Alfven eigenmodes (TAE; f {approx} 50-250 kHz) and global and compressional Alfven eigenmodes (GAE and CAE; f >{approx} 400 kHz)-play critical roles in many aspects of plasma performance. Their investigation on NSTX is aided by an array of fixed-frequency quadrature reflectometers used to determine their radial density perturbation structure. The array has been recently upgraded to 16 channels spanning 30-75 GHz (n{sub cutoff} = (1.1-6.9) x 10{sup 19} m{sup -3} in O-mode), improving spatial sampling and access to the core of H-mode plasmas. The upgrade has yielded significant new results that advance the understanding of global modes in NSTX. The GAE and CAE structures have been measured for the first time in the core of an NSTX high-power (6 MW) beam-heated H-mode plasma. The CAE structure is strongly core-localized, which has important implications for electron thermal transport. The TAE structure has been measured with greatly improved spatial sampling, and measurements of the TAE phase, the first in NSTX, show strong radial variation near the midplane, indicating radial propagation caused by non-ideal MHD effects. Finally, the tearing mode structure measurements provide unambiguous evidence of coupling to an external kink.

  2. Progress towards Steady State on NSTX

    International Nuclear Information System (INIS)

    Gates, D.A.; Kessel, C.; Menard, J.; Taylor, G.; Wilson, J.R.

    2005-01-01

    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 the National Spherical Torus Experiment (NSTX) has been raised from κ ∼ 2.1 to κ ∼ 2.6--approximately a 25% increase. This increase in elongation has lead to a doubling 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 β t 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 second. Data is 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 during and to delay the onset of MHD instabilities. A modeled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be presented. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity ((delta) ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement for steady state on NSTX is the ability to drive a fraction of the total plasma current with radio-frequency waves. The results of High Harmonic Fast Wave heating and current drive studies as well as electron Bernstein Wave emission studies will be presented

  3. Power exhaust scenarios and control for projected high-power NSTX-U operation

    Science.gov (United States)

    Menard, Jonathan; Gerhardt, S. P.; Myers, C. E.; Reinke, M. L.; Brooks, A.; Mardenfeld, M.; NSTX Upgrade Team

    2017-10-01

    An important goal of the NSTX Upgrade (NSTX-U) research program is to characterize energy confinement in the low-aspect-ratio spherical tokamak configuration over a significantly expanded range of plasma current, toroidal field, and heating power, while increasing flattop durations up to 5 seconds. However, the narrowing of the scrape-off layer at higher current combined with an improved understanding of expected halo-current loads has motivated a significant re-design of NSTX-U plasma facing components in the high-heat-flux regions of the divertor. In order to reduce the expected divertor heat flux to acceptable levels, a combination of mitigation techniques will be used: increased divertor poloidal flux expansion, increased divertor radiation, and controlled strike-point sweeping. The machine requirements for these various mitigation techniques are studied here using a newly implemented reduced heat-flux model. Systematic equilibrium scans are used to quantify the required divertor coil currents and to verify vertical stability for a range of plasma shapes. Free-boundary control schemes to constrain the strike-point location and field-line angle-of-incidence will also be discussed. Work supported by DOE contract DE-AC02- 09CH11466.

  4. Making of the NSTX Facility

    International Nuclear Information System (INIS)

    Neumeyer, C.; Ono, M.; Kaye, S.M.; Peng, Y.-K.M.

    1999-01-01

    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

  5. Construction of a high beta plasma source

    International Nuclear Information System (INIS)

    Naraghi, M.; Torabi-Fard, A.

    1976-02-01

    A high beta plasma source has been designed and constructed. This source will serve as a means of developing and exercising different diagnostic techniques as required for ALVAND I, linear theta pinch experiment. Also, it will serve to acquaint the technicians with some of the techniques and safety rules of high voltage and capacitor discharge experiments. The operating parameters of the theta pinch and Z-pinch preionization is presented and the program of diagnostic measurements on the high beta plasma source is discussed

  6. National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Masayuki Ono

    2000-01-01

    The main aim of National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the innovative spherical torus (ST) concept. Physics outcome of the NSTX research program is relevant to near-term applications such as the Volume Neutron Source (VNS) and burning plasmas, and future applications such as the pilot and power plants. The NSTX device began plasma operations in February 1999 and the plasma current was successfully ramped up to the design value of 1 million amperes (MA) on December 14, 1999. The CHI (Coaxial Helicity Injection) and HHFW (High Harmonic Fast Wave) experiments have also started. Stable CHI discharges of up to 133 kA and 130-msec duration have been produced using 20 kA of injected current. Using eight antennas connected to two transmitters, up to 2 MW of HHFW power was successfully coupled to the plasma. The Neutral-beam Injection (NBI) heating system and associated NBI-based diagnostics such as the Charge-exchange Recombination Spectrometer (CHERS) will be operational in October 2000

  7. Overview of impurity control and wall conditioning in NSTX

    International Nuclear Information System (INIS)

    Kugel, H.W.; Maingi, R.; Wampler, W.; Barry, R.E.; Bell, M.; Blanchard, W.; Gates, D.; Johnson, D.; Kaita, R.; Kaye, S.; Maqueda, R.; Menard, J.; Menon, M.M.; Mueller, D.; Ono, M.; Paul, S.; Peng, Y-K.M.; Raman, R.; Roquemore, A.; Skinner, C. H.; Sabbagh, S.; Stratton, B.; Stutman, D.; Wilson, J. R.; Zweben, S.

    2000-01-01

    The National Spherical Torus Experiment (NSTX) started plasma operations i n February 1999. In the first extended period of experiments, NSTX achieved high current, inner wall limited, double null, and single null plasma discharges, initial Coaxial Helicity Injection, and High Harmonic Fast Wave results. As expected, discharge reproducibility and performance were strongly affected by wall conditions. In this paper, the authors describe the internal geometry, and initial plasma discharge, impurity control, wall conditioning, erosion, and deposition results

  8. ECRH/EBWH system for NSTX-U

    Directory of Open Access Journals (Sweden)

    Hosea J.C.

    2012-09-01

    Full Text Available The National Spherical Torus Experiment Upgrade (NSTX-U will operate at an axial toroidal field of up to 1 T, about twice the field available on NSTX. A 28 GHz electron cylotron resonance heating (ECRH system is currently being planned for NSTX-U. A 1 MW 28 GHz gyrotron will be employed. Intially the system will use short, 10-50 ms, 1 MW pulses for ECRH-assisted discharge start-up. Later the pulse length will be extended to 1-5 s to study electron Bernstein wave heating (EBWH during the plasma current flat top. A mirror launcher will be used to couple microwave power to the plasma via O-mode to the slow X-mode to EBW (O-X-B double mode conversion. This paper presents a pre-conceptual design for the ECRH/EBWH system proposed for NSTX-U and includes ray tracing and Fokker-Planck modeling results for 28 GHz ECRH during plasma start-up and EBW heating and current drive during the plasma current flattop of a NSTX-U advanced H-mode plasma scenario.

  9. H-Mode Turbulence, Power Threshold, ELM, and Pedestal Studies in NSTX

    International Nuclear Information System (INIS)

    Maingi, R.; Bush, C.E.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Menard, J.E.; Meyer, H.; Mueller, D.; Nishino, N.; Roquemore, A.L.; Sabbagh, S.A.; Tritz, K.; Zweben, S.J.; Bell, M.G.; Bell, R.E.; Biewer, T.; Boedo, J.A.; Johnson, D.W.; Kaita, R.; Kugel, H.W.; Maqueda, R.J.; Munsat, T.; Raman, R.; Soukhanovskii, V.A.; Stevenson, T.; Stutman, D.

    2004-01-01

    High-confinement mode (H-mode) operation plays a crucial role in NSTX [National Spherical Torus Experiment] research, allowing higher beta limits due to reduced plasma pressure peaking, and long-pulse operation due to high bootstrap current fraction. Here, new results are presented in the areas of edge localized modes (ELMs), H-mode pedestal physics, L-H turbulence, and power threshold studies. ELMs of several other types (as observed in conventional aspect ratio tokamaks) are often observed: (1) large, Type I ELMs, (2) ''medium'' Type II/III ELMs, and (3) giant ELMs which can reduce stored energy by up to 30% in certain conditions. In addition, many high-performance discharges in NSTX have tiny ELMs (newly termed Type V), which have some differences as compared with ELM types in the published literature. The H-mode pedestal typically contains between 25-33% of the total stored energy, and the NSTX pedestal energy agrees reasonably well with a recent international multi-machine scaling. We find that the L-H transition occurs on a ∼100 (micro)sec timescale as viewed by a gas puff imaging diagnostic, and that intermittent quiescent periods precede the final transition. A power threshold identity experiment between NSTX and MAST shows comparable loss power at the L-H transition in balanced double-null discharges. Both machines require more power for the L-H transition as the balance is shifted toward lower single null. High field side gas fueling enables more reliable H-mode access, but does not always lead to a lower power threshold e.g., with a reduction of the duration of early heating. Finally the edge plasma parameters just before the L-H transition were compared with theories of the transition. It was found that while some theories can separate well-developed L- and H-mode data, they have little predictive value

  10. Effect of Boronization on Ohmic Plasmas in NSTX

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Maingi, R.; Wampler, W.R.; Blanchard, W.; Bell, M.; Bell, R.; LeBlanc, B.; Gates, D.; Kaye, S.; LaMarche, P.; Menard, J.; Mueller, D.; Na, H.K.; Nishino, N.; Paul, S.; Sabbagh, S.; Soukhanovskii, V.

    2001-01-01

    Boronization of the National Spherical Torus Experiment (NSTX) has enabled access to higher density, higher confinement plasmas. A glow discharge with 4 mTorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B+C areal density of 10 to the 18 (B+C) cm to the -2 corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by x15, carbon emission reduced by two and copper reduced to undetectable levels. After boronization, the plasma current flattop time increased by 70% enabling access to higher density, higher confinement plasmas

  11. Edge Ion Heating by Launched High Harmonic Fast Waves in NSTX

    International Nuclear Information System (INIS)

    Biewer, T.M.; Bell, R.E.; Diem, S.J.; Phillips, C.K.; Wilson, J.R.; Ryan, P.M.

    2004-01-01

    A new spectroscopic diagnostic on the National Spherical Torus Experiment (NSTX) measures the velocity distribution of ions in the plasma edge simultaneously along both poloidal and toroidal views. An anisotropic ion temperature is measured during 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 of ions are present and have temperatures of typically 500 eV and 50 eV with rotation velocities of -50 km/s and -10 km/s, respectively (predominantly perpendicular to the local magnetic field). This bi-modal distribution is observed in both the toroidal and poloidal views (for both He + and C 2+ ions), and is well correlated with the period of rf power application to the plasma. The temperature of the hot component is observed to increase with the applied rf power, which was scanned between 0 and 4.3 MW . The 30 MHz HHFW launched by the NSTX antenna is expected and observed to heat core electrons, but plasma ions do not resonate with the launched wave, which is typically at >10th harmonic of the ion cyclotron frequency in the region of observation. A likely ion heating mechanism is parametric decay of the launched HHFW into an Ion Bernstein Wave (IBW). The presence of the IBW in NSTX plasmas during HHFW application has been directly confirmed with probe measurements. IBW heating occurs in the perpendicular ion distribution, consistent with the toroidal and poloidal observations. Calculations of IBW propagation indicate that multiple waves could be created in the parametric decay process, and that most of the IBW power would be absorbed in the outer 10 to 20 cm of the plasma, predominantly on fully stripped ions. These predictions are in qualitative agreement with the observations, and must be accounted for when calculating the energy budget of the plasma

  12. Tokamak Simulation Code modeling of NSTX

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.

    2000-01-01

    The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption

  13. High beta plasmas in the PBX tokamak

    International Nuclear Information System (INIS)

    Bol, K.; Buchenauer, D.; Chance, M.

    1986-04-01

    Bean-shaped configurations favorable for high β discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present β limit

  14. Investigation of EBW Thermal Emission and Mode Conversion Physics in H-Mode Plasmas on NSTX

    International Nuclear Information System (INIS)

    Diem, S.J.; Taylor, G.; Efthimion, P.C.; Kugel, H.W.; LeBlanc, B.P.; Phillips, C.K.; Caughman, J.B.; Wilgen, J.B.; Harvey, R.W.; Preinhaelter, J.; Urban, J.; Sabbagh, S.A.

    2008-01-01

    High β plasmas in the National Spherical Torus Experiment (NSTX) operate in the overdense regime, allowing the electron Bernstein wave (EBW) to propagate and be strongly absorbed/emitted at the electron cyclotron resonances. As such, EBWs may provide local electron heating and current drive. For these applications, efficient coupling between the EBWs and electromagnetic waves outside the plasma is needed. Thermal EBW emission (EBE) measurements, via oblique B-X-O double mode conversion, have been used to determine the EBW transmission efficiency for a wide range of plasma conditions on NSTX. Initial EBE measurements in H-mode plasmas exhibited strong emission before the L-H transition, but the emission rapidly decayed after the transition. EBE simulations show that collisional damping of the EBW prior to the mode conversion (MC) layer can significantly reduce the measured EBE for T e < 20 eV, explaining the observations. Lithium evaporation was used to reduce EBE collisional damping near the MC layer. As a result, the measured B-X-O transmission efficiency increased from < 10% (no Li) to 60% (with Li), consistent with EBE simulations.

  15. Edge Plasma Simulations in NSTX and CTF: Synergy of Lithium Coating, Non-Diffusive Anomalous Transport and Drifts. Final Technical Report

    International Nuclear Information System (INIS)

    Pigarov, Alexander

    2012-01-01

    This is the final report for the Research Grant DE-FG02-08ER54989 'Edge Plasma Simulations in NSTX and CTF: Synergy of Lithium Coating, Non-Diffusive Anomalous Transport and Drifts'. The UCSD group including: A.Yu. Pigarov (PI), S.I. Krasheninnikov and R.D. Smirnov, was working on modeling of the impact of lithium coatings on edge plasma parameters in NSTX with the multi-species multi-fluid code UEDGE. The work was conducted in the following main areas: (i) improvements of UEDGE model for plasma-lithium interactions, (ii) understanding the physics of low-recycling divertor regime in NSTX caused by lithium pumping, (iii) study of synergistic effects with lithium coatings and non-diffusive ballooning-like cross-field transport, (iv) simulation of experimental multi-diagnostic data on edge plasma with lithium pumping in NSTX via self-consistent modeling of D-Li-C plasma with UEDGE, and (v) working-gas balance analysis. The accomplishments in these areas are given in the corresponding subsections in Section 2. Publications and presentations made under the Grant are listed in Section 3.

  16. High-beta plasma blobs in the morningside plasma sheet

    Directory of Open Access Journals (Sweden)

    G. Haerendel

    1999-12-01

    Full Text Available Equator-S frequently encountered, i.e. on 30% of the orbits between 1 March and 17 April 1998, strong variations of the magnetic field strength of typically 5–15-min duration outside about 9RE during the late-night/early-morning hours. Very high-plasma beta values were found, varying between 1 and 10 or more. Close conjunctions between Equator-S and Geotail revealed the spatial structure of these "plasma blobs" and their lifetime. They are typically 5–10° wide in longitude and have an antisymmetric plasma or magnetic pressure distribution with respect to the equator, while being altogether low-latitude phenomena  (≤ 15°. They drift slowly sunward, exchange plasma across the equator and have a lifetime of at least 15–30 min. While their spatial structure may be due to some sort of mirror instability, little is known about the origin of the high-beta plasma. It is speculated that the morningside boundary layer somewhat further tailward may be the source of this plasma. This would be consistent with the preference of the plasma blobs to occur during quiet conditions, although they are also found during substorm periods. The relation to auroral phenomena in the morningside oval is uncertain. The energy deposition may be mostly too weak to generate a visible signature. However, patchy aurora remains a candidate for more disturbed periods.Key words. Magnetospheric physics (plasma convection; plasma sheet; plasma waves and instabilities

  17. Characteristics of MHD stability of high beta plasmas in LHD

    International Nuclear Information System (INIS)

    Sato, M.; Nakajima, N.; Watanabe, K.Y.; Todo, Y.; Suzuki, Y.

    2012-11-01

    In order to understand characteristics of the MHD stability of high beta plasmas obtained in the LHD experiments, full MHD simulations have been performed for the first time. Since there is a magnetic hill in a plasma peripheral region, the ballooning modes extending into the plasma peripheral region with a chaotic magnetic field are destabilized. However, in the nonlinear phase, the core region comes under the in influence of the instabilities and the central pressure decreases. There is a tendency that modes are suppressed as the beta value and/or magnetic Reynolds number increase, which is consistent with a result that high beta plasmas enter the second stable region of the ideal ballooning modes as beta increases and remaining destabilized ballooning modes are considered to be resistive type. (author)

  18. Lithium coatings on NSTX plasma facing components and its effects on boundary control, core plasma performance, and operation

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Schneider, H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Raman, R. [University of Washington, Seattle, WA 98195 (United States); Sabbagh, S. [Columbia University, New York, NY 10027 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2010-11-15

    NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.

  19. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    Energy Technology Data Exchange (ETDEWEB)

    H.W.Kugel, M.G.Bell, H.Schneider, J.P.Allain, R.E.Bell, R Kaita, J.Kallman, S. Kaye, B.P. LeBlanc, D. Mansfield, R.E. Nygen, R. Maingi, J. Menard, D. Mueller, M. Ono, S. Paul, S.Gerhardt, R.Raman, S.Sabbagh, C.H.Skinner, V.Soukhanovskii, J.Timberlake, L.E.Zakharov, and the NSTX Research Team

    2010-01-25

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  20. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Schneider, H.; Allain, J.P.; Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D.; Nygen, R.E.; Maingi, R.; Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S.; Raman, R.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Timberlake, J.; Zakharov, L.E.; NSTX Research Team

    2010-01-01

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  1. Analysis of NSTX TF Joint Voltage Measurements

    International Nuclear Information System (INIS)

    Woolley R

    2005-01-01

    This report presents findings of analyses of recorded current and voltage data associated with 72 electrical joints operating at high current and high mechanical stress. The analysis goal was to characterize the mechanical behavior of each joint and thus evaluate its mechanical supports. The joints are part of the toroidal field (TF) magnet system of the National Spherical Torus Experiment (NSTX) pulsed plasma device operating at the Princeton Plasma Physics Laboratory (PPPL). Since there is not sufficient space near the joints for much traditional mechanical instrumentation, small voltage probes were installed on each joint and their voltage monitoring waveforms have been recorded on sampling digitizers during each NSTX ''shot''

  2. Diagnostics of ST Plasmas in NSTX: Challenges and Opportunities

    International Nuclear Information System (INIS)

    Johnson, D.; Efthimion, P.; Foley, J.; Jones, B.; Mazzucato, E.; Park, H.; Taylor, G.; Levinton, F.; Luhmann, N.

    2001-01-01

    This paper will highlight some of the challenges and opportunities present in the diagnosis of spherical torus (ST) plasmas on the National Spherical Torus Experiment (NSTX) and discuss the corresponding diagnostic development that is presently underway. After a brief description of diagnostic systems currently installed, examples of ST-specific diagnostic challenges will be highlighted, as will another case, where the ST configuration offers opportunities for new measurements

  3. Plasma Start-up in HIT-II and NSTX using Transient Coaxial Helicity Injection

    International Nuclear Information System (INIS)

    Raman, R.; Jarboe, T.R.; Nelson, B.A.; Mueller, D.; Bell, M.G.; Ono, M.

    2008-01-01

    The method of transient coaxial helicity injection (CHI) has previously been used in the HITII experiment at the University of Washington to produce 100 kA of closed flux current. The generation of the plasma current by CHI involves the process of magnetic reconnection, which has been experimentally controlled in the National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Laboratory to allow this potentially unstable phenomenon to reorganize the magnetic field lines to form closed, nested magnetic surfaces carrying a plasma current up to 160 kA. This is a world record for non-inductive closed-flux current generation, and demonstrates the high current capability of this method

  4. RF heating and current drive on NSTX with high harmonic fast waves

    International Nuclear Information System (INIS)

    Ryan, P.M.

    2002-01-01

    NSTX is a small aspect ratio tokamak with a large dielectric constant (50-100); under these conditions high harmonic fast waves (HHFW) will readily damp on electrons via Landau damping and TTMP. The HHFW system is a 30 MHz, 12-element array capable of launching both symmetric and directional wave spectra for plasma heating and non-inductive current drive. It has delivered up to 6 MW for short pulses and has routinely operated at ∼3-4 MW for 100-200 ms pulses. Results include strong, centrally-peaked electron heating in both D and He plasmas, for both high and low phase velocity spectra. H-modes were obtained with application of HHFW power alone, with stored energy doubling after the L-H transition. Beta poloidal as large as unity has been obtained with large fractions (0.4) of bootstrap current. A fast ion tail with energies extending up to 140 keV has been observed when HHFW interacts with 80 keV neutral beams; neutron rate and lost ion measurements, as well as modeling, indicate significant power absorption by the fast ions. Radial power deposition profiles are being calculated with ray tracing and kinetic full-wave codes and benchmarked against measurements. (author)

  5. Confinement and Local Transport in the National Spherical Torus Experiment NSTX

    International Nuclear Information System (INIS)

    Kaye, S.M.; Levinton, F.M.; Stutman, D.; Tritz, K.; Yuh, H.; Bell, M.G.; Bell, R.E.; Domier, C.W.; Gates, D.; Horton, W.; Kim, J.; LeBlanc, B.P.; Luhmann, N.C. Jr.; Maingi, T.; Mazzucato, E.; Menard, J.E.; Mikkelsen, D.; Mueller, D; Park, H.; Rewoldt, G.; Sabbagh, S.A.; Smith, D.R.; Wang, W.

    2007-01-01

    NSTX operates at low aspect ratio (R/a∼1.3) and high beta (up to 40%), allowing tests of global confinement and local transport properties that have been established from higher aspect ratio devices. NSTX plasmas are heated by up to 7 MW of deuterium neutral beams with preferential electron heating as expected for ITER. Confinement scaling studies indicate a strong B T dependence, with a current dependence that is weaker than that observed at higher aspect ratio. Dimensionless scaling experiments indicate a strong increase of confinement with decreasing collisionality and a weak degradation with beta. The increase of confinement with B T is due to reduced transport in the electron channel, while the improvement with plasma current is due to reduced transport in the ion channel related to the decrease in the neoclassical transport level. Improved electron confinement has been observed in plasmas with strong reversed magnetic shear, showing the existence of an electron internal transport barrier (eITB). The development of the eITB may be associated with a reduction in the growth of microtearing modes in the plasma core. Perturbative studies show that while L-mode plasmas with reversed magnetic shear and an eITB exhibit slow changes of L Te across the profile after the pellet injection, H-mode plasmas with a monotonic q-profile and no eITB show no change in this parameter after pellet injection, indicating the existence of a critical gradient that may be related to the q-profile. Both linear and non-linear simulations indicate the potential importance of ETG modes at the lowest B T . Localized measurements of high-k fluctuations exhibit a sharp decrease in signal amplitude levels across the L-H transition, associated with a decrease in both ion and electron transport, and a decrease in calculated linear microinstability growth rates across a wide k-range, from the ITG/TEM regime up to the ETG regime

  6. High-beta plasma blobs in the morningside plasma sheet

    Directory of Open Access Journals (Sweden)

    G. Haerendel

    Full Text Available Equator-S frequently encountered, i.e. on 30% of the orbits between 1 March and 17 April 1998, strong variations of the magnetic field strength of typically 5–15-min duration outside about 9RE during the late-night/early-morning hours. Very high-plasma beta values were found, varying between 1 and 10 or more. Close conjunctions between Equator-S and Geotail revealed the spatial structure of these "plasma blobs" and their lifetime. They are typically 5–10° wide in longitude and have an antisymmetric plasma or magnetic pressure distribution with respect to the equator, while being altogether low-latitude phenomena 
    (≤ 15°. They drift slowly sunward, exchange plasma across the equator and have a lifetime of at least 15–30 min. While their spatial structure may be due to some sort of mirror instability, little is known about the origin of the high-beta plasma. It is speculated that the morningside boundary layer somewhat further tailward may be the source of this plasma. This would be consistent with the preference of the plasma blobs to occur during quiet conditions, although they are also found during substorm periods. The relation to auroral phenomena in the morningside oval is uncertain. The energy deposition may be mostly too weak to generate a visible signature. However, patchy aurora remains a candidate for more disturbed periods.

    Key words. Magnetospheric physics (plasma convection; plasma sheet; plasma waves and instabilities

  7. Solenoid-free Plasma Start-up in NSTX using Transient CHI

    International Nuclear Information System (INIS)

    R. Raman, B.A. Nelson, D. Mueller, T.R. Jarboe, M.G. Bell, B. LeBlanc, R. Maqueda, J. Menard, M. Ono, M. Nagata, L. Roquemore, and V. Soukhanovskii

    2008-01-01

    Experiments in NSTX have now unambiguously demonstrated the coupling of toroidal plasmas produced by the technique of CHI to inductive sustainment and ramp-up of the toroidal plasma current. This is an important step because an alternate method for plasma startup is essential for developing a fusion reactor based on the spherical torus concept. Elimination of the central solenoid would also allow greater flexibility in the choice of the aspect ratio in tokamak designs now being considered. The transient CHI method for spherical torus startup was originally developed on the HIT-II experiment at the University of Washington

  8. CHI Research on NSTX-U

    Science.gov (United States)

    Lay, W.-S.; Raman, R.; Jarboe, T. R.; Nelson, B. A.; Mueller, D.; Ebrahimi, F.; Ono, M.; Jardin, S. C.; Taylor, G.

    2017-10-01

    At present about 20% of the total plasma current required for sustained operation has been generated by transient CHI. The present understanding suggests that it may be possible to generate all of the needed current in a ST / tokamak using transient CHI. In such a scenario, one could transition directly from a CHI produced plasma to a non-inductively sustained plasma, without the difficult intermediate step that involves non-inductive current ramp-up. STs based on this new configuration would take advantage of evolving developments in high-temperature superconductor technology to develop a simpler design ST that relies primarily on CHI for plasma current generation. Motivated by the very good results from NSTX and HIT-II, we are examining the potential application of transient CHI for reactor configurations through these studies. (1) Study of the maximum levels of start-up currents that could be generated on NSTX-U, (2) application of a single biased electrode configuration on QUEST to protect the insulator from neutron damage in a CHI reactor installation, and (3) QUEST-like, but a double biased electrode configuration for PEGASUS and NSTX-U. Results from these on-going studies will be described. This work is supported by U.S. DOE Contracts: DE-AC02-09CH11466, DE-FG02-99ER54519 AM08, and DE-SC0006757.

  9. Be Foil ''Filter Knee Imaging'' NSTX Plasma with Fast Soft X-ray Camera

    International Nuclear Information System (INIS)

    B.C. Stratton; S. von Goeler; D. Stutman; K. Tritz; L.E. Zakharov

    2005-01-01

    A fast soft x-ray (SXR) pinhole camera has been implemented on the National Spherical Torus Experiment (NSTX). This paper presents observations and describes the Be foil Filter Knee Imaging (FKI) technique for reconstructions of a m/n=1/1 mode on NSTX. The SXR camera has a wide-angle (28 o ) field of view of the plasma. The camera images nearly the entire diameter of the plasma and a comparable region in the vertical direction. SXR photons pass through a beryllium foil and are imaged by a pinhole onto a P47 scintillator deposited on a fiber optic faceplate. An electrostatic image intensifier demagnifies the visible image by 6:1 to match it to the size of the charge-coupled device (CCD) chip. A pair of lenses couples the image to the CCD chip

  10. NSTX Overview

    International Nuclear Information System (INIS)

    M. Ono; M. Bell; R.E. Bell; M. Bitter; C. Bourdelle; D. Darrow; D. Gates; J. Hosea; S.M. Kaye; R. Kaita; H. Kugel; D. Johnson; B. LeBlanc; S. Medley

    2001-01-01

    The National Spherical Torus Experiment (NSTX) has had a very productive period of plasma operations since the last ST Workshop in Seattle, WA, in November 1999. A number of new research tools have become available and the plasma parameters have improved significantly. These advances are describe in this paper

  11. Stability Limits of High-Beta Plasmas in DIII-D

    International Nuclear Information System (INIS)

    Strait, E.J.

    2005-01-01

    Stability at high beta is an important requirement for a compact, economically attractive fusion reactor. DIII-D experiments have shown that ideal magnetohydrodynamic (MHD) theory is an accurate predictor of the ultimate stability limits for tokamaks, and the Troyon scaling law has provided a useful approximation of ideal stability limits for discharges with 'conventional' profiles. However, variation of the discharge shape, pressure profile, and current density profile can lead to ideal MHD beta limits that differ significantly from simple Troyon scaling. The need for profiles consistent with steady-state operation places an important additional constraint on plasma stability. Nonideal effects can also be important and must be taken into account. For example, neoclassical tearing modes (NTMs), resulting from plasma resistivity and the nonlinear effects of the bootstrap current, can become unstable at beta values well below the ideal MHD limit. DIII-D experiments are now entering a new era of unprecedented control over plasma stability, including suppression of NTMs by localized current drive at the island location, and direct feedback stabilization of kink modes with a resistive wall. The continuing development of physics understanding and control tools holds the potential for stable, steady-state fusion plasmas at high beta

  12. NSTX plasma operation with a Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Ellis, R.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Maingi, R.; McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M.; Paul, S.F. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); and others

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer NSTX 2010 experiments tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium molybdenum divertor surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. Black-Right-Pointing-Pointer Noteworthy improvements in plasma performance with the plasma strike point on the liquid lithium molybdenum divertor were obtained similar to those obtained previously with lithiated graphite. The role of lithium impurities in this result is discussed. Black-Right-Pointing-Pointer Inspection of the liquid lithium molybdenum divertor after the Campaign indicated mechanical damage to supports, and other hardware resulting from forces following plasma current disruptions. - Abstract: NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the

  13. Profile Modifications Resulting from Early High-harmonic Fast Wave heating in NSTX

    International Nuclear Information System (INIS)

    Mendard, J.E.; LeBlanc, Wilson J.R.; Sabbagh, S.A.; Stutman, D.; Swain, D.W.

    2001-01-01

    Experiments have been performed in the National Spherical Torus Experiment (NSTX) to inject high harmonic fast wave (HHFW) power early during the plasma current ramp-up in an attempt to reduce the current penetration rate to raise the central safety factor during the flattop phase of the discharge. To date, up to 2 MW of HHFW power has been coupled to deuterium plasmas as early as t = 50 ms using the slowest interstrap phasing of k|| approximately equals 14 m(superscript)-1 (nf = 24). Antenna-plasma gap scans have been performed and find that for small gaps (5-8 cm), electron heating is observed with relatively small density rises and modest reductions in current penetration rate. For somewhat larger gaps (10-12 cm), weak electron heating is observed but with a spontaneous density rise at the plasma edge similar to that observed in NSTX H-modes. In the larger gap configuration, EFIT code reconstructions (without MSE [motional Stark effect]) find that resistive flux consumption is reduced as much as 30%, the internal inductance is maintained below 0.6 at 1 MA into the flattop, q(0) is increased significantly, and the MHD stability character of the discharges is strongly modified

  14. An overview of recent physics results from NSTX

    Science.gov (United States)

    Kaye, S. M.; Abrams, T.; Ahn, J.-W.; Allain, J. P.; Andre, R.; Andruczyk, D.; Barchfeld, R.; Battaglia, D.; Bhattacharjee, A.; Bedoya, F.; Bell, R. E.; Belova, E.; Berkery, J.; Berry, L.; Bertelli, N.; Beiersdorfer, P.; Bialek, J.; Bilato, R.; Boedo, J.; Bonoli, P.; Boozer, A.; Bortolon, A.; Boyer, M. D.; Boyle, D.; Brennan, D.; Breslau, J.; Brooks, J.; Buttery, R.; Capece, A.; Canik, J.; Chang, C. S.; Crocker, N.; Darrow, D.; Davis, W.; Delgado-Aparicio, L.; Diallo, A.; D'Ippolito, D.; Domier, C.; Ebrahimi, F.; Ethier, S.; Evans, T.; Ferraro, N.; Ferron, J.; Finkenthal, M.; Fonck, R.; Fredrickson, E.; Fu, G. Y.; Gates, D.; Gerhardt, S.; Glasser, A.; Gorelenkov, N.; Gorelenkova, M.; Goumiri, I.; Gray, T.; Green, D.; Guttenfelder, W.; Harvey, R.; Hassanein, A.; Heidbrink, W.; Hirooka, Y.; Hooper, E. B.; Hosea, J.; Humphreys, D.; Jaeger, E. F.; Jarboe, T.; Jardin, S.; Jaworski, M. A.; Kaita, R.; Kessel, C.; Kim, K.; Koel, B.; Kolemen, E.; Kramer, G.; Ku, S.; Kubota, S.; LaHaye, R. J.; Lao, L.; LeBlanc, B. P.; Levinton, F.; Liu, D.; Lore, J.; Lucia, M.; Luhmann, N., Jr.; Maingi, R.; Majeski, R.; Mansfield, D.; Maqueda, R.; McKee, G.; Medley, S.; Meier, E.; Menard, J.; Mueller, D.; Munsat, T.; Muscatello, C.; Myra, J.; Nelson, B.; Nichols, J.; Ono, M.; Osborne, T.; Park, J.-K.; Peebles, W.; Perkins, R.; Phillips, C.; Podesta, M.; Poli, F.; Raman, R.; Ren, Y.; Roszell, J.; Rowley, C.; Russell, D.; Ruzic, D.; Ryan, P.; Sabbagh, S. A.; Schuster, E.; Scotti, F.; Sechrest, Y.; Shaing, K.; Sizyuk, T.; Sizyuk, V.; Skinner, C.; Smith, D.; Snyder, P.; Solomon, W.; Sovenic, C.; Soukhanovskii, V.; Startsev, E.; Stotler, D.; Stratton, B.; Stutman, D.; Taylor, C.; Taylor, G.; Tritz, K.; Walker, M.; Wang, W.; Wang, Z.; White, R.; Wilson, J. R.; Wirth, B.; Wright, J.; Yuan, X.; Yuh, H.; Zakharov, L.; Zweben, S. J.

    2015-10-01

    The National Spherical Torus Experiment (NSTX) is currently being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive operation in high-performance plasmas, assessing plasma-material interface (PMI) solutions to handle the high heat loads expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection (CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet-Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Toroidal Alfvén eigenmode avalanches and higher frequency Alfvén eigenmodes can affect NB-driven current through energy loss and redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat

  15. NSTX High Temperature Sensor Systems

    International Nuclear Information System (INIS)

    McCormack, B.; Kugel, H.W.; Goranson, P.; Kaita, R.

    1999-01-01

    The design of the more than 300 in-vessel sensor systems for the National Spherical Torus Experiment (NSTX) has encountered several challenging fusion reactor diagnostic issues involving high temperatures and space constraints. This has resulted in unique miniature, high temperature in-vessel sensor systems mounted in small spaces behind plasma facing armor tiles, and they are prototypical of possible high power reactor first-wall applications. In the Center Stack, Divertor, Passive Plate, and vessel wall regions, the small magnetic sensors, large magnetic sensors, flux loops, Rogowski Coils, thermocouples, and Langmuir Probes are qualified for 600 degrees C operation. This rating will accommodate both peak rear-face graphite tile temperatures during operations and the 350 degrees C bake-out conditions. Similar sensor systems including flux loops, on other vacuum vessel regions are qualified for 350 degrees C operation. Cabling from the sensors embedded in the graphite tiles follows narrow routes to exit the vessel. The detailed sensor design and installation methods of these diagnostic systems developed for high-powered ST operation are discussed

  16. RF heating and current drive on NSTX with high harmonic fast waves

    International Nuclear Information System (INIS)

    Ryan, P.M.; Swain, D.W.; Rosenberg, A.L.

    2003-01-01

    NSTX is a small aspect ratio tokamak (R = 0.85 m, a = 0.65 m). The High Harmonic Fast Wave (HHFW) system is a 30 MHz, 12-element array capable of launching both symmetric and directional wave spectra for plasma heating and non-inductive current drive. It has delivered up to 6 MW for short pulses and has routinely operated at ∼3 MW for 100-400 ms pulses. Results include strong, centrally-peaked electron heating in both D and He plasmas for both high and low phase velocity spectra. H-modes were obtained with application of HHFW power alone, with stored energy doubling after the L-H transition. Beta poloidal as large as unity has been obtained with significant fractions (0.4) of bootstrap current. Differences in the loop voltage are observed depending on whether the array is phased to drive current in the co- or counter-current directions. A fast ion tail with energies extending up to 140 keV has been observed when HHFW interacts with 80 keV neutral beams; neutron rate and lost ion measurements, as well as modeling, indicate significant power absorption by the fast ions. Radial rf power deposition and driven current profiles have been calculated with ray tracing and kinetic full-wave codes and compared with measurements. (author)

  17. Overview of the NSTX Control System

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.; Oliaro, G.; Roney, P.

    2001-01-01

    The National Spherical Torus Experiment (NSTX) is an innovative magnetic fusion device that was constructed by the Princeton Plasma Physics Laboratory (PPPL) in collaboration with the Oak Ridge National Laboratory, Columbia University, and the University of Washington at Seattle. Since achieving first plasma in 1999, the device has been used for fusion research through an international collaboration of more than twenty institutions. The NSTX is operated through a collection of control systems that encompass a wide range of technology, from hardwired relay controls to real-time control systems with giga-FLOPS of capability. This paper presents a broad introduction to the control systems used on NSTX, with an emphasis on the computing controls, data acquisition, and synchronization systems

  18. Snowflake Divertor Configuration in NSTX

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Ahn, Joonwook; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.; Maingi, Rajesh; Maqueda, R.J.; McLean, Adam G.; Menard, J.E.; Mueller, D.; Paul, S.F.; Raman, R.; Roquemore, L.; Ryutov, D.D.; Scott, H.A.

    2011-01-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  19. 'Snowflake' divertor configuration in NSTX

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Ahn, J.-W.; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J.E.; Mueller, D.M.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Ryutov, D.D.; Scott, H.A.

    2011-01-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  20. "Snowflake" divertor configuration in NSTX

    Science.gov (United States)

    Soukhanovskii, V. A.; Ahn, J.-W.; Bell, R. E.; Gates, D. A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H. W.; Leblanc, B. P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J. E.; Mueller, D. M.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Ryutov, D. D.; Scott, H. A.

    2011-08-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel "snowflake" divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  1. The national spherical torus experiment (NSTX) research programme and progress towards high beta, long pulse operating scenarios

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Bell, M.G.; Bell, R.E.

    2003-01-01

    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 β T ≡ /(B T0 2 /2μ 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 comparison 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

  2. High Speed Images of Edge Plasmas in NSTX and Alcator C-Mod

    International Nuclear Information System (INIS)

    Maqueda, R.J.; Grulke, O.; Terry, J.L.; Zweben, S.J.

    2007-01-01

    This talk will describe the high speed imaging diagnostics on NSTX and Alcator C-Mod and show movies of various edge phenomena, including turbulence during L-modes and H modes, L-H and H-L transitions, effects of MHD activity and ELMs of various types, and wide angle views of the toroidal vs. poloidal structure of these edge '' filaments ''. Issues concerning the interpretation of these images will be discussed. (author)

  3. Wall stabilization of high beta plasmas in DIII-D

    International Nuclear Information System (INIS)

    Taylor, T.S.; Strait, E.J.; Lao, L.L.; Turnbull, A.D.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Groebner, R.J.; La Haye, R.J.; Mauel, M.

    1995-02-01

    Detailed analysis of recent high beta discharges in the DIII-D tokamak demonstrates that the resistive vacuum vessel can provide stabilization of low n magnetohydrodynamic (MHD) modes. The experimental beta values reaching up to β T = 12.6% are more than 30% larger than the maximum stable beta calculated with no wall stabilization. Plasma rotation is essential for stabilization. When the plasma rotation slows sufficiently, unstable modes with the characteristics of the predicted open-quotes resistive wallclose quotes mode are observed. Through slowing of the plasma rotation between the q = 2 and q = 3 surfaces with the application of a non-axisymmetric field, the authors have determined that the rotation at the outer rational surfaces is most important, and that the critical rotation frequency is of the order of Ω/2π = 1 kHz

  4. Fast Neutral Pressure Gauges in NSTX

    International Nuclear Information System (INIS)

    Raman, R.; Kugel, H.W.; Gernhardt, R.; Provost, T.; Jarboe, T.R.; Soukhanovskii, V.

    2004-01-01

    Successful operation in NSTX of two prototype fast-response micro ionization gauges during plasma operations has motivated us to install five gauges at different toroidal and poloidal locations to measure the edge neutral pressure and its dependence on the type of discharge (L-mode, H-mode, CHI) and the fueling method and location. The edge neutral pressure is also used as an input to the transport analysis codes TRANSP and DEGAS-2. The modified PDX-type Penning gauges are well suited for pressure measurements in the NSTX divertor where the toroidal field is relatively high. Behind the NSTX outer divertor plates where the field is lower, an unshielded fast ion gauge of a new design has been installed. This gauge was developed after laboratory testing of several different designs in a vacuum chamber with applied magnetic fields

  5. Development of NSTX Particle Control Techniques

    International Nuclear Information System (INIS)

    Kugel, H.W.; Maingi, R.; Bell, M.; Gates, D.; Hill, K.; LeBlanc, B.; Mueller, D.; Kaita, R.; Paul, S.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Stratton, B.; Raman, R.

    2004-01-01

    The National Spherical Torus Experiment (NSTX) High Harmonic Fast Wave (HHFW) current-drive discharges will require density control for acceptable efficiency. In NSTX, this involves primarily controlling impurity influxes and recycling. We have compared boronization on hot and cold surfaces, varying helium glow discharge conditioning (HeGDC) durations, helium discharge cleaning, brief daily boronization, and between discharge boronization to reduce and control spontaneous density rises. Access to Ohmic H-modes was enabled by boronization on hot surfaces, however, the duration of the effectiveness of hot and cold boronization was comparable. A 15 minute HeGDC between discharges was needed for reproducible L-H transitions. Helium discharge conditioning yielded slower density rises than 15 minutes of HeGDC. Brief daily boronization followed by a comparable duration of applied HeGDC restored and enhanced good conditions. Additional brief boronizations between discharges did not improve plasma performance (reduced recycling, reduced impurity luminosities, earlier L-H transitions, longer plasma current flattops, higher stored energies) if conditions were already good. Between discharge boronization required increases in the NSTX duty cycle due to the need for additional HeGDC to remove codeposited D

  6. Interaction of a spheromak-like compact toroid with a high beta spherical tokamak plasma

    International Nuclear Information System (INIS)

    Hwang, D.Q.; McLean, H.S.; Baker, K.L.; Evans, R.W.; Horton, R.D.; Terry, S.D.; Howard, S.; Schmidt, G.L.

    2000-01-01

    Recent experiments using accelerated spheromak-like compact toroids (SCTs) to fuel tokamak plasmas have quantified the penetration mechanism in the low beta regime; i.e. external magnetic field pressure dominates plasma thermal pressure. However, fusion reactor designs require high beta plasma and, more importantly, the proper plasma pressure profile. Here, the effect of the plasma pressure profile on SCT penetration, specifically, the effect of diamagnetism, is addressed. It is estimated that magnetic field pressure dominates penetration even up to 50% local beta. The combination of the diamagnetic effect on the toroidal magnetic field and the strong poloidal field at the outer major radius of a spherical tokamak will result in a diamagnetic well in the total magnetic field. Therefore, the spherical tokamak is a good candidate to test the potential trapping of an SCT in a high beta diamagnetic well. The diamagnetic effects of a high beta spherical tokamak discharge (low aspect ratio) are computed. To test the penetration of an SCT into such a diamagnetic well, experiments have been conducted of SCT injection into a vacuum field structure which simulates the diamagnetic field effect of a high beta tokamak. The diamagnetic field gradient length is substantially shorter than that of the toroidal field of the tokamak, and the results show that it can still improve the penetration of the SCT. Finally, analytic results have been used to estimate the effect of plasma pressure on penetration, and the effect of plasma pressure was found to be small in comparison with the magnetic field pressure. The penetration condition for a vacuum field only is reported. To study the diamagnetic effect in a high beta plasma, additional experiments need to be carried out on a high beta spherical tokamak. (author)

  7. Conceptual design for the NSTX Central Instrumentation and Control System

    International Nuclear Information System (INIS)

    Bashore, D.; Oliaro, G.; Roney, P.; Sichta, P.; Tindall, K.

    1997-01-01

    The design and construction phase for the National Spherical Torus Experiment (NSTX) is under way at the Princeton Plasma Physics Laboratory (PPPL). Operation is scheduled to begin on April 30, 1999. This paper describes the conceptual design for the NSTX Central Instrumentation and Control (I and C) System. Major elements of the Central I and C System include the Process Control System, Plasma Control System, Network System, Data Acquisition System, and Synchronization System to support the NSTX experimental device

  8. Model-based Optimization and Feedback Control of the Current Density Profile Evolution in NSTX-U

    Science.gov (United States)

    Ilhan, Zeki Okan

    Nuclear fusion research is a highly challenging, multidisciplinary field seeking contributions from both plasma physics and multiple engineering areas. As an application of plasma control engineering, this dissertation mainly explores methods to control the current density profile evolution within the National Spherical Torus eXperiment-Upgrade (NSTX-U), which is a substantial upgrade based on the NSTX device, which is located in Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ. Active control of the toroidal current density profile is among those plasma control milestones that the NSTX-U program must achieve to realize its next-step operational goals, which are characterized by high-performance, long-pulse, MHD-stable plasma operation with neutral beam heating. Therefore, the aim of this work is to develop model-based, feedforward and feedback controllers that can enable time regulation of the current density profile in NSTX-U by actuating the total plasma current, electron density, and the powers of the individual neutral beam injectors. Motivated by the coupled, nonlinear, multivariable, distributed-parameter plasma dynamics, the first step towards control design is the development of a physics-based, control-oriented model for the current profile evolution in NSTX-U in response to non-inductive current drives and heating systems. Numerical simulations of the proposed control-oriented model show qualitative agreement with the high-fidelity physics code TRANSP. The next step is to utilize the proposed control-oriented model to design an open-loop actuator trajectory optimizer. Given a desired operating state, the optimizer produces the actuator trajectories that can steer the plasma to such state. The objective of the feedforward control design is to provide a more systematic approach to advanced scenario planning in NSTX-U since the development of such scenarios is conventionally carried out experimentally by modifying the tokamak's actuator

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

    International Nuclear Information System (INIS)

    Lyons, B.C.; Scotti, F.; Zweben, S.J.; Gray, T.K.; Hosea, J.; Kaita, R.; Kugel, H.W.; Maqueda, R.J.; McLean, A.G.; Roquemore, A.L.; Soukhanovskii, V.A.; Taylor, G.

    2011-01-01

    Fast cameras already installed on the National Spherical Torus Experiment (NSTX) have be equipped with near-infrared (NIR) filters in order to measure the surface temperature in the lower divertor region. Such a system provides a unique combination of high speed (> 50 kHz) and wide fi eld-of-view (> 50% of the divertor). Benchtop calibrations demonstrated the system's ability to measure thermal emission down to 330 oC. There is also, however, signi cant plasma light background in NSTX. Without improvements in background reduction, the current system is incapable of measuring signals below the background equivalent temperature (600 - 700 oC). Thermal signatures have been detected in cases of extreme divertor heating. It is observed that the divertor can reach temperatures around 800 oC when high harmonic fast wave (HHFW) heating is used. These temperature profiles were fi t using a simple heat diffusion code, providing a measurement of the heat flux to the divertor. Comparisons to other infrared thermography systems on NSTX are made.

  10. ELMs and the H-mode pedestal in NSTX

    International Nuclear Information System (INIS)

    Maingi, R.; Sabbagh, S.A.; Bush, C.E.; Fredrickson, E.D.; Menard, J.E.; Stutman, D.; Tritz, K.; Bell, M.G.; Bell, R.E.; Boedo, J.A.; Gates, D.A.; Johnson, D.W.; Kaita, R.; Kaye, S.M.; Kugel, H.W.; LeBlanc, B.P.; Mueller, D.; Raman, R.; Roquemore, A.L.; Soukhanovskii, V.A.; Stevenson, T.

    2005-01-01

    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 line-average 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 II/III ELMs have also been observed in NSTX, as well as a high-performance regime with small ELMs which we designate Type V. The Type V ELMs are characterized by an intermittent n 1 magnetic pre-cursor oscillation rotating counter to the plasma current; the mode vanishes between Type V ELMs crashes. Without active pumping, the density rises continuously through the Type V phase, albeit at a slower rate than ELM-free discharges

  11. Calculation of the non-inductive current profile in high-performance NSTX plasmas

    Science.gov (United States)

    Gerhardt, S. P.; Fredrickson, E.; Gates, D.; Kaye, S.; Menard, J.; Bell, M. G.; Bell, R. E.; Le Blanc, B. P.; Kugel, H.; Sabbagh, S. A.; Yuh, H.

    2011-03-01

    The constituents of the current profile have been computed for a wide range of high-performance plasmas in NSTX (Ono et al 2000 Nucl. Fusion 40 557); these include cases designed to maximize the non-inductive fraction, pulse length, toroidal-β or stored energy. In the absence of low-frequency MHD activity, good agreement is found between the reconstructed current profile and that predicted by summing the independently calculated inductive, pressure-driven and neutral beam currents, without the need to invoke any anomalous beam ion diffusion. Exceptions occur, for instance, when there are toroidal Alfvén eigenmode avalanches or coupled m/n = 1/1 + 2/1 kink-tearing modes. In these cases, the addition of a spatially and temporally dependent fast-ion diffusivity can reduce the core beam current drive, restoring agreement between the reconstructed profile and the summed constituents, as well as bringing better agreement between the simulated and measured neutron emission rate. An upper bound on the fast-ion diffusivity of ~0.5-1 m2 s-1 is found in 'MHD-free' discharges, based on the neutron emission, the time rate of change in the neutron signal when a neutral beam is stepped and reconstructed on-axis current density.

  12. Calculation of the Non-Inductive Current Profile in High-Performance NSTX Plasmas

    International Nuclear Information System (INIS)

    Gerhardt, S.P.; Fredrickson, E.; Gates, D.; Kaye, S.; Menard, J.; Bell, M.G.; Bell, R.E.; Le Blanc, B.P.; Kugel, H.; Sabbagh, S.A.; Yuh, H.

    2011-01-01

    The constituents of the current profile have been computed for a wide range of high-performance plasmas in NSTX [M. Ono, et al., Nuclear Fusion 40, 557 (2000)]; these include cases designed to maximize the non-inductive fraction, pulse length, toroidal-β, or stored energy. In the absence of low-frequency MHD activity, good agreement is found between the reconstructed current profile and that predicted by summing the independently calculated inductive, pressure-driven, and neutral beam currents, without the need to invoke any anomalous beam ion diffusion. Exceptions occur, for instance, when there are toroidal Alfven eigenmode avalanches or coupled m/n=1/1+2/1 kink-tearing modes. In these cases, the addition of a spatially and temporally dependent fast ion diffusivity can reduce the core beam current drive, restoring agreement between the reconstructed profile and the summed constituents, as well as bringing better agreement between the simulated and measured neutron emission rate. An upper bound on the fast ion diffusivity of ∼0.5-1 m 2 /sec is found in 'MHD-free' discharges, based on the neutron emission, time rate of change of the neutron signal when a neutral beam is stepped, and reconstructed on-axis current density.

  13. Equilibrium of high beta plasma in closed magnetic line system (MBT)

    International Nuclear Information System (INIS)

    Gesso, H.; Shiina, S.; Saito, K.; Nogi, Y.; Osaniai, Y.; Yoshimura, H.; Todoroki, J.; Hamada, S.; Nihon Univ., Tokyo. Atomic Energy Research Inst.)

    1985-01-01

    The beta effects on the plasma equilibrium in Modified Bumpy Torus (MBT) sector, which is an asymmetric closed line system with l = 0 and fairly large l = +- 1 field distortions, are studied. For this purpose, the equilibrium of high beta plasma produced by theta-pinch is compared with that of betaless plasma numerically calculated from the measured magnetic field profiles in device. The equilibrium condition depends weakly on beta value, but the plasma cross-section is vertically elongated as the beta value increases. The m = 1 long wavelength MHD instability is not observed during the observation time of approx. 15 μs. These experimental results are compared with MHD theory based on the new ordering taking the finiteness of l = +- 1 field distortion (deltasub(+-1) > or approx. 1) into account, which suggests significant stabilizing effects due to self formation of magnetic well and also due to the conducting wall. (author)

  14. Non-inductive Solenoid-less Plasma Current Start-up in NSTX Using Transient CHI

    International Nuclear Information System (INIS)

    Raman, R.; Mueller, D.; Jarboe, T.R.; Nelson, B.A.; Bell, M.G.; Ono, M.; Bigelow, T.; Kaita, R.; LeBlanc, B.; Lee, K.C.; Maqueda, R.; Menard, J.; Paul, S.; Roquemore, L.

    2007-01-01

    Coaxial Helicity Injection (CHI) has been successfully used in the National Spherical Torus Experiment (NSTX) for a demonstration of closed flux current generation without the use of the central solenoid. The favorable properties of the Spherical Torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. CHI is a promising candidate for solenoid-free plasma startup in a ST. The method has now produced closed flux current up to 160 kA verifying the high current capability of this method in a large ST built with conventional tokamak components.

  15. Achieving a long-lived high-beta plasma state by energetic beam injection

    Science.gov (United States)

    Guo, H. Y.; Binderbauer, M. W.; Tajima, T.; Milroy, R. D.; Steinhauer, L. C.; Yang, X.; Garate, E. G.; Gota, H.; Korepanov, S.; Necas, A.; Roche, T.; Smirnov, A.; Trask, E.

    2015-04-01

    Developing a stable plasma state with high-beta (ratio of plasma to magnetic pressures) is of critical importance for an economic magnetic fusion reactor. At the forefront of this endeavour is the field-reversed configuration. Here we demonstrate the kinetic stabilizing effect of fast ions on a disruptive magneto-hydrodynamic instability, known as a tilt mode, which poses a central obstacle to further field-reversed configuration development, by energetic beam injection. This technique, combined with the synergistic effect of active plasma boundary control, enables a fully stable ultra-high-beta (approaching 100%) plasma with a long lifetime.

  16. Overview of innovative PMI research on NSTX-U and associated PMI facilities at PPPL

    International Nuclear Information System (INIS)

    Ono, M.; Jaworski, M.; Kaita, R.; Skinner, C. N.; Allain, J. P.; Maingi, R.; Scotti, F.; Soukhanovskii, V. A.

    2013-01-01

    Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ∼15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2

  17. The H-mode Pedestal and Edge Localized Modes in NSTX

    International Nuclear Information System (INIS)

    Maingi, R.; Fredrickson, E.D.; Menard, J.E.; Nishino, N.; Roquemore, A.L.; Sabbagh, S.A.; Tritz, K.

    2004-01-01

    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

  18. Production and study of high-beta plasma confined by a superconducting dipole magnet

    International Nuclear Information System (INIS)

    Garnier, D.T.; Hansen, A.; Mauel, M.E.; Ortiz, E.; Boxer, A.C.; Ellsworth, J.; Karim, I.; Kesner, J.; Mahar, S.; Roach, A.

    2006-01-01

    The Levitated Dipole Experiment (LDX) [J. Kesner et al., in Fusion Energy 1998, 1165 (1999)] is a new research facility that is exploring the confinement and stability of plasma created within the dipole field produced by a strong superconducting magnet. Unlike other configurations in which stability depends on curvature and magnetic shear, magnetohydrodynamic stability of a dipole derives from plasma compressibility. Theoretically, the dipole magnetic geometry can stabilize a centrally peaked plasma pressure that exceeds the local magnetic pressure (β>1), and the absence of magnetic shear allows particle and energy confinement to decouple. In initial experiments, long-pulse, quasi-steady-state microwave discharges lasting more than 10 s have been produced that are consistent with equilibria having peak beta values of 20%. Detailed measurements have been made of discharge evolution, plasma dynamics and instability, and the roles of gas fueling, microwave power deposition profiles, and plasma boundary shape. In these initial experiments, the high-field superconducting floating coil was supported by three thin supports. The plasma is created by multifrequency electron cyclotron resonance heating at 2.45 and 6.4 GHz, and a population of energetic electrons, with mean energies above 50 keV, dominates the plasma pressure. Creation of high-pressure, high-beta plasma is possible only when intense hot electron interchange instabilities are stabilized by sufficiently high background plasma density. A dramatic transition from a low-density, low-beta regime to a more quiescent, high-beta regime is observed when the plasma fueling rate and confinement time become sufficiently large

  19. Characterization of fueling NSTX H-mode plasmas diverted to a liquid lithium divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kugel, H.W.; Abrams, T. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kallman, J. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Mansfield, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); and others

    2013-07-15

    Deuterium fueling experiments were conducted with the NSTX Liquid Lithium Divertor (LLD). Lithium evaporation recoated the LLD surface to approximate flowing liquid Li to sustain D retention. In the first experiment with the diverted outer strike point on the LLD, the difference between the applied D gas input and the plasma D content reached very high values without disrupting the plasma, as would normally occur in the absence of Li pumping, and there was also little change in plasma D content. In the second experiment, constant fueling was applied, as the LLD temperature was varied to change the surface from solid to liquid. The D retention was relatively constant, and about the same as that for solid Li coatings on graphite, or twice that achieved without Li PFC coatings. Contamination of the LLD surface was also possible due to compound formation and erosion and redeposition from carbon PFCs.

  20. NSTX Electrical Power Systems

    International Nuclear Information System (INIS)

    A. Ilic; E. Baker; R. Hatcher; S. Ramakrishnan; et al

    1999-01-01

    The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physic Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. The design of the NSTX electrical power system was tailored to suit the available infrastructure and electrical equipment on site. Components were analyzed to verify their suitability for use in NSTX. The total number of circuits and the location of the NSTX device drove the major changes in the Power system hardware. The NSTX has eleven (11) circuits to be fed as compared to the basic three power loops for TFTR. This required changes in cabling to insure that each cable tray system has the positive and negative leg of cables in the same tray. Also additional power cabling had to be installed to the new location. The hardware had to b e modified to address the need for eleven power loops. Power converters had to be reconnected and controlled in anti-parallel mode for the Ohmic heating and two of the Poloidal Field circuits. The circuit for the Coaxial Helicity Injection (CHI) System had to be carefully developed to meet this special application. Additional Protection devices were designed and installed for the magnet coils and the CHI. The thrust was to making the changes in the most cost-effective manner without compromising technical requirements. This paper describes the changes and addition to the Electrical Power System components for the NSTX magnet systems

  1. Snowflake divertor configuration studies for NSTX-Upgrade

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.

    2011-01-01

    Snowflake divertor experiments in NSTX provide basis for PMI development toward NSTX-Upgrade. Snowflake configuration formation was followed by radiative detachment. Significant reduction of steady-state divertor heat flux observed in snowflake divertor. Impulsive heat loads due to Type I ELMs are partially mitigated in snowflake divertor. Magnetic control of snowflake divertor configuration is being developed. Plasma material interface development is critical for NSTX-U success. Four divertor coils should enable flexibility in boundary shaping and control in NSTX-U. Snowflake divertor experiments in NSTX provide good basis for PMI development in NSTX-Upgrade. FY 2009-2010 snowflake divertor experiments in NSTX: (1) Helped understand control of magnetic properties; (2) Core H-mode confinement unchanged; (3) Core and edge carbon concentration reduced; and (4) Divertor heat flux significantly reduced - (a) Steady-state reduction due to geometry and radiative detachment, (b) Encouraging results for transient heat flux handling, (c) Combined with impurity-seeded radiative divertor. Outlook for snowflake divertor in NSTX-Upgrade: (1) 2D fluid modeling of snowflake divertor properties scaling - (a) Edge and divertor transport, radiation, detachment threshold, (b) Compatibility with cryo-pump and lithium conditioning; (2) Magnetic control development; and (3) PFC development - PFC alignment and PFC material choice.

  2. National Spherical Torus Experiment (NSTX) Center Stack Upgrade

    International Nuclear Information System (INIS)

    Neumeyer, C.; Avasarala, S.; Chrzanowski, J.; Dudek, L.; Fan, H.; Hatcher, H.; Heitzenroeder, P.; Menard, J.; Ono, M.; Ramakrishnan, S.; Titus, P.; Woolley, R.; Zhan, H.

    2009-01-01

    The purpose of the NSTX Center Stack Upgrade project is to expand the NSTX operational space and thereby the physics basis for next-step ST facilities. The plasma aspect ratio (ratio of plasma major to minor radius) of the upgrade is increased to 1.5 from the original value of 1.26, which increases the cross sectional area of the center stack by a factor of ∼ 3 and makes possible higher levels of performance and pulse duration.

  3. Advances in boronization on NSTX-Upgrade

    Directory of Open Access Journals (Sweden)

    C. H Skinner

    2017-08-01

    Full Text Available Boronization has been effective in reducing plasma impurities and enabling access to higher density, higher confinement plasmas in many magnetic fusion devices. The National Spherical Torus eXperiment, NSTX, has recently undergone a major upgrade to NSTX-U in order to develop the physics basis for a ST-based Fusion Nuclear Science Facility (FNSF with capability for double the toroidal field, plasma current, and NBI heating power and increased pulse duration from 1–1.5s to 5–8s. A new deuterated tri-methyl boron conditioning system was implemented together with a novel surface analysis diagnostic. We report on the spatial distribution of the boron deposition versus discharge pressure, gas injection and electrode location. The oxygen concentration of the plasma facing surface was measured by in-vacuo XPS and increased both with plasma exposure and with exposure to trace residual gases. This increase correlated with the rise of oxygen emission from the plasma.

  4. High beta experiments in CHS

    International Nuclear Information System (INIS)

    Okamura, S.; Matsuoka, K.; Nishimura, K.

    1994-09-01

    High beta experiments were performed in the low-aspect-ratio helical device CHS with the volume-averaged equilibrium beta up to 2.1 %. These values (highest for helical systems) are obtained for high density plasmas in low magnetic field heated with two tangential neutral beams. Confinement improvement given by means of turning off gas puffing helped significantly to make high betas. Magnetic fluctuations increased with increasing beta, but finally stopped to increase in the beta range > 1 %. The coherent modes appearing in the magnetic hill region showed strong dependence on the beta values. The dynamic poloidal field control was applied to suppress the outward plasma movement with the plasma pressure. Such an operation gave fixed boundary operations of high beta plasmas in helical systems. (author)

  5. NSTX Protection And Interlock Systems For Coil And Powers Supply Systems

    International Nuclear Information System (INIS)

    Zhao, X.; Ramakrishnan, S.; Lawson, J.; Neumeyer, C.; Marsala, R.; Schneider, H.

    2009-01-01

    NSTX at Princeton Plasma Physics Laboratory (PPPL) requires sophisticated plasma positioning control system for stable plasma operation. TF magnetic coils and PF magnetic coils provide electromagnetic fields to position and shape the plasma vertically and horizontally respectively. NSTX utilizes twenty six coil power supplies to establish and initiate electromagnetic fields through the coil system for plasma control. A power protection and interlock system is utilized to detect power system faults and protect the TF coils and PF coils against excessive electromechanical forces, overheating, and over current. Upon detecting any fault condition the power system is restricted, and it is either prevented from initializing or suppressed to de-energize coil power during pulsing. Power fault status is immediately reported to the computer system. This paper describes the design and operation of NSTX's protection and interlocking system and possible future expansion.

  6. Recent progress of NSTX lithium program and opportunities for magnetic fusion research

    Energy Technology Data Exchange (ETDEWEB)

    Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Kaita, R.; Kugel, H.W. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Ahn, J.-W. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Allain, J.P.; Battaglia, D. [Purdue University, West Lafayette, IN 47907 (United States); Bell, R.E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Canik, J.M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Ding, S. [Academy of Science Institute of Plasma Physics, Hefei (China); Gerhardt, S. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Gray, T.K. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Guttenfelder, W.; Hosea, J.; Jaworski, M.A.; Kallman, J.; Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States); Mansfield, D.K. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); and others

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer In this paper, we review the recent progress on the NSTX lithium research. Black-Right-Pointing-Pointer We summarize positive features of lithium effects on plasma. Black-Right-Pointing-Pointer We also point out unresolved issues and unanswered questions on the lithium research. Black-Right-Pointing-Pointer We describe a possible closed liquid lithium divertor tray concept. Black-Right-Pointing-Pointer We note opportunities and challenges of lithium applications for magnetic fusion. - Abstract: Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up to {approx}160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium, 1300 g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1], we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R and D required

  7. NSTX-U Control System Upgrades

    International Nuclear Information System (INIS)

    Erickson, K.G.; Gates, D.A.; Gerhardt, S.P.; Lawson, J.E.; Mozulay, R.; Sichta, P.; Tchilinguirian, G.J.

    2014-01-01

    The National Spherical Tokamak Experiment (NSTX) is undergoing a wealth of upgrades (NSTX-U). These upgrades, especially including an elongated pulse length, require broad changes to the control system that has served NSTX well. A new fiber serial Front Panel Data Port input and output (I/O) stream will supersede the aging copper parallel version. Driver support for the new I/O and cyber security concerns require updating the operating system from Redhat Enterprise Linux (RHEL) v4 to RedHawk (based on RHEL) v6. While the basic control system continues to use the General Atomics Plasma Control System (GA PCS), the effort to forward port the entire software package to run under 64-bit Linux instead of 32-bit Linux included PCS modifications subsequently shared with GA and other PCS users. Software updates focused on three key areas: (1) code modernization through coding standards (C99/C11), (2) code portability and maintainability through use of the GA PCS code generator, and (3) support of 64-bit platforms. Central to the control system upgrade is the use of a complete real time (RT) Linux platform provided by Concurrent Computer Corporation, consisting of a computer (iHawk), an operating system and drivers (RedHawk), and RT tools (NightStar). Strong vendor support coupled with an extensive RT toolset influenced this decision. The new real-time Linux platform, I/O, and software engineering will foster enhanced capability and performance for NSTX-U plasma control

  8. NSTX-U Control System Upgrades

    Energy Technology Data Exchange (ETDEWEB)

    Erickson, K.G., E-mail: kerickso@pppl.gov; Gates, D.A.; Gerhardt, S.P.; Lawson, J.E.; Mozulay, R.; Sichta, P.; Tchilinguirian, G.J.

    2014-06-15

    The National Spherical Tokamak Experiment (NSTX) is undergoing a wealth of upgrades (NSTX-U). These upgrades, especially including an elongated pulse length, require broad changes to the control system that has served NSTX well. A new fiber serial Front Panel Data Port input and output (I/O) stream will supersede the aging copper parallel version. Driver support for the new I/O and cyber security concerns require updating the operating system from Redhat Enterprise Linux (RHEL) v4 to RedHawk (based on RHEL) v6. While the basic control system continues to use the General Atomics Plasma Control System (GA PCS), the effort to forward port the entire software package to run under 64-bit Linux instead of 32-bit Linux included PCS modifications subsequently shared with GA and other PCS users. Software updates focused on three key areas: (1) code modernization through coding standards (C99/C11), (2) code portability and maintainability through use of the GA PCS code generator, and (3) support of 64-bit platforms. Central to the control system upgrade is the use of a complete real time (RT) Linux platform provided by Concurrent Computer Corporation, consisting of a computer (iHawk), an operating system and drivers (RedHawk), and RT tools (NightStar). Strong vendor support coupled with an extensive RT toolset influenced this decision. The new real-time Linux platform, I/O, and software engineering will foster enhanced capability and performance for NSTX-U plasma control.

  9. Electron Bernstein wave simulations and comparison to preliminary NSTX emission data

    International Nuclear Information System (INIS)

    Preinhaelter, Josef; Urban, Jakub; Pavlo, Pavol; Taylor, Gary; Diem, Steffi; Vahala, Linda; Vahala, George

    2006-01-01

    Simulations indicate that during flattop current discharges the optimal angles for the aiming of the National Spherical Torus Experiment (NSTX) antennae are quite rugged and basically independent of time. The time development of electron Bernstein wave emission (EBWE) at particular frequencies as well as the frequency spectrum of EBWE as would be seen by the recently installed NSTX antennae are computed. The simulation of EBWE at low frequencies (e.g., 16 GHz) agrees well with the recent preliminary EBWE measurements on NSTX. At high frequencies, the sensitivity of EBWE to magnetic field variations is understood by considering the Doppler broadened electron cyclotron harmonics and the cutoffs and resonances in the plasma. Significant EBWE variations are seen if the magnetic field is increased by as little as 2% at the plasma edge. The simulations for the low frequency antenna are compared to preliminary experimental data published separately by Diem et al. [Rev. Sci. Instrum.77 (2006)

  10. The use of MDSplus on NSTX at PPPL

    International Nuclear Information System (INIS)

    Davis, W.; Roney, P.; Carroll, T.; Gibney, T.; Mastrovito, D.

    2002-01-01

    The MDSplus data acquisition system has been used successfully since the 1999 startup of NSTX for control, data acquisition and analysis for diagnostic subsystems. For each plasma 'shot' on NSTX about 75 MBs of data is acquired and loaded into MDSplus hierarchical data structures in 2-3 min. Physicists adapted to the MDSplus software tools with no real difficulty. Some locally developed tools are described. The support from the developers at MIT was timely and insightful. The use of MDSplus has resulted in significant cost savings for NSTX

  11. Rogowski Loop design for NSTX

    International Nuclear Information System (INIS)

    McCormack, B.; Kaita, R.; Kugel, H.; Hatcher, R.

    2000-01-01

    The Rogowski Loop is one of the most basic diagnostics for tokamak operations. On the National Spherical Torus Experiment (NSTX), the plasma current Rogowski Loop had the constraints of the very limited space available on the center stack, 5,000 volt isolation, flexibility requirements as it remained a part of the Center Stack assembly after the first phase of operation, and a +120 C temperature requirement. For the second phase of operation, four Halo Current Rogowski Loops under the Center Stack tiles will be installed having +600 C and limited space requirements. Also as part of the second operational phase, up to ten Rogowski Loops will installed to measure eddy currents in the Passive Plate support structures with +350 C, restricted space, and flexibility requirements. This presentation will provide the details of the material selection, fabrication techniques, testing, and installation results of the Rogowski Loops that were fabricated for the high temperature operational and bakeout requirements, high voltage isolation requirements, and the space and flexibility requirements imposed upon the Rogowski Loops. In the future operational phases of NSTX, additional Rogowski Loops could be anticipated that will measure toroidal plasma currents in the vacuum vessel and in the Passive Plate assemblies

  12. Boronization on NSTX using Deuterated Trimethylboron

    International Nuclear Information System (INIS)

    Blanchard, W.R.; Gernhardt, R.C.; Kugel, H.W.; LaMarche, P.H.

    2002-01-01

    Boronization on the National Spherical Torus Experiment (NSTX) has proved to be quite beneficial with increases in confinement and density, and decreases in impurities observed in the plasma. The boron has been applied to the interior surfaces of NSTX, about every 2 to 3 weeks of plasma operation, by producing a glow discharge in the vacuum vessel using deuterated trimethylboron (TMB) in a 10% mixture with helium. Special NSTX requirements restricted the selection of the candidate boronization method to the use of deuterated boron compounds. Deuterated TMB met these requirements, but is a hazardous gas and special care in the execution of the boronization process is required. This paper describes the existing GDC, Gas Injection, and Torus Vacuum Pumping System hardware used for this process, the glow discharge process, and the automated control system that allows for remote operation to maximize both the safety and efficacy of applying the boron coating. The administrative requirements and the detailed procedure for the setup, operation and shutdown of the process are also described

  13. Effect of progressively increasing lithium conditioning on edge transport and stability in high triangularity NSTX H-modes

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R., E-mail: rmaingi@pppl.gov [Princeton Plasma Physics Laboratory, 100 Stellarator Road, Princeton, NJ 08543 (United States); Canik, J.M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Bell, R.E. [Princeton Plasma Physics Laboratory, 100 Stellarator Road, Princeton, NJ 08543 (United States); Boyle, D.P. [Princeton University, Princeton, NJ (United States); Diallo, A.; Kaita, R.; Kaye, S.M.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, 100 Stellarator Road, Princeton, NJ 08543 (United States); Sabbagh, S.A. [Columbia University, New York, NY (United States); Scotti, F.; Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

    2017-04-15

    A sequence of H-mode discharges with increasing levels of pre-discharge lithium evaporation (‘dose’) was conducted in high triangularity and elongation boundary shape in NSTX. Energy confinement increased, and recycling decreased with increasing lithium dose, similar to a previous lithium dose scan in medium triangularity and elongation plasmas. Data-constrained SOLPS interpretive modeling quantified the edge transport change: the electron particle diffusivity decreased by 10–30x. The electron thermal diffusivity decreased by 4x just inside the top of the pedestal, but increased by up to 5x very near the separatrix. These results provide a baseline expectation for lithium benefits in NSTX-U, which is optimized for a boundary shape similar to the one in this experiment.

  14. Hydrogen retention in lithium on metallic walls from “in vacuo” analysis in LTX and implications for high-Z plasma-facing components in NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Lucia, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Allain, J.P.; Bedoya, F. [Department of Nuclear, Plasma, & Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL (United States); Bell, R.; Boyle, D. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Capece, A. [Department of Physics, The College of New Jersey, Ewing, NJ (United States); Jaworski, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Koel, B.E. [Department of Chemical & Biological Engineering, Princeton University, Princeton, NJ (United States); Majeski, R. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Roszell, J. [Department of Chemical & Biological Engineering, Princeton University, Princeton, NJ (United States); Schmitt, J. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

    2017-04-15

    The application of lithium to plasma-facing components (PFCs) has long been used as a technique for wall conditioning in magnetic confinement devices to improve plasma performance. Determining the characteristics of PFCs at the time of exposure to the plasma, however, is difficult because they can only be analyzed after venting the vacuum vessel and removing them at the end of an operational period. The Materials Analysis and Particle Probe (MAPP) addresses this problem by enabling PFC samples to be exposed to plasmas, and then withdrawn into an analysis chamber without breaking vacuum. The MAPP system was used to introduce samples that matched the metallic PFCs of the Lithium Tokamak Experiment (LTX). Lithium that was subsequently evaporated onto the walls also covered the MAPP samples, which were then subject to LTX discharges. In vacuo extraction and analysis of the samples indicated that lithium oxide formed on the PFCs, but improved plasma performance persisted in LTX. The reduced recycling this suggests is consistent with separate surface science experiments that demonstrated deuterium retention in the presence of lithium oxide films. Since oxygen decreases the thermal stability of the deuterium in the film, the release of deuterium was observed below the lithium deuteride dissociation temperature. This may explain what occurred when lithium was applied to the surface of the NSTX Liquid Lithium Divertor (LLD). The LLD had segments with individual heaters, and the deuterium-alpha emission was clearly lower in the cooler regions. The plan for NSTX-U is to replace the graphite tiles with high-Z PFCs, and apply lithium to their surfaces with lithium evaporation. Experiments with lithium coatings on such PFCs suggest that deuterium could still be retained if lithium compounds form, but limiting their surface temperatures may be necessary.

  15. Overview of the initial NSTX experimental results

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.

    2001-01-01

    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 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ=1:6-2.2 and triangularity δ=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 3.5x10 13 cm 3 , increasing the plasma energy to 59 kJ and the toroidal β, β 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 β T ∼18% at a plasma current of 1.1 MA. (author)

  16. Control System for the NSTX Lithium Pellet Injector

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.; Gernhardt, R.; Gettelfinger, G.; Kugel, H.

    2003-01-01

    The Lithium Pellet Injector (LPI) is being developed for the National Spherical Torus Experiment (NSTX). The LPI will inject ''pellets'' of various composition into the plasma in order to study wall conditioning, edge impurity transport, liquid limiter simulations, and other areas of research. The control system for the NSTX LPI has incorporated widely used advanced technologies, such as LabVIEW and PCI bus I/O boards, to create a low-cost control system which is fully integrated into the NSTX computing environment. This paper will present the hardware and software design of the computer control system for the LPI

  17. The Use of MDSplus on NSTX at PPPL

    International Nuclear Information System (INIS)

    Davis, W.; Roney, P.; Carroll, T.; Gibney, T.; Mastrovito, D.

    2002-01-01

    The MDSplus data acquisition system has been used successfully since the 1999 startup of NSTX [National Spherical Torus Experiment] for control, data acquisition, and analysis for diagnostic subsystems. For each plasma ''shot'' on NSTX about 75 MBs of data is acquired and loaded into MDSplus hierarchical data structures in 2-3 minutes. Physicists adapted to the MDSplus software tools with no real difficulty. Some locally developed tools are described. The support from the developers at MIT [Massachusetts Institute of Technology] was timely and insightful. The use of MDSplus has resulted in a significant cost savings for NSTX

  18. Transport in Auxiliary Heated NSTX Discharges

    International Nuclear Information System (INIS)

    LeBlanc, B.P.; Bell, M.G.; Bell, R.E.; Bitte, M.L.; Bourdelle, C.; Gates, D.A.; Kaye, S.M.; Maingi, R.; Menard, J.E.; Mueller, D.; Ono, M.; Paul, S.F.; Redi, M.H.; Roquemore, A.L.; Rosenberg, A.; Sabbagh, S.A.; Stutman, D.; Synakowski, E.J.; Soukhanovskii, V.A.; Wilson, J.R.

    2003-01-01

    The NSTX spherical torus (ST) provides a unique platform to investigate magnetic confinement in auxiliary-heated plasmas at low aspect ratio. Auxiliary power is routinely coupled to ohmically heated plasmas by deuterium neutral-beam injection (NBI) and by high-harmonic fast waves (HHFW) launch. While theory predicts both techniques to preferentially heat electrons, experiment reveals the electron temperature is greater than the ion temperature during HHFW, but the electron temperature is less than the ion temperature during NBI. In the following we present the experimental data and the results of transport analyses

  19. Far-infrared tangential interferometer/polarimeter design and installation for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Scott, E. R., E-mail: evrscott@ucdavis.edu [Department of Mechanical and Aerospace Engineering, University of California, Davis, California 95616 (United States); Barchfeld, R. [Department of Applied Science, University of California, Davis, California 95616 (United States); Riemenschneider, P.; Domier, C. W.; Sohrabi, M.; Luhmann, N. C. [Department of Electrical and Computer Engineering, University of California, Davis, California 95616 (United States); Muscatello, C. M. [General Atomics, San Diego, California 92121 (United States); Kaita, R.; Ren, Y. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)

    2016-11-15

    The Far-infrared Tangential Interferometer/Polarimeter (FIReTIP) system has been refurbished and is being reinstalled on the National Spherical Torus Experiment—Upgrade (NSTX-U) to supply real-time line-integrated core electron density measurements for use in the NSTX-U plasma control system (PCS) to facilitate real-time density feedback control of the NSTX-U plasma. Inclusion of a visible light heterodyne interferometer in the FIReTIP system allows for real-time vibration compensation due to movement of an internally mounted retroreflector and the FIReTIP front-end optics. Real-time signal correction is achieved through use of a National Instruments CompactRIO field-programmable gate array.

  20. Experimental study of high beta toroidal plasmas

    International Nuclear Information System (INIS)

    Kellman, A.G.

    1983-09-01

    Experiments on the Wisconsin Levitated Toroidal Octupole have produced a wide range of stable high β plasmas with β significantly above single fluid MHD theory predictions. A stable β approx. 8% plasma, twice the fluid limit, is obtained with 5 rho/sub i/ approx. L/sub n/ and tau/sub β/ approx. = 6000 tau/sub Alfven/ = 600 μsec. The enhanced stability is explained with a kinetic treatment that includes the effect of finite ion gyroradius which couples the ballooning mode to an ion drift wave. In a more collisional, large gyroradius (2 rho/sub i/ approx. L/sub n/) regime, a stable β approx. 35% plasma is obtained with a decay time of 1000 Alfven times. Measurement of the equilibrium magnetic field in this regime indicates that the diamagnetic current density is five times smaller than predicted by ideal MHD, probably due to ion gyroviscosity. Particle transport is anomalous and ranges from agreement with the classical diffusion rate at the highest beta, lowest field plasma (B/sub P/ = 200 G), to thirteen times the classical rate in a β=11%, high field plasma (B/sub P/ = 860 G) where the level of enhancement increase with magnetic field. Fluctuations in density, electrostatic potential, and magnetic field have been studied in plasmas with β from 0.1% to 40%

  1. Lithium Wall Conditioning And Surface Dust Detection On NSTX

    International Nuclear Information System (INIS)

    Skinner, C.H.; Allain, J.P.; Bell, M.G.; Friesen, F.Q.L.; Heim, B.; Jaworski, M.A.; Kugel, H.; Maingi, R.; Rais, B.; Taylor, C.N.

    2011-01-01

    Lithium evaporation onto NSTX plasma facing components (PFC) has resulted in improved energy confinement, and reductions in the number and amplitude of edge-localized modes (ELMs) up to the point of complete ELM suppression. The associated PFC surface chemistry has been investigated with a novel plasma material interface probe connected to an in-vacuo surface analysis station. Analysis has demonstrated that binding of D atoms to the polycrystalline graphite material of the PFCs is fundamentally changed by lithium - in particular deuterium atoms become weakly bonded near lithium atoms themselves bound to either oxygen or the carbon from the underlying material. Surface dust inside NSTX has been detected in real-time using a highly sensitive electrostatic dust detector. In a separate experiment, electrostatic removal of dust via three concentric spiral-shaped electrodes covered by a dielectric and driven by a high voltage 3-phase waveform was evaluated for potential application to fusion reactors

  2. Operational Characteristics of Liquid Lithium Divertor in NSTX

    Science.gov (United States)

    Kaita, R.; Kugel, H.; Abrams, T.; Bell, M. G.; Bell, R. E.; Gerhardt, S.; Jaworski, M. A.; Kallman, J.; Leblanc, B.; Mansfield, D.; Mueller, D.; Paul, S.; Roquemore, A. L.; Scotti, F.; Skinner, C. H.; Timberlake, J.; Zakharov, L.; Maingi, R.; Nygren, R.; Raman, R.; Sabbagh, S.; Soukhanovskii, V.

    2010-11-01

    Lithium coatings on plasma-facing components (PFC's) have resulted in improved plasma performance on NSTX in deuterium H-mode plasmas with neutral beam heating.^ Salient results included improved electron confinement and ELM suppression. In CDX-U, the use of lithium-coated PFC's and a large-area liquid lithium limiter resulted in a six-fold increase in global energy confinement time. A Liquid Lithium Divertor (LLD) has been installed in NSTX for the 2010 run campaign. The LLD PFC consists of a thin film of lithium on a temperature-controlled substrate to keep the lithium liquefied between shots, and handle heat loads during plasmas. This capability was demonstrated when the LLD withstood a strike point on its surface during discharges with up to 4 MW of neutral beam heating.

  3. Images of Edge Turbulence in NSTX

    International Nuclear Information System (INIS)

    Zweben, S.J.; Bush, C.E.; Maqueda, R.; Munsat, T.; Stotler, D.; Lowrance, J.; Mastracola, V.; Renda, G.

    2004-01-01

    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

  4. Using LGI experiments to achieve better understanding of pedestal-edge coupling in NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zhehui [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-23

    PowerPoint presentation. Latest advances in granule or dust injection technologies, fast and high-resolution imaging, together with micro-/nano-structured material fabrication, provide new opportunities to examine plasma-material interaction (PMI) in magnetic fusion environment. Some of our previous work in these areas is summarized. The upcoming LGI experiments in NSTX-U will shed new light on granular matter transport in the pedestal-edge region. In addition to particle control, these results can also be used for code validation and achieving better understanding of pedestal-edge coupling in fusion plasmas in both NSTX-U and others.

  5. Analysis of vertical stability limits and vertical displacement event behavior on NSTX-U

    Science.gov (United States)

    Boyer, Mark; Battaglia, Devon; Gerhardt, Stefan; Menard, Jonathan; Mueller, Dennis; Myers, Clayton; Sabbagh, Steven; Smith, David

    2017-10-01

    The National Spherical Torus Experiment Upgrade (NSTX-U) completed its first run campaign in 2016, including commissioning a larger center-stack and three new tangentially aimed neutral beam sources. NSTX-U operates at increased aspect ratio due to the larger center-stack, making vertical stabilization more challenging. Since ST performance is improved at high elongation, improvements to the vertical control system were made, including use of multiple up-down-symmetric flux loop pairs for real-time estimation, and filtering to remove noise. Similar operating limits to those on NSTX (in terms of elongation and internal inductance) were achieved, now at higher aspect ratio. To better understand the observed limits and project to future operating points, a database of vertical displacement events and vertical oscillations observed during the plasma current ramp-up on NSTX/NSTX-U has been generated. Shots were clustered based on the characteristics of the VDEs/oscillations, and the plasma parameter regimes associated with the classes of behavior were studied. Results provide guidance for scenario development during ramp-up to avoid large oscillations at the time of diverting, and provide the means to assess stability of target scenarios for the next campaign. Results will also guide plans for improvements to the vertical control system. Work supported by U.S. D.O.E. Contract No. DE-AC02-09CH11466.

  6. Divertor scenario development for NSTX Upgrade

    Science.gov (United States)

    Soukhanovskii, V. A.; McLean, A. G.; Meier, E. T.; Rognlien, T. D.; Ryutov, D. D.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; Kaita, R.; Kolemen, E.; Leblanc, B. P.; Menard, J. E.; Podesta, M.; Scotti, F.

    2012-10-01

    In the NSTX-U tokamak, initial plans for divertor plasma-facing components (PFCs) include lithium and boron coated graphite, with a staged transition to molybdenum. Steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m^2 in 2 MA, 12 MW NBI-heated discharges of up to 5 s duration, thus challenging PFC thermal limits. Based on the recent NSTX divertor experiments and modeling with edge transport code UEDGE, a favorable basis for divertor power handling in NSTX-U is developed. The snowflake divertor geometry and feedback-controlled divertor impurity seeding applied to the lower and upper divertors are presently envisioned. In the NSTX snowflake experiments with lithium-coated graphite PFCs, the peak divertor heat fluxes from Type I ELMs and between ELMs were significantly reduced due to geometry effects, increased volumetric losses and null-point convective redistribution between strike points. H-mode core confinement was maintained at H98(y,2)<=1 albeit the radiative detachment. Additional CD4 seeding demonstrated potential for a further increase of divertor radiation.

  7. High-beta plasma effects in a low-pressure helicon plasma

    International Nuclear Information System (INIS)

    Corr, C. S.; Boswell, R. W.

    2007-01-01

    In this work, high-beta plasma effects are investigated in a low-pressure helicon plasma source attached to a large volume diffusion chamber. When operating above an input power of 900 W and a magnetic field of 30 G a narrow column of bright blue light (due to Ar II radiation) is observed along the axis of the diffusion chamber. With this blue mode, the plasma density is axially very uniform in the diffusion chamber; however, the radial profiles are not, suggesting that a large diamagnetic current might be induced. The diamagnetic behavior of the plasma has been investigated by measuring the temporal evolution of the magnetic field (B z ) and the plasma kinetic pressure when operating in a pulsed discharge mode. It is found that although the electron pressure can exceed the magnetic field pressure by a factor of 2, a complete expulsion of the magnetic field from the plasma interior is not observed. In fact, under our operating conditions with magnetized ions, the maximum diamagnetism observed is ∼2%. It is observed that the magnetic field displays the strongest change at the plasma centre, which corresponds to the maximum in the plasma kinetic pressure. These results suggest that the magnetic field diffuses into the plasma sufficiently quickly that on a long time scale only a slight perturbation of the magnetic field is ever observed

  8. Ideal-MHD beta limits: scaling laws and comparison with Doublet III high-beta plasmas

    International Nuclear Information System (INIS)

    Bernard, L.C.; Bhadra, D.K.; Helton, F.J.; Lao, L.L.; Todd, T.N.

    1983-06-01

    Doublet III (DIII) recently has achieved a value for #betta#, the ratio of volume averaged plasma to magnetic pressure, of 4.5%. This #betta# value is in the range required for an economically attractive tokamak reactor, and also close to the relevant limit predicted by ideal-MHD theory. It is therefore of great interest to assess the validity of the theory by comparison with experiment and thus to have a basis for the prediction of future reactor performance. A large variety of plasma shapes have been obtained in DIII. These shapes can be divided into two classes: (1) limiter discharges, and (2) diverted discharges, which are of great interest because of their good confinement in the H-mode operation. We derive simple scaling laws from the variation of optimized ideal-MHD beta limits (#betta#/sub c/) with plasma shape parameters. The current profile is optimized for fixed plasma shapes, separately for the high-n (ballooning) and the low-n (kink) modes. Results are presented in the form of suitability normalized curves of #betta# versus poloidal beta, #betta#/sub p/, for both ballooning and kink modes in order to simultaneously compare all the DIII experimental data

  9. Testing Gyrokinetics on C-Mod and NSTX

    International Nuclear Information System (INIS)

    Redi, M.H.; Dorland, W.; Fiore, C.L.; Stutman, D.; Baumgaertel, J.A.; Davis, B.; Kaye, S.M.; McCune, D.C.; Menard, J.; Rewoldt, G.

    2005-01-01

    Quantitative benchmarks of computational physics codes against experiment are essential for the credible application of such codes. Fluctuation measurements can provide necessary critical tests of nonlinear gyrokinetic simulations, but such require extraordinary computational resources. Linear micro-stability calculations with the GS2 [1] gyrokinetic code have been carried out for tokamak and ST experiments which exhibit internal transport barriers (ITB) and good plasma confinement. Qualitative correlation is found for improved confinement before and during ITB plasmas on Alcator C-Mod [2] and NSTX [3], with weaker long wavelength micro-instabilities in the plasma core regions. Mixing length transport models are discussed. The NSTX L-mode is found to be near marginal stability for kinetic ballooning modes. Fully electromagnetic, linear, gyrokinetic calculations of the Alcator C-Mod ITB during off-axis rf heating, following four plasma species and including the complete electron response show ITG/TEM microturbulence is suppressed in the plasma core and in the barrier region before barrier formation, without recourse to the usual requirements of velocity shear or reversed magnetic shear [4-5]. No strongly growing long or short wavelength drift modes are found in the plasma core but strong ITG/TEM and ETG drift wave turbulence is found outside the barrier region. Linear microstability analysis is qualitatively consistent with the experimental transport analysis, showing low transport inside and high transport outside the ITB region before barrier formation, without consideration of ExB shear stabilization

  10. The Use of MDSplus on NSTX at PPPL; TOPICAL

    International Nuclear Information System (INIS)

    W. Davis; P. Roney; T. Carroll; T. Gibney; D. Mastrovito

    2002-01-01

    The MDSplus data acquisition system has been used successfully since the 1999 startup of NSTX[National Spherical Torus Experiment] for control, data acquisition, and analysis for diagnostic subsystems. For each plasma ''shot'' on NSTX about 75 MBs of data is acquired and loaded into MDSplus hierarchical data structures in 2-3 minutes. Physicists adapted to the MDSplus software tools with no real difficulty. Some locally developed tools are described. The support from the developers at MIT[Massachusetts Institute of Technology] was timely and insightful. The use of MDSplus has resulted in a significant cost savings for NSTX

  11. Modeling and control of plasma rotation for NSTX using neoclassical toroidal viscosity and neutral beam injection

    Energy Technology Data Exchange (ETDEWEB)

    Goumiri, I. R. [Princeton Univ., NJ (United States). Mechanical and Aerospace Dept.; Rowley, C. W. [Princeton Univ., NJ (United States). Mechanical and Aerospace Dept.; Sabbagh, S. A. [Columbia Univ., New York, NY (United States). Dept. of Applied Physics and Applied Mathematics; Gates, D. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Gerhardt, S. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Boyer, M. D. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Andre, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kolemen, E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Taira, K. [Florida State Univ, Dept Mech Engn, Tallahassee, FL USA.

    2016-02-19

    A model-based feedback system is presented to control plasma rotation in a magnetically confined toroidal fusion device, to maintain plasma stability for long-pulse operation. This research uses experimental measurements from the National Spherical Torus Experiment (NSTX) and is aimed at controlling plasma rotation using two different types of actuation: momentum from injected neutral beams and neoclassical toroidal viscosity generated by three-dimensional applied magnetic fields. Based on the data-driven model obtained, a feedback controller is designed, and predictive simulations using the TRANSP plasma transport code show that the controller is able to attain desired plasma rotation profiles given practical constraints on the actuators and the available measurements of rotation.

  12. Mode-converted electron Bernstein wave emission research on CDX-U and NSTX

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C; Jones, B.; Munsat, T.; Hosea, J.C; Kaita, R.; Majeski, R.; Spaleta, J.; Wilson, J.R.; Wilgen, J.B.; Bell, G.L.; Rasmussen, D.A.; Ram, A.K.; Bers, A.; Harvey, R.W.; Smirnov, A.P.

    2003-01-01

    Electron Bernstein waves (EBWs) may enable electron temperature profile measurements and local electron heating and current drive in high β overdense (ω pe /ω ce >>1) plasmas. Significant results are presented from the measurement of X-mode radiation, converted from EBWs observed normal to the magnetic field on the mid-plane of overdense plasmas in CDX-U and NSTX. A radially scannable, in-vessel, quad-ridged antenna and Langmuir probe array on CDX-U studied EBW to X-mode conversion. A local limiter optimized the conversion efficiency by modifying the density scale length at the mode conversion layer. The fundamental EBW conversion efficiency increased, by an order of magnitude, to ∼100% when the local limiter and antenna were inserted near the conversion layer. This technique can be extended to large, high temperature devices. Another significant observation was that the EBW emission source was localized near the electron cyclotron resonance. As a result, mode-converted EBW radiometry has measured radial transport in CDX-U. In addition, a threefold increase in conversion efficiency was observed at the L to H transition in NSTX. Measured conversion efficiency agreed well with theoretical predictions. EBW ray tracing and bounce-averaged Fokker-Planck codes are being used to model EBW heating and current drive scenarios for NSTX equilibria with β up to 40%. So far, results show that it is possible to drive localized currents on the high field side of the magnetic axis in NSTX at β ∼ 12% with current drive efficiency which compares favorably with ECCD. (authors)

  13. Deuterium-tritium TFTR plasmas in the high poloidal beta regime

    International Nuclear Information System (INIS)

    Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.

    1995-03-01

    Deuterium-tritium plasmas with enhanced energy confinement and stability have been produced in the high poloidal beta, advanced tokamak regime in TFTR. Confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.46 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasma,s exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Fusion power exceeding 6.7 MW with a fusion power gain Q DT = 0.22 has been produced with reduced alpha particle first orbit loss provided by the increased l i

  14. Recent Physics Results from NSTX

    International Nuclear Information System (INIS)

    Menard, J E; Bell, M G; Bell, R E; Bialek, J M; Boedo, J A; Bush, C E; Crocker, N A; Diem, S; Ferron, J R; Fredrickson, E D; Gates, D A; Hill, K W; Hosea, J C; Kaye, S M; Kessel, C E; Kubota, S; Kugel, H W; LeBlanc, B P; Lee, K C; Levinton, F M; Maingi, R; Mansfield, D K; Majeski, R P; Maqueda, R J; Mazzucato, E; Medley, S S; Mueller, D; Park, H K; Paul, S F; Peebles, W A; Raman, R; Sabbagh, S A; Skinner, C H; Smith, D R; Sontag, A C; Soukhanovskii, V A; Stratton, B C; Stutman, D; Taylor, G; Tritz, K; Wilson, J R; Yuh, H; Zhu, W; Zweben, S J

    2006-01-01

    The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for ITER and future low-aspect-ratio Spherical Torus (ST) devices. Plasma durations up to 1.6s (5 current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while achieving β T and β N values of 16% and 5.7 (%mT/MA), respectively. Newly available Motional Stark Effect data has allowed systematic study and validation of current drive sources and improved the understanding of ''hybrid''-like scenarios. In MHD research, six mid-plane ex-vessel radial field coils have been utilized to infer and correct intrinsic error fields, provide rotation control, and actively stabilize the n=1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence, the low aspect ratio and wide range of achievable β in NSTX provide unique data for confinement scaling studies. A new high-k scattering diagnostic is investigating turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In the area of energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large Toroidal Alfven Eigenmodes (TAEs) similar to the ''sea-of-TAE'' modes predicted for ITER. Three wave coupling processes between energetic particle modes and TAEs have also been observed for the first time. In boundary physics, advanced shape control has been utilized to study the role of magnetic balance in H-mode access and ELM stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode compatible radiative divertor, and Lithium conditioning has demonstrated particle pumping and improved thermal confinement. Finally, non-solenoidal plasma start-up research is particularly important for the ST, and Coaxial Helicity Injection has now produced 160kA plasma

  15. Mechanical Design of the NSTX High-k Scattering Diagnostic

    International Nuclear Information System (INIS)

    Feder, R.; Mazzucato, E.; Munsat, T.; Park, H.; Smith, D.R.; Ellis, R.; Labik, G.; Priniski, C.

    2005-01-01

    The NSTX High-k Scattering Diagnostic measures small-scale density fluctuations by the heterodyne detection of waves scattered from a millimeter wave probe beam at 280 GHz and λ = 1.07 mm. To enable this measurement, major alterations were made to the NSTX vacuum vessel and Neutral Beam armor. Close collaboration between the PPPL physics and engineering staff resulted in a flexible system with steerable launch and detection optics that can position the scattering volume either near the magnetic axis (ρ ∼ .1) or near the edge (ρ ∼ .8). 150 feet of carefully aligned corrugated waveguide was installed for injection of the probe beam and collection of the scattered signal in to the detection electronics

  16. Mechanical Design of the NSTX High-k Scattering Diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Feder, R.; Mazzucato, E.; Munsat, T.; Park, H,; Smith, D. R.; Ellis, R.; Labik, G.; Priniski, C.

    2005-09-26

    The NSTX High-k Scattering Diagnostic measures small-scale density fluctuations by the heterodyne detection of waves scattered from a millimeter wave probe beam at 280 GHz and {lambda}=1.07 mm. To enable this measurement, major alterations were made to the NSTX vacuum vessel and Neutral Beam armor. Close collaboration between the PPPL physics and engineering staff resulted in a flexible system with steerable launch and detection optics that can position the scattering volume either near the magnetic axis ({rho} {approx} .1) or near the edge ({rho} {approx} .8). 150 feet of carefully aligned corrugated waveguide was installed for injection of the probe beam and collection of the scattered signal in to the detection electronics.

  17. Recent Progress on the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.; Bell, M.G.; Bell, R.E.; Bialek, J.; Bigelow, T.; Bitter, M.; Bonoli, P.; Darrow, D.; Efthimion, P.

    2002-01-01

    Recent upgrades to the NSTX facility have led to improved plasma performance. Using 5MW of neutral beam injection, plasmas with toroidal β T (= 2(micro) 0 /B T 2 where B T is the vacuum toroidal field at the plasma geometric center) > 30% have been achieved with normalized β N (= β T aB I /I p ) ∼ 6% · m · T/MA.. The highest β discharge exceeded the calculated no-wall β limit for several wall times. The stored energy has reached 390kJ at higher toroidal field (0.55T) corresponding to β T ∼ 20% and β N = 5.4. Long pulse (∼1s) high β p (∼1.5) discharges have also been obtained at higher β φ (0.5T) with up to 6MW NBI power. The highest energy confinement times, up to 120ms, were observed during H-mode operation which is now routine. Confinement times of ∼1.5 times ITER98pby2 for several τ E are observed during both H-Mode and non-H-Mode discharges. Calculations indicate that many NSTX discharges have very good ion confinement, approaching neoclassical levels. High Harmonic Fast Wave current drive has been demonstrated by comparing discharges with waves launched parallel and anti-parallel to the plasma current

  18. Resistive wall mode stabilization in slowly rotating high beta plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Reimerdes, H [Columbia University, New York, NY 10027 (United States); Garofalo, A M [Columbia University, New York, NY 10027 (United States); Okabayashi, M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Strait, E J [General Atomics, San Diego, CA 92186-5608 (United States); Betti, R [University of Rochester, Rochester, NY 14627 (United States); Chu, M S [General Atomics, San Diego, CA 92186-5608 (United States); Hu, B [University of Rochester, Rochester, NY 14627 (United States); In, Y [FAR-TECH, Inc., San Diego, CA 92121 (United States); Jackson, G L [General Atomics, San Diego, CA 92186-5608 (United States); La Haye, R J [General Atomics, San Diego, CA 92186-5608 (United States); Lanctot, M J [Columbia University, New York, NY 10027 (United States); Liu, Y Q [Chalmers University of Technology, S-412 96 Goeteborg (Sweden); Navratil, G A [Columbia University, New York, NY 10027 (United States); Solomon, W M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Takahashi, H [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Groebner, R J [General Atomics, San Diego, CA 92186-5608 (United States)

    2007-12-15

    DIII-D experiments show that the resistive wall mode (RWM) can remain stable in high {beta} scenarios despite a low net torque from nearly balanced neutral beam injection heating. The minimization of magnetic field asymmetries is essential for operation at the resulting low plasma rotation of less than 20 krad s{sup -1} (measured with charge exchange recombination spectroscopy using C VI emission) corresponding to less than 1% of the Alfven velocity or less than 10% of the ion thermal velocity. In the presence of n = 1 field asymmetries the rotation required for stability is significantly higher and depends on the torque input and momentum confinement, which suggests that a loss of torque-balance can lead to an effective rotation threshold above the linear RWM stability threshold. Without an externally applied field the measured rotation can be too low to neglect the diamagnetic rotation. A comparison of the instability onset in plasmas rotating with and against the direction of the plasma current indicates the importance of the toroidal flow driven by the radial electric field in the stabilization process. Observed rotation thresholds are compared with predictions for the semi-kinetic damping model, which generally underestimates the rotation required for stability. A more detailed modeling of kinetic damping including diamagnetic and precession drift frequencies can lead to stability without plasma rotation. However, even with corrected error fields and fast plasma rotation, plasma generated perturbations, such as edge localized modes, can nonlinearly destabilize the RWM. In these cases feedback control can increase the damping of the magnetic perturbation and is effective in extending the duration of high {beta} discharges.

  19. Progress Towards High Performance, Steady-state Spherical Torus

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.; Bigelow, T.; Bitter, M.; Blanchard, W.; Boedo, J.; Bourdelle, C.; Bush, C.; Choe, W.; Chrzanowski, J.; Darrow, D.S.; Diem, S.J.; Doerner, R.; Efthimion, P.C.; Ferron, J.R.; Fonck, R.J.; Fredrickson, E.D.; Garstka, G.D.; Gates, D.A.; Gray, T.; Grisham, L.R.; Heidbrink, W.; Hill, K.W.; Hoffman, D.; Jarboe, T.R.; Johnson, D.W.; Kaita, R.; Kaye, S.M.; Kessel, C.; Kim, J.H.; Kissick, M.W.; Kubota, S.; Kugel, H.W.; LeBlanc, B.P.; Lee, K.; Lee, S.G.; Lewicki, B.T.; Luckhardt, S.; Maingi, R.; Majeski, R.; Manickam, J.; Maqueda, R.; Mau, T.K.; Mazzucato, E.; Medley, S.S.; Menard, J.; Mueller, D.; Nelson, B.A.; Neumeyer, C.; Nishino, N.; Ostrander, C.N.; Pacella, D.; Paoletti, F.; Park, H.K.; Park, W.; Paul, S.F.; Peng, Y.-K. M.; Phillips, C.K.; Pinsker, R.; Probert, P.H.; Ramakrishnan, S.; Raman, R.; Redi, M.; Roquemore, A.L.; Rosenberg, A.; Ryan, P.M.; Sabbagh, S.A.; Schaffer, M.; Schooff, R.J.; Seraydarian, R.; Skinner, C.H.; Sontag, A.C.; Soukhanovskii, V.; Spaleta, J.; Stevenson, T.; Stutman, D.; Swain, D.W.; Synakowski, E.; Takase, Y.; Tang, X.; Taylor, G.; Timberlake, J.; Tritz, K.L.; Unterberg, E.A.; Von Halle, A.; Wilgen, J.; Williams, M.; Wilson, J.R.; Xu, X.; Zweben, S.J.; Akers, R.; Barry, R.E.; Beiersdorfer, P.; Bialek, J.M.; Blagojevic, B.; Bonoli, P.T.; Carter, M.D.; Davis, W.; Deng, B.; Dudek, L.; Egedal, J.; Ellis, R.; Finkenthal, M.; Foley, J.; Fredd, E.; Glasser, A.; Gibney, T.; Gilmore, M.; Goldston, R.J.; Hatcher, R.E.; Hawryluk, R.J.; Houlberg, W.; Harvey, R.; Jardin, S.C.; Hosea, J.C.; Ji, H.; Kalish, M.; Lowrance, J.; Lao, L.L.; Levinton, F.M.; Luhmann, N.C.; Marsala, R.; Mastravito, D.; Menon, M.M.; Mitarai, O.; Nagata, M.; Oliaro, G.; Parsells, R.; Peebles, T.; Peneflor, B.; Piglowski, D.; Porter, G.D.; Ram, A.K.; Rensink, M.; Rewoldt, G.; Roney, P.; Shaing, K.; Shiraiwa, S.; Sichta, P.; Stotler, D.; Stratton, B.C.; Vero, R.; Wampler, W.R.; Wurden, G.A.

    2003-01-01

    Research on the Spherical Torus (or Spherical Tokamak) 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 Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size 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 U.S., 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 (B), noninductive 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 bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction (∼60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted

  20. Gyrokinetic Calculations of Microturbulence and Transport for NSTX and Alcator-CMOD H-modes

    International Nuclear Information System (INIS)

    Redi, M.H.; Dorland, W.; Bell, R.; Bonoli, P.; Bourdelle, C.; Candy, J.; Ernst, D.; Fiore, C.; Gates, D.; Hammett, G.; Hill, K.; Kaye, S.; LeBlanc, B.; Menard, J.; Mikkelsen, D.; Rewoldt, G.; Rice, J.; Waltz, R.; Wukitch, S.

    2003-01-01

    Recent H-mode experiments on NSTX [National Spherical Torus Experiment] and experiments on Alcator-CMOD, which also exhibit internal transport barriers (ITB), have been examined with gyrokinetic simulations with the GS2 and GYRO codes to identify the underlying key plasma parameters for control of plasma performance and, ultimately, the successful operation of future reactors such as ITER [International Thermonuclear Experimental Reactor]. On NSTX the H-mode is characterized by remarkably good ion confinement and electron temperature profiles highly resilient in time. On CMOD, an ITB with a very steep electron density profile develops following off-axis radio-frequency heating and establishment of H-mode. Both experiments exhibit ion thermal confinement at the neoclassical level. Electron confinement is also good in the CMOD core

  1. Operation of the ultrasoft x-ray system on NSTX (abstract)

    International Nuclear Information System (INIS)

    Stutman, D.; Iovea, M.; Finkenthal, M.; Kaita, R.; Johnson, D.; Roquemore, L.; Roney, P.

    2001-01-01

    The ultrasoft x-ray imaging system on National Spherical Torus Experiment (NSTX) became operational and provided the first data in the filtered diode slow bow tie configuration. Using different band pass filters on each of three arrays allows an approximate spectroscopic estimate of the plasma impurity content, as well as of the electron temperature. Magnetohydrodynamics (MHD) activity from different plasma regions is also observed. The soft x-ray emission profiles are well behaved until an Internal Reconnection Event occurs. Examples of NSTX MHD phenomena seen in the ultrasoft x-ray emission under different operational regimes will be presented. From a technical point of view, we point out that the industrial PC based data acquisition system was not adversely affected by stray magnetic fields due to its close proximity to the NSTX device. Also, the surface barrier diodes withstood baking to 100 o C relatively well

  2. The NSTX Central Instrumentation and Control System

    International Nuclear Information System (INIS)

    G. Oliaro; J. Dong; K. Tindall; P. Sichta

    1999-01-01

    Earlier this year the National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Laboratory achieved ''first plasma''. The Central Instrumentation and Control System was used to support plasma operations. Major elements of the system include the Process Control System, Plasma Control System, Network System, Data Acquisition System, and Synchronization System. This paper will focus on the Process Control System. Topics include the architecture, hardware interface, operator interface, data management, and system performance

  3. Increase in beta limit in tokamak plasmas

    International Nuclear Information System (INIS)

    Kamada, Yutaka

    2003-01-01

    This paper reviews recent studies of tokamak MHD stability towards the achievement of a high beta steady-state, where the profile control of current, pressure, and rotation, and the optimization of the plasma shape play fundamental roles. The key instabilities include the neoclassical tearing mode, the resistive wall mode, the edge localized mode, etc. In order to demonstrate an economically attractive tokamak reactor, it is necessary to increase the beta value simultaneously with a sufficiently high integrated plasma performance. Towards this goal, studies of stability control in self-regulating plasma systems are essential. (author)

  4. Exploration of High Harmonic Fast Wave Heating on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bell, R.E.; Bernabei, S.; Bitter, M.; Bonoli, P.; Gates, D.; Hosea, J.; LeBlanc, B.; Mau, T.K.; Medley, S.; Menard, J.; Mueller, D.; Ono, M.; Phillips, C.K.; Pinsker, R.I.; Raman, R.; Rosenberg, A.; Ryan, P.; Sabbagh, S.; Stutman, D.; Swain, D.; Takase, Y.; Wilgen, J.

    2003-01-01

    High Harmonic Fast Wave (HHFW) heating has been proposed as a particularly attractive means for plasma heating and current drive in the high-beta plasmas that are achievable in spherical torus (ST) devices. The National Spherical Torus Experiment (NSTX) [Ono, M., Kaye, S.M., Neumeyer, S., et al., Proceedings, 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, 1999, (IEEE, Piscataway, NJ (1999), p. 53.)] is such a device. An radio-frequency (rf) heating system has been installed on NSTX to explore the physics of HHFW heating, current drive via rf waves and for use as a tool to demonstrate the attractiveness of the ST concept as a fusion device. To date, experiments have demonstrated many of the theoretical predictions for HHFW. In particular, strong wave absorption on electrons over a wide range of plasma parameters and wave parallel phase velocities, wave acceleration of energetic ions, and indications of current drive for directed wave spectra have been observed. In addition HHFW heating has been used to explore the energy transport properties of NSTX plasmas, to create H-mode (high-confinement mode) discharges with a large fraction of bootstrap current and to control the plasma current profile during the early stages of the discharge

  5. Recent EBW Emission Results on NSTX

    Czech Academy of Sciences Publication Activity Database

    Diem, S.J.; Taylor, G.; Efthimion, P.C.; LeBlanc, B.P.; Caughman, J.B.; Bigelow, T.S.; Wilgen, J.B.; Harvey, R.W.; Preinhaelter, Josef; Urban, Jakub; Sabbagh, S.A.

    2007-01-01

    Roč. 52, č. 16 (2007), s. 63-63 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/49th./. Orlando , Florida, 12.11.2007-16.11.2007] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://meetings.aps.org/Meeting/DPP07/Content/901

  6. Interactions of Deuterium Plasma with Lithiated and Boronized Surfaces in NSTX-U

    Science.gov (United States)

    Krstic, Predrag

    2015-09-01

    The main research goal of the presented research has been to understand the changes in surface composition and chemistry at the nanoscopic temporal and spatial scales for long pulse Plasma Facing Components (PFCs) and link these to the overall machine performance of the National Spherical Torus Experiment Upgrade (NSTX-U). A study is presented of the lithium surface science, with atomic spatial and temporal resolutions. The dynamic surface responds and evolves in a mixed material environments (D, Li, C, B, O, Mo, W) with impingement of plasma particles in the energy range below 100 eV. The results, obtained by quantum-classical molecular dynamics, include microstructure changes, erosion, surface chemistry, deuterium implantation and permeation. Main objectives of the research are i) a comparison of Li and B deposition on carbon, ii) the role of oxygen and other impurities e.g. boron, carbon in the lithium performance, and iii) how this performance will change when lithium is applied to a high-Z refractory metal substrate (Mo, W). In addition to predicting and understanding the phenomenology of the processes, we will show plasma induced erosion of PFCs, including chemical and physical sputtering yields at various temperatures (300-700 K) as well as deuterium uptake/recycling. This work is supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Science, Award Number DE-SC0013752.

  7. Edge Recycling and Heat Fluxes in L- and H-mode NSTX Plasmas

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Maingi, R.; Raman, R.; Kugel, H.; LeBlanc, B.; Roquemore, A.L.; Lasnier, C.J.

    2003-01-01

    Introduction Edge characterization experiments have been conducted in NSTX to provide an initial survey of the edge particle and heat fluxes and their scaling with input power and electron density. The experiments also provided a database of conditions for the analyses of the NSTX global particle sources, core fueling, and divertor operating regimes

  8. Raman Spectroscopy of Carbon Dust Samples from NSTX

    International Nuclear Information System (INIS)

    Raitses, Y.; Skinner, C.H.; Jiang, F.; Duffy, T.S.

    2008-01-01

    The Raman spectrum of dust particles exposed to the NSTX plasma is different from the spectrum of unexposed particles scraped from an unused graphite tile. For the unexposed particles, the high energy G-mode peak (Raman shift ∼1580 cm -1 ) is much stronger than the defect-induced D-mode peak (Raman shift ∼1350 cm -1 ), a pattern that is consistent with Raman spectrum for commercial graphite materials. For dust particles exposed to the plasma, the ratio of G-mode to D-mode peaks is lower and becomes even less than 1. The Raman measurements indicate that the production of carbon dust particles in NSTX involves modifications of the physical and chemical structure of the original graphite material. These modifications are shown to be similar to those measured for carbon deposits from atmospheric pressure helium arc discharge with an ablating anode electrode made from a graphite tile material. We also demonstrate experimentally that heating to 2000-2700 K alone can not explain the observed structural modifications indicating that they must be due to higher temperatures needed for graphite vaporization, which is followed either by condensation or some plasma-induced processes leading to the formation of more disordered forms of carbon material than the original graphite.

  9. Conceptual design of a divertor Thomson scattering diagnostic for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    McLean, A. G., E-mail: mclean@fusion.gat.com; Soukhanovskii, V. A.; Allen, S. L. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States); Carlstrom, T. N. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); LeBlanc, B. P.; Ono, M.; Stratton, B. C. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2014-11-15

    A conceptual design for a divertor Thomson scattering (DTS) diagnostic has been developed for the NSTX-U device to operate in parallel with the existing multipoint Thomson scattering system. Higher projected peak heat flux in NSTX-U will necessitate application of advanced magnetics geometries and divertor detachment. Interpretation and modeling of these divertor scenarios will depend heavily on local measurement of electron temperature, T{sub e}, and density, n{sub e}, which DTS provides in a passive manner. The DTS design for NSTX-U adopts major elements from the successful DIII-D DTS system including 7-channel polychromators measuring T{sub e} to 0.5 eV. If implemented on NSTX-U, the divertor TS system would provide an invaluable diagnostic for the boundary program to characterize the edge plasma.

  10. Exploration of high harmonic fast wave heating on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bell, R.E.; Bernabei, S.; Bitter, M.; Gates, D.; Hosea, J.; Le Blanc, B.; Medley, S.; Menard, J.; Mueller, D.; Ono, M.; Phillips, C.K.; Rosenberg, A.; Bonoli, P.; Mau, T.K.; Pinsker, R.I.; Raman, R.; Ryan, P.; Swain, D.; Wilgen, J.

    2003-01-01

    High harmonic fast wave (HHFW) heating has been proposed as a particularly attractive means for plasma heating and current drive in the high beta plasmas that are achievable in spherical torus (ST) devices. The National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, S. Neumeyer et al., in Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, 1999 (IEEE, Piscataway, NJ, 1999), p. 53] is such a device. An rf heating system has been installed on the NSTX to explore the physics of HHFW heating, current drive via rf waves and for use as a tool to demonstrate the attractiveness of the ST concept as a fusion device. To date, experiments have demonstrated many of the theoretical predictions for HHFW. In particular, strong wave absorption on electrons over a wide range of plasma parameters and wave parallel phase velocities, wave acceleration of energetic ions, and indications of current drive for directed wave spectra have been observed. In addition HHFW heating has been used to explore the energy transport properties of NSTX plasmas, to create H-mode discharges with a large fraction of bootstrap current and to control the plasma current profile during the early stages of the discharge

  11. Final Scientific/Technical Report, USDOE Award DE-FG-02ER54684, Recipient: CompX, Project Title: Fokker-Planck/Ray Tracing for Electron Bernstein and Fast Wave Modeling in Support of NSTX

    International Nuclear Information System (INIS)

    Harvey, R.W.

    2009-01-01

    This DOE grant supported fusion energy research, a potential long-term solution to the world's energy needs. Magnetic fusion, exemplified by confinement of very hot ionized gases, i.e., plasmas, in donut-shaped tokamak vessels is a leading approach for this energy source. Thus far, a mixture of hydrogen isotopes has produced 10's of megawatts of fusion power for seconds in a tokamak reactor at Princeton Plasma Physics Laboratory in New Jersey. The research grant under consideration, ER54684, uses computer models to aid in understanding and projecting efficacy of heating and current drive sources in the National Spherical Torus Experiment, a tokamak variant, at PPPL. The NSTX experiment explores the physics of very tight aspect ratio, almost spherical tokamaks, aiming at producing steady-state fusion plasmas. The current drive is an integral part of the steady-state concept, maintaining the magnetic geometry in the steady-state tokamak. CompX further developed and applied models for radiofrequency (rf) heating and current drive for applications to NSTX. These models build on a 30 year development of rf ray tracing (the all-frequencies GENRAY code) and higher dimensional Fokker-Planck rf-collisional modeling (the 3D collisional-quasilinear CQL3D code) at CompX. Two mainline current-drive rf modes are proposed for injection into NSTX: (1) electron Bernstein wave (EBW), and (2) high harmonic fast wave (HHFW) modes. Both these current drive systems provide a means for the rf to access the especially high density plasma--termed high beta plasma--compared to the strength of the required magnetic fields. The CompX studies entailed detailed modeling of the EBW to calculate the efficiency of the current drive system, and to determine its range of flexibility for driving current at spatial locations in the plasma cross-section. The ray tracing showed penetration into NSTX bulk plasma, relatively efficient current drive, but a limited ability to produce current over the whole

  12. Development of Laser Based Plasma Diagnostics for Fusion Research on NSTX-U

    Science.gov (United States)

    Barchfeld, Robert Adam

    plasma diagnostics. Plasma diagnostics collect data from fusion reactors in a number of different ways. Among these are far infrared (FIR) laser based systems. By probing a fusion plasma with FIR lasers, many properties can be measured, such as density and density fluctuations. This dissertation discusses the theory and design of two laser based diagnostic instruments: 1) the Far Infrared Tangential Interferometer and Polarimeter (FIReTIP) systems, and 2) the High-ktheta Scattering System. Both of these systems have been designed and fabricated at UC Davis for use on the National Spherical Torus Experiment - Upgrade (NSTX-U), located at Princeton Plasma Physics Laboratory (PPPL). These systems will aid PPPL scientists in fusion research. The FIReTIP system uses 119 ?m methanol lasers to pass through the plasma core to measure a chord averaged plasma density through interferometry. It can also measure the toroidal magnetic field strength by the way of polarimetery. The High-ktheta Scattering System uses a 693 GHz formic acid laser to measure electron scale turbulence. Through collective Thomson scattering, as the probe beam passes through the plasma, collective electron motion will scatter power to a receiver with the angle determined by the turbulence wavenumber. This diagnostic will measure ktheta from 7 to 40 cm-1 with a 4-channel receiver array. The High-ktheta Scattering system was designed to facilitate research on electron temperature gradient (ETG) modes, which are believed to be a major contributor to anomalous transport on NSTX-U. The design and testing of these plasma diagnostics are described in detail. There are a broad range of components detailed including: optically pumped gas FIR lasers, overmoded low loss waveguide, launching and receiving optical designs, quasi-optical mixers, electronics, and monitoring and control systems. Additionally, details are provided for laser maintenance, alignment techniques, and the fundamentals of nano-CNC-machining.

  13. Power and Particle Balance Calculations with Impurities in NSTX

    Science.gov (United States)

    Holland, C. G.; Maingi, R.; Owen, L. W.; Kaye, S. M.

    1998-11-01

    We reported the development C. Holland, et. al., Bull. Am. Phys. Soc. 42 (1997) 1927. and application R. Maingi et al., Proc. 3rd International Workshop on Spherical Tori, Sept. 3-5, 1997, St. Petersburg, Russia. of a Graphical User Interface to assess the important terms for edge and divertor plasma calculations for NSTX with the b2.5 edge plasma transport code B. Braams, Contrib. Plasma Phys. 36 (1996) 276.. The goals of those calculations were to estimate the worst case peak heat flux for plasma-facing component design, and the radiation requirements to reduce the peak heat flux. In this study we present the first simulations with intrinsic carbon impurity radiation. We find in general that the intrinsic carbon radiation should be sufficient to provide a wide operation window for the NSTX device. Details of the relative importance of heat flux transport mechanisms as determined with the GUI will be presented.

  14. National Spherical Torus Experiment (NSTX) Torus Design, Fabrication and Assembly

    International Nuclear Information System (INIS)

    Neumeyer, C.; Barnes, G.; Chrzanowski, J.H.; Heitzenroeder, P.

    1999-01-01

    The National Spherical Torus Experiment (NSTX) is a low aspect ratio spherical torus (ST) located at Princeton Plasma Physics Laboratory (PPPL). Fabrication, assembly, and initial power tests were completed in February of 1999. The majority of the design and construction efforts were constructed on the Torus system components. The Torus system includes the centerstack assembly, external Poloidal and Toroidal coil systems, vacuum vessel, torus support structure and plasma facing components (PFC's). NSTX's low aspect ratio required that the centerstack be made with the smallest radius possible. This, and the need to bake NSTXs carbon-carbon composite plasma facing components at 350 degrees C, was major drivers in the design of NSTX. The Centerstack Assembly consists of the inner legs of the Toroidal Field (TF) windings, the Ohmic Heating (OH) solenoid and its associated tension cylinder, three inner Poloidal Field (PF) coils, thermal insulation, diagnostics and an Inconel casing which forms the inner wall of the vacuum vessel boundary. It took approximately nine months to complete the assembly of the Centerstack. The tight radial clearances and the extreme length of the major components added complexity to the assembly of the Centerstack components. The vacuum vessel was constructed of 304-stainless steel and required approximately seven months to complete and deliver to the Test Cell. Several of the issues associated with the construction of the vacuum vessel were control of dimensional stability following welding and controlling the permeability of the welds. A great deal of time and effort was devoted to defining the correct weld process and material selection to meet our design requirements. The PFCs will be baked out at 350 degrees C while the vessel is maintained at 150 degrees C. This required care in designing the supports so they can accommodate the high electromagnetic loads resulting from plasma disruptions and the resulting relative thermal expansions

  15. Electron Bernstein Wave Research on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Taylor, G.; Bers, A.; Bigelow, T.S.; Carter, M.D.; Caughman, J.B.; Decker, J.; Diem, S.; Efthimion, P.C.; Ershov, N.M.; Fredd, E.; Harvey, R.W.; Hosea, J.; Jaeger, F.; Preinhaelter, J.; Ram, A.K.; Rasmussen, D.A.; Smirnov, A.P.; Wilgen, J.B.; Wilson, J.R.

    2005-01-01

    Off-axis electron Bernstein wave current drive (EBWCD) may be critical for sustaining noninductive high-beta National Spherical Torus Experiment (NSTX) plasmas. Numerical modeling results predict that the ∼100 kA of off-axis current needed to stabilize a solenoid-free high-beta NSTX plasma could be generated via Ohkawa current drive with 3 MW of 28 GHz EBW power. In addition, synergy between EBWCD and bootstrap current may result in a 10% enhancement in current-drive efficiency with 4 MW of EBW power. Recent dual-polarization EBW radiometry measurements on NSTX confirm that efficient coupling to EBWs can be readily accomplished by launching elliptically polarized electromagnetic waves oblique to the confining magnetic field, in agreement with numerical modeling. Plans are being developed for implementing a 1 MW, 28 GHz proof-of-principle EBWCD system on NSTX to test the EBW coupling, heating and current-drive physics at high radio-frequency power densities

  16. Design and characterization of a prototype divertor viewing infrared video bolometer for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Eden, G. G. van; Morgan, T. W. [Dutch Institute for Fundamental Energy Research, 5612 AJ Eindhoven (Netherlands); Reinke, M. L.; Gray, T. K.; Lore, J. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Peterson, B. J.; Mukai, K. [National Institute for Fusion Science, Toki 509-5292 (Japan); Delgado-Aparicio, L. F.; Jaworski, M. A. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543 (United States); Sano, R. [National Institutes for Quantum and Radiological Science and Technology, Naka 311-0193 (Japan); Pandya, S. N. [Institute for Plasma Research, Bhat Village, Gandhinagar, 382428 Gujarat (India)

    2016-11-15

    The InfraRed Video Bolometer (IRVB) is a powerful tool to measure radiated power in magnetically confined plasmas due to its ability to obtain 2D images of plasma emission using a technique that is compatible with the fusion nuclear environment. A prototype IRVB has been developed and installed on NSTX-U to view the lower divertor. The IRVB is a pinhole camera which images radiation from the plasma onto a 2.5 μm thick, 9 × 7 cm{sup 2} Pt foil and monitors the resulting spatio-temporal temperature evolution using an IR camera. The power flux incident on the foil is calculated by solving the 2D+time heat diffusion equation, using the foil’s calibrated thermal properties. An optimized, high frame rate IRVB, is quantitatively compared to results from a resistive bolometer on the bench using a modulated 405 nm laser beam with variable power density and square wave modulation from 0.2 Hz to 250 Hz. The design of the NSTX-U system and benchtop characterization are presented where signal-to-noise ratios are assessed using three different IR cameras: FLIR A655sc, FLIR A6751sc, and SBF-161. The sensitivity of the IRVB equipped with the SBF-161 camera is found to be high enough to measure radiation features in the NSTX-U lower divertor as estimated using SOLPS modeling. The optimized IRVB has a frame rate up to 50 Hz, high enough to distinguish radiation during edge-localized-modes (ELMs) from that between ELMs.

  17. Specific radioimmunoassay of human. beta. -endorphin in unextracted plasma

    Energy Technology Data Exchange (ETDEWEB)

    Wiedemann, E. (Univ. of California, Berkeley); Saito, T.; Linfoot, J.A.; Li, C.H.

    1979-09-01

    With an antiserum against human ..beta..-endorphin (..beta..-EP) crossreacting <2% with human ..beta..-lipotropin (..beta..-LPH) by weight we have developed a radioimmunoassay that can detect 1 pg ..beta..-EP in diluted raw plasma. In a.m. fasting plasma of 14 normal subjects ..beta..-EP ranged from <5 to 45 pg/ml. ..beta..-EP was elevated in untreated, but normal in successfully treated Cushing's disease; undetectable in a patient with adrenal adenoma; extremely high in Nelson's syndrome; and elevated in a patient with bronchogenic carcinoma before, but undetectable after tumor resection. In subjects with intact hypothalamic-pituitary-adrenal axis, ..beta..-EP was undectectable after dexamethasone and increased after metyrapone administration and insulin-induced hypoglycemia. ..beta..-EP concentration was considerably lower in serum than in simultaneously collected plasma, but increased in serum left unfrozen for several hours after clot removal. Thus, ..beta..-EP behaves like a hormone responding to the same stimuli as ACTH and ..beta..-LPH and blood appears to contain enzymes both generating and destroying immunoreactive ..beta..-EP.

  18. Electron Bernstein Wave Research on CDX-U and NSTX

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C.; Jones, B.; Hosea, J.C.; Kaita, R.; LeBlanc, B.P.; Majeski, R.; Munsat, T.; Phillips, C.K.; Spaleta, J.; Wilson, J.R.; Rasmussen, D.; Bell, G.; Bigelow, T.S.; Carter, M.D.; Swain, D.W.; Wilgen, J.B.; Ram, A.K.; Bers, A.; Harvey, R.W.; Forest, C.B.

    2001-01-01

    Mode-converted electron Bernstein waves (EBWs) potentially allow the measurement of local electron temperature (Te) and the implementation of local heating and current drive in spherical torus (ST) devices, which are not directly accessible to low harmonic electron cyclotron waves. This paper reports on the measurement of X-mode radiation mode-converted from EBWs observed normal to the magnetic field on the midplane of the Current Drive Experiment-Upgrade (CDX-U) and the National Spherical Torus Experiment (NSTX) spherical torus plasmas. The radiation temperature of the EBW emission was compared to Te measured by Thomson scattering and Langmuir probes. EBW mode-conversion efficiencies of over 20% were measured on both CDX-U and NSTX. Sudden increases of mode-conversion efficiency, of over a factor of three, were observed at high-confinement-mode transitions on NSTX, when the measured edge density profile steepened. The EBW mode-conversion efficiency was found to depend on the density gradient at the mode-conversion layer in the plasma scrape-off, consistent with theoretical predictions. The EBW emission source was determined by a perturbation technique to be localized at the electron cyclotron resonance layer and was successfully used for radial transport studies. Recently, a new in-vessel antenna and Langmuir probe array were installed on CDX-U to better characterize and enhance the EBW mode-conversion process. The probe incorporates a local adjustable limiter to control and maximize the mode-conversion efficiency in front of the antenna by modifying the density profile in the plasma scrape-off where fundamental EBW mode conversion occurs. Initial results show that the mode-conversion efficiency can be increased to ∼100% when the local limiter is inserted near the mode-conversion layer. Plans for future EBW research, including EBW heating and current-drive studies, are discussed

  19. Multi-wavelength imaging of solar plasma. High-beta disruption model of solar flares

    International Nuclear Information System (INIS)

    Shibasaki, Kiyoto

    2007-01-01

    Solar atmosphere is filled with plasma and magnetic field. Activities in the atmosphere are due to plasma instabilities in the magnetic field. To understand the physical mechanisms of activities / instabilities, it is necessary to know the physical conditions of magnetized plasma, such as temperature, density, magnetic field, and their spatial structures and temporal developments. Multi-wavelength imaging is essential for this purpose. Imaging observations of the Sun at microwave, X-ray, EUV and optical ranges are routinely going on. Due to free exchange of original data among solar physics and related field communities, we can easily combine images covering wide range of spectrum. Even under such circumstances, we still do not understand the cause of activities in the solar atmosphere well. The current standard model of solar activities is based on magnetic reconnection: release of stored magnetic energy by reconnection is the cause of solar activities on the Sun such as solar flares. However, recent X-ray, EUV and microwave observations with high spatial and temporal resolution show that dense plasma is involved in activities from the beginning. Based on these observations, I propose a high-beta model of solar activities, which is very similar to high-beta disruptions in magnetically confined fusion experiments. (author)

  20. Analysis Efforts Supporting NSTX Upgrades

    International Nuclear Information System (INIS)

    Zhang, H.; Titus, P.; Rogoff, P.; Zolfaghari, A.; Mangra, D.; Smith, M.

    2010-01-01

    The National Spherical Torus Experiment (NSTX) is a low aspect ratio, spherical torus (ST) configuration device which is located at Princeton Plasma Physics Laboratory (PPPL) This device is presently being updated to enhance its physics by doubling the TF field to 1 Tesla and increasing the plasma current to 2 Mega-amperes. The upgrades include a replacement of the centerstack and addition of a second neutral beam. The upgrade analyses have two missions. The first is to support design of new components, principally the centerstack, the second is to qualify existing NSTX components for higher loads, which will increase by a factor of four. Cost efficiency was a design goal for new equipment qualification, and reanalysis of the existing components. Showing that older components can sustain the increased loads has been a challenging effort in which designs had to be developed that would limit loading on weaker components, and would minimize the extent of modifications needed. Two areas representing this effort have been chosen to describe in more details: analysis of the current distribution in the new TF inner legs, and, second, analysis of the out-of-plane support of the existing TF outer legs.

  1. Status and Plans for the National Spherical Torus Experimental Research Facility

    International Nuclear Information System (INIS)

    Ono, M.; Bell, M.G.; Bell, R.E.; Bialek, J.M.; Bigelow, T.; Bitter, M.

    2005-01-01

    An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions

  2. Modification of the magnetic field structure of high-beta plasmas with a perturbation field in the Large Helical Device

    International Nuclear Information System (INIS)

    Sakakibara, S; Suzuki, Y; Narushima, Y; Watanabe, K Y; Ohdachi, S; Ida, K; Yoshinuma, M; Narihara, K; Yamada, I; Tanaka, K; Tokuzawa, T; Yamada, H; Takemura, Y

    2013-01-01

    The effect of resonant magnetic perturbation (RMP) on MHD characteristics is investigated in high-beta plasmas of the Large Helical Device. The ramp-up and static m/n = 1/1 RMP field are applied in medium- (∼2%) and high- (∼4%) beta plasmas in order to find beta dependences of mode penetration, MHD activities and confinement. The results show that the threshold of mode penetration linearly increases with the beta value and/or plasma collisionality. The threshold of mode penetration in the RMP ramp-up experiments is roughly consistent with the static RMP case. The beta value gradually decreases with the RMP field strength before mode penetration, which is caused by a reduction in the pressure inside the ι/2π = 1 resonance. The width of the magnetic island after the penetration becomes larger than the given RMP field, and it is further enhanced by the increment of the beta value. (paper)

  3. Evaporated Lithium Surface Coatings in NSTX

    International Nuclear Information System (INIS)

    Kugel, H.W.; Mansfield, D.; Maingi, R.; Bel, M.G.; Bell, R.E.; Allain, J.P.; Gates, D.; Gerhardt, S.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.; Majeski, R.; Menard, J.; Mueller, D.; Ono, M.

    2009-01-01

    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

  4. Evaporated Lithium Surface Coatings in NSTX

    International Nuclear Information System (INIS)

    Kugel, H.W.; Mansfield, D.; Maingi, Rajesh; Bell, M.G.; Bell, R.E.; Allain, J.P.; Gates, D.; Gerhardt, S.P.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Majeski, R.; Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Raman, R.; Roquemore, A.L.; Ross, P.W.; Sabbagh, S.A.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.; Stevenson, T.; Timberlake, J.; Wampler, W.R.; Wilgen, John B.; Zakharov, L.E.

    2009-01-01

    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.

  5. Lithium Pellet Injector Development for NSTX

    International Nuclear Information System (INIS)

    Gettelfinger, G.; Dong, J.; Gernhardt, R.; Kugel, H.; Sichta, P.; Timberlake, J.

    2003-01-01

    A pellet injector suitable for the injection of lithium and other low-Z pellets of varying mass into plasmas at precise velocities from 5 to 500 m/s is being developed for use on NSTX (National Spherical Torus Experiment). The ability to inject low-Z impurities will significantly expand NSTX experimental capability for a broad range of diagnostic and operational applications. The architecture employs a pellet-carrying cartridge propelled through a guide tube by deuterium gas. Abrupt deceleration of the cartridge at the end of the guide tube results in the pellet continuing along its intended path, thereby giving controlled reproducible velocities for a variety of pellets materials and a reduced gas load to the torus. The planned injector assembly has four hundred guide tubes contained in a rotating magazine with eight tubes provided for injection into plasmas. A PC-based control system is being developed as well and will be described elsewhere in these Proceedings. The development path and mechanical performance of the injector will be described

  6. Status of the Experimental Physics and Industrial Control System at NSTX

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.

    2002-01-01

    The NSTX achieved first plasma in 1999. The Experimental Physics and Industrial Control System (EPICS) is used to provide data-integration services for monitoring and control of all NSTX engineering subsystems. EPICS is a set of software initially developed at U.S. DOE laboratories. It is currently used and maintained through a global collaboration of hundreds of scientists and engineers. This paper will relate some of our experiences using and supporting the EPICS software. Topics include reliability and maintainability, lessons learned, recently added engineering subsystems, new EPICS software tools, and a review of our first EPICS software upgrade. Steps to modernize the technical infrastructure of EPICS to ensure effective support for NSTX will also be described

  7. Accounting of the Power Balance for Neutral-beam heated H-Mode Plasmas in NSTX

    International Nuclear Information System (INIS)

    Paul, S.F.; Maingi, R.; Soukhanovskii, V.; Kaye, S.M.; Kugel, H.

    2004-01-01

    A survey of the dependence of power balance on input power, shape, and plasma current was conducted for neutral-beam-heated plasmas in the National Spherical Torus Experiment (NSTX). Measurements of heat to the divertor strike plates and divertor and core radiation were taken over a wide range of plasma conditions. The different conditions were obtained by inducing a L-mode to H-mode transition, changing the divertor configuration [lower single null (LSN) vs. double-null (DND)] and conducting a NBI power scan in H-mode. 60-70% of the net input power is accounted for in the LSN discharges with 20% of power lost as fast ions, 30-45% incident on the divertor plates, up to 10% radiated in the core, and about 12% radiated in the divertor. In contrast, the power accountability in DND is 85-90%. A comparison of DND and LSN data show that the remaining power in the LSN is likely to be directed to the upper divertor

  8. MHD analysis of LHD high beta plasma

    International Nuclear Information System (INIS)

    Ichiguchi, K.; Nakajima, N.; Carreras, B.A.

    2003-01-01

    The nonlinear behavior of the interchange mode in the LHD plasma depends on the overlap of the vortices with different helicity. If the vortices are separated in the radial direction, each mode saturates mildly with generating the local flat regions in the pressure profile. In the case of the significant overlap of the vortices, the convection is enhanced and the sudden global reduction of the pressure occurs. Succession of the saturated pressure profile in the increase of beta can suppress the overlap. Self-organization of the pressure profile to suppress the overlap of the vortices can be the stabilizing mechanism in the LHD plasma. (orig.)

  9. High beta tokamaks

    International Nuclear Information System (INIS)

    Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.

    1978-01-01

    MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby

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

    International Nuclear Information System (INIS)

    Mueller, D.; Gates, D.A.; Menard, J.E.; Ferron, J.R.; Sabbagh, S.A.

    2004-01-01

    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 κ control in a variety of experiments

  11. Electron Bernstein Wave Research on NSTX and PEGASUS

    International Nuclear Information System (INIS)

    Diem, S. J.; LeBlanc, B. P.; Taylor, G.; Caughman, J. B.; Bigelow, T.; Wilgen, J. B.; Garstka, G. D.; Harvey, R. W.; Preinhaelter, J.; Urban, J.; Sabbagh, S. A.

    2007-01-01

    Spherical tokamaks (STs) routinely operate in the overdense regime (ω pe >>ω ce ), prohibiting the use of standard ECCD and ECRH. However, the electrostatic electron Bernstein wave (EBW) can propagate in the overdense regime and is strongly absorbed and emitted at the electron cyclotron resonances. As such, EBWs offer the potential for local electron temperature measurements and local electron heating and current drive. A critical challenge for these applications is to establish efficient coupling between the EBWs and electromagnetic waves outside the cutoff layer. Two STs in the U.S., the National Spherical Tokamak Experiment (NSTX, at Princeton Plasma Physics Laboratory) and PEGASUS Toroidal Experiment (University of Wisconsin-Madison) are focused on studying EBWs for heating and current drive. On NSTX, two remotely steered, quad-ridged antennas have been installed to measure 8-40 GHz (fundamental, second and third harmonics) thermal EBW emission (EBE) via the oblique B-X-O mode conversion process. This diagnostic has been successfully used to map the EBW mode conversion efficiency as a function of poloidal and toroidal angles on NSTX. Experimentally measured mode conversion efficiencies of 70±20% have been measured for 15.5 GHz (fundamental) emission in L-mode discharges, in agreement with a numerical EBE simulation. However, much lower mode conversion efficiencies of 25±10% have been measured for 25 GHz (second harmonic) emission in L-mode plasmas. Numerical modeling of EBW propagation and damping on the very-low aspect ratio PEGASUS Toroidal Experiment has been performed using the GENRAY ray-tracing code and CQL3D Fokker-Planck code in support of planned EBW heating and current drive (EBWCD) experiments. Calculations were performed for 2.45 GHz waves launched with a 10 cm poloidal extent for a variety of plasma equilibrium configurations. Poloidal launch scans show that driven current is maximum when the poloidal launch angle is between 10 and 25 degrees

  12. Reduced and high molecular weight barley beta-glucans decrease plasma total and non-HDL-cholesterol in hypercholesterolemic Syrian golden hamsters.

    Science.gov (United States)

    Wilson, Thomas A; Nicolosi, Robert J; Delaney, Bryan; Chadwell, Kim; Moolchandani, Vikas; Kotyla, Timothy; Ponduru, Sridevi; Zheng, Guo-Hua; Hess, Richard; Knutson, Nathan; Curry, Leslie; Kolberg, Lore; Goulson, Melanie; Ostergren, Karen

    2004-10-01

    Consumption of concentrated barley beta-glucan lowers plasma cholesterol because of its soluble dietary fiber nature. The role of molecular weight (MW) in lowering serum cholesterol is not well established. Prior studies showed that enzymatic degradation of beta-glucan eliminates the cholesterol-lowering activity; however, these studies did not evaluate the MW of the beta-glucan. The current study was conducted to evaluate whether barley beta-glucan concentrates, partially hydrolyzed to reduce MW, possess cholesterol-lowering and antiatherogenic activities. The reduced MW fraction was compared with a high MW beta-glucan concentrate from the same barley flour. Concentrated beta-glucan preparations were evaluated in Syrian Golden F(1)B hamsters fed a hypercholesterolemic diet (HCD) with cholesterol, hydrogenated coconut oil, and cellulose. After 2 wk, hamsters were fed HCD or diets that contained high or reduced MW beta-glucan at a concentration of 8 g/100 g at the expense of cellulose. Decreases in plasma total cholesterol (TC) and non-HDL-cholesterol (non-HDL-C) concentrations occurred in the hamsters fed reduced MW and high MW beta-glucan diets. Plasma HDL-C concentrations did not differ. HCD-fed hamsters had higher plasma triglyceride concentrations. Liver TC, free cholesterol, and cholesterol ester concentrations did not differ. Aortic cholesterol ester concentrations were lower in the reduced MW beta-glucan-fed hamsters. Consumption of either high or reduced MW beta-glucan increased concentrations of fecal total neutral sterols and coprostanol, a cholesterol derivative. Fecal excretion of cholesterol was greater than in HCD-fed hamsters only in those fed the reduced MW beta-glucan. Study results demonstrate that the cholesterol-lowering activity of barley beta-glucan may occur at both lower and higher MW.

  13. Beta limit of resistive plasma in Torsatron / Heliotron

    International Nuclear Information System (INIS)

    Itoh, K.; Ichiguchi, K.; Itoh, S.I.

    1992-02-01

    Stability against the interchange mode in the Torsatron / Heliotron device is investigated, taking into account the effects of the resistivity, current diffusivity, ion viscosity and thermal diffusivity. Critical beta for the low-mode-number stability is found at the finite beta value. For the range of plasma parameters of the present experiments, the resistive mode and current diffusive mode set comparable critical beta values, which are consistent with experimental observations. In future high temperature plasmas, the current diffusive mode determines the stability limit for the global mode. (author)

  14. Parametric Decay during HHFW on NSTX

    International Nuclear Information System (INIS)

    Wilson, J.R.; Bernabei, S.; Biewer, T.; Diem, S.; Hosea, J.; LeBlanc, B.; Phillips, C.K.; Ryan, P.; Swain, D.W.

    2005-01-01

    High Harmonic Fast Wave (HHFW) heating experiments on NSTX have been observed to be accompanied by significant edge ion heating (T i >> T e ). This heating is found to be anisotropic with T perp > T par . Simultaneously, coherent oscillations have been detected with an edge Langmuir probe. The oscillations are consistent with parametric decay of the incident fast wave (ω > 13ω ci ) into ion Bernstein waves and an unobserved ion-cyclotron quasi-mode. The observation of anisotropic heating is consistent with Bernstein wave damping, and the Bernstein waves should completely damp in the plasma periphery as they propagate toward a cyclotron harmonic resonance. The number of daughter waves is found to increase with rf power, and to increase as the incident wave's toroidal wavelength increases. The frequencies of the daughter wave are separated by the edge ion cyclotron frequency. Theoretical calculations of the threshold for this decay in uniform plasma indicate an extremely small value of incident power should be required to drive the instability. While such decays are commonly observed at lower harmonics in conventional ICRF heating scenarios, they usually do not involve the loss of significant wave power from the pump wave. On NSTX an estimate of the power loss can be found by calculating the minimum power required to support the edge ion heating (presumed to come from the decay Bernstein wave). This calculation indicates at least 20-30% of the incident rf power ends up as decay waves

  15. Proceedings of the US-Japan workshop and the satellite meeting of ITC-9 on physics of high beta plasma confinement in innovative fusion system

    Energy Technology Data Exchange (ETDEWEB)

    Goto, Seiichi; Yoshimura, Satoru [eds.

    1999-04-01

    The US-Japan Workshop on Physics of High Beta Plasma Confinement in Innovative Fusion System was held jointly with the Satellite Meeting of ITC-9 at National Institute for Fusion Science (NIFS), Toki-city during December 14-15, 1998. This proceedings book includes the papers of the talks given at the workshop. These include: Theoretical analysis on the stability of field reversed configuration (FRC) plasmas; Theory and Modeling of high {beta} plasmas; Recent progressive experiments in high {beta} systems; Formation of high {beta} plasmas using merging phenomenon; Theory and Modeling of a FRC Fusion Reactor. The 15 papers are indexed individually. (J.P.N.)

  16. Modeling of Low Frequency MHD Induced Beam Ion Transport In NSTX

    International Nuclear Information System (INIS)

    Gorelenkov, N.N.; Medley, S.S.

    2004-01-01

    Beam ion transport in the presence of low frequency MHD activity in National Spherical Tokamak Experiment (NSTX) plasma is modeled numerically and analyzed theoretically in order to understand basic underlying physical mechanisms responsible for the observed fast ion redistribution and losses. Numerical modeling of the beam ions flux into the NPA in NSTX shows that after the onset of low frequency MHD activity high energy part of beam ion distribution, E b > 40keV, is redistributed radially due to stochastic diffusion. Such diffusion is caused by high order harmonics of the transit frequency resonance overlap in the phase space. Large drift orbit radial width induces such high order resonances. Characteristic confinement time is deduced from the measured NPA energy spectrum and is typically ∼ 4msec. Considered MHD activity may induce losses on the order of 10% at the internal magnetic field perturbation (delta)B/B = Ο (10 -3 ), which is comparable to the prompt orbit losses

  17. Occurrence of high-beta superthermal plasma events in the close environment of Jupiter's bow shock as observed by Ulysses

    International Nuclear Information System (INIS)

    Marhavilas, P. K.; Sarris, E. T.; Anagnostopoulos, G. C.

    2011-01-01

    The ratio of the plasma pressure to the magnetic field pressure (or of their energy densities) which is known as the plasma parameter 'beta'(β) has important implications to the propagation of energetic particles and the interaction of the solar wind with planetary magnetospheres. Although in the scientific literature the contribution of the superthermal particles to the plasma pressure is generally assumed negligible, we deduced, by analyzing energetic particles and magnetic field measurements recorded by the Ulysses spacecraft, that in a series of events, the energy density contained in the superthermal tail of the particle distribution is comparable to or even higher than the energy density of the magnetic field, creating conditions of high-beta plasma. More explicitly, in this paper we analyze Ulysses/HI-SCALE measurements of the energy density ratio (parameter β ep ) of the energetic ions'(20 keV to ∼5 MeV) to the magnetic field's in order to find occurrences of high-beta (β ep >1) superthermal plasma conditions in the environment of the Jovian magnetosphere, which is an interesting plasma laboratory and an important source of emissions in our solar system. In particular, we examine high-beta ion events close to Jupiter's bow shock, which are produced by two processes: (a) bow shock ion acceleration and (b) ion leakage from the magnetosphere.

  18. Comparison of neutral density profiles measured using Dα and C5+ in NSTX-U

    Science.gov (United States)

    Bell, R. E.; Scotti, F.; Diallo, A.; Leblanc, B. P.; Podesta, M.; Sabbagh, S. A.

    2017-10-01

    Edge neutral density profiles determined from two different measurements are compared on NSTX-U plasmas. Neutral density measurements were not typical on NSTX plasmas. An array of fibers dedicated to the measurement of passive emission of C5+, used to subtract background emission for charge exchange recombination spectroscopy (CHERS), can be used to infer deuterium neutral density near the plasma edge. The line emission from C5+ is dominated by charge exchange with neutral deuterium near the plasma edge. An edge neutral density diagnostic consisting of a camera with a Dα filter was installed on NSTX-U. The line-integrated measurements from both diagnostics are inverted to obtain local emissivity profiles. Neutral density is then inferred using atomics rates from ADAS and profile measurements from Thomson scattering and CHERS. Comparing neutral density profiles from the two diagnostic measurements helps determine the utility of using the more routinely available C5+ measurements for neutral density profiles. Initial comparisons show good agreement between the two measurements inside the separatrix. Supported by US DoE Contracts DE-AC02-09CH11466 and DE-AC52-07NA27344.

  19. Overview of physics results from NSTX

    Czech Academy of Sciences Publication Activity Database

    Raman, R.; Ahn, J-W.; Allain, J.P.; Andre, R.; Bastasz, R.; Battaglia, D.; Beiersdorfer, P.; Bell, M.; Bell, R.; Belova, E.; Berkery, J.; Betti, R.; Bialek, J.; Bigelow, T.; Bitter, M.; Boedo, J.; Bonoli, P.; Boozer, A.; Bortolon, A.; Brennan, D.; Breslau, J.; Buttery, R.; Canik, J.; Caravelli, G.; Chang, C.; Crocker, N.A.; Darrow, D.; Davis, W.; Delgado-Aparicio, L.; Diallo, A.; Ding, S.; D’Ippolito, D.; Domier, C.; Dorland, W.; Ethier, S.; Evans, T.; Ferron, J.; Finkenthal, M.; Foley, J.; Fonck, R.; Frazin, R.; Fredrickson, E.; Fu, G.; Gates, D.; Gerhardt, S.; Glasser, A.; Gorelenkov, N.; Gray, T.; Guo, Y.; Guttenfelder, W.; Hahm, T.; Harvey, R.; Hassanein, A.; Heidbrink, W.; Hill, K.; Hirooka, Y.; Hooper, E.B.; Hosea, J.; Hu, B.; Humphreys, D.; Indireshkumar, K.; Jaeger, F.; Jarboe, T.; Jardin, S.; Jaworski, M.; Kaita, R.; Kallman, J.; Katsuro-Hopkins, O.; Kaye, S.; Kessel, C.; Kim, J.; Kolemen, E.; Krasheninnikov, S.; Kubota, S.; Kugel, H.; La Haye, R.; Lao, L.; LeBlanc, B.; Lee, W.; Lee, K.; Leuer, J.; Levinton, F.; Liang, Y.; Liu, D.; Luhmann Jr, N.; Maingi, R.; Majeski, R.; Manickam, J.; Mansfield, D.; Maqueda, R.; Mazzucato, E.; McLean, A.; McCune, D.; McGeehan, B.; McKee, G.; Medley, S.; Menard, J.; Menon, M.; Meyer, H.; Mikkelsen, D.; Miloshevsky, G.; Mueller, D.; Munsat, T.; Myra, J.; Nelson, B.; Nishino, N.; Nygren, R.; Ono, M.; Osborne, T.; Park, H.; Park, J.; Paul, S.; Peebles, W.; Penaflor, B.; Phillips, C.; Pigarov, A.; Podesta, M.; Preinhaelter, Josef; Ren, Y.; Reimerdes, H.; Ross, P.; Rowley, C.; Ruskov, E.; Russell, D.; Ruzic, D.; Ryan, P.; Sabbagh, S.A.; Schaffer, M.; Schuster, E.; Scotti, F.; Shaing, K.; Shevchenko, V.; Shinohara, K.; Sizyuk, V.; Skinner, C.H.; Smirnov, A.; Smith, D.; Snyder, P.; Solomon, W.; Sontag, A.; Soukhanovskii, V.; Stoltzfus-Dueck, T.; Stotler, D.; Stratton, B.; Stutman, D.; Takahashi, H.; Takase, Y.; Tamura, N.; Tang, X.; Taylor, C.N.; Taylor, G.; Taylor, C.; Tritz, K.; Tsarouhas, D.; Umansky, M.; Urban, Jakub; Walker, M.; Wampler, W.; Wang, W.; Whaley, J.; White, R.; Wilgen, J.; Wilson, R.; Wong, K.L.; Wright, J.; Xia, Z.; Youchison, D.; Yu, H.; Yuh, H.; Zakharov, L.; Zemlyanov, D.; Zimmer, G.; Zweben, S.J.

    2011-01-01

    Roč. 51, č. 9 (2011), 094011-094011 ISSN 0029-5515. [Fusion Energy Conference (FEC 2010)/23rd./. Daejon, 11.10.2010-16.10.2010] R&D Projects: GA ČR GA202/08/0419; GA MŠk 7G09042 Institutional research plan: CEZ:AV0Z20430508 Keywords : NSTX * Spherical tokamaks * Overdense plasma * Conversion * Emission * Tokamaks * Elektron Bernstein waves Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.090, year: 2011 http://iopscience.iop.org/0029-5515/51/9/094011/pdf/0029-5515_51_9_094011.pdf

  20. Fast Neutral Pressure Measurements in NSTX

    International Nuclear Information System (INIS)

    R. Raman; H.W. Kugel; T. Provost; R. Gernhardt; T.R. Jarboe; M.G. Bell

    2002-01-01

    Several fast neutral pressure gauges have been installed on NSTX [National Spherical Torus Experiment] to measure the vessel and divertor pressure during inductive and coaxial helicity injected (CHI) plasma operations. Modified, PDX [Poloidal Divertor Experiment]-type Penning gauges have been installed on the upper and lower divertors. Neutral pressure measurements during plasma operations from these and from two shielded fast Micro ion gauges at different toroidal locations on the vessel mid-plane are described. A new unshielded ion gauge, referred to as the In-vessel Neutral Pressure (INP) gauge is under development

  1. Electron Bernstein Wave Coupling and Emission Measurements on NSTX

    Czech Academy of Sciences Publication Activity Database

    Taylor, G.; Diem, S.J.; Caughman, J.; Efthimion, P.; Harvey, R.W.; LeBlanc, B.P.; Philips, C.K.; Preinhaelter, Josef; Urban, Jakub

    2006-01-01

    Roč. 51, č. 7 (2006), s. 177 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/48th./. Philadelphia, Pennsylvania , 30.10.2006-3.11.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * MAST * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://www.aps.org/meet/DPP06/baps/all_DPP06.pdf

  2. Effect of Various EFIT NSTX Equilibria on EBW Simulations

    Czech Academy of Sciences Publication Activity Database

    Urban, Jakub; Preinhaelter, Josef; Sabbagh, S.; Pavlo, Pavol; Vahala, L.; Vahala, G.

    2006-01-01

    Roč. 51, č. 7 (2006), QPI.00027 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/48th./. Philadelphia, Pennsylvania , 30.10.2006-3.11.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * MAST * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://www.aps.org/meet/DPP06/baps/all_DPP06.pdf

  3. Transport and stability analyses supporting disruption prediction in high beta KSTAR plasmas

    Science.gov (United States)

    Ahn, J.-H.; Sabbagh, S. A.; Park, Y. S.; Berkery, J. W.; Jiang, Y.; Riquezes, J.; Lee, H. H.; Terzolo, L.; Scott, S. D.; Wang, Z.; Glasser, A. H.

    2017-10-01

    KSTAR plasmas have reached high stability parameters in dedicated experiments, with normalized beta βN exceeding 4.3 at relatively low plasma internal inductance li (βN/li>6). Transport and stability analyses have begun on these plasmas to best understand a disruption-free path toward the design target of βN = 5 while aiming to maximize the non-inductive fraction of these plasmas. Initial analysis using the TRANSP code indicates that the non-inductive current fraction in these plasmas has exceeded 50 percent. The advent of KSTAR kinetic equilibrium reconstructions now allows more accurate computation of the MHD stability of these plasmas. Attention is placed on code validation of mode stability using the PEST-3 and resistive DCON codes. Initial evaluation of these analyses for disruption prediction is made using the disruption event characterization and forecasting (DECAF) code. The present global mode kinetic stability model in DECAF developed for low aspect ratio plasmas is evaluated to determine modifications required for successful disruption prediction of KSTAR plasmas. Work supported by U.S. DoE under contract DE-SC0016614.

  4. PDX modification to produce a bean-shaped high-beta plasma

    International Nuclear Information System (INIS)

    Materna, P.; Chrzanowski, J.; Heitzenroeder, P.; Lee, K.; Pereira, M.

    1983-01-01

    Princeton's PDX tokamak is being converted to produce bean-shaped plasmas which hopefully will reach beta of 10%. The work, which is nearly complete, involves repositioning active coils, adding passive coils, and making external modifications

  5. An In-situ materials analysis particle probe (MAPP) diagnostic to study particle density control and hydrogenic fuel retention in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Allain, Jean-Paul [Purdue Univ., West Lafayette, IN (United States)

    2014-09-05

    A new materials analysis particle probe (MAPP) was designed, constructed and tested to develop understanding of particle control and hydrogenic fuel retention in lithium-based plasma-facing surfaces in NSTX. The novel feature of MAPP is an in-situ tool to probe the divertor NSTX floor during LLD and lithium-coating shots with subsequent transport to a post-exposure in-vacuo surface analysis chamber to measure D retention. In addition, the implications of a lithiated graphite-dominated plasma-surface environment in NSTX on LLD performance, operation and ultimately hydrogenic pumping and particle control capability are investigated in this proposal. MAPP will be an invaluable tool for erosion/redeposition simulation code validation.

  6. The NSTX Trouble Reporting System

    International Nuclear Information System (INIS)

    Sengupta, S.; Oliaro, G.

    2002-01-01

    An online Trouble Reporting System (TRS) has been introduced at the National Spherical Torus Experiment (NSTX). The TRS is used by NSTX operators to report problems that affect NSTX operations. The purpose of the TRS is to enhance NSTX reliability and maintainability by identifying components, occurrences, and trends that contribute to machine downtime. All NSTX personnel have access to the TRS. The user interface is via a web browser, such as Netscape or Internet Explorer. This web-based feature permits any X-terminal, PC, or MAC access to the TRS. The TRS is based upon a trouble reporting system developed at the DIII-D Tokamak, at General Atomics Technologies. This paper will provide a detailed description of the TRS software architecture, user interface, MS SQL server interface and operational experiences. In addition, sample data from the TRS database will be summarized and presented

  7. Simultaneous determination of F-beta-alanine and beta-alanine in plasma and urine with dual-column reversed-phase high-performance liquid chromatography

    NARCIS (Netherlands)

    van Kuilenburg, A. B.; Stroomer, A. E.; Peters, G. J.; van Gennip, A. H.

    2001-01-01

    F-beta-Alanine and beta-alanine were detected in plasma and urine samples with fluorescence detection of orthophthaldialdehyde derivatives of F-beta-alanine and beta-alanine after separation with dual-column reversed-phase HPLC. The detection limits of F-beta-alanine and beta-alanine in the HPLC

  8. Results of using the NSTX-U Plasma Control System for scenario development

    Science.gov (United States)

    Boyer, M. D.; Battaglia, D. J.; Gates, D. A.; Gerhardt, S.; Menard, J.; Mueller, D.; Myers, C. E.; Ferron, J.; Sabbagh, S.; NSTX-U Team

    2016-10-01

    To best use the new capabilities of NSTX-U (e.g., higher toroidal field and additional, more distributed heating and current drive sources) and to achieve the operational goals of the program, major upgrades to the Plasma Control System have been made. These include improvements to vertical control, real-time equilibrium reconstruction, and plasma boundary shape control and the addition of flexible algorithms for beam modulation and gas injection to control the upgraded actuators in real-time, enabling their use in algorithms for stored energy and profile control. Control system commissioning activities have so far focused on vertical position and shape control. The upgraded controllers have been used to explore the vertical stability limits in inner wall limited and diverted discharges, and control of X-point and strike point locations has been demonstrated and is routinely used. A method for controlling the mid-plane inner gap, a challenge for STs, has also been added to improve reproducible control of diverted discharges. A supervisory shutdown handling algorithm has also been commissioned to ramp the plasma down and safely turn off actuators after an event such as loss of vertical control. Use of the upgrades has contributed to achieving 1MA, 0.65T scenarios with greater than 1s pulse length. Work supported by U.S. D.O.E. Contract No. DE-AC02-09CH11466.

  9. Recent Fast Wave Coupling and Heating Studies on NSTX, with Possible Implications for ITER

    International Nuclear Information System (INIS)

    Hosea, J.C.; Bell, R.E.; Feibush, E.; Harvey, R.W.; Jaeger, E.F.; LeBlanc, B.P; Maingi, R.; Phillips, C.K.; Roquemore, L.; Ryan, P.M.; Taylor, G.; Tritz, K.; Valeo, E.J.; Wilgen, J.; Wilson, J.R.

    2009-01-01

    The goal of the high harmonic fast wave (HHFW) research on NSTX is to maximize the coupling of RF power to the core of the plasma by minimizing the coupling of RF power to edge loss processes. HHFW core plasma heating efficiency in helium and deuterium L-mode discharges is found to improve markedly on NSTX when the density 2 cm in front of the antenna is reduced below that for the onset of perpendicular wave propagation (n onset ∝ B*k # parallel# 2 /ω). In NSTX, the observed RF power losses in the plasma edge are driven in the vicinity of the antenna as opposed to resulting from multi-pass edge damping. PDI surface losses through ion-electron collisions are estimated to be significant. Recent spectroscopic measurements suggest that additional PDI losses could be caused by the loss of energetic edge ions on direct loss orbits and perhaps result in the observed clamping of the edge rotation. Initial deuterium H-mode heating studies reveal that core heating is degraded at lower k φ (- 8 m -1 relative to 13 m -1 ) as for the Lmode case at elevated edge density. Fast visible camera images clearly indicate that a major edge loss process is occurring from the plasma scrape off layer (SOL) in the vicinity of the antenna and along the magnetic field lines to the lower outer divertor plate. Large type I ELMs, which are observed at both k φ values, appear after antenna arcs caused by precursor blobs, low level ELMs, or dust. For large ELMs without arcs, the source reflection coefficients rise on a 0.1 ms time scale, which indicates that the time derivative of the reflection coefficient can be used to discriminate between arcs and ELMs.

  10. Flux consumption optimization and the achievement of 1 MA discharges on NSTX

    International Nuclear Information System (INIS)

    Menard, J.; LeBlanc, B.; Sabbagh, S.A.

    2001-01-01

    The spherical tokamak (ST), because of its slender central column, has very limited volt-second capability relative to a standard aspect ratio tokamak of similar plasma cross-section. Recent experiments on the National Spherical Torus Experiment (NSTX) have begun to quantify and optimize the ohmic current drive efficiency in a MA-class ST device. Sustainable ramp-rates in excess of 5MA/sec during the current rise phase have been achieved on NSTX, while faster ramps generate significant MHD activity. Discharges with I P exceeding 1MA have been achieved in NSTX with nominal parameters: aspect ratio A=1.3-1.4, elongation κ=2-2.2, triangularity δ=0.4, internal inductance l i =0.6, and Ejima coefficient C E =0.35. Flux consumption efficiency results, performance improvements associated with first boronization, and comparisons to neoclassical resistivity are described. (author)

  11. Achieving high fusion reactivity in high poloidal beta discharges in TFTR

    International Nuclear Information System (INIS)

    Manuel, M.E.; Navratil, G.A.; Sabbagh, S.A.; Batha, S.; Bell, M.G.; Bell, R.; Budny, R.V.; Bush, C.E.; Cavallo, A.; Chance, M.S.; Cheng, C.Z.; Efthimion, P.C.; Fredrickson, E.D.; Fu, G.Y.; Hawryluk, R.J.; Janos, A.C.; Jassby, D.L.; Levinton, F.; Mikkelsen, D.R.; Manickam, J.; McCune, D.C.; McGuire, K.M.; Medley, S.S.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Park, H.K.; Ramsey, A.T.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Wieland, R.M.; Yamada, M.; Zarnstorff, M.C.: Zweben, S.; Kesner, J.; Marmar, E.; Snipes, J.; Terry, J.

    1993-04-01

    High poloidal beta discharges have been produced in TFTR that achieved high fusion reactivities at low plasma currents. By rapidly decreasing the plasma current just prior to high-power neutral beam injection, relatively peaked current profiles were created having high l i > 2, high Troyon-normalized beta, βN > 3, and high poloidal beta. β p ≥ 0.7 R/a. The global energy confinement time after the current ramp was comparable to supershots, and the combination of improved MHD stability and good confinement produced a new high εβ p high Q DD operating mode for TFTR. Without steady-state current profile control, as the pulse lengths of high βp discharges were extended, l i decreased, and the improved stability produced immediately after by the current ramp deteriorated. In four second, high εβ p discharges, the current profile broadened under the influence of bootstrap and beam-drive currents. When the calculated voltage throughout the plasma nearly vanished, MHD instabilities were observed with β N as low as 1.4. Ideal MHD stability calculations showed this lower beta limit to be consistent with theoretical expectations

  12. EBW simulation for MAST and NSTX experiments

    International Nuclear Information System (INIS)

    Preinhaelter, J.; Urban, J.; Pavlo, P.; Taylor, G.; Shevchenko, V.; Valovic, M.; Vahala, L.; Vahala, G.

    2005-01-01

    The interpretation of EBW emission from spherical tokamaks is nontrivial. We report on a 3D simulation model of this process that incorporates Gaussian beams for the antenna, a full wave solution of EBW-X and EBW-X-O conversions using adaptive finite elements, and EBW ray tracing to determine the radiative temperature. This model is then used to interpret the experimental results from MAST and NSTX. EBW for ELM free H-modes in MAST suggests that the magnetic equilibrium determined by the EFIT code does not adequately represent the B-field within the transport barrier. Using the EBW signal for the reconstruction of the radial profile of the magnetic field, we determine a new equilibrium and see that the EBW simulation now yields better agreement with experimental results. EBW simulations yield excellent results for the time development of the plasma temperature as measured by the EBW radiometer on NSTX

  13. Ballooning stable high beta tokamak equilibria

    International Nuclear Information System (INIS)

    Tuda, Takashi; Azumi, Masafumi; Kurita, Gen-ichi; Takizuka, Tomonori; Takeda, Tatsuoki

    1981-04-01

    The second stable regime of ballooning modes is numerically studied by using the two-dimensional tokamak transport code with the ballooning stability code. Using the simple FCT heating scheme, we find that the plasma can locally enter this second stable regime. And we obtained equilibria with fairly high beta (β -- 23%) stable against ballooning modes in a whole plasma region, by taking into account of finite thermal diffusion due to unstable ballooning modes. These results show that a tokamak fusion reactor can operate in a high beta state, which is economically favourable. (author)

  14. Numerical models for high beta magnetohydrodynamic flow

    International Nuclear Information System (INIS)

    Brackbill, J.U.

    1987-01-01

    The fundamentals of numerical magnetohydrodynamics for highly conducting, high-beta plasmas are outlined. The discussions emphasize the physical properties of the flow, and how elementary concepts in numerical analysis can be applied to the construction of finite difference approximations that capture these features. The linear and nonlinear stability of explicit and implicit differencing in time is examined, the origin and effect of numerical diffusion in the calculation of convective transport is described, and a technique for maintaining solenoidality in the magnetic field is developed. Many of the points are illustrated by numerical examples. The techniques described are applicable to the time-dependent, high-beta flows normally encountered in magnetically confined plasmas, plasma switches, and space and astrophysical plasmas. 40 refs

  15. Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

    Energy Technology Data Exchange (ETDEWEB)

    Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Jaworski, M.A.; Kaita, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Hirooka, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Gray, T.K. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2017-04-15

    Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety concerns (Federici et al., 2001) . It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance (Ono et al., 2013, 2014) . The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium (LL) divertor (RLLD) concept (Ono et al., 2013) and its variant, the active liquid lithium divertor concept (ARLLD) (Ono et al., 2014) , taking advantage of the enhanced non-coronal Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/s of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤450 °C than the first wall ∼600–700 °C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ∼1 l/s (l/s) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust/impurities are removed by relatively simple filter and cold/hot trap systems. Using a

  16. Diagnostics for Evaluating Performance of NSTX Liquid Lihium Divertor

    Science.gov (United States)

    Kaita, R.; Kugel, H.; Kallman, J.; Leblanc, B.; Paul, S.; Roquemore, A. L.; Skinner, C.; Soukhanovskii, V.; Maingi, R.; Ahn, J.-W.; Wilgen, J.; Allain, J.-P.; Taylor, C.

    2009-11-01

    A Liquid Lithium Divertor (LLD) is being installed on NSTX to investigate particle control and power handling with liquid lithium as plasma-facing component (PFC). The LLD is expected to provide a low-recycling plasma-facing component (PFC). To study the effects of such a PFC on plasma performance, a variety of edge measurements are required. Since its surface is highly reflective at visible wavelengths, a Lyman-alpha detector array will be used to monitor the recycling. To understand changes in edge transport, electron temperature and density measurements will be made with Langmuir probes mounted in PFC's near the LLD, and the edge sightlines of a multipoint Thomson scattering system. A frequency-scanning reflectometer will also provide scrapeoff layer electron density profiles. The LLD response to heat loads will be examined with infrared cameras and thermocouples. Diagnostics are also needed to measure the erosion and codeposition of lithium. They include quartz deposition monitors and a retractable probe for exposing samples to the plasma.

  17. Te(R,t) Measurements using Electron Bernstein Wave Thermal Emission on NSTX

    International Nuclear Information System (INIS)

    Diem, S.J.; Taylor, G.; Efthimion, P.C.; LeBlanc, B.P.; Carter, M.; Caughman, J.; Wilgen, J.B.; Harvey, R.W.; Preinhaelter, J.; Urban, J.

    2006-01-01

    The National Spherical Torus Experiment (NSTX) routinely studies overdense plasmas with n e of (1-5) x 10 19 m -3 and total magnetic field of e measurement. A significant upgrade to the previous NSTX EBW emission diagnostic to measure thermal EBW emission via the oblique B-X-O mode conversion process has been completed. The new EBW diagnostic consists of two remotely steerable, quad-ridged horn antennas, each of which is coupled to a dual channel radiometer. Fundamental (8-18 GHz) and second and third harmonic (18-40 GHz) thermal EBW emission and polarization measurements can be obtained simultaneously.

  18. Impurity analysis of NSTX using a transmission grating-based imaging spectrometer

    International Nuclear Information System (INIS)

    Kumar, Deepak; Finkenthal, Michael; Stutman, Dan; Clayton, Daniel J; Tritz, Kevin; Bell, Ronald E; Diallo, Ahmed; LeBlanc, Ben P; Podesta, Mario

    2012-01-01

    A transmission grating-based imaging spectrometer has recently been installed and operated on the National Spherical Torus Experiment (NSTX) at PPPL. This paper describes the spectral and spatial characteristics of impurity emission under different operating conditions of the experiment—neutral beam heated, ohmic heated and RF heated plasma. A typical spectrum from each scenario is analyzed to provide quantitative estimates of impurity fractions in the plasma. (paper)

  19. The NSTX Trouble Reporting System; TOPICAL

    International Nuclear Information System (INIS)

    S. Sengupta; G. Oliaro

    2002-01-01

    An online Trouble Reporting System (TRS) has been introduced at the National Spherical Torus Experiment (NSTX). The TRS is used by NSTX operators to report problems that affect NSTX operations. The purpose of the TRS is to enhance NSTX reliability and maintainability by identifying components, occurrences, and trends that contribute to machine downtime. All NSTX personnel have access to the TRS. The user interface is via a web browser, such as Netscape or Internet Explorer. This web-based feature permits any X-terminal, PC, or MAC access to the TRS. The TRS is based upon a trouble reporting system developed at the DIII-D Tokamak, at General Atomics Technologies. This paper will provide a detailed description of the TRS software architecture, user interface, MS SQL server interface and operational experiences. In addition, sample data from the TRS database will be summarized and presented

  20. Design and Construction of the NSTX Bakeout, Cooling and Vacuum Systems

    International Nuclear Information System (INIS)

    Dudek, L.E.; Kalish, M.; Gernhardt, R.; Parsells, R.F.; Blanchard, W.

    1999-01-01

    This paper will describe the design, construction and initial operation of the NSTX bakeout, water cooling and vacuum systems. The bakeout system is designed for two modes of operation. The first mode allows heating of the first wall components to 350 degrees C while the external vessel is cooled to 150 degrees C. The second mode cools the first wall to 150 degrees C and the external vessel to 50 degrees C. The system uses a low viscosity heat transfer oil which is capable of high temperature low pressure operation. The NSTX Torus Vacuum Pumping System (TVPS) is designed to achieve a base pressure of approximately 1x10 (superscript -8) Torr and to evacuate the plasma fuel gas loads in less than 5 minutes between discharges. The vacuum pumping system is capable of a pumping speed of approximately 3400 l/s for deuterium. The hardware consists of two turbo molecular pumps (TMPs) and a mechanical pump set consisting of a mechanical and a Roots blower pump. A PLC is used as the control system to provide remote monitoring, control and software interlock capability. The NSTX cooling water provides chilled, de ionized water for heat removal in the TF, OH and PF, power supplies, bus bar systems, and various diagnostics. The system provides flow monitoring via a PLC to prevent damage due to loss of flow

  1. Sawtooth crashes at high beta on JET

    Energy Technology Data Exchange (ETDEWEB)

    Alper, B; Huysmans, G T.A.; Sips, A C.C. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Nave, M F.F. [Universidade Tecnica, Lisbon (Portugal). Inst. Superior Tecnico

    1994-07-01

    The sawtooth crashes on JET display features which depend on beta. The main observation is a transient bulging of flux surfaces (duration inferior to 30 microsec.), which is predominantly on the low field side and extends to larger radii as beta increases. This phenomenon reaches the plasma boundary when beta{sub N} exceeds 0.5 and in these cases is followed by an ELM within 50 microsec. These sawtooth/ELM events limit plasma performance. Modelling of mode coupling shows qualitative agreement between observations of the structure of the sawtooth precursor and the calculated internal kink mode at high beta. (authors). 6 refs., 5 figs.

  2. Development of a low-energy and high-current pulsed neutral beam injector with a washer-gun plasma source for high-beta plasma experiments.

    Science.gov (United States)

    Ii, Toru; Gi, Keii; Umezawa, Toshiyuki; Asai, Tomohiko; Inomoto, Michiaki; Ono, Yasushi

    2012-08-01

    We have developed a novel and economical neutral-beam injection system by employing a washer-gun plasma source. It provides a low-cost and maintenance-free ion beam, thus eliminating the need for the filaments and water-cooling systems employed conventionally. In our primary experiments, the washer gun produced a source plasma with an electron temperature of approximately 5 eV and an electron density of 5 × 10(17) m(-3), i.e., conditions suitable for ion-beam extraction. The dependence of the extracted beam current on the acceleration voltage is consistent with space-charge current limitation, because the observed current density is almost proportional to the 3/2 power of the acceleration voltage below approximately 8 kV. By optimizing plasma formation, we successfully achieved beam extraction of up to 40 A at 15 kV and a pulse length in excess of 0.25 ms. Its low-voltage and high-current pulsed-beam properties enable us to apply this high-power neutral beam injection into a high-beta compact torus plasma characterized by a low magnetic field.

  3. Central safety factor and β N control on NSTX-U via beam power and plasma boundary shape modification, using TRANSP for closed loop simulations

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, M. D.; Andre, R.; Gates, D. A.; Gerhardt, S.; Goumiri, I. R.; Menard, J.

    2015-04-24

    The high-performance operational goals of NSTX-U will require development of advanced feedback control algorithms, including control of ßN and the safety factor profile. In this work, a novel approach to simultaneously controlling ßN and the value of the safety factor on the magnetic axis, q0, through manipulation of the plasma boundary shape and total beam power, is proposed. Simulations of the proposed scheme show promising results and motivate future experimental implementation and eventual integration into a more complex current profile control scheme planned to include actuation of individual beam powers, density, and loop voltage. As part of this work, a flexible framework for closed loop simulations within the high-fidelity code TRANSP was developed. The framework, used here to identify control-design-oriented models and to tune and test the proposed controller, exploits many of the predictive capabilities of TRANSP and provides a means for performing control calculations based on user-supplied data (controller matrices, target waveforms, etc.). The flexible framework should enable high-fidelity testing of a variety of control algorithms, thereby reducing the amount of expensive experimental time needed to implement new control algorithms on NSTX-U and other devices.

  4. Radioimmunoassay of human. beta. -lipotropin in unextracted plasma. [/sup 125/I tracer technique

    Energy Technology Data Exchange (ETDEWEB)

    Wiedemann, E. (Univ. of California, Berkeley); Saito, T.; Linfoot, J.A.; Li, C.H.

    1977-11-01

    A sensitive and specific radioimmunoassay for human ..beta..-lipotropin (..beta../sub h/-LPH) in unextracted plasma was developed using pure ..beta../sub h/-LPH as tracer and standard and an antiserum not cross-reacting with human ..beta..-MSH and hACTH. In healthy volunteers plasma ..beta../sub h/-LPH ranged from <20 to 150 pg/ml at 8:00 a.m. and rose after metyrapone administration. ..beta../sub h/-LPH was very low in panhypopituitarism, normal in most patients with untreated Cushing's disease, elevated in acromegaly and extremely high in Nelson's syndrome.

  5. Equilibrium and stability of high-beta plasma in a finite l=+-1 toroidal system

    International Nuclear Information System (INIS)

    Shiina, S.; Saito, K.; Todoroki, J.; Hamada, S.; Gesso, H.; Nogi, Y.; Osanai, Y.; Yoshimura, H.

    1983-01-01

    The equilibrium and stability are theoretically and experimentally investigated of high-beta plasma in the Modified Bumpy Torus, which is an asymmetric closed-line system with fairly large l=0 and l=+-1 field components. The finiteness of the l=+-1 component induces significant stabilizing effects due both to self formation of a magnetic well and to the conducting wall. (author)

  6. High-k Scattering Receiver Mixer Performance for NSTX-U

    Science.gov (United States)

    Barchfeld, Robert; Riemenschneider, Paul; Domier, Calvin; Luhmann, Neville; Ren, Yang; Kaita, Robert

    2016-10-01

    The High-k Scattering system detects primarily electron-scale turbulence k θ spectra for studying electron thermal transport in NSTX-U. A 100 mW, 693 GHz probe beam passes through plasma, and scattered power is detected by a 4-pixel quasi optical, mixer array. Remotely controlled receiving optics allows the scattering volume to be located from core to edge with a k θ span of 7 to 40 cm-1. The receiver array features 4 RF diagonal input horns, where the electric field polarization is aligned along the diagonal of a square cross section horn, at 30 mm channel spacing. The local oscillator is provided by a 14.4 GHz source followed by a x48 multiplier chain, giving an intermediate frequency of 1 GHz. The receiver optics receive 4 discreet scattering angles simultaneously, and then focus the signals as 4 parallel signals to their respective horns. A combination of a steerable probe beam, and translating receiver, allows for upward or downward scattering which together can provide information about 2D turbulence wavenumber spectrum. IF signals are digitized and stored for later computer analysis. The performance of the receiver mixers is discussed, along with optical design features to enhance the tuning and performance of the mixers. Work supported in part by U.S. DOE Grant DE-FG02-99ER54518 and DE-AC02-09CH1146.

  7. High-beta studies with beam-heated, non-circular plasmas in ISX-B

    International Nuclear Information System (INIS)

    Lazarus, E.A.; Bates, S.C.; Bush, C.E.

    1981-01-01

    In this paper we describe some preliminary results of high beta studies on ISX-B for mildly D shaped discharges. ISX-B is a modest size tokamak (R 0 = 93 cm, a = 27 cm) equipped with two tangantially-aligned neutral beam injectors giving a total power up to 3 MW. The poloidal coil system allows choice of plasma boundary shapes from circular to elongated (kappa less than or equal to 1.8), with D, elliptical, or inverse D cross sections. The non-circular work discussed here is for kappa approx. = 1.5

  8. Visible imaging of edge turbulence in NSTX

    International Nuclear Information System (INIS)

    Zweben, S.; Maqueda, R.; Hill, K.; Johnson, D.

    2000-01-01

    Edge plasma turbulence in tokamaks and stellarators is believed to cause the radical heat and particle flux across the separatrix and into the scrape-off-layers of these devices. This paper describes initial measurements of 2-D space-time structure of the edge density turbulence made using a visible imaging diagnostic in the National Spherical Torus Experiment (NSTX). The structure of the edge turbulence is most clearly visible using a method of gas puff imaging to locally illuminate the edge density turbulence

  9. Visible imaging of edge turbulence in NSTX

    International Nuclear Information System (INIS)

    S. Zweben; R. Maqueda; K. Hill; D. Johnson; S. Kaye; H. Kugel; F. Levinton; R. Maingi; L. Roquemore; S. Sabbagh; G. Wurden

    2000-01-01

    Edge plasma turbulence in tokamaks and stellarators is believed to cause the radial heat and particle flux across the separatrix and into the scrape-off-layers of these devices. This paper describes initial measurements of 2-D space-time structure of the edge density turbulence made using a visible imaging diagnostic in the National Spherical Torus Experiment (NSTX). The structure of the edge turbulence is most clearly visible using a method of ''gas puff imaging'' to locally illuminate the edge density turbulence

  10. Plasma beta-endorphin, beta-lipotropin and corticotropin in polycystic ovarian disease.

    Science.gov (United States)

    Laatikainen, T; Salminen, K; Virtanen, T; Apter, D

    1987-04-01

    In 9 women with polycystic ovarian disease (PCOD) and in 11 control subjects at the follicular phase of the normal cycle, blood samples were collected at 15-min intervals during a 2 h period of bed rest for the assay of beta-endorphin, beta-lipotropin, corticotropin, cortisol and prolactin. During the study period, the plasma levels of these hormones decreased more significantly in the PCOD than in the control group, suggesting that the PCOD patients had a more significant stress response to the puncture of the vein than the control subjects. The second hour of the study period was considered to represent resting levels of hormones. The mean resting levels (+/- S.E.) of the hormones between the PCOD and control groups, respectively, were as follows: beta-E, 2.0 +/- 0.4 vs. 1.1 +/- 0.1 pmol/l, p less than 0.05; beta-LPH, 3.4 +/- 0.6 vs. 2.1 +/- 0.5 pmol/l, N.S.; corticotropin, 2.0 +/- 0.3 vs. 1.1 +/- 0.5 pmol/l, p less than 0.05; cortisol, 176 +/- 24 vs. 128 +/- 16, N.S.; and prolactin; 3.9 +/- 0.6 vs. 5.6 +/- 1.2 ng/ml, N.S. These results confirm the previous findings on increased circulating levels of beta-E in PCOD. A concomitant increase of the plasma level of corticotropin suggests that the basal secretion of both beta-E and corticotropin from the anterior pituitary gland is increased in women with PCOD.

  11. Overview of results from the National Spherical Torus Experiment (NSTX)

    Czech Academy of Sciences Publication Activity Database

    Gates, D.A.; Ahn, J.; Allain, J.; Andre, R.; Bastasz, R.; Bell, M.; Bell, R.; Belova, E.; Berkery, J.; Betti, R.; Bialek, J.; Biewer, T.; Bigelow, T.; Bitter, M.; Boedo, J.; Bonoli, P.; Boozer, A.; Brennan, D.; Breslau, J.; Brower, D.; Bush, C.; Canik, J.; Caravelli, G.; Carter, M.; Caughman, J.; Chang, C.; Crocker, N.; Darrow, D.; Delgado-Aparicio, L.; Diem, S.; D’Ippolito, D.; Domier, C.; Dorland, W.; Efthimion, P.; Ejiri, A.; Ershov, N.; Evans, T.; Feibush, E.; Fenstermacher, M.; Ferron, J.; Finkenthal, M.; Foley, J.; Frazin, R.; Fredrickson, E.; Fu, G.; Funaba, H.; Gerhardt, S.; Glasser, A.; Gorelenkov, N.; Grisham, L.; Hahm, T.; Harvey, R.; Hassanein, A.; Heidbrink, W.; Hill, K.; Hillesheim, J.; Hillis, D.; Hirooka, Y.; Hosea, J.; Hu, B.; Humphreys, D.; Idehara, T.; Indireshkumar, K.; Ishida, A.; Jaeger, F.; Jarboe, T.; Jardin, S.; Jaworski, M.; Ji, H.; Kaita, R.; Kallman, J.; Katsuro-Hopkins, O.; Kawahata, K.; Kawamori, E.; Kaye, S.; Kessel, C.; Kimura, H.; Kolemen, E.; Krasheninnikov, H.; Krstic, P.; Ku, S.; Kubota, S.; Kugel, H.; La Haye, R.; Lao, L.; LeBlanc, B.; Lee, K.; Leuer, J.; Levinton, F.; Liang, Y.; Liu, D.; Luhmann Jr, N.; Maingi, R.; Majeski, R.; Manickam, J.; Mansfield, D.; Maqueda, R.; Mazzucato, E.; McCune, D.; McGeehan, B.; McKee, G.; Medley, S.; Menard, J.; Menon, M.; Meyer, H.; Mikkelsen, D.; Miloshevsky, G.; Mitarai, O.; Mueller, D.; Mueller, S.; Munsat, T.; Myra, J.; Nagayama, Y.; Nelson, B.; Nguyen, X.; Nishino, N.; Nishiura, M.; Nygren, R.; Ono, M.; Osborne, T.; Pacella, D.; Park, J.; Paul, S.; Peebles, W.; Penaflor, B.; Peng, M.; Phillips, C.; Pigarov, A.; Podesta, M.; Preinhaelter, Josef; Ram, A.; Raman, R.; Rasmussen, D.; Redd, A.; Reimerdes, H.; Rewoldt, G.; Ross, P.; Rowley, C.; Ruskov, E.; Russell, D.; Ruzic, D.; Ryan, P.; Sabbagh, S.; Schaffer, M.; Schuster, E.; Scott, S.; Shaing, K.; Sharpe, P.; Shevchenko, V.; Shinohara, K.; Sizyuk, V.; Skinner, C.; Smirnov, A.; Smith, D.; Snyder, P.; Solomon, W.; Sontag, A.; Soukhanovskii, V.; Stoltzfus-Dueck, T.; Stotler, D.; Strait, T.; Stratton, B.; Stutman, D.; Takahashi, R.; Takase, Y.; Tamura, N.; Tang, X.; Taylor, G.; Taylor, C.; Ticos, C.; Tritz, K.; Tsarouhas, D.; Turrnbull, A.; Tynan, G.; Ulrickson, M.; Umansky, M.; Urban, Jakub; Utergberg, E.; Walker, M.; Wampler, M.; Wang, J.; Wang, W.; Welander, A.; Whaley, J.; White, R.; Wilgen, J.; Wilson, R.; Wong, K.; Wright, J.; Xia, Z.; Xu, X.; Youchison, D.; Yu, G.; Yuh, H.; Zakharov, L.; Zemlyanov, D.; Zweben, S.; Choe, W.; Jung, H.; Kim, J.; Lee, W.; Park, H.

    2009-01-01

    Roč. 49, č. 10 (2009), s. 104016-104016 ISSN 0029-5515. [IAEA Fusion Energy Conference/22nd./. Geneva, 13.10.2008-18.10.2008] R&D Projects: GA ČR GA202/08/0419 Institutional research plan: CEZ:AV0Z20430508 Keywords : NSTX * Spherical tokamaks * Overdense plasma * Conversion * Emission * Tokamaks * Elektron Bernstein waves Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://www.iop.org/EJ/article/0029-5515/49/10/104016/nf9_10_104016

  12. High beta and confinement studies on TFTR

    International Nuclear Information System (INIS)

    Navratil, G.A.; Bhattacharjee, A.; Iacono, R.; Mauel, M.E.; Sabbagh, S.A.; Kesner, J.

    1992-01-01

    A new regime of high poloidal beta operation in TFTR was developed in the course of the first two years of this project (9/25/89 to 9/24/91). Our proposal to continue this successful collaboration between Columbia University and the Massachusetts Institute of Technology with the Princeton Plasma Physics Laboratory for a three year period (9/25/91 to 9/24/94) to continue to investigate improved confinement and tokamak performance in high poloidal beta plasmas in TFTR through the DT phase of operation was approved by the DOE and this is a report of our progress during the first 9 month budget period of the three year grant (9/25/91 to 6/24/92). During the approved three year project period we plan to (1) extend and apply the low current, high QDD discharges to the operation of TFTR using Deuterium and Tritium plasma; (2) continue the analysis and plan experiments on high poloidal beta phenomena in TFTR including: stability properties, enhanced global confinement, local transport, bootstrap current, and divertor formation; (3) plan and carry out experiments on TFTR which attempt to elevate the central q to values > 2 where entry to the second stability regime is predicted to occur; and (4) collaborate on high beta experiments using bean-shaped plasmas with a stabilizing conducting shell in PBX-M. In the seven month period covered by this report we have made progress in each of these four areas through the submission of 4 TFTR Experimental Proposals and the partial execution of 3 of these using a total of 4.5 run days during the August 1991 to February 1992 run

  13. Investigation of collisional EBW damping and its importance to EBW emission from NSTX

    Czech Academy of Sciences Publication Activity Database

    Urban, Jakub; Preinhaelter, Josef; Diem, S.J.; Taylor, G.; Vahala, L.; Vahala, G.

    2007-01-01

    Roč. 52, č. 16 (2007), s. 304-304 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/49th./. Orlando , Florida, 12.11.2007-16.11.2007] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://meetings.aps.org/Meeting/DPP07/Content/901

  14. High-resolution Tangential AXUV Arrays for Radiated Power Density Measurements on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Delgado-Aparicio, L [PPPL; Bell, R E [PPPL; Faust, I [MIT; Tritz, K [The Johns Hopkins University, Baltimore, MD, 21209, USA; Diallo, A [PPPL; Gerhardt, S P [PPPL; Kozub, T A [PPPL; LeBlanc, B P [PPPL; Stratton, B C [PPPL

    2014-07-01

    Precise measurements of the local radiated power density and total radiated power are a matter of the uttermost importance for understanding the onset of impurity-induced instabilities and the study of particle and heat transport. Accounting of power balance is also needed for the understanding the physics of various divertor con gurations for present and future high-power fusion devices. Poloidal asymmetries in the impurity density can result from high Mach numbers and can impact the assessment of their flux-surface-average and hence vary the estimates of P[sub]rad (r, t) and (Z[sub]eff); the latter is used in the calculation of the neoclassical conductivity and the interpretation of non-inductive and inductive current fractions. To this end, the bolometric diagnostic in NSTX-U will be upgraded, enhancing the midplane coverage and radial resolution with two tangential views, and adding a new set of poloidally-viewing arrays to measure the 2D radiation distribution. These systems are designed to contribute to the near- and long-term highest priority research goals for NSTX-U which will integrate non-inductive operation at reduced collisionality, with high-pressure, long energy-confinement-times and a divertor solution with metal walls.

  15. Plasma beta-endorphin-like immunoreactivity and its variations in baboons

    Energy Technology Data Exchange (ETDEWEB)

    Golanov, E.V.; Fufacheva, A.A.; Parin, S.B.

    1986-04-01

    This paper determines the level of beta-endorphin-like immunoreactivity (beta-elir) in the blood plasma of baboons and studies its changes in certain situations. For radioimmunoassay of beta-ELIR in the blood plasma, a standard kit and the appropriate technique were used. The background plasma beta-ELIR level of the baboons, in a state of quiet wakefulness, was 0.0 = 1.0 fmoles/ml. The total level of b-ELIR was 134 plus or minus 24 pg/ml. The data show that elevation of the plasma b-ELIR level accompanies stress formation, including the development of a state of shock in baboons. A definite role in the regulation of the plasma b-endorphin level may be played by the paraventricular-perifornical region of the hypothalamus.

  16. An Edge Rotation and Temperature Diagnostic on NSTX

    International Nuclear Information System (INIS)

    Biewer, T.M.; Bell, R.E.; Feder, R.; Johnson, D.W.; Palladino, R.W.

    2003-01-01

    A new diagnostic for the National Spherical Torus Experiment (NSTX) is described whose function is to measure ion rotation and temperature at the plasma edge. The diagnostic is sensitive to C III, C IV, and He II intrinsic emission, covering a radial region of 15 cm at the extreme edge of the outboard midplane. Thirteen chords are distributed between toroidal and poloidal views, allowing the toroidal and poloidal rotation and temperature of the plasma edge to be simultaneously measured with 10 ms resolution. Combined with the local pressure gradient and the EFIT code reconstructed magnetic field profile, the edge flow gives a measure of the local radial electric field

  17. Resonant magnetohydrodynamic waves in high-beta plasmas

    International Nuclear Information System (INIS)

    Ruderman, M. S.

    2009-01-01

    When a global magnetohydrodynamic (MHD) wave propagates in a weakly dissipative inhomogeneous plasma, the resonant interaction of this wave with either local Alfven or slow MHD waves is possible. This interaction occurs at the resonant position where the phase velocity of the global wave coincides with the phase velocity of either Alfven or slow MHD waves. As a result of this interaction a dissipative layer embracing the resonant position is formed, its thickness being proportional to R -1/3 , where R>>1 is the Reynolds number. The wave motion in the resonant layer is characterized by large amplitudes and large gradients. The presence of large gradients causes strong dissipation of the global wave even in very weakly dissipative plasmas. Very often the global wave motion is characterized by the presence of both Alfven and slow resonances. In plasmas with small or moderate plasma beta β, the resonance positions corresponding to the Alfven and slow resonances are well separated, so that the wave motion in the Alfven and slow dissipative layers embracing the Alfven and slow resonant positions, respectively, can be studied separately. However, when β > or approx. R 1/3 , the two resonance positions are so close that the two dissipative layers overlap. In this case, instead of two dissipative layers, there is one mixed Alfven-slow dissipative layer. In this paper the wave motion in such a mixed dissipative layer is studied. It is shown that this motion is a linear superposition of two motions, one corresponding to the Alfven and the other to the slow dissipative layer. The jump of normal velocity across the mixed dissipative layer related to the energy dissipation rate is equal to the sum of two jumps, one that occurs across the Alfven dissipative layer and the other across the slow dissipative layer.

  18. Comparison of the dynamical processes in plasma turbulence observed in the high- and low-{beta} regions of the terrestrial foreshock

    Energy Technology Data Exchange (ETDEWEB)

    Coca, D.; Balikhin, M.; Billings, S

    2001-06-01

    This paper highlights the fact that the dynamical processes that characterise plasma turbulence observed in the high-{beta} region of the terrestrial foreshock are significantly different from the dynamical processes identified in the low-{beta} region. The study is based on a time-domain model identified from measurements taken by AMPTE-UKS and AMPTE-IRM satellites. (author)

  19. Deposition Measurements in NSTX

    Science.gov (United States)

    Skinner, C. H.; Kugel, H. W.; Hogan, J. T.; Wampler, W. R.

    2004-11-01

    Two quartz microbalances have been used to record deposition on the National Spherical Torus Experiment. The experimental configuration mimics a typical diagnostic window or mirror. An RS232 link was used to acquire the quartz crystal frequency and the deposited thickness was recorded continuously with 0.01 nm resolution. Nuclear Reaction Analysis of the deposit was consistent with the measurement of the total deposited mass from the change in crystal frequency. We will present measurements of the variation of deposition with plasma conditions. The transport of carbon impurities in NSTX has been modelled with the BBQ code. Preliminary calculations indicated a negligible fraction of carbon generated at the divertor plates in quiescent discharges directly reaches the outer wall, and that transient events are responsible for the deposition.

  20. MHD-induced Energetic Ion Loss during H-mode Discharges in the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Medley, S.S.; Gorelenkov, N.N.; Andre, R.; Bell, R.E.; Darrow, D.S.; Fredrickson, E.D.; Kaye, S.M.; LeBlanc, B.P.; Roquemore, A.L.

    2004-01-01

    MHD-induced energetic ion loss in neutral-beam-heated H-mode [high-confinement mode] discharges in NSTX [National Spherical Torus Experiment] is discussed. A rich variety of energetic ion behavior resulting from magnetohydrodynamic (MHD) activity is observed in the NSTX using a horizontally scanning Neutral Particle Analyzer (NPA) whose sightline views across the three co-injected neutral beams. For example, onset of an n = 2 mode leads to relatively slow decay of the energetic ion population (E ∼ 10-100 keV) and consequently the neutron yield. The effect of reconnection events, sawteeth, and bounce fishbones differs from that observed for low-n, low-frequency, tearing-type MHD modes. In this case, prompt loss of the energetic ion population occurs on a time scale of less than or equal to 1 ms and a precipitous drop in the neutron yield occurs. This paper focuses on MHD-induced ion loss during H-mode operation in NSTX. After H-mode onset, the NPA charge-exchange spectrum usually exhibits a significant loss of energetic ions only for E > E(sub)b/2 where E(sub)b is the beam injection energy. The magnitude of the energetic ion loss was observed to decrease with increasing tangency radius, R(sub)tan, of the NPA sightline, increasing toroidal field, B(sub)T, and increasing neutral-beam injection energy, E(sub)b. TRANSP modeling suggests that MHD-induced ion loss is enhanced during H-mode operation due to an evolution of the q and beam deposition profiles that feeds both passing and trapped ions into the region of low-n MHD activity. ORBIT code analysis of particle interaction with a model magnetic perturbation supported the energy selectivity of the MHD-induced loss observed in the NPA measurements. Transport analysis with the TRANSP code using a fast-ion diffusion tool to emulate the observed MHD-induced energetic ion loss showed significant modifications of the neutral- beam heating as well as the power balance, thermal diffusivities, energy confinement times, and

  1. MHD-induced Energetic Ion Loss during H-mode Discharges in the National Spherical Torus Experiment (NSTX)

    Energy Technology Data Exchange (ETDEWEB)

    S.S. Medley; N.N. Gorelenkov; R. Andre; R.E. Bell; D.S. Darrow; E.D. Fredrickson; S.M. Kaye; B.P. LeBlanc; A.L. Roquemore; and the NSTX Team

    2004-03-15

    MHD-induced energetic ion loss in neutral-beam-heated H-mode [high-confinement mode] discharges in NSTX [National Spherical Torus Experiment] is discussed. A rich variety of energetic ion behavior resulting from magnetohydrodynamic (MHD) activity is observed in the NSTX using a horizontally scanning Neutral Particle Analyzer (NPA) whose sightline views across the three co-injected neutral beams. For example, onset of an n = 2 mode leads to relatively slow decay of the energetic ion population (E {approx} 10-100 keV) and consequently the neutron yield. The effect of reconnection events, sawteeth, and bounce fishbones differs from that observed for low-n, low-frequency, tearing-type MHD modes. In this case, prompt loss of the energetic ion population occurs on a time scale of less than or equal to 1 ms and a precipitous drop in the neutron yield occurs. This paper focuses on MHD-induced ion loss during H-mode operation in NSTX. After H-mode onset, the NPA charge-exchange spectrum usually exhibits a significant loss of energetic ions only for E > E(sub)b/2 where E(sub)b is the beam injection energy. The magnitude of the energetic ion loss was observed to decrease with increasing tangency radius, R(sub)tan, of the NPA sightline, increasing toroidal field, B(sub)T, and increasing neutral-beam injection energy, E(sub)b. TRANSP modeling suggests that MHD-induced ion loss is enhanced during H-mode operation due to an evolution of the q and beam deposition profiles that feeds both passing and trapped ions into the region of low-n MHD activity. ORBIT code analysis of particle interaction with a model magnetic perturbation supported the energy selectivity of the MHD-induced loss observed in the NPA measurements. Transport analysis with the TRANSP code using a fast-ion diffusion tool to emulate the observed MHD-induced energetic ion loss showed significant modifications of the neutral- beam heating as well as the power balance, thermal diffusivities, energy confinement times

  2. Exercise- and cold-induced changes in plasma beta-endorphin and beta-lipotropin in men and women.

    Science.gov (United States)

    Viswanathan, M; Van Dijk, J P; Graham, T E; Bonen, A; George, J C

    1987-02-01

    The plasma beta-endorphin (beta-EP) and beta-lipotropin (beta-LPH) response of men, eumenorrheic women, and amenorrheic women (n = 6) to 1 h of rest or to a bicycle ergometer test [20 min at 30% maximum O2 uptake (VO2max), 20 min at 60% VO2max, and at 90% VO2max to exhaustion] was studied in both normal (22 degrees C) and cold (5 degrees C) environments. beta-EP and beta-LPH was measured by radioimmunoassay in venous samples collected every 20 min during rest or after each exercise bout. Exhaustive exercise at ambient temperature (Ta) 22 degrees C induced significant increases in plasma beta-EP and beta-LPH in all subjects as did work at 60% VO2max in amenorrheic and eumenorrheic women. During work at Ta 5 degrees C, the relative increase in beta-EP and beta-LPH was suppressed in eumenorrheic women and completely prevented in amenorrheic women. Although significant lowering of beta-EP and beta-LPH was observed in men and eumenorrheic women during rest at 5 degrees C, amenorrheic women maintained precold exposure levels. These findings suggest that plasma beta-EP and beta-LPH may reflect a thermoregulatory response to heat load. There appears to be a sexual dimorphism in exercise- and cold-induced release of beta-EP and beta-LPH and amenorrhea may be accompanied by alterations in these responses.

  3. Recent Developments in High-Harmonic Fast Wave Physics in NSTX

    International Nuclear Information System (INIS)

    LeBlanc, B.P.; Bell, R.E.; Bonoli, P.; Harvey, R.; Heidbrink, W.W.; Hosea, J.C.; Kaye, S.M.; Liu, D.; Maingi, R.; Medley, S.S.; Ono, M.; Podesta, M.; Phillips, C.K.; Ryan, P.M.; Roquemore, A.L.; Taylor, G.; Wilson, J.R.

    2010-01-01

    Understanding the interaction between ion cyclotron range of frequency (ICRF) fast waves and the fast-ions created by neutral beam injection (NBI) is critical for future devices such as ITER, which rely on a combination ICRF and NBI. Experiments in NSTX which use 30 MHz High-Harmonic Fast-Wave (HHFW) ICRF and NBI heating show a competition between electron heating via Landau damping and transit-time magnetic pumping, and radio-frequency wave acceleration of NBI generated fast ions. Understanding and mitigating some of the power loss mechanisms outside the last closed flux surface (LCFS) has resulted in improved HHFW heating inside the LCFS. Nevertheless a significant fraction of the HHFW power is diverted away from the enclosed plasma. Part of this power is observed locally on the divertor. Experimental observations point toward the radio-frequency (RF) excitation of surface waves, which disperse wave power outside the LCFS, as a leading loss mechanism. Lithium coatings lower the density at the antenna, thereby moving the critical density for perpendicular fast-wave propagation away from the antenna and surrounding material surfaces. Visible and infrared imaging reveal flows of RF power along open field lines into the divertor region. In L-mode -- low average NBI power -- conditions, the fast-ion D-alpha (FIDA) diagnostic measures a near doubling and broadening of the density profile of the upper energetic level of the fast ions concurrent with the presence of HHFW power launched with k// = -8m-1. We are able to heat NBI-induced H-mode plasmas with HHFW. The captured power is expected to be split between absorption by the electrons and absorption by the fast ions, based on TORIC calculation. In the case discussed here the Te increases over the whole profile when ∼2MW of HHFW power with antenna k// = 13m-1 is applied after the H-mode transition. But somewhat unexpectedly fast-ion diagnostics do not observe a change between the HHFW heated NBI discharge and the

  4. Chosen Solutions to the Engineering Challenges of the National Spherical Torus Experiment (NSTX) Magnets

    International Nuclear Information System (INIS)

    Neumeyer, C.; Fan, H.M.; Chrzanowski, J.; Heitzenroeder, P.

    1999-01-01

    NSTX is one of the largest of a new class of magnetic plasma research devices known as spherical toroids (STs). The plasma in a ST is characterized by its almost spherical shape with a slender cylindrical region through its vertical axis. The so-called 'center stack' is located in this region. It contains magnetic windings for confining the plasma, induce the plasma current, and shape the plasma. This paper will describe the engineering challenges of designing the center stack magnets to meet their operational requirements within this constrained space

  5. Two approaches to the reactor-relevant high-beta plasmas with profile control in the Large Helical Device

    International Nuclear Information System (INIS)

    Ohdachi, S.; Watanabe, K.Y.; Sakakibara, S.

    2008-10-01

    From detailed optimization of configuration, volume averaged beta ∼ 5% has been achieved in the Large Helical Device(LHD). While the heating efficiency was the main point to be optimized in this approach, to form a more peaked pressure profile is another promising approach towards the high beta regime. A higher electron density profile with a steeper pressure gradient has been formed by pellet injection. From the MHD stability analysis, this peaked pressure profile is stable against the ideal MHD modes. By both approaches, the central plasma β 0 reaches about 10%. (author)

  6. Observations of plasma rotation in the high-beta tokamak Torus II

    International Nuclear Information System (INIS)

    Kostek, C.; Marshall, T.C.

    1982-01-01

    Toroidal and poloidal plasma rotation are measured in a high Beta tokamak device by studying the Doppler shift of the 4686 A He II line. The toroidal flow motion is in the same direction as the plasma current at an average velocity of 1.6 x 10 6 cm/sec, a small fraction of the ion thermal speed. The poloidal flow follows the ion diamagnetic direction, also at an average speed of 1.6 x 10 6 cm/sec. In view of certain ordering parameters, the toroidal flow is compared with the predictions of neoclassical transport theory in the collisional regime. For the poloidal motion, however, it appears that an (E/sub r/ x B)/B 2 drift in a positive radial electric field, approaching a stable ambipolar state (STRINGER, 1970) is responsible. Mechanisms for the time evolution of the rotation are also examined. The radial electric field responsible for the (E/sub r/ x B)/B 2 drift is determined from the theory using the measured poloidal velocity

  7. Extended steady-state and high-beta regimes of net-current free heliotron plasmas in the Large Helical Device

    International Nuclear Information System (INIS)

    Motojima, O.; Yamada, H.; Komori, A.; Ohyabu, N.; Mutoh, T.; Kaneko, O.; Kawahata, K.; Mito, T.; Ida, K.; Imagawa, S.; Nagayama, Y.; Shimozuma, T.; Watanabe, K.Y.; Masuzaki, S.; Miyazawa, J.; Morisaki, T.; Morita, S.; Ohdachi, S.; Ohno, N.; Saito, K.; Sakakibara, S.; Takeiri, Y.; Tamura, N.; Toi, K.; Tokitani, M.; Yokoyama, M.; Yoshinuma, M.; Ikeda, K.; Isayama, A.; Ishii, K.; Kubo, S.; Murakami, S.; Nagasaki, K.; Seki, T.; Takahata, K.; Takenaga, H.

    2007-01-01

    The performance of net-current free heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fuelling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an internal diffusion barrier (IDB) by a combination of efficient pumping of the local island divertor function and core fuelling by pellet injection has realized a super dense core as high as 5 x 10 20 m -3 , which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5% and a discharge duration of 54 min with a total input energy of 1.6 GJ (490 kW on average) are also highlighted. The progress of LHD experiments in these two years is overviewed by highlighting IDB, high β and long pulse

  8. Progress towards Steady State at Low Aspect Ratio on the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.; Menard, J.; Maingi, R.; Kaye, S.; Sabbagh, S.A.; Diem, S.; Wilson, J.R.; Bell, M.G.; Bell, R.E.; Ferron, J.; Fredrickson, E.D.; Kessel, C.E.; LeBlanc, B.P.; Levinton, F.; Manickam, J.; Mueller, D.; Raman, R.; Stevenson, T.; Stutman, D.; Taylor, G.; Tritz, K.; Yu, H.

    2007-01-01

    Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S (triple b ond) q 95 (I p /aB t )) of ∼ 41 (MA m -1 T -1 ) simultaneous with a record plasma elongation of κ (triple b ond) b/a ∼ 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as well as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance.

  9. High plasma concentration of beta-D-glucan after administration of sizofiran for cervical cancer

    Directory of Open Access Journals (Sweden)

    Hirokazu Tokuyasu

    2010-09-01

    Full Text Available Hirokazu Tokuyasu1, Kenichi Takeda1, Yuji Kawasaki1, Yasuto Sakaguchi2, Noritaka Isowa2, Eiji Shimizu3, Yasuto Ueda31Divisions of Respiratory Medicine, 2Thoracic Surgery, Matsue Red Cross Hospital, 200 Horomachi, Matsue, Shimane; 3Division of Medical Oncology and Molecular Respirology, Department of Multidisciplinary Internal Medicine, Faculty of Medicine, Tottori University, Yonago, JapanAbstract: A 69-year-old woman with a history of cervical cancer was admitted to our hospital for further investigation of abnormal shadows on her chest roentgenogram. Histologic examination of transbronchial lung biopsy specimens revealed epithelioid cell granuloma, and Mycobacterium intracellulare was detected in the bronchial lavage fluid. The plasma level of (1→3-beta-D-glucan was very high, and this elevated level was attributed to administration of sizofiran for treatment of cervical cancer 18 years previously. Therefore, in patients with cervical cancer, it is important to confirm whether or not sizofiran has been administered before measuring (1→3-beta-D-glucan levels.Keywords: (1→3-beta-D-glucan, cervical cancer, Mycobacterium intracellulare, sizofiran

  10. Plasma beta HCG determination

    International Nuclear Information System (INIS)

    Amaral, L.B.D.; Pinto, J.C.M.; Linhares, E.; Linhares, Estevao

    1981-01-01

    There are three important indications for the early diagnosis of pregnancy through the determination of the beta sub-unit of chorionic gonadotrophin using radioimmunoassay: 1) some patient's or doctor's anxiety to discover the problem; 2) when it will be necessary to employ diagnostic or treatment procedures susceptible to affect the ovum; and 3) in the differential diagnosis of amenorrhoea, uterine hemorrhage and abdominal tumors. Other user's are the diagnosis of missed absortion, and the diagnosis and follow-up of chrorioncarcinoma. The AA. studied 200 determinations of plasma beta-HCG, considering the main difficulties occuring in the clinical use of this relevant laboratory tool in actual Obstetrics. (author) [pt

  11. Simulation of the time development of EBW emission from NSTX

    Czech Academy of Sciences Publication Activity Database

    Preinhaelter, Josef; Urban, Jakub; Taylor, G.; Diem, S.; Vahala, L.; Vahala, G.

    2006-01-01

    Roč. 51, č. 4 (2006), K1.00024 ISSN 0003-0503. [International Sherwood Fusion Theory Conference/2006./. Dallas, Texas , 22.4.2006-25.4.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * MAST * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://www.aps.org/meet/APR06/baps/all_APR06.pdf http://meetings.aps.org/Meeting/APR06/Event/47670

  12. Plasma. beta. -endorphin and stress hormone levels during adaptation and stress

    Energy Technology Data Exchange (ETDEWEB)

    Lishmanov, Yu.B.; Trifonova, Zh.V.; Tsibin, A.N.; Maslova, L.V.; Dement' eva, L.A.

    1987-09-01

    This paper describes a comparative study of ..beta..-endorphin and stress hormone levels in the blood plasma of rats during stress and adaptation. Immunoreactive ..beta..-endorphin in the blood plasma was assayed by means of a kit after preliminary isolation of the ..beta..-endorphin fraction by affinity chromatography on sepharose; ACTH was assayed with a kit and cortisol, insulin, thyroxine and tri-iodothyronine by means of kits from Izotop. Determination of plasma levels of ..beta..-endorphin and other opioids could evidently be an important method of assessing the state of resistance of the organism to stress.

  13. RF Rectification on LAPD and NSTX: the relationship between rectified currents and potentials

    Science.gov (United States)

    Perkins, R. J.; Carter, T.; Caughman, J. B.; van Compernolle, B.; Gekelman, W.; Hosea, J. C.; Jaworski, M. A.; Kramer, G. J.; Lau, C.; Martin, E. H.; Pribyl, P.; Tripathi, S. K. P.; Vincena, S.

    2017-10-01

    RF rectification is a sheath phenomenon important in the fusion community for impurity injection, hot spot formation on plasma-facing components, modifications of the scrape-off layer, and as a far-field sink of wave power. The latter is of particular concern for the National Spherical Torus eXperiment (NSTX), where a substantial fraction of the fast-wave power is lost to the divertor along scrape-off layer field lines. To assess the relationship between rectified currents and rectified voltages, detailed experiments have been performed on the Large Plasma Device (LAPD). An electron current is measured flowing out of the antenna and into the limiters, consistent with RF rectification with a higher RF potential at the antenna. The scaling of this current with RF power will be presented. The limiters are also floated to inhibit this DC current; the impact of this change on plasma-potential and wave-field measurements will be shown. Comparison to data from divertor probes in NSTX will be made. These experiments on a flexible mid-sized experiment will provide insight and guidance into the effects of ICRF on the edge plasma in larger fusion experiments. Funded by the DOE OFES (DE-FC02-07ER54918 and DE-AC02-09CH11466), NSF (NSF- PHY 1036140), and the Univ. of California (12-LR- 237124).

  14. Equilibrium and stability of high-beta plasma in Modified Bumpy Torus (MBT)

    International Nuclear Information System (INIS)

    Todoroki, J.; Shiina, S.; Saito, K.; Osanai, Y.; Nogi, Y.; Gesso, H.; Yagi, I.; Yokoyama, K.; Yoshimura, H.; Nihon Univ., Tokyo. Atomic Energy Research Inst.)

    1977-01-01

    The equilibrium and stability properties of the plasma in Modified Bumpy Torus, which is an asymmetric system with closed magnetic lines of force, is reported. For small beta value, the growth rate of m=1 mode instability in MBT can be smaller than that of Scyllac configuration. The results of 1/4 toroidal sector experiment are reported. (author)

  15. Transition to ELM-free Improved H-mode by Lithium Deposition on NSTX Graphite Divertor Surfaces

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Kugel, H.W.; Maingi, R.; Bell, M.G.; Bell, R.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.; Mueller, D.; Paul, S.; Raman, R.; Roquemore, L.; Sabbagh, S.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.; Timberlake, J.; Wilgen, J.; Zakharov, L.

    2009-01-01

    Lithium evaporated onto plasma facing components in the NSTX lower divertor has made dramatic improvements in discharge performance. As lithium accumulated, plasmas previously exhibiting robust Type 1 ELMs gradually transformed into discharges with intermittent ELMs and finally into continuously evolving ELM-free discharges. During this sequence, other discharge parameters changed in a complicated manner. As the ELMs disappeared, energy confinement improved and remarkable changes in edge and scrape-off layer plasma properties were observed. These results demonstrate that active modification of plasma surface interactions can preempt large ELMs.

  16. Diagnostics for the Biased Electrode Experiment on NSTX

    International Nuclear Information System (INIS)

    Roquemore, A.L.; Zweben, S.J.; Bush, C.E.; Kaita, R.; Marsalsa, R.J.; Maqueda, R.J.

    2009-01-01

    A linear array of four small biased electrodes was installed in NSTX in an attempt to control the width of the scrape-off layer (SOL) by creating a strong local poloidal electric field. The set of electrodes were separated poloidally by a 1 cm gap between electrodes and were located slightly below the midplane of NSTX, 1 cm behind the RF antenna and oriented so that each electrode is facing approximately normal to the magnetic field. Each electrode can be independently biased to ± 100 volts. Present power supplies limit the current on two electrodes to 30 amps the other two to 10 amps each. The effect of local biasing was measured with a set of Langmuir probes placed between the electrodes and another set extending radially outward from the electrodes, and also by the gas puff imaging diagnostic (GPI) located 1 m away along the magnetic field lines intersecting the electrodes. Two fast cameras were also aimed directly at the electrode array. The hardware and controls of the biasing experiment will be presented and the initial effects on local plasma parameters will be discussed

  17. NSTX-U Digital Coil Protection System Software Detailed Design

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-06-01

    The National Spherical Torus Experiment (NSTX) currently uses a collection of analog signal processing solutions for coil protection. Part of the NSTX Upgrade (NSTX-U) entails replacing these analog systems with a software solution running on a conventional computing platform. The new Digital Coil Protection System (DCPS) will replace the old systems entirely, while also providing an extensible framework that allows adding new functionality as desired.

  18. Plasma beta-endorphin levels in obese and non-obese patients with polycystic ovary disease.

    Science.gov (United States)

    Martínez-Guisasola, J; Guerrero, M; Alonso, F; Díaz, F; Cordero, J; Ferrer, J

    2001-02-01

    The aim of this study was to determine the influence of body weight on circulating plasma levels of beta-endorphin and insulin in women with polycystic ovary disease (PCOD), as well as the correlation between the plasma levels of beta-endorphin and insulin. One-hundred and sixty-seven consecutive subjects with PCOD were recruited, 117 of whom had normal weight (body mass index (BMI) 25). A venous blood sample was taken and plasma concentrations of beta-endorphin, insulin, gonadotropins, prolactin, progesterone, 17 beta-estradiol, estrone, androgens, dehydroepiandrosterone sulfate and sex hormone-binding globulin (SHBG) were measured. Mean beta-endorphin and insulin plasma levels were significantly higher (p PCOD women than in non-obese ones. Correlation analysis showed a positive association between insulin and beta-endorphin, beta-endorphin and BMI (and weight), insulin and BMI (and weight), and a negative correlation was found between insulin and SHBG. A weak association was found between beta-endorphin and luteinizing hormone (LH) in peripheral plasma. Stratified and linear regression analysis showed that plasma beta-endorphin concentrations correlate more with BMI than with insulinemia.

  19. The impact of lithium wall coatings on NSTX discharges and the engineering of the Lithium Tokamak eXperiment (LTX)

    International Nuclear Information System (INIS)

    Majeski, R.; Kugel, H.; Kaita, R.; Avasarala, S.; Bell, M.G.; Bell, R.E.; Berzak, L.; Beiersdorfer, P.; Gerhardt, S.P.; Gransted, E.; Gray, T.; Jacobson, C.; Kallman, J.; Kaye, S.; Kozub, T.; LeBlanc, B.P.; Lepson, J.; Lundberg, D.P.; Maingi, R.; Mansfield, D.; Paul, S.F.; Pereverzev, G.V.; Schneider, H.; Soukhanovskii, V.; Strickler, T.; Stotler, D.; Timberlake, J.; Zakharov, L.E.

    2010-01-01

    Recent experiments on the National Spherical Torus eXperiment (NSTX) have shown the benefits of solid lithium coatings on carbon PFC's to diverted plasma performance, in both L- and H-mode confinement regimes. Better particle control, with decreased inductive flux consumption, and increased electron temperature, ion temperature, energy confinement time, and DD neutron rate were observed. Successive increases in lithium coverage resulted in the complete suppression of ELM activity in H-mode discharges. A liquid lithium divertor (LLD), which will employ the porous molybdenum surface developed for the LTX shell, is being installed on NSTX for the 2010 run period, and will provide comparisons between liquid walls in the Lithium Tokamak eXperiment (LTX) and liquid divertor targets in NSTX. LTX, which recently began operations at the Princeton Plasma Physics Laboratory, is the world's first confinement experiment with full liquid metal plasma-facing components (PFCs). All materials and construction techniques in LTX are compatible with liquid lithium. LTX employs an inner, heated, stainless steel-faced liner or shell, which will be lithium-coated. In order to ensure that lithium adheres to the shell, it is designed to operate at up to 500-600 degrees C to promote wetting of the stainless by the lithium, providing the first hot wall in a tokamak to Operate at reactor-relevant temperatures. The engineering of LTX will be discussed.

  20. Mechanical Design of the NSTX Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

    2009-02-19

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  1. Mechanical Design of the NSTX Liquid Lithium Divertor

    International Nuclear Information System (INIS)

    Ellis, R.; Kaita, R.; Kugel, H.; Paluzzi, G.; Viola, M.; Nygren, R.

    2009-01-01

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuum compatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  2. ACTH, corticosterone, and beta-endorphin in rat blood plasma after prolonged immobilization stress

    Energy Technology Data Exchange (ETDEWEB)

    Kiyatkin, E.A.; Amiragova, M.G.; Kushlinskii, N.E.; Polyntsev, Yu. V.

    1986-01-01

    To assess functional relations between changes in ACTH, beta-endorphin (BE), and corticosterone (CS) levels, plasma concentrations of these hormones were studied in rats during the development of prolonged immobilization stress. Plasma hormone concentrations were determined by radioimmunoassay. The results were analyzed by standard statistical methods on a microcomputer. A particular feature about the kit used to determine BE was the presence of 50% cross-reactivity of the antiserum against beta-lipotrophin. To determine CS a highly specific antiserum produced by a laboratory was used.

  3. [High beta tokamak research and plasma theory

    International Nuclear Information System (INIS)

    1990-01-01

    Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report

  4. NSTX Tangential Divertor Camera

    International Nuclear Information System (INIS)

    Roquemore, A.L.; Ted Biewer; Johnson, D.; Zweben, S.J.; Nobuhiro Nishino; Soukhanovskii, V.A.

    2004-01-01

    Strong magnetic field shear around the divertor x-point is numerically predicted to lead to strong spatial asymmetries in turbulence driven particle fluxes. To visualize the turbulence and associated impurity line emission near the lower x-point region, a new tangential observation port has been recently installed on NSTX. A reentrant sapphire window with a moveable in-vessel mirror images the divertor region from the center stack out to R 80 cm and views the x-point for most plasma configurations. A coherent fiber optic bundle transmits the image through a remotely selected filter to a fast camera, for example a 40500 frames/sec Photron CCD camera. A gas puffer located in the lower inboard divertor will localize the turbulence in the region near the x-point. Edge fluid and turbulent codes UEDGE and BOUT will be used to interpret impurity and deuterium emission fluctuation measurements in the divertor

  5. High beta, sawtooth-free tokamak operation using energetic trapped particles

    International Nuclear Information System (INIS)

    White, R.B.; Bussac, M.N.; Romanelli, F.

    1988-08-01

    It is shown that a population of high energy trapped particles, such as that produced by ion cyclotron heating in tokamaks, can result in a plasma completely stable to both sawtooth oscillations and the fishbone mode. The stable window of operation increases in size with plasma temperature and with trapped particle energy, and provides a means of obtaining a stable plasma with high current and high beta. 13 refs., 2 figs

  6. Maternal plasma concentrations of beta-lipotrophin, beta-endorphin and gamma-lipotrophin throughout pregnancy.

    Science.gov (United States)

    Browning, A J; Butt, W R; Lynch, S S; Shakespear, R A

    1983-12-01

    Plasma beta-LPH, beta-EP and gamma-LPH concentrations were measured by radioimmunoassay in 10 pregnant women from 12 weeks gestation until term and in nine women in the early follicular phase of the cycle. There was a progressive and significant rise in the concentration of all three peptides throughout pregnancy and by 32 weeks the concentrations of beta-LPH and beta-EP were greater than the corresponding concentrations in the follicular phase: gamma-LPH was greater than in the follicular phase by the end of pregnancy in those women who were delivered after 40 weeks. The ratio of beta-LPH to gamma-LPH did not change significantly throughout pregnancy, but there was a progressive fall in the beta-LPH/beta-EP ratio. The possible presence of a 'big LPH' to explain this finding is discussed.

  7. Towards identifying the mechanisms underlying field-aligned edge-loss of HHFW power on NSTX

    International Nuclear Information System (INIS)

    Perkins, R. J.; Bell, R. E.; Bertelli, N.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; LeBlanc, B. P.; Kramer, G. J.; Maingi, R.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Scotti, F.; Taylor, G.; Wilson, J. R.; Ahn, J-W.; Gray, T. K.; Green, D. L.; McLean, A.

    2014-01-01

    Fast-wave heating will be a major heating scheme on ITER, as it can heat ions directly and is relatively unaffected by the large machine size unlike neutral beams. However, fast-wave interactions with the plasma edge can lead to deleterious effects such as, in the case of the high-harmonic fast-wave (HHFW) system on NSTX, large losses of fast-wave power in the scrape off layer (SOL) under certain conditions. In such scenarios, a large fraction of the lost HHFW power is deposited on the upper and lower divertors in bright spiral shapes. The responsible mechanism(s) has not yet been identified but may include fast-wave propagation in the scrape off layer, parametric decay instability, and RF currents driven by the antenna reactive fields. Understanding and mitigating these losses is important not only for improving the heating and current-drive on NSTX-Upgrade but also for understanding fast-wave propagation across the SOL in any fast-wave system. This talk summarizes experimental results demonstrating that the flow of lost HHFW power to the divertor regions largely follows the open SOL magnetic field lines. This lost power flux is relatively large close to both the antenna and the last closed flux surface with a reduced level in between, so the loss mechanism cannot be localized to the antenna. At the same time, significant losses also occur along field lines connected to the inboard edge of the bottom antenna plate. The power lost within the spirals is roughly estimated, showing that these field-aligned losses to the divertor are significant but may not account for the total HHFW loss. To elucidate the role of the onset layer for perpendicular fast-wave propagation with regards to fast-wave propagation in the SOL, a cylindrical cold-plasma model is being developed. This model, in addition to advanced RF codes such as TORIC and AORSA, is aimed at identifying the underlying mechanism(s) behind these SOL losses, to minimize their effects in NSTX-U, and to predict

  8. Towards identifying the mechanisms underlying field-aligned edge-loss of HHFW power on NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, R. J.; Bell, R. E.; Bertelli, N.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; LeBlanc, B. P.; Kramer, G. J.; Maingi, R.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Scotti, F.; Taylor, G.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Ahn, J-W.; Gray, T. K.; Green, D. L.; McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); and others

    2014-02-12

    Fast-wave heating will be a major heating scheme on ITER, as it can heat ions directly and is relatively unaffected by the large machine size unlike neutral beams. However, fast-wave interactions with the plasma edge can lead to deleterious effects such as, in the case of the high-harmonic fast-wave (HHFW) system on NSTX, large losses of fast-wave power in the scrape off layer (SOL) under certain conditions. In such scenarios, a large fraction of the lost HHFW power is deposited on the upper and lower divertors in bright spiral shapes. The responsible mechanism(s) has not yet been identified but may include fast-wave propagation in the scrape off layer, parametric decay instability, and RF currents driven by the antenna reactive fields. Understanding and mitigating these losses is important not only for improving the heating and current-drive on NSTX-Upgrade but also for understanding fast-wave propagation across the SOL in any fast-wave system. This talk summarizes experimental results demonstrating that the flow of lost HHFW power to the divertor regions largely follows the open SOL magnetic field lines. This lost power flux is relatively large close to both the antenna and the last closed flux surface with a reduced level in between, so the loss mechanism cannot be localized to the antenna. At the same time, significant losses also occur along field lines connected to the inboard edge of the bottom antenna plate. The power lost within the spirals is roughly estimated, showing that these field-aligned losses to the divertor are significant but may not account for the total HHFW loss. To elucidate the role of the onset layer for perpendicular fast-wave propagation with regards to fast-wave propagation in the SOL, a cylindrical cold-plasma model is being developed. This model, in addition to advanced RF codes such as TORIC and AORSA, is aimed at identifying the underlying mechanism(s) behind these SOL losses, to minimize their effects in NSTX-U, and to predict

  9. National Spherical Torus Experiment Real Time Plasma Control Data Acquisition Hardware

    International Nuclear Information System (INIS)

    R.J. Marsala; J. Schneider

    2002-01-01

    The National Spherical Torus Experiment (NSTX) is currently providing researchers data on low aspect-ratio toroidal plasmas. NSTX's Plasma Control System adjusts the firing angles of thyristor rectifier power supplies, in real time, to control plasma position, shape and density. A Data Acquisition system comprised of off-the-shelf and custom hardware provides the magnetic diagnostics data required in calculating firing angles. This VERSAmodule Eurocard (VME) bus-based system utilizes Front Panel Data Port (FPDP) for high-speed data transfer. Data coming from physically different locations is referenced to several different ground potentials necessitating the need for a custom FPDP multiplexer. This paper discusses the data acquisition system configuration, the in-house designed 4-to-1 FPDP Input Multiplexing Module (FIMM), and future expansion plans

  10. Access to high beta advanced inductive plasmas at low injected torque

    International Nuclear Information System (INIS)

    Solomon, W.M.; Grierson, B.A.; Okabayashi, M.; Politzer, P.A.; Buttery, R.J.; Ferron, J.R.; Garofalo, A.M.; Jackson, G.L.; Kinsey, J.E.; La Haye, R.J.; Luce, T.C.; Petty, C.C.; Welander, A.S.; Holcomb, C.T.; Lanctot, M.J.; Hanson, J.M.; Turco, F.; In, Y.

    2013-01-01

    Recent experiments on DIII-D demonstrate that advanced inductive (AI) discharges with high equivalent normalized fusion gain can be accessed and sustained with very low amounts (∼1 N m) of externally injected torque, a level of torque that is anticipated to drive a similar amount of rotation as the beams on ITER, via simple consideration of the scaling of the moment of inertia and confinement time. The AI regime is typically characterized by high confinement, and high β N , allowing the possibility for high performance, high gain operation at reduced plasma current. Discharges achieved β N ∼ 3.1 with H 98(y,2) ∼ 1 at q 95 ∼ 4, and are sustained for the maximum duration of the counter neutral beams (NBs). In addition, plasmas using zero net NB torque from the startup all the way through to the high β N phase have been created. AI discharges are found to become increasingly susceptible to m/n = 2/1 neoclassical tearing modes as the torque is decreased, which if left unmitigated, generally slow and lock, terminating the high performance phase of the discharge. Access is not notably different whether one ramps the torque down at high β N , or ramps β N up at low torque. The use of electron cyclotron heating (ECH) and current drive proved to be an effective method of avoiding such modes, enabling stable operation at high beta and low torque, a portion of phase space that has otherwise been inaccessible. Thermal confinement is significantly reduced at low rotation, a result that is reproduced using the TGLF transport model. Although it is thought that stiffness is increased in regions of low magnetic shear, in these AI plasmas, the reduced confinement occurs at radii outside the low shear, and in fact, higher temperature gradients can be found in the low shear region at low rotation. Momentum transport is also larger at low rotation, but a significant intrinsic torque is measured that is consistent with a previous scaling considering the role of the

  11. Access to high beta advanced inductive plasmas at low injected torque

    Science.gov (United States)

    Solomon, W. M.; Politzer, P. A.; Buttery, R. J.; Holcomb, C. T.; Ferron, J. R.; Garofalo, A. M.; Grierson, B. A.; Hanson, J. M.; In, Y.; Jackson, G. L.; Kinsey, J. E.; La Haye, R. J.; Lanctot, M. J.; Luce, T. C.; Okabayashi, M.; Petty, C. C.; Turco, F.; Welander, A. S.

    2013-09-01

    Recent experiments on DIII-D demonstrate that advanced inductive (AI) discharges with high equivalent normalized fusion gain can be accessed and sustained with very low amounts (∼1 N m) of externally injected torque, a level of torque that is anticipated to drive a similar amount of rotation as the beams on ITER, via simple consideration of the scaling of the moment of inertia and confinement time. The AI regime is typically characterized by high confinement, and high βN, allowing the possibility for high performance, high gain operation at reduced plasma current. Discharges achieved βN ∼ 3.1 with H98(y,2) ∼ 1 at q95 ∼ 4, and are sustained for the maximum duration of the counter neutral beams (NBs). In addition, plasmas using zero net NB torque from the startup all the way through to the high βN phase have been created. AI discharges are found to become increasingly susceptible to m/n = 2/1 neoclassical tearing modes as the torque is decreased, which if left unmitigated, generally slow and lock, terminating the high performance phase of the discharge. Access is not notably different whether one ramps the torque down at high βN, or ramps βN up at low torque. The use of electron cyclotron heating (ECH) and current drive proved to be an effective method of avoiding such modes, enabling stable operation at high beta and low torque, a portion of phase space that has otherwise been inaccessible. Thermal confinement is significantly reduced at low rotation, a result that is reproduced using the TGLF transport model. Although it is thought that stiffness is increased in regions of low magnetic shear, in these AI plasmas, the reduced confinement occurs at radii outside the low shear, and in fact, higher temperature gradients can be found in the low shear region at low rotation. Momentum transport is also larger at low rotation, but a significant intrinsic torque is measured that is consistent with a previous scaling considering the role of the turbulent

  12. Beta-endorphin and ACTH in plasma during attacks of common and classic migraine

    DEFF Research Database (Denmark)

    Bach, F W; Jensen, K; Blegvad, N

    1985-01-01

    Plasma levels of beta-endorphin and ACTH were measured during and outside migraine attacks in 17 patients with common migraine and 11 patients with classic migraine. Specific radioimmunoassays for beta-endorphin and ACTH were used. The beta-endorphin assay did not cross-react with beta......-lipotropin. In common migraine, median plasma beta-endorphin was 3.3 pmol/l (95% confidence limits: 2.5-4.0 pmol/l) during attacks and 2.9 (2.4-3.2) pmol/l in the headache-free period. In classic migraine, plasma beta-endorphin was 3.2 (1.4-4.3) pmol/l during attacks and 2.4 (1.1-3.6) pmol/l outside attacks. ACTH...

  13. Paramagnetism and plasma beta in a screw-pinch

    International Nuclear Information System (INIS)

    Lehnert, B.; Scheffel, J.

    1991-02-01

    Anisotropic resistivity causes paramagnetic effects (B z ' (r) less then 0) in a screw pinch, being basically different to the self-relaxation described by Taylor. We compute, analytically and numerically, the resulting effect on equilibrium in a 1-D straight cylindrical plasma. In particular we compute paramagnetic effects on the plasma radius and on plasma beta. Ohm's law also contains diamagnetic terms; in this paper we consider radial particle diffusion and the Nernst effect. In a Tokamak or rector plasma these effects are shown to be negligible, whereas they may contribute in present ULQ, Extrap and RFP experiments. A basic result is an expression for the coupling between the poloidal and axial magnetic field components with the above effects included. A result of specific importance to the Extrap programme is that plasma current limitation can arise from lack of equilibrium when the plasma radius tends to exceed its upper limit, being defined by a magnetic or material limiter. The paramagnetic effect described in this work lowers the plasma beta further, making D-D reactor depending on safety factors q(a) bigger than 1 seems less attainable. (au)

  14. Simulation Of Microtearing Turbulence In NSTX

    International Nuclear Information System (INIS)

    Guttenfelder, W.; Candy, J.; Kaye, S.M.; Nevins, W.M.; Wanag, E.; Zhang, J.; Bell, R.E.; Crocker, N.A.; Hammett, G.W.; LeBlanc, B.P.; Mikkelsen, D.R.; Ren, Y.; Yuh, H.

    2012-01-01

    Thermal energy confinement times in NSTX dimensionless parameter scans increase with decreasing collisionality. While ion thermal transport is neoclassical, the source of anomalous electron thermal transport in these discharges remains unclear, leading to considerable uncertainty when extrapolating to future ST devices at much lower collisionality. Linear gyrokinetic simulations find microtearing modes to be unstable in high collisionality discharges. First non-linear gyrokinetic simulations of microtearing turbulence in NSTX show they can yield experimental levels of transport. Magnetic flutter is responsible for almost all the transport (∼98%), perturbed field line trajectories are globally stochastic, and a test particle stochastic transport model agrees to within 25% of the simulated transport. Most significantly, microtearing transport is predicted to increase with electron collisionality, consistent with the observed NSTX confinement scaling. While this suggests microtearing modes may be the source of electron thermal transport, the predictions are also very sensitive to electron temperature gradient, indicating the scaling of the instability threshold is important. In addition, microtearing turbulence is susceptible to suppression via sheared E-B flows as experimental values of E-B shear (comparable to the linear growth rates) dramatically reduce the transport below experimental values. Refinements in numerical resolution and physics model assumptions are expected to minimize the apparent discrepancy. In cases where the predicted transport is strong, calculations suggest that a proposed polarimetry diagnostic may be sensitive to the magnetic perturbations associated with the unique structure of microtearing turbulence.

  15. Effect of lithium PFC coatings on NSTX density control

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Bell, R.; Bush, C.; Gates, D.; Gray, T.; Kaita, R.; Leblanc, B.; Maingi, R.; Majeski, R.; Mansfield, D.; Mueller, D.; Paul, S.; Raman, R.; Roquemore, A.L.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Stevenson, T.; Zakharov, L.

    2007-01-01

    Lithium coatings on the graphite plasma facing components (PFCs) in NSTX are being investigated as a tool for density profile control and reducing the recycling of hydrogen isotopes. Repeated lithium pellet injection into Center Stack Limited and Lower Single Null ohmic helium discharges were used to coat graphite surfaces that had been pre-conditioned with ohmic helium discharges of the same shape to reduce their contribution to hydrogen isotope recycling. The following deuterium NBI reference discharges exhibited a reduction in density by a factor of about 3 for limited and 2 for diverted plasmas, respectively, and peaked density profiles. Recently, a lithium evaporator has been used to apply thin coatings on conditioned and unconditioned PFCs. Effects on the plasma density and the impurities were obtained by pre-conditioning the PFCs with ohmic helium discharges, and performing the first deuterium NBI discharge as soon as possible after applying the lithium coating

  16. Three new extreme ultraviolet spectrometers on NSTX-U for impurity monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Weller, M. E., E-mail: weller4@llnl.gov; Beiersdorfer, P.; Soukhanovskii, V. A.; Magee, E. W.; Scotti, F. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2016-11-15

    Three extreme ultraviolet (EUV) spectrometers have been mounted on the National Spherical Torus Experiment–Upgrade (NSTX-U). All three are flat-field grazing-incidence spectrometers and are dubbed X-ray and Extreme Ultraviolet Spectrometer (XEUS, 8–70 Å), Long-Wavelength Extreme Ultraviolet Spectrometer (LoWEUS, 190–440 Å), and Metal Monitor and Lithium Spectrometer Assembly (MonaLisa, 50–220 Å). XEUS and LoWEUS were previously implemented on NSTX to monitor impurities from low- to high-Z sources and to study impurity transport while MonaLisa is new and provides the system increased spectral coverage. The spectrometers will also be a critical diagnostic on the planned laser blow-off system for NSTX-U, which will be used for impurity edge and core ion transport studies, edge-transport code development, and benchmarking atomic physics codes.

  17. Proceedings of the US-Japan workshop and the satellite meeting of ITC-9 on physics of high beta plasma confinement in innovative fusion system

    International Nuclear Information System (INIS)

    Goto, Seiichi; Yoshimura, Satoru

    1999-04-01

    The US-Japan Workshop on Physics of High Beta Plasma Confinement in Innovative Fusion System was held jointly with the Satellite Meeting of ITC-9 at National Institute for Fusion Science (NIFS), Toki-city during December 14-15, 1998. This proceedings book includes the papers of the talks given at the workshop. These include: Theoretical analysis on the stability of field reversed configuration (FRC) plasmas; Theory and Modeling of high β plasmas; Recent progressive experiments in high β systems; Formation of high β plasmas using merging phenomenon; Theory and Modeling of a FRC Fusion Reactor. The 15 papers are indexed individually. (J.P.N.)

  18. Calculations of Neutral Beam Ion Confinement for the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Redi, M.H.; Darrow, D.S.; Egedal, J.; Kaye, S.M.; White, R.B.

    2002-01-01

    The spherical torus (ST) concept underlies several contemporary plasma physics experiments, in which relatively low magnetic fields, high plasma edge q, and low aspect ratio combine for potentially compact, high beta and high performance fusion reactors. An important issue for the ST is the calculation of energetic ion confinement, as large Larmor radius makes conventional guiding center codes of limited usefulness and efficient plasma heating by RF and neutral beam ion technology requires minimal fast ion losses. The National Spherical Torus Experiment (NSTX) is a medium-sized, low aspect ratio ST, with R=0.85 m, a=0.67 m, R/a=1.26, Ip*1.4 MA, Bt*0.6 T, 5 MW of neutral beam heating and 6 MW of RF heating. 80 keV neutral beam ions at tangency radii of 0.5, 0.6 and 0.7 m are routinely used to achieve plasma betas above 30%. Transport analyses for experiments on NSTX often exhibit a puzzling ion power balance. It will be necessary to have reliable beam ion calculations to distinguish among the source and loss channels, and to explore the possibilities for new physics phenomena, such as the recently proposed compressional Alfven eigenmode ion heating

  19. Infrared Camera Diagnostic for Heat Flux Measurements on NSTX

    International Nuclear Information System (INIS)

    D. Mastrovito; R. Maingi; H.W. Kugel; A.L. Roquemore

    2003-01-01

    An infrared imaging system has been installed on NSTX (National Spherical Torus Experiment) at the Princeton Plasma Physics Laboratory to measure the surface temperatures on the lower divertor and center stack. The imaging system is based on an Indigo Alpha 160 x 128 microbolometer camera with 12 bits/pixel operating in the 7-13 (micro)m range with a 30 Hz frame rate and a dynamic temperature range of 0-700 degrees C. From these data and knowledge of graphite thermal properties, the heat flux is derived with a classic one-dimensional conduction model. Preliminary results of heat flux scaling are reported

  20. Concept of a charged fusion product diagnostic for NSTX.

    Science.gov (United States)

    Boeglin, W U; Valenzuela Perez, R; Darrow, D S

    2010-10-01

    The concept of a new diagnostic for NSTX to determine the time dependent charged fusion product emission profile using an array of semiconductor detectors is presented. The expected time resolution of 1-2 ms should make it possible to study the effect of magnetohydrodynamics and other plasma activities (toroidal Alfvén eigenmodes (TAE), neoclassical tearing modes (NTM), edge localized modes (ELM), etc.) on the radial transport of neutral beam ions. First simulation results of deuterium-deuterium (DD) fusion proton yields for different detector arrangements and methods for inverting the simulated data to obtain the emission profile are discussed.

  1. The study of non-axisymmetric control coil applications in NSTX-U

    Science.gov (United States)

    Park, J.-K.; Menard, J. E.; Kim, K.; Gerhardt, S. P.; Maingi, R.; Bialek, J. M.; Sabbagh, S. A.; Berkery, J. W.; Boozer, A. H.; Canik, J. M.; Evans, T. E.

    2013-10-01

    As expanded 3D field capability is essential to meet NSTX-U programmatic goals and support ITER, non-axisymmetric control coil (NCC) configurations have been proposed and studied to assess potential physics applications. IPEC-NTV, POCA, and TRIP-3D code analysis show that NCC can provide a range of non-resonant error field control while minimizing resonant error field, and enhance NTV variability to better control rotation and shear, and also largely vary stochastic layers in the edge while maintaining similar plasma response characteristics. VALEN-3D analysis shows that RWM control performance increases with NCC and indicates even the possibility of operation near the ideal-wall limit. In addition, 3D analysis using stellarator codes such as COBRA indicates that NCC can directly broaden ballooning unstable region across radius and thus can be used to improve ELM pacing in NSTX-U. Relevant figures-of-merit are defined and used to quantify these NCC physics capabilities, as will be presented with future analysis plans. This work was supported by DOE Contract DE-AC02-09CH11466.

  2. Plasma TGF beta level in rats after hemithoracic irradiation

    NARCIS (Netherlands)

    Vujaskovic, Z; Down, JD; vanWaarde, MAWH; vanAssen, AJ; Szabo, BG; Konings, AWT

    Changes in TGF-beta plasma levels were observed 18 weeks after hemithoracic irradiation in rats. This coincides with an increase in the breathing frequency, being most pronounced between 22 and 28 weeks after irradiation. The correlation suggests a potential role of the circulating TGF-beta in the

  3. Recent EBW Emission Results and Plans for a 350 kW 28 GHz EC/EBW Heating System on NSTX

    Czech Academy of Sciences Publication Activity Database

    Taylor, G.; Diem, S.J.; Ellis, R.A.; Fredd, E.; Greenough, N.I.; Hosea, J.C.; Bigelow, T.S.; Caughman, J.B.; Rasmussen, D.A.; Ryan, P.; Wilgen, J.B.; Harvey, R.W.; Smirnov, A.P.; Ershov, N.M.; Preinhaelter, Josef; Urban, Jakub; Ram, A.K.

    2007-01-01

    Roč. 52, č. 16 (2007), s. 304-304 ISSN 0003-0503. [Annual Meeting of the Division of Plasma Physics/49th./. Orlando, Florida, 12.11.2007-16.11.2007] Institutional research plan: CEZ:AV0Z20430508 Keywords : Conversion * Emission * Tokamaks * Electron Bernstein waves * Simulation * NSTX Subject RIV: BL - Plasma and Gas Discharge Physics http://meetings.aps.org/Meeting/DPP07/Content/901

  4. Characteristics of Energy Transport of Li-conditioned and non-Li-conditioned Plasmas in the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Ding, S.; Kaye, S.M.; Bell, R.E.; Kaita, R.; Kugel, H.; LeBlanc, B.P.; Paul, S.; Wan, B.

    2009-01-01

    The transport properties of NSTX plasmas obtained during the 2008 experimental campaign have been studied and are reported here. Transport trends and dependences have been isolated, and it is found that both electron and ion energy transport coefficients have strong dependences on local values of n(del)T, which in turn is strongly dependent on local current density profile. Without identifying this dependence, it is difficult to identify others, such as the dependence of transport coefficients on B p (or q), I p and P heat . In addition, a comparison between discharges with and without Lithium wall conditioning has been made. While the trends in the two sets of data are similar, the thermal transport loss, especially in the electron channel, is found to strongly depend on the amount of Lithium deposited, decreasing by up to 50% of its no-Lithium value.

  5. Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

    International Nuclear Information System (INIS)

    Kawashima, H.; Sakurai, S.; Shimizu, K.; Takizuka, T.; Tamai, H.; Matsukawa, M.; Fujita, T.; Miura, Y.

    2006-01-01

    The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2 . Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design

  6. Effect of bootstrap current on MHD equilibrium beta limit in heliotron plasmas

    International Nuclear Information System (INIS)

    Watanabe, K.Y.

    2001-01-01

    The effect of bootstrap current on the beta limit of MHD equilibria is studied systematically by an iterative calculation of MHD equilibrium and the consistent bootstrap current in high beta heliotron plasmas. The LHD machine is treated as a standard configuration heliotron with an L=2 planar axis. The effects of vacuum magnetic configurations, pressure profiles and the vertical field control method are studied. The equilibrium beta limit with consistent bootstrap current is quite sensitive to the magnetic axis location for finite beta, compared with the currentless cases. For a vacuum configuration with the magnetic axis shifted inwards in the torus, even in the high beta regimes, the bootstrap current flows to increase the rotational transform, leading to an increase in the equilibrium beta limit. On the contrary, for a vacuum configuration with the magnetic axis shifted outwards in the torus, even in the low beta regimes, the bootstrap current flows so as to reduce the rotational transform; therefore, there is an acceleration of the Shafranov shift increase as beta increases, leading to a decrease in the equilibrium beta limit. The pressure profiles and vertical field control methods influence the equilibrium beta limit through the location of the magnetic axis for finite beta. These characteristics are independent of both device parameters, such as magnetic field strength, and device size in the low collisional regime. (author)

  7. Toroidal asymmetries in divertor impurity influxes in NSTX

    Directory of Open Access Journals (Sweden)

    F. Scotti

    2017-08-01

    Full Text Available Toroidal asymmetries in divertor carbon and lithium influxes were observed in NSTX, due to toroidal differences in surface composition, tile leading edges, externally-applied three-dimensional (3D fields and toroidally-localized edge plasma modifications due to radio frequency heating. Understanding toroidal asymmetries in impurity influxes is critical for the evaluation of total impurity sources, often inferred from measurements with a limited toroidal coverage. The toroidally-asymmetric lithium deposition induced asymmetries in divertor lithium influxes. Enhanced impurity influxes at the leading edge of divertor tiles were the main cause of carbon toroidal asymmetries and were enhanced during edge localized modes. Externally-applied 3D fields led to strike point splitting and helical lobes observed in divertor impurity emission, but marginal changes to the toroidally-averaged impurity influxes. Power coupled to the scrape-off layer SOL plasma during radio frequency (RF heating of H-mode discharges enhanced impurity influxes along the non-axisymmetric divertor footprint of flux tubes connecting to plasma in front of the RF antenna.

  8. OEDGE modeling of outer wall erosion in NSTX and the effect of changes in neutral pressure

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, J.H., E-mail: jnichols@pppl.gov; Jaworski, M.A.; Kaita, R.; Abrams, T.; Skinner, C.H.; Stotler, D.P.

    2015-08-15

    Gross erosion from the outer wall is expected to be a major source of impurities for high power fusion devices due to the low redeposition fraction. Scaling studies of sputtering from the all-carbon outer wall of NSTX are reported. It is found that wall erosion decreases with divertor plasma pressure in low/mid temperature regimes, due to increasing divertor neutral opacity. Wall erosion is found to consistently decrease with reduced recycling coefficient, with outer target recycling providing the largest contribution. Upper and lower bounds are calculated for the increase in wall erosion due to a low-field-side gas puff.

  9. Maternal and cord plasma concentrations of beta-lipotrophin, beta-endorphin and gamma-lipotrophin at delivery; effect of analgesia.

    Science.gov (United States)

    Browning, A J; Butt, W R; Lynch, S S; Shakespear, R A; Crawford, J S

    1983-12-01

    Maternal venous plasma concentrations of beta-LPH, beta-EP and gamma-LPH were compared in (i) patients undergoing vaginal delivery, 11 with an epidural block and 13 with pethidine and nitrous oxide or no analgesics; (ii) patients delivered by caesarean section, 7 under epidural block and 8 under general anaesthesia. Patients delivered by either method under epidural block had significantly lower levels of all three peptides than those receiving no epidural. There were significant negative correlations between umbilical vein beta-LPH, beta-EP and gamma-LPH concentrations and umbilical artery pH and positive correlations between beta-LPH and beta-EP but not gamma-LPH and cord PCO2 in 29 patients. There was no relation between cord levels of any of the three peptides and the method of analgesia or the route of delivery. Although concentrations of all three peptides were closely correlated to one another in either maternal or cord plasma, there was no relationship between maternal and fetal levels.

  10. Explosion of optimal high-beta operation regime by magnetic axis swing in the Large Helical Device

    International Nuclear Information System (INIS)

    Sakakibara, S.; Ohdachi, S.; Watanabe, K.Y.

    2010-11-01

    In Large Helical Device (LHD), the volume averaged beta value dia > as high as 5.1% was achieved in FY2007-2008 experiments. High beta operation regime was explorated by the programmed control of magnetic axis position, which characterizes MHD equilibrium, stability and transport. This control became enable by increasing capability of poloidal coil power supply. The experiments made clear the effect of magnetic hill on MHD activities in high-beta plasmas with more than 4%. Also it enabled to access the ideal stability boundary with keeping high-beta state. The strong m/n=2/1 mode leading minor collapse in core plasma appeared with the inward shift of the magnetic axis. (author)

  11. Control and data acquisition upgrades for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W.M., E-mail: bdavis@pppl.gov; Tchilinguirian, G.J., E-mail: gtchilin@pppl.gov; Carroll, T., E-mail: tcarroll@pppl.gov; Erickson, K.G., E-mail: kerickson@pppl.gov; Gerhardt, S.P., E-mail: sgerhardt@pppl.gov; Henderson, P., E-mail: phenderson@pppl.gov; Kampel, S.H., E-mail: skampel@pppl.gov; Sichta, P., E-mail: psichta@pppl.gov; Zimmer, G.N., E-mail: gzimmer@pppl.gov

    2016-11-15

    Highlights: • The NSTX-U upgrade is nearing completion, and various control and data acquisition upgrades are needed. • The Digital Coil Protection System is a major addition which provides hardware and software to protect the magnetic coils from the complex, increased, stresses added from the upgrade. • The increased computational requirements for the upgrade have largely followed Moore’s Law, and enhancements to the infrastructure and computer hardware should maintain or exceed the previous functionality. • Data requirements for Fast 2-D cameras have exceeded those of “conventional” time-varying signals. There has been a particular emphasis and increase in data from IR cameras. - Abstract: The extensive NSTX Upgrade (NSTX-U) Project includes major components which allow a doubling of the toroidal field strength to 1 T, of the Neutral Beam heating power to 12 MW, and the plasma current to 2 MA, and substantial structural enhancements to withstand the increased electromagnetic loads. The maximum pulse length will go from 1.5 to 5 s. The larger and more complex forces on the coils will be protected by a Digital Coil Protection System, which requires demanding real-time data input rates, calculations and responses. The amount of conventional digitized data for a given pulse is expected to increase from 2.5 to 5 GB per second of pulse. 2-D Fast Camera data is expected to go from 2.5 GB/pulse to 10, and another 2 GB/pulse is expected from new IR cameras. Our network capacity will be increased by a factor of 10, with 10 Gb/s fibers used for the major trunks. 32-core Linux systems will be used for several functions, including between-shot data processing, MDSplus data serving, between-shot EFIT analysis, real-time processing, and for a new capability, between-shot TRANSP. Improvements to the MDSplus events subsystem will be made through the use of both UDP and TCP/IP based methods and the addition of a dedicated “event server”.

  12. Erosion of lithium coatings on TZM molybdenum and graphite during high-flux plasma bombardment

    NARCIS (Netherlands)

    Abrams, T.; Jaworski, M. A.; Kaita, R.; Stotler, D. P.; De Temmerman, G.; Morgan, T. W.; van den Berg, M. A.; van der Meiden, H. J.

    2014-01-01

    Abstract The rate at which Li films will erode under plasma bombardment in the NSTX-U divertor is currently unknown. It is important to characterize this erosion rate so that the coatings can be replenished before they are completely depleted. An empirical formula for the Li erosion rate as a

  13. Development of a Universal Networked Timer at NSTX

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.; Lawson, J.E.; Oliaro, G.; Wertenbaker, J.

    2005-01-01

    A new Timing and Synchronization System component, the Universal Networked Timer (UNT), is under development at the National Spherical Torus Experiment (NSTX). The UNT is a second-generation multifunction timing device that emulates the timing functionality and electrical interfaces originally provided by various CAMAC modules. Using Field Programmable Gate Array (FPGA) technology, each of the UNT's eight channels can be dynamically programmed to emulate a specific CAMAC module type. The timer is compatible with the existing NSTX timing and synchronization system and will also support a (future) clock system with extended performance. To assist system designers and collaborators, software will be written to integrate the UNT with EPICS, MDSplus, and LabVIEW. This paper will describe the timing capabilities, hardware design, programming/software support, and the current status of the Universal Networked Timer at NSTX

  14. ZORNOC: a 1 1/2-D tokamak data analysis code for studying noncircular high beta plasmas

    International Nuclear Information System (INIS)

    Zurro, B.; Wieland, R.M.; Murakami, M.; Swain, D.W.

    1980-03-01

    A new tokamak data analysis code, ZORNOC, was developed to study noncircular, high beta plasmas in the Impurity Study Experiment (ISX-B). These plasmas exhibit significant flux surface shifts and elongation in both ohmically heated and beam-heated discharges. The MHD equilibrium flux surface geometry is determined by solving the Grad-Shafranov equation based on: (1) the shape of the outermost flux surface, deduced from the magnetic loop probes; (2) a pressure profile, deduced by means of Thomson scattering data (electrons), charge exchange data (ions), and a Fokker-Planck model (fast ions); and (3) a safety factor profile, determined from the experimental data using a simple model (Z/sub eff/ = const) that is self-consistently altered while the plasma equilibrium is iterated. For beam-heated discharches the beam deposition profile is determined by means of a Monte Carlo scheme and the slowing down of the fast ions by means of an analytical solution of the Fokker-Planck equation. The code also carries out an electron power balance and calculates various confinement parameters. The code is described and examples of its operation are given

  15. Numerical Study of Instabilities Driven by Energetic Neutral Beam Ions in NSTX

    International Nuclear Information System (INIS)

    Belova, E.V.; Gorelenkov, N.N.; Cheng, C.Z.; Fredrickson, E.D.

    2003-01-01

    Recent experimental observations from NSTX [National Spherical Torus Experiment] suggest that many modes in a subcyclotron frequency range are excited during neutral-beam injection (NBI). These modes have been identified as Compressional Alfven Eigenmodes (CAEs) and Global Alfven Eigenmodes (GAEs), which are driven unstable through the Doppler-shifted cyclotron resonance with the beam ions. The injection velocities of the NBI ions in NSTX are large compared to Alfven velocity, V(sub)0 > 3V(sub)A, and a strong anisotropy in the fast-ion pitch-angle distribution provides the energy source for the instabilities. Recent interest in the excitation of Alfven Eigenmodes in the frequency range omega less than or approximately equal to omega(sub)ci, where omega(sub)ci is the ion cyclotron frequency, is related to the possibility that these modes can provide a mechanism for direct energy transfer from super-Alfvenic beam ions to thermal ions. Numerical simulations are required in order to find a self-consistent mode structure, and to include the effects of finite-Larmor radius (FLR), the nonlinear effects, and the thermal plasma kinetic effects

  16. Integrated plasma control for high performance tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.

    2005-01-01

    Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)

  17. Recent reflectometry results from the UCLA plasma diagnostics group

    International Nuclear Information System (INIS)

    Gilmore, M.; Doyle, E.J.; Kubota, S.; Nguyen, X.V.; Peebles, W.A.; Rhodes, T.L.; Zeng, L.

    2001-01-01

    The UCLA Plasma Diagnostics Group has an active ongoing reflectometry program. The program is threefold, including 1) profile and 2) fluctuation measurements on fusion devices (DIII-D, NSTX, and others), and 3) basic reflectometry studies in linear and laboratory plasmas that seek to develop new measurement capabilities and increase the physics understanding of reflectometry. Recent results on the DIII-D tokamak include progress toward the implementation of FM reflectometry as a standard density profile diagnostic, and correlation length measurements in QDB discharges that indicate a very different scaling than normally observed in L-mode plasmas. The first reflectometry measurements in a spherical torus (ST) have also been obtained on NSTX. Profiles in NSTX show good agreement with those of Thomson scattering. Finally, in a linear device, a local magnetic field strength measurement based on O-X correlation reflectometry has been demonstrated to proof of principle level, and correlation lengths measured by reflectometry are in good agreement with probes. (author)

  18. High beta plasma confinement and neoclassical effects in a small aspect ratio reversed field pinch

    International Nuclear Information System (INIS)

    Hayase, K.; Sugimoto, H.; Ashida, H.

    2003-01-01

    The high β equilibrium and stability of a reversed field pinch (RFP) configuration with a small aspect ratio are theoretically studied. The equilibrium profile, high beta limit and the bootstrap current effect on those are calculated. The Mercier stable critical β decreases with 1/A, but β∼0.2 is permissible at A=2 with help of edge current profile modification. The effect of bootstrap current is evaluated for various pressure and current profiles and cross-sectional shapes of plasma by a self-consistent neoclassical PRSM equilibrium formulation. The high bootstrap current fraction (F bs ) increases the shear stabilization effect in the core region, which enhances significantly the stability β limit compared with that for the classical equilibrium. These features of small aspect ratio RFP, high β and high F bs , and a possibly easier access to the quasi-single helicity state beside the intrinsic compact structure are attractive for the feasible economical RFP reactor concept. (author)

  19. EBW-Bootstrap Current Synergy in the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Harvey, R.W.; Taylor, G.

    2005-01-01

    Current driven by electron Bernstein waves (EBW) and by the electron bootstrap effect are calculated separately and concurrently with a kinetic code, to determine the degree of synergy between them. A target β = 40% NSTX plasma is examined. A simple bootstrap model in the CQL3D Fokker-Planck code is used in these studies: the transiting electron distributions are connected in velocity-space at the trapped-passing boundary to trapped-electron distributions which are displaced radially by a half-banana width outwards/inwards for the co-/counter-passing regions. This model agrees well with standard bootstrap current calculations, over the outer 60% of the plasma radius. Relatively small synergy net bootstrap current is obtained for EBW power up to 4 MW. Locally, bootstrap current density increases in proportion to increased plasma pressure, and this effect can significantly affect the radial profile of driven current

  20. On the ordinary mode instability for low beta plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Hadi, F.; Qamar, A. [Institute of Physics and Electronics, University of Peshawar, Peshawar (Pakistan); Bashir, M. F. [Department of Physics, G. C. University, Lahore (Pakistan); Salam Chair in Physics, G. C. University, Lahore (Pakistan); Yoon, P. H. [Institute for Physical Science and Technology, University of Maryland, College Park, Maryland 20742-2431 (United States); School of Space Research, Kyung Hee University, Yongin (Korea, Republic of); Schlickeiser, R. [Institut für Theoretische Physik, Lehrstuhl IV: Weltraum- and Astrophysik, Ruhr-Universität, Bochum (Germany)

    2014-05-15

    The purely growing ordinary (O) mode instability, first discussed by Davidson and Wu [Phys. Fluids 13, 1407 (1970)], has recently received renewed attention owing to its potential applicability to the solar wind plasma. In a series of papers, Ibscher, Schlickeiser, and their colleagues [Phys. Plasmas 19, 072116 (2012); ibid. 20, 012103 (2013); ibid. 20, 042121 (2013); ibid. 21, 022110 (2014)] revisited the O mode instability and extended it to the low-beta plasma regime by considering a counter-streaming bi-Maxwellian model. However, the O-mode instability is, thus, far discussed only on the basis of the marginal stability condition rather than actual numerical solutions of the dispersion relation. The present paper revisits the O-mode instability by considering the actual complex roots. The marginal stability condition as a function of the (electron) temperature anisotropy and beta naturally emerges in such a scheme.

  1. Beta-lipotropin is the major component of the plasma opioid response to surgical stress in humans

    Energy Technology Data Exchange (ETDEWEB)

    Porro, C.A.; Facchinetti, F.; Bertellini, E.; Petraglia, F.; Stacca, R.; Barbieri, G.C.; Genazzani, A.R.

    1987-12-07

    There is growing experimental evidence that beta-endorphin immunoreactivity is raised by surgical stress in patients undergoing general anesthesia. As the assay methods employed to date did not allow to fully discriminate between beta-endorphin and its immediate precursor, beta-lipotropin, the authors have investigated in the present study plasma levels of these two peptides by separating them by chromatography on plasma extracts prior to radioimmunoassay. Beta-lipotropin, but not beta-endorphin, plasma levels were found to be significantly elevated during surgery in the general anesthesia group, while no change was found in either peptide concentration in the spinal one. Cortisol plasma levels also increased significantly 90 minutes after the beginning of surgery. Although the sampling time they adopted may have prevented them from detecting an early peak of beta-endorphin during the first 30 minutes of surgery, the major component of the pituitary opioid response to surgical stress appears to be related to beta-lipotropin. This is in agreement with results of experimental work on various kinds of stress in animals and humans and seems to rule out a role for plasma beta-endorphin in post-operative analgesia. 38 references, 1 figure, 1 table.

  2. Beta II plasma-gun mechanical design and construction

    International Nuclear Information System (INIS)

    Pedrotti, L.; Deis, G.; Wong, R.; Calderon, M.; Chargin, A.; Garner, D.

    1979-01-01

    The magnetized coaxial plasma gun (located on the east end of the Beta II facility at the Lawrence Livermore Laboratory) will be used to test a new method of initiating a field reversed mirror plasma. The field-reversed mirror is expected to improve the mirror-fusion reactor by enhancing the ratio of fusion power to injected power. This paper concentrates on the mechanical design and construction of the magnetized coaxial plasma gun and also discusses the diagnostic devices necessary to demonstrate the formation of field-reversed rings

  3. Global kink and ballooning modes in high-beta systems and stability of toroidal drift modes

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Goedbloed, J.P.; Rem, J.; Sakanaka, P.H.; Schep, T.J.; Venema, M.

    1983-01-01

    A numerical code (HBT) has been developed which solves for the equilibrium, global stability and high-n stability of plasmas with arbitrary cross-section. Various plasmas are analysed for their stability to these modes in the high-beta limit. Screw-pinch equilibria are stable to high-n ballooning modes up to betas of 18%. The eigenmode equation for drift waves is analysed numerically. The toroidal branch is shown to be destabilized by the non-adiabatic response of trapped and circulating particles. (author)

  4. Plasma confinement of Nagoya high-beta toroidal-pinch experiments

    International Nuclear Information System (INIS)

    Hirano, K.; Kitagawa, S.; Wakatani, M.; Kita, Y.; Yamada, S.; Yamaguchi, S.; Sato, K.; Aizawa, T.; Osanai, Y.; Noda, N.

    1977-01-01

    Two different types of high-β toroidal pinch experiments, STP [1] and CCT [2,3], have been done to study the confinement of the plasma produced by a theta-pinch. The STP is an axisymmetric toroidal pinch of high-β tokamak type, while the CCT consists of multiply connected periodic toroidal traps. Internal current-carrying copper rings are essential to the CCT. Since both apparatuses use the same fast capacitor bank system, they produce rather similar plasma temperatures and densities. The observed laser scattering temperature and density is about 50 eV and 4x10 15 cm -3 , respectively, when the filling pressure is 5 mtorr. In the STP experiment, strong correlations are found between the βsub(p) value and the amplitude of m=2 mode. It has a minimum around the value of βsub(p) of 0.8. The disruptive instability is observed to expand the pinched plasma column without lowering the plasma temperature. Just before the disruption begins, the q value around the magnetic axis becomes far less than 1 and an increase of the amplitude of m=2 mode is seen. The CCT also shows rapid plasma expansion just before the magnetic field reaches its maximum. Then the trap is filled up with the plasma by this irreversible expansion and stable plasma confinement is achieved. The energy confinement time of the CCT is found to be about 35 μs. (author)

  5. Status and Plans for NSTX-U Recovery

    Science.gov (United States)

    Hawryluk, R. J.; Gerhardt, S.; Menard, J.; Neumeyer, C.

    2017-10-01

    The NSTX-U device experienced a series of technical problems; the most recent of which was the failure of one of the poloidal magnetic field coils, which has rendered the device inoperable and in need of significant repair. As a result of these incidents, the Laboratory performed a very comprehensive analysis of all of the systems on NSTX-U. Through an integrated system's analysis approach, this process identified which actions need to be taken to form a corrective action plan to ensure reliable and predictable operation. The actions required to address the deficiencies were reviewed by external experts who made recommendations on four high-level programmatic decisions regarding the inner poloidal field coils, limitations to the required bakeout temperature needed for conditioning of the vacuum vessel, divertor and wall protection tiles and coaxial helicity injection. The plans for addressing the recommendations from the external review panels will be presented. This research was sponsored by the U.S. Dept. of Energy under contract DE-AC02-09CH11466.

  6. Sustainment of high confinement in JT-60U reversed shear plasmas

    International Nuclear Information System (INIS)

    Fujita, T.; Kamada, Y.; Ide, S.; Takeji, S.; Sakamoto, Y.; Isayama, A.; Suzuki, T.; Oikawa, T.; Fukuda, T.

    2001-01-01

    confinement is achieved owing to strong internal transport barriers (ITBs), are reported. In a high current plasma with an L-mode edge, deuterium-tritium-equivalent fusion power gain, Q DT eq =0.5 was sustained for 0.8 s (∼ energy confinement time) by adjusting plasma beta precisely using feedback control of stored energy. In a high triangularity plasma with an ELMy H-mode edge, the shrinkage of reversed shear region was suppressed and quasi steady sustainment of high confinement was achieved by raising the poloidal beta and enhancing the bootstrap current peaked at the ITB layer. High bootstrap current fraction (∼80%) was obtained in a high q regime (q 95 ∼9), which leaded to full non-inductive current drive condition. The normalized beta (β N ) of ∼ 2 and H-factor of H 89 ∼3.5 (HH 98y2 ∼2.2) were sustained for 2.7 s (∼ 6 times energy confinement time). (author)

  7. Initial results from the NSTX Real-Time Velocity diagnostic

    Science.gov (United States)

    Podesta, M.; Bell, R. E.

    2011-10-01

    A new diagnostic for fast measurements of plasma rotation through active charge-exchange recombination spectroscopy (CHERS) was installed on NSTX. The diagnostic infers toroidal rotation from carbon ions undergoing charge-exchange with neutrals from a heating Neutral Beam (NB). Each of the 4 channels, distributed along the outer major radius, includes active views intercepting the NB and background views missing the beam. Estimated uncertainties in the measured velocity are system. Signals are acquired on 2 CCD detectors, each controlled by a dedicated PC. Spectra are fitted in real-time through a C++ processing code and velocities are made available to the Plasma Control System for future implementation of feedback on velocity. Results from the initial operation during the 2011 run are discussed, emphasizing the fast dynamics of toroidal rotation, e . g . during L-H mode transition and breaking caused by instabilities and by externally-imposed magnetic perturbations. Work supported by USDOE Contract No. DE-AC02-09CH11466.

  8. Estimation of boswellic acids from market formulations of Boswellia serrata extract and 11-keto beta-boswellic acid in human plasma by high-performance thin-layer chromatography.

    Science.gov (United States)

    Shah, Shailesh A; Rathod, Ishwarsinh S; Suhagia, Bhanubhai N; Patel, Dharmesh A; Parmar, Vijay K; Shah, Bharat K; Vaishnavi, Vikas M

    2007-04-01

    A rapid and sensitive high-performance thin-layer chromatographic (HPTLC) method was developed and validated for the quantitative estimation of boswellic acids in formulation containing Boswellia serrata extract (BSE) and 11-keto beta-boswellic acid in human plasma. Simple extraction method was used for isolation of boswellic acid from formulation sample and acidified plasma sample. The isolated samples were chromatographed on silica gel 60F(254)-TLC plates, developed using ternary-solvent system (hexane-chloroform-methanol, 5:5:0.5, v/v) and scanned at 260 nm. The linearity range for 11-KBA spiked in 1 ml of plasma was 29.15-145.75 ng with average recovery of 91.66%. The limit of detection and limit of quantification for 11-KBA in human plasma were found to be 8.75 ng/ml and 29.15 ng/ml. The developed method was successfully applied for the assay of market formulations containing BSE and to determine plasma level of 11-keto beta-boswellic acid in a clinical pilot study.

  9. The effect of plasma beta on high-n ballooning stability at low magnetic shear

    Science.gov (United States)

    Connor, J. W.; Ham, C. J.; Hastie, R. J.

    2016-08-01

    An explanation of the observed improvement in H-mode pedestal characteristics with increasing core plasma pressure or poloidal beta, {β\\text{pol}} , as observed in MAST and JET, is sought in terms of the impact of the Shafranov shift, {{Δ }\\prime} , on ideal ballooning MHD stability. To illustrate this succinctly, a self-consistent treatment of the low magnetic shear region of the ‘s-α ’ stability diagram is presented using the large aspect ratio Shafranov equilibrium, but enhancing both α and {{Δ }\\prime} so that they compete with each other. The method of averaging, valid at low s, is used to simplify the calculation and demonstrates how α , {{Δ }\\prime} , plasma shaping and ‘average favourable curvature’ all contribute to stability.

  10. Fast Soft X-ray Images of MHD Phenomena in NSTX

    International Nuclear Information System (INIS)

    Bush, C.E.; Stratton, B.C.; Robinson, J.; Zakharov, L.E.; Fredrickson, E.D.; Stutman, D.; Tritz, K.

    2008-01-01

    A variety of magnetohydrodynamic (MHD) phenomena have been observed on the National Spherical Torus Experiment (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 PIN diode arrays for soft x-ray emission from the plasma core. Data reported here are from an 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 CCD 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 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 PIN diode array results.

  11. Plasma Shape Control on the National Spherical Torus Experiment using Real-time Equilibrium Reconstruction

    International Nuclear Information System (INIS)

    Gates, D.A.; Ferron, J.R.; Bell, M.; Gibney, T.; Johnson, R.; Marsala, R.J.; Mastrovito, D.; Menard, J.E.; Mueller, D.; Penaflor, B.; Sabbagh, S.A.; Stevenson, T.

    2005-01-01

    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

  12. Beta-induced Alfven-acoustic Eigenmodes in NSTX and DIII-D Driven by Beam Ions

    International Nuclear Information System (INIS)

    Gorelenkov, N.N.; Van Zeeland, M.A.; Berk, H.L.; Crocker, N.A.; Darrow, D.; Fredrickson, E.; Fu, G.-Y.; Heidbrink, W.W.; Menard, J.; Nazikian, R.

    2009-01-01

    Kinetic theory and experimental observations of a special class of energetic particle driven instabilities called here Beta-induced Alfven-Acoustic Eigenmodes (BAAE) are reported confirming previous results [N.N. Gorelenkov H.L. Berk, N.A. Crocker et. al. Plasma Phys. Control. Fusion 49 B371 (2007)] The kinetic theory is based on the ballooning dispersion relation where the drift frequency effects are retained. BAAE gaps are recovered in kinetic theory. It is shown that the observed certain low-frequency instabilities on DIII-D [J.L. Luxon, Nucl. Fusion 42 614 (2002)] and National Spherical Torus Experiment [M. Ono, S.M. Kaye, Y.-K M. Peng et. al., Nucl. Fusion 40 3Y 557 (2000)] are consistent with their identification as BAAEs. BAAEs deteriorated the fast ion confinement in DIII-D and can have a similar effect in next-step fusion plasmas, especially if excited together with multiple global Toroidicity-induced shear Alfven Eigenmode (TAE) instabilities. BAAEs can also be used to diagnose safety factor profiles, a technique known as magnetohydrodynamic spectroscopy

  13. Beta-induced Alfven-acousti Eigenmodes in NSTX and DIII-D Driven by Beam Ions

    Energy Technology Data Exchange (ETDEWEB)

    Gorelenkov, N. N.; Van Zeeland, M. A.; Berk, H. L.; Crocker, N. A.; Darrow, D.; Fredrickson, E.; Fu, G. Y.; Heidbrink, W. W.; Menard, J.; Nazikian, R.

    2009-03-06

    Kinetic theory and experimental observations of a special class of energetic particle driven instabilities called here Beta-induced Alfven-Acoustic Eigenmodes (BAAE) are reported confirming previous results [N.N. Gorelenkov H.L. Berk, N.A. Crocker et. al. Plasma Phys. Control. Fusion 49 B371 (2007)] The kinetic theory is based on the ballooning dispersion relation where the drift frequency effects are retained. BAAE gaps are recovered in kinetic theory. It is shown that the observed certain low-frequency instabilities on DIII-D [J.L. Luxon, Nucl. Fusion 42 614 (2002)] and National Spherical Torus Experiment [M. Ono, S.M. Kaye, Y.-K M. Peng et. al., Nucl. Fusion 40 3Y 557 (2000)] are consistent with their identification as BAAEs. BAAEs deteriorated the fast ion confinement in DIII-D and can have a similar effect in next-step fusion plasmas, especially if excited together with multiple global Toroidicity-induced shear Alfven Eigenmode (TAE) instabilities. BAAEs can also be used to diagnose safety factor profiles, a technique known as magnetohydrodynamic spectroscopy.

  14. Startup of the experimental physics industrial control system at NSTX

    International Nuclear Information System (INIS)

    Sichta, P.; Dong, J.

    1999-01-01

    The Experimental Physics Industrial Control System (EPICS) is a set of software which is being used as the basis of the National Spherical Torus Experiment's (NSTX) Process Control System, a major element of the NSTX's Central Instrumentation and Control System. EPICS is a result of a co-development effort started by several US Department of Energy National Laboratories. EPICS is actively supported through an international collaboration made up of government and industrial users. EPICS' good points include portability, scalability, and extensibility. A drawback for small experiments is that a wide range of software skills are necessary to get the software tools running for the process engineers. The authors' experience in designing, developing, operating, and maintaining NSTX's EPICS (system) will be reviewed

  15. Preliminary design of experiment ALVAND 1 and low beta plasma source

    International Nuclear Information System (INIS)

    Anvari, A.; Azodi, H.; Naraghi, M.; Taherzadeh, M.; Torabi-Fard, A.

    1975-12-01

    The ultimate goal of a two years program, namely ''ALVAND PROJECT'' and problems concerned with the achievement of controlled thermonuclear reactions in Iran is covered. The report consists of six sections. The introduction deals with the advantages of a linear theta pinch and its comparison with toroidal and mirror devices. Attention has been paid to two important properties of the ALVAND 1 device, namely, its property of carrying out important missions in support of high beta controlled thermonuclear research and also its acting as a source for developing different diagnostic apparatus and excercising different measurement techniques. Expected plasma parameters for ALVAND 1 are given in the third section. In the fourth section the low beta plasma source and the important diagnostic techniques that may be developed are discussed. References are given in the fifth section. The calculation of the minimum required radius of a toroidal theta pinch reactor has been given in an appendix, which covers at the same time 2 schematic figures. Finally a program schedule is presented in section six

  16. Development of slow and fast wave coupling and heating from the C-Stellarator to NSTX

    Directory of Open Access Journals (Sweden)

    Hosea Joel

    2017-01-01

    Full Text Available A historical perspective on key discoveries which contributed to understanding the properties of coupling both slow and fast waves and the effects on plasma heating and current drive will be presented. Important steps made include the demonstration that the Alfven resonance was in fact a mode conversion on the C-stellarator, that toroidal m = -1 eigenmodes were excited in toroidal geometry and impurity influx caused the Z mode on the ST tokamak, that the H minority regime provided strong heating and that 3He minority could be used as well on PLT, that the 2nd harmonic majority tritium regime was viable on TFTR, and that high harmonic fast wave heating was efficient when the SOL losses were avoided on NSTX.

  17. Different pituitary. beta. -endorphin and adrenal cortisol response to ethanol in individuals with high and low risk for future development of alcoholism

    Energy Technology Data Exchange (ETDEWEB)

    Gianoulakis, C.G.; Beliveau, D.; Angelogianni, P.; Meaney, M.; Thavundayil, J.; Tawar, V.; Dumas, M. (McGill Univ., Quebec (Canada))

    1989-01-01

    The purpose of the present studies was to investigate the activity of the adrenal gland and the pituitary {beta}-endorphin system in individuals from families with a 3 generation history of alcoholism, High Risk group, or from families without history of alcoholism, Low Risk group. On the day of testing, blood sample was taken at 9:00 a.m., then the subject drank a placebo drink or an ethanol solution. Additional blood samples were taken at 15, 45 and 120 minutes post-drink. Results indicated that individuals of the High Risk group had lower basal levels of {beta}-endorphin like immunoreactivity ({beta}-EPLIR) than individuals of the Low Risk group. The dose of 0.5 g ethanol/kg B.Wt. induced an induce an increase in the plasma content of {beta}-EPLIR of the High Risk group, but not of the Low Risk group. In the Low Risk group ethanol did not induce an increase above the 9:00 a.m. levels, however, it attenuated the {beta}-endorphin decrease overtime, observed following the placebo drink. Analysis of {beta}-endorphin-like peptides in the plasma of the High Risk group, with Sephadex G-75 chromatography indicated that the major component of the plasma {beta}-EPLIR was {beta}-lipotropin. Plasma cortisol levels, following ethanol intake, presented a small increase in the High Risk group but not in the Low Risk group.

  18. Status of Far Infrared Tangential Interferometry/Polarimetry (FIReTIP) on NSTX

    International Nuclear Information System (INIS)

    Park, H.K.; Edwards, S.; Guttadora, L.; Deng, B.; Domier, C.W.; Lee, K.C.; Johnson, M.; Luhmann, N.C. Jr.

    2000-01-01

    The Influence of paramagnetism and diamagnetism will significantly alter the vacuum toroidal magnetic field in the spherical torus. Therefore, plasma parameters dependent upon BT such as the q-profile and the local b value need an independent measurement of BT(r,t). The multi-chord Tangential Far Infrared Interferometer/Polarimeter (FIReTIP) system [1] currently under development for the National Spherical Torus Experiment (NSTX) will provide temporally and radially resolved toroidal field profile [BT(r,t)] and 2-D electron density profile [ne(r,t)] data. A two-channel interferometer will be operational this year and the full system will be ready by 2002

  19. Progress toward fully noninductive, high beta conditions in DIII-D

    International Nuclear Information System (INIS)

    Murakami, M.; Wade, M.R.; Greenfield, C.M.; Luce, T.C.; Ferron, J.R.; St John, H.E.; DeBoo, J.C.; Osborne, T.H.; Petty, C.C.; Politzer, P.A.; Burrell, K.H.; Gohil, P.; Gorelov, I.A.; Groebner, R.J.; Hyatt, A.W.; Kajiwara, K.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lohr, J.

    2006-01-01

    The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, β T =3.6%, normalized beta, β N =3.5, and confinement factor, H 89 =2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1 s or half of a current relaxation time (τ CR ). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2 s duration (1τ CR ) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have

  20. Ideal magnetohydrodynamic simulations of unmagnetized dense plasma jet injection into a hot strongly magnetized plasma

    OpenAIRE

    Liu, Wei; Hsu, Scott C.

    2010-01-01

    We present results from three-dimensional ideal magnetohydrodynamic simulations of unmagnetized dense plasma jet injection into a uniform hot strongly magnetized plasma, with the aim of providing insight into core fueling of a tokamak with parameters relevant for ITER and NSTX (National Spherical Torus Experiment). Unmagnetized dense plasma jet injection is similar to compact toroid injection but with much higher plasma density and total mass, and consequently lower required injection velocit...

  1. Diagnostic Development on NSTX

    International Nuclear Information System (INIS)

    A.L. Roquemore; D. Johnson; R. Kaita; et al

    1999-01-01

    Diagnostics are described which are currently installed or under active development for the newly commissioned NSTX device. The low aspect ratio (R/a less than or equal to 1.3) and low toroidal field (0.1-0.3T) used in this device dictate adaptations in many standard diagnostic techniques. Technical summaries of each diagnostic are given, and adaptations, where significant, are highlighted

  2. Early incorporation of cell-derived cholesterol into pre-beta-migrating high-density lipoprotein

    International Nuclear Information System (INIS)

    Castro, G.R.; Fielding, C.J.

    1988-01-01

    Cultures of human skin fibroblasts were labeled to high cholesterol specific activity with [ 3 H]cholesterol and incubated briefly (1-3 min) with normal human plasma. The plasma was fractionated by two-dimensional agarose-polyacrylamide gel electrophoresis and the early appearance of cholesterol label among plasma lipoproteins determined. A major part of the label at 1-min incubation was in a pre-beta-migrating apo A-I lipoprotein fraction with a molecular weight of ca. 70,000. Label was enriched about 30-fold in this fraction relative to its content of apo A-I (1-2% of total apo A-I). The proportion of label in this lipoprotein was strongly correlated with its concentration in plasma. Further incubation (2 min) in the presence of unlabeled cells demonstrated transfer of label from this fraction to a higher molecular weight pre-beta apo A-I species, to low-density lipoprotein, and to the alpha-migrating apo A-I that made up the bulk (96%) of total apo A-I in plasma. The data suggest that a significant part of cell-derived cholesterol is transferred specifically to a pre-beta-migrating lipoprotein A-I species as part of a cholesterol transport transfer sequence in plasma

  3. Center for the Study of Plasma Microturbulence

    International Nuclear Information System (INIS)

    Parker, Scott E.

    2012-01-01

    We have discovered a possible 'natural fueling' mechanism in tokamak fusion reactors using large scale gyrokinetic turbulence simulation. In the presence of a heat flux dominated tokamak plasma, cold ions naturally pinch radially inward. If cold DT fuel is introduced near the edge using shallow pellet injection, the cold fuel will pinch inward, at the expense of hot helium ash going radially outward. By adjusting the cold DT fuel concentration, the core DT density profiles can be maintained. We have also shown that cold source ions from edge recycling of cold neutrals are pinched radially inward. This mechanism may be important for fully understanding the edge pedestal buildup after an ELM crash. Work includes benchmarking the gyrokinetic turbulence codes in the electromagnetic regime. This includes cyclone base case parameters with an increasing plasma beta. The code comparisons include GEM, GYRO and GENE. There is good linear agreement between the codes using the Cyclone base case, but including electromagnetics and scanning the plasma beta. All the codes have difficulty achieving nonlinear saturation as the kinetic ballooning limit is approached. GEM does not saturate well when beta gets above about 1/2 of the ideal ballooning limit. We find that the lack of saturation is due to the long wavelength k y modes being nonlinearly pumped to high levels. If the fundamental k y mode is zeroed out, higher values of beta nonlinearly saturate well. Additionally, there have been studies to better understand CTEM nonlinear saturation and the importance of zonal flows. We have continued our investigation of trapped electron mode (TEM) turbulence. More recently, we have focused on the nonlinear saturation of TEM turbulence. An important feature of TEM is that in many parameter regimes, the zonal flow is unimportant. We find that when zonal flows are unimportant, zonal density is the dominant saturation mechanism. We developed a simple theory that agrees with the simulation

  4. Beta-endorphin and ACTH in plasma during attacks of common and classic migraine

    DEFF Research Database (Denmark)

    Bach, F W; Jensen, K; Blegvad, N

    1985-01-01

    Plasma levels of beta-endorphin and ACTH were measured during and outside migraine attacks in 17 patients with common migraine and 11 patients with classic migraine. Specific radioimmunoassays for beta-endorphin and ACTH were used. The beta-endorphin assay did not cross-react with beta-lipotropin...... migraine. Accordingly, we could not add evidence to the theory of a dysfunction of the endogenous opioid system in migraine....

  5. Approximate model for toroidal force balance in the high-beta stellarator

    International Nuclear Information System (INIS)

    Barnes, D.C.

    1979-03-01

    A simple model for estimating the body force acting on a diffuse plasma confined in a three-dimensional, high-beta stellarator geometry is given. The equilibrium is treated by an asymptotic expansion about a straight theta pinch with diffuse, circular cross section. The expansion parameter delta is the strength of the applied helical fields. This expansion leads to an inconsistent set of equations for the equilibrium in second order. Nevertheless, by averaging the equilibrium equations over the volume of the confined plasma, a unique condition for toroidal equilibrium is obtained. When the results are compared with the predictions of previous equilibrium theory, which is based on the sharp-boundary model, a large deviation is found. This correction is especially large for l = 0,1 systems at high beta and must be accounted for in any confinement experiment

  6. Preliminary design of a tangentially viewing imaging bolometer for NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, B. J., E-mail: peterson@LHD.nifs.ac.jp; Mukai, K. [National Institute for Fusion Science, Toki 509-5292 (Japan); SOKENDAI (The Graduate University for Advance Studies), Toki 509-5292 (Japan); Sano, R. [National Institutes for Quantum and Radiological Science and Technology, Naka, Ibaraki 311-0193 (Japan); Reinke, M. L.; Canik, J. M.; Lore, J. D.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Delgado-Aparicio, L. F.; Jaworski, M. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Eden, G. G. van [FOM Institute DIFFER, 5612 AJ Eindhoven (Netherlands)

    2016-11-15

    The infrared imaging video bolometer (IRVB) measures plasma radiated power images using a thin metal foil. Two different designs with a tangential view of NSTX-U are made assuming a 640 × 480 (1280 × 1024) pixel, 30 (105) fps, 50 (20) mK, IR camera imaging the 9 cm × 9 cm × 2 μm Pt foil. The foil is divided into 40 × 40 (64 × 64) IRVB channels. This gives a spatial resolution of 3.4 (2.2) cm on the machine mid-plane. The noise equivalent power density of the IRVB is given as 113 (46) μW/cm{sup 2} for a time resolution of 33 (20) ms. Synthetic images derived from Scrape Off Layer Plasma Simulation data using the IRVB geometry show peak signal levels ranging from ∼0.8 to ∼80 (∼0.36 to ∼26) mW/cm{sup 2}.

  7. SOLPS simulations of X-divertor in NSTX-U

    Science.gov (United States)

    Chen, Zhongping; Kotschenreuther, Mike; Mahajan, Swadesh

    2017-10-01

    The X-divertor (XD) geometry in NSTX-U has demonstrated, in SOLPS simulations, a better performance than the standard divertor (SD) regarding detachment: achieving detachment with a lower upstream density and stabilizing the detachment front near the target. The benefits of such a localized front is that the power exhaust requirement can be satisfied without the radiation front encroaching on the core plasma. It is also found by our simulations that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures. These advantages are attributed to the unique geometric characteristics of XD - poloidal flaring near the target. The detailed physical mechanisms behind the better XD performance that is found in the simulations will be examined. Work supported by US DOE under DE-FG02-04ER54742 and SC 0012956.

  8. ICRF [Ion Cyclotron Range of Frequencies] heating and antenna coupling in a high beta tokamak

    International Nuclear Information System (INIS)

    Elet, R.S.

    1988-01-01

    Maxwell's Equations are solved in two-dimensions for the electromagnetic fields in a toroidal cavity using the cold plasma fluid dielectric tensor in the Ion Cyclotron Range of Frequencies (ICRF). The Vector Wave Equation is transformed to a set of two, coupled second-order partial differential equations with inhomogeneous forcing functions which model a wave launcher. The resulting equations are finite differenced and solved numerically with a complex banded matrix algorithm on a Cray-2 computer using a code described in this report. This code is used to study power coupling characteristics of a wave launcher for low and high beta tokamaks. The low and high beta equilibrium tokamak magnetic fields applied in this model are determined from analytic solutions to the Grad-Shafranov equation. The code shows good correspondence with the results of low field side ICRF heating experiments performed on the Tokamak of Fontenay-Aux-Roses (TFR). Low field side and high field side antenna coupling properties for ICRF heating in the Columbia High Beta Tokamak (HBT) experiment are calculated with this code. Variations of antenna position in the tokamak, ionic concentration and plasma density, and volume-averaged beta have been analyzed for HBT. It is found that the location of the antenna with respect to the plasma has the dominant role in the design of an ICRF heating experiment in HBT. 10 refs., 52 figs., 13 tabs

  9. Suppressed gross erosion of high-temperature lithium via rapid deuterium implantation

    NARCIS (Netherlands)

    Abrams, T.; Jaworski, M. A.; Chen, M.; Carter, E. A.; Kaita, R.; Stotler, D. P.; De Temmerman, G.; Morgan, T. W.; van den Berg, M. A.; van der Meiden, H. J.

    2016-01-01

    Lithium-coated high- Z substrates are planned for use in the NSTX-U divertor and are a candidate plasma facing component (PFC) for reactors, but it remains necessary to characterize the gross Li erosion rate under high plasma fluxes (>10 23 m −2 s −1 ), typical for the divertor region. In this

  10. Operational Regimes of the National Spherical Torus Experiment; TOPICAL

    International Nuclear Information System (INIS)

    D. Mueller; M.G. Bell; R.E. Bell; M. Bitter; T. Bigelow; P. Bonoli; M. Carter; J. Ferron; E. Fredrickson; D. Gates; L. Grisham; J.C. Hosea; D. Johnson; R. Kaita; S.M. Kaye; H. Kugel; B.P. LeBlanc; R. Maingi; R. Majeski; R. Maqueda; J. Menard; M. Ono; F. Paoletti; S. Paul; C.K. Phillips; R. Pinsker; R. Raman; S.A. Sabbagh; C.H. Skinner; V.A. Soukhanovskii; D. Stutman; D. Swain; Y. Takase; J. Wilgen; J.R. Wilson; G.A. Wurden; S. Zweben

    2002-01-01

    The National Spherical Torus Experiment (NSTX) is a proof-of-principle experiment designed to study the physics of Spherical Tori (ST), i.e., low-aspect-ratio toroidal plasmas. Important issues for ST research are whether the high-eta stability and reduced transport theoretically predicted for this configuration can be realized experimentally. In NSTX, the commissioning of a digital real-time plasma control system, the provision of flexible heating systems, and the application of wall conditioning techniques were instrumental in achieving routine operation with good confinement. NSTX has produced plasmas with R/a(approx) 0.85 m/0.68 m, A(approx) 1.25, Ip* 1.1 MA, BT= 0.3-0.45 T, k* 2.2, d* 0.5, with auxiliary heating by up to 4 MW of High Harmonic Fast Waves, and 5 MW of 80 keV D0 Neutral Beam Injection (NBI). The energy confinement time in plasmas heated by NBI has exceeded 100 ms and a toroidal beta (bT= 2m0 and lt;p and gt;/BT02, where BT0 is the central vacuum toroidal magnetic field) up to 22% has be en achieved. HHFW power of 2.3 MW has increased the electron temperature from an initial 0.4 keV to 0.9 keV both with and without producing a significant density rise in the plasma. The early application of both NBI and HHFW heating has slowed the penetration of the inductively produced plasma current, modifying the current profile and, thereby, the observed MHD stability

  11. Operational Regimes of the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Mueller, D.; Bell, M.G.; Bell, R.E.; Bitter, M.; Bigelow, T.; Bonoli, P.; Carter, M.; Ferron, J.; Fredrickson, E.; Gates, D.; Grisham, L.; Hosea, J.C.; Johnson, D.; Kaita, R.; Kaye, S.M.; Kugel, H.; LeBlanc, B.P.; Maingi, R.; Majeski, R.; Maqueda, R.; Menard, J.; Ono, M.; Paoletti, F.; Paul, S.; Phillips, C.K.; Pinsker, R.; Raman, R.; Sabbagh, S.A.; Skinner, C.H.; Soukhanovskii, V.A; Stutman, D.; Swain, D.; Takase, Y.; Wilgen, J.; Wilson, J.R.; Wurden, G.A.; Zweben, S.

    2002-01-01

    The National Spherical Torus Experiment (NSTX) is a proof-of-principle experiment designed to study the physics of Spherical Tori (ST), i.e., low-aspect-ratio toroidal plasmas. Important issues for ST research are whether the high-eta stability and reduced transport theoretically predicted for this configuration can be realized experimentally. In NSTX, the commissioning of a digital real-time plasma control system, the provision of flexible heating systems, and the application of wall conditioning techniques were instrumental in achieving routine operation with good confinement. NSTX has produced plasmas with R/a ∼ 0.85 m/0.68 m, A ∼ 1.25, Ip * 1.1 MA, BT = 0.3-0.45 T, k * 2.2, d * 0.5, with auxiliary heating by up to 4 MW of High Harmonic Fast Waves, and 5 MW of 80 keV D0 Neutral Beam Injection (NBI). The energy confinement time in plasmas heated by NBI has exceeded 100 ms and a toroidal beta (bT = 2m0 /BT02, where BT0 is the central vacuum toroidal magnetic field) up to 22% has be en achieved. HHFW power of 2.3 MW has increased the electron temperature from an initial 0.4 keV to 0.9 keV both with and without producing a significant density rise in the plasma. The early application of both NBI and HHFW heating has slowed the penetration of the inductively produced plasma current, modifying the current profile and, thereby, the observed MHD stability

  12. Scenario development during commissioning operations on the National Spherical Torus Experiment Upgrade

    Science.gov (United States)

    Battaglia, D. J.; Boyer, M. D.; Gerhardt, S.; Mueller, D.; Myers, C. E.; Guttenfelder, W.; Menard, J. E.; Sabbagh, S. A.; Scotti, F.; Bedoya, F.; Bell, R. E.; Berkery, J. W.; Diallo, A.; Ferraro, N.; Kaye, S. M.; Jaworski, M. A.; LeBlanc, B. P.; Ono, M.; Park, J.-K.; Podesta, M.; Raman, R.; Soukhanovskii, V.; NSTX-U Research, the; Operations; Engineering Team

    2018-04-01

    The National Spherical Torus Experiment Upgrade (NSTX-U) will advance the physics basis required for achieving steady-state, high-beta, and high-confinement conditions in a tokamak by accessing high toroidal fields (1 T) and plasma currents (1.0-2.0 MA) in a low aspect ratio geometry (A  =  1.6-1.8) with flexible auxiliary heating systems (12 MW NBI, 6 MW HHFW). This paper describes the progress in the development of L- and H-mode discharge scenarios and the commissioning of operational tools in the first ten weeks of operation that enable the scientific mission of NSTX-U. Vacuum field calculations completed prior to operations supported the rapid development and optimization of inductive breakdown at different values of ohmic solenoid current. The toroidal magnetic field (B T0  =  0.65 T) exceeded the maximum values achieved on NSTX and novel long-pulse L-mode discharges with regular sawtooth activity exceeded the longest pulses produced on NSTX (t pulse  >  1.8 s). The increased flux of the central solenoid facilitated the development of stationary L-mode discharges over a range of density and plasma current (I p). H-mode discharges achieved similar levels of stored energy, confinement (H98y,2  >  1) and stability (β N/β N-nowall  >  1) compared to NSTX discharges for I p  ⩽  1 MA. High-performance H-mode scenarios require an L-H transition early in the I p ramp-up phase in order to obtain low internal inductance (l i) throughout the discharge, which is conducive to maintaining vertical stability at high elongation (κ  >  2.2) and achieving long periods of MHD quiescent operations. The rapid progress in developing L- and H-mode scenarios in support of the scientific program was enabled by advances in real-time plasma control, efficient error field identification and correction, effective conditioning of the graphite wall and excellent diagnostic availability.

  13. Implementation of a 3D halo neutral model in the TRANSP code and application to projected NSTX-U plasmas

    Science.gov (United States)

    Medley, S. S.; Liu, D.; Gorelenkova, M. V.; Heidbrink, W. W.; Stagner, L.

    2016-02-01

    A 3D halo neutral code developed at the Princeton Plasma Physics Laboratory and implemented for analysis using the TRANSP code is applied to projected National Spherical Torus eXperiment-Upgrade (NSTX-U plasmas). The legacy TRANSP code did not handle halo neutrals properly since they were distributed over the plasma volume rather than remaining in the vicinity of the neutral beam footprint as is actually the case. The 3D halo neutral code uses a ‘beam-in-a-box’ model that encompasses both injected beam neutrals and resulting halo neutrals. Upon deposition by charge exchange, a subset of the full, one-half and one-third beam energy components produce first generation halo neutrals that are tracked through successive generations until an ionization event occurs or the descendant halos exit the box. The 3D halo neutral model and neutral particle analyzer (NPA) simulator in the TRANSP code have been benchmarked with the Fast-Ion D-Alpha simulation (FIDAsim) code, which provides Monte Carlo simulations of beam neutral injection, attenuation, halo generation, halo spatial diffusion, and photoemission processes. When using the same atomic physics database, TRANSP and FIDAsim simulations achieve excellent agreement on the spatial profile and magnitude of beam and halo neutral densities and the NPA energy spectrum. The simulations show that the halo neutral density can be comparable to the beam neutral density. These halo neutrals can double the NPA flux, but they have minor effects on the NPA energy spectrum shape. The TRANSP and FIDAsim simulations also suggest that the magnitudes of beam and halo neutral densities are relatively sensitive to the choice of the atomic physics databases.

  14. Implementation of a 3D halo neutral model in the TRANSP code and application to projected NSTX-U plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Medley, S. S. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Liu, D. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy; Gorelenkova, M. V. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Heidbrink, W. W. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy; Stagner, L. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy

    2016-01-12

    A 3D halo neutral code developed at the Princeton Plasma Physics Laboratory and implemented for analysis using the TRANSP code is applied to projected National Spherical Torus eXperiment-Upgrade (NSTX-U plasmas). The legacy TRANSP code did not handle halo neutrals properly since they were distributed over the plasma volume rather than remaining in the vicinity of the neutral beam footprint as is actually the case. The 3D halo neutral code uses a 'beam-in-a-box' model that encompasses both injected beam neutrals and resulting halo neutrals. Upon deposition by charge exchange, a subset of the full, one-half and one-third beam energy components produce first generation halo neutrals that are tracked through successive generations until an ionization event occurs or the descendant halos exit the box. The 3D halo neutral model and neutral particle analyzer (NPA) simulator in the TRANSP code have been benchmarked with the Fast-Ion D-Alpha simulation (FIDAsim) code, which provides Monte Carlo simulations of beam neutral injection, attenuation, halo generation, halo spatial diffusion, and photoemission processes. When using the same atomic physics database, TRANSP and FIDAsim simulations achieve excellent agreement on the spatial profile and magnitude of beam and halo neutral densities and the NPA energy spectrum. The simulations show that the halo neutral density can be comparable to the beam neutral density. These halo neutrals can double the NPA flux, but they have minor effects on the NPA energy spectrum shape. The TRANSP and FIDAsim simulations also suggest that the magnitudes of beam and halo neutral densities are relatively sensitive to the choice of the atomic physics databases.

  15. A thin foil Faraday collector as a lost alpha detector for high yield d-t tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Cecil, F. Ed

    2011-01-01

    This report summarizes the accomplishment of sixteen years of work toward the development of thin foil Faraday collectors as a lost energetic ion diagnostic for high temperature magnetic confinement fusion plasmas. Following initial, proof of principle accelerator based studies, devices have been tested on TFTR, NSTX, ALCATOR, DIII-D, and JET (KA-1 and KA-2). The reference numbers refer to the attached list of publications. The JET diagnostic KA-2 continues in operation and hopefully will provide valuable diagnostic information during a possible d-t campaign on JET in the coming years. A thin Faraday foil spectrometer, by virtue of its radiation hardness, may likewise provide a solution to the very challenging problem of lost alpha particle measurements on ITER and other future burning plasma machines.

  16. Reversed magnetic shear suppression of electron-scale turbulence on NSTX

    Science.gov (United States)

    Yuh, Howard Y.; Levinton, F. M.; Bell, R. E.; Hosea, J. C.; Kaye, S. M.; Leblanc, B. P.; Mazzucato, E.; Smith, D. R.; Domier, C. W.; Luhmann, N. C.; Park, H. K.

    2009-11-01

    Electron thermal internal transport barriers (e-ITBs) are observed in reversed (negative) magnetic shear NSTX discharges^1. These e-ITBs can be created with either neutral beam heating or High Harmonic Fast Wave (HHFW) RF heating. The e-ITB location occurs at the location of minimum magnetic shear determined by Motional Stark Effect (MSE) constrained equilibria. Statistical studies show a threshold condition in magnetic shear for e-ITB formation. High-k fluctuation measurements at electron turbulence wavenumbers^3 have been made under several different transport regimes, including a bursty regime that limits temperature gradients at intermediate magnetic shear. The growth rate of fluctuations has been calculated immediately following a change in the local magnetic shear, resulting in electron temperature gradient relaxation. Linear gyrokinetic simulation results for NSTX show that while measured electron temperature gradients exceed critical linear thresholds for ETG instability, growth rates can remain low under reversed shear conditions up to high electron temperatures gradients. ^1H. Yuh, et. al., PoP 16, 056120 ^2D.R. Smith, E. Mazzucato et al., RSI 75, 3840 ^3E. Mazzucato, D.R. Smith et al., PRL 101, 075001

  17. Collisional Damping of Electron Bernstein Waves and its Mitigation by Evaporated Lithium Conditioning in Spherical-Tokamak Plasmas

    International Nuclear Information System (INIS)

    Diem, S. J.; Caughman, J. B.; Taylor, G.; Efthimion, P. C.; Kugel, H.; LeBlanc, B. P.; Phillips, C. K.; Preinhaelter, J.; Urban, J.; Sabbagh, S. A.

    2009-01-01

    The first experimental verification of electron Bernstein wave (EBW) collisional damping, and its mitigation by evaporated Li conditioning, in an overdense spherical-tokamak plasma has been observed in the National Spherical Torus Experiment (NSTX). Initial measurements of EBW emission, coupled from NSTX plasmas via double-mode conversion to O-mode waves, exhibited <10% transmission efficiencies. Simulations show 80% of the EBW energy is dissipated by collisions in the edge plasma. Li conditioning reduced the edge collision frequency by a factor of 3 and increased the fundamental EBW transmission to 60%.

  18. NSTX-U Advances in Real-Time C++11 on Linux

    Science.gov (United States)

    Erickson, Keith G.

    2015-08-01

    Programming languages like C and Ada combined with proprietary embedded operating systems have dominated the real-time application space for decades. The new C++11 standard includes native, language-level support for concurrency, a required feature for any nontrivial event-oriented real-time software. Threads, Locks, and Atomics now exist to provide the necessary tools to build the structures that make up the foundation of a complex real-time system. The National Spherical Torus Experiment Upgrade (NSTX-U) at the Princeton Plasma Physics Laboratory (PPPL) is breaking new ground with the language as applied to the needs of fusion devices. A new Digital Coil Protection System (DCPS) will serve as the main protection mechanism for the magnetic coils, and it is written entirely in C++11 running on Concurrent Computer Corporation's real-time operating system, RedHawk Linux. It runs over 600 algorithms in a 5 kHz control loop that determine whether or not to shut down operations before physical damage occurs. To accomplish this, NSTX-U engineers developed software tools that do not currently exist elsewhere, including real-time atomic synchronization, real-time containers, and a real-time logging framework. Together with a recent (and carefully configured) version of the GCC compiler, these tools enable data acquisition, processing, and output using a conventional operating system to meet a hard real-time deadline (that is, missing one periodic is a failure) of 200 microseconds.

  19. NSTX-U Advances in Real-Time C++11 on Linux

    International Nuclear Information System (INIS)

    Erickson, Keith G.

    2015-01-01

    Programming languages like C and Ada combined with proprietary embedded operating systems have dominated the real-time application space for decades. The new C++11standard includes native, language-level support for concurrency, a required feature for any nontrivial event-oriented real-time software. Threads, Locks, and Atomics now exist to provide the necessary tools to build the structures that make up the foundation of a complex real-time system. The National Spherical Torus Experiment Upgrade (NSTX-U) at the Princeton Plasma Physics Laboratory (PPPL) is breaking new ground with the language as applied to the needs of fusion devices. A new Digital Coil Protection System (DCPS) will serve as the main protection mechanism for the magnetic coils, and it is written entirely in C++11 running on Concurrent Computer Corporation's real-time operating system, RedHawk Linux. It runs over 600 algorithms in a 5 kHz control loop that determine whether or not to shut down operations before physical damage occurs. To accomplish this, NSTX-U engineers developed software tools that do not currently exist elsewhere, including real-time atomic synchronization, real-time containers, and a real-time logging framework. Together with a recent (and carefully configured) version of the GCC compiler, these tools enable data acquisition, processing, and output using a conventional operating system to meet a hard real-time deadline (that is, missing one periodic is a failure) of 200 microseconds

  20. Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

    Science.gov (United States)

    McLean, A. G.; Gan, K. F.; Ahn, J.-W.; Gray, T. K.; Maingi, R.; Abrams, T.; Jaworski, M. A.; Kaita, R.; Kugel, H. W.; Nygren, R. E.; Skinner, C. H.; Soukhanovskii, V. A.

    2013-07-01

    Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of Tsurface near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q⊥,peak = 5 MW/m2 inter-ELM and up to 10 MW/m2 during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

  1. Stabilizing effect of resistivity towards ELM-free H-mode discharge in lithium-conditioned NSTX

    Science.gov (United States)

    Banerjee, Debabrata; Zhu, Ping; Maingi, Rajesh

    2017-07-01

    Linear stability analysis of the national spherical torus experiment (NSTX) Li-conditioned ELM-free H-mode equilibria is carried out in the context of the extended magneto-hydrodynamic (MHD) model in NIMROD. The purpose is to investigate the physical cause behind edge localized mode (ELM) suppression in experiment after the Li-coating of the divertor and the first wall of the NSTX tokamak. Besides ideal MHD modeling, including finite-Larmor radius effect and two-fluid Hall and electron diamagnetic drift contributions, a non-ideal resistivity model is employed, taking into account the increase of Z eff after Li-conditioning in ELM-free H-mode. Unlike an earlier conclusion from an eigenvalue code analysis of these equilibria, NIMROD results find that after reduced recycling from divertor plates, profile modification is necessary but insufficient to explain the mechanism behind complete ELMs suppression in ideal two-fluid MHD. After considering the higher plasma resistivity due to higher Z eff, the complete stabilization could be explained. A thorough analysis of both pre-lithium ELMy and with-lithium ELM-free cases using ideal and non-ideal MHD models is presented, after accurately including a vacuum-like cold halo region in NIMROD to investigate ELMs.

  2. High beta and second stability region transport and stability analysis

    International Nuclear Information System (INIS)

    1991-01-01

    This document describes ideal and resistive MHD studies of high-beta plasmas and of the second stability region. Significant progress is reported on the resistive stability properties of high beta poloidal ''supershot'' discharges. For these studies initial profiles were taken from the TRANSP code which is used extensively to analyze experimental data. When an ad hoc method of removing the finite pressure stabilization of tearing modes is implemented it is shown that there is substantial agreement between MHD stability computation and experiment. In particular, the mode structures observed experimentally are consistent with the predictions of the resistive MHD model. We also report on resistive stability near the transition to the second region in TFTR. Tearing modes associated with a nearby infernal mode may explain the increase in MHD activity seen in high beta supershots and which impede the realization of Q∼1. We also report on a collaborative study with PPPL involving sawtooth stabilization with ICRF

  3. Preliminary Study of Ideal Operational MHD Beta Limit in HL-2A Tokamak Plasmas

    International Nuclear Information System (INIS)

    Shen Yong; Dong Jiaqi; He Hongda; Turnbull, A. D.

    2009-01-01

    Magnetohydrodynamic (MHD) n = 1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q 0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of β c N = 2.56 for a typical HL-2A discharge with a plasma current I p = 0.245 MA, and the scaling of β c N ∼constant is confirmed. (magnetically confined plasma)

  4. Electrostatic fluctuation in Low-{beta} plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Krane, B

    1997-11-01

    The thesis gives an overview, which by no means is complete, to two dimensional plasma flows. The plasma is taken to be strongly magnetized such that magnetic fields generated by internal currents are small compared to the background magnetic field. This requires that the magnetic pressure exerted by the ambient magnetic field is large compared to the pressure due to thermal fluctuations, i.e. low-{beta} plasma. The author also assume low frequency electrostatic fluctuations with {omega}<<{omega}{sub c}i where {omega}{sub c}i is the ion gyro frequency. A brief introduction to nonlinear phenomena in two dimensional plasma flows has been presented. Particular attention was given to simple models describing flute and drift modes. Although the derivations of the model equations are based on different assumptions regarding the plasma conditions, the resulting equations exhibit similar behavior in some respects. For instance, both the simple guiding center model and the Hasegawa-Mima model have stable dipolar structures. The inverse cascade was also found in both models. However, it is evident that there are significant differences, first of all the Hasegawa-Mima model assumes a background density gradient which makes it an inhomogeneous model. Secondly, in this model the electrons respond instantaneously to variations in the ion density by moving along the magnetic field, thereby introducing Debye shielding.

  5. Influence of insulin on beta-endorphin plasma levels in obese and normal weight subjects.

    Science.gov (United States)

    Brunani, A; Pincelli, A I; Pasqualinotto, L; Tibaldi, A; Baldi, G; Scacchi, M; Fatti, L M; Cavagnini, F

    1996-08-01

    To establish the possible role of hyperinsulinemia in the elevation of plasma beta-endorphin (beta-EP) levels observed in obese patients after an oral glucose load. Oral glucose tolerance test (OGTT) and euglycemic-hyperinsulinemic clamp. Two groups of six (age: 22-39 y, BMI: 30-48 kg/m2) and eight obese men (age: 18-37 y, BMI: 35-45 kg/m2), respectively, and five normal weight healthy men (age: 22-30 y, BMI 22-23 kg/m2). Glucose, insulin and beta-EP levels at baseline and every 30 min until 180 min during the OGTT; glucose, insulin, C-peptide and beta-EP concentrations at baseline and in steady state condition (i.e. during the last 30 min of insulin infusion) in the euglycemic-hyperinsulinemic clamp studies. In the six obese patients undergoing the OGTT a significant elevation of beta-EP plasma levels was observed between 60 and 90 min after glucose ingestion. In the clamp studies no significant differences in beta-EP plasma levels, blood glucose and serum insulin were observed between obese and normal weight subjects both at baseline and at steady state. A markedly diminished insulin sensitivity along with a lower inhibition of C-peptide during insulin infusion was observed in obese patients compared to control subjects. A rise in serum insulin levels unaccompanied by a concomitant increase in blood glucose concentration is unable to elicit a beta-EP response in obese patients.

  6. Diurnal variation of. beta. -endorphin like immunoreactivity in rat brain, pituitary gland, and plasma

    Energy Technology Data Exchange (ETDEWEB)

    Izquierdo, I.A.; Perry, M.L.S.; Carrasco, M.A.; Dias, R.D. (Rio Grande do Sul Univ., Porto Alegre (Brazil). Inst. de Biociencias); Orsingher, O.A. (Universidad Nacional de Cordoba (Argentina))

    1984-09-01

    ..beta..-endorphin like immunoreactivity was measured in the brain, pituitary gland and plasma of rats at 2 A.M, 8 A.M, 2 P.M and 8 P.M. Values were higher in the brain and pituitary gland at 8 P.M and in the plasma at 8 A.M and 2 P.M. The findings suggest a circadian rhythm in the production and release of ..beta..-endorphin immunoreactive material.

  7. High beta and second stability region transport and stability analysis

    International Nuclear Information System (INIS)

    Hughes, M.H.; Phillps, M.W.; Todd, A.M.M.; Krishnaswami, J.; Hartley, R.

    1992-09-01

    This report describes ideal and resistive studies of high-beta plasmas and of the second stability region. Emphasis is focused on ''supershot'' plasmas in TFIR where MHD instabilities are frequently observed and which spoil their confinement properties. Substantial results are described from the analysis of these high beta poloidal plasmas. During these studies, initial pressure and safety factor profiles were obtained from the TRANSP code, which is used extensively to analyze experimental data. Resistive MBD stability studies of supershot equilibria show that finite pressure stabilization of tearing modes is very strong in these high βp plasmas. This has prompted a detailed re-examination of linear tearing mode theory in which we participated in collaboration with Columbia University and General Atomics. This finite pressure effect is shown to be highly sensitive to small scale details of the pressure profile. Even when an ad hoc method of removing this stabilizing mechanism is implemented, however, it is shown that there is only superficial agreement between resistive MBD stability computation and the experimental data. While the mode structures observed experimentally can be found computationally, there is no convincing correlation with the experimental observations when the computed results are compared with a large set of supershot data. We also describe both the ideal and resistive stability properties of TFIR equilibria near the transition to the second region. It is shown that the highest β plasmas, although stable to infinite-n ideal ballooning modes, can be unstable to the so called ''infernal'' modes associated with small shear. The sensitivity of these results to the assumed pressure and current density profiles is discussed. Finally, we describe results from two collaborative studies with PPPL. The first involves exploratory studies of the role of the 1/1 mode in tokamaks and, secondly, a study of sawtooth stabilization using ICRF

  8. Recent advancement in research and planning toward high beta steady state operation in KSTAR

    International Nuclear Information System (INIS)

    Park, Hyeon Keo; Hong, S.; Humphreys, D.

    2015-01-01

    The goal of Korean Superconducting Tokamak Advanced Research (KSTAR) research is to explore stable improved confinement regimes and technical challenge for superconducting tokamak operation and thus, to establish the basis for predictable high beta steady state tokamak plasma operation. To fulfil the goal, the current KSTAR research program is composed of three elements: 1) Exploration of anticipated engineering and technology for a stable long pulse operation of high beta plasmas including Edge Localized Mode (ELM) control with the low n (=1, 2) Resonant Magnetic Perturbation (RMP) using in-vessel control coils and innovative non-inductive current drives. The achieved long pulse operation up to ∼50s and fully non-inductive current drive will be combined in the future. Study of efficient heat exhaust will be combined with an innovative divertor design/operation. 2) Exploration of the operation boundary through establishment of true stability limits of the harmful MagnetoHydroDynamic (MHD) instabilities and confinement of the tokamak plasmas in KSTAR, making use of the lowest error field and magnetic ripple simultaneously achieved among all tokamaks ever built. The intrinsic machine error field has a long history of research as the source of MHD instabilities and magnetic ripple is known to be a cause of energy loss in the plasma. The achieved high beta discharges at β N ∼4 and stable discharges at q 95 (∼2) will be further improved. 3) Validation of theoretical modeling of MHD instabilities and turbulence toward predictive capability of stable high beta plasmas. In support of these research goals, the state of the art diagnostic systems, such as Electron Cyclotron Emission Imaging (ECEI) system in addition to accurate profile diagnostics, are deployed not only to provide precise 2D/3D information of the MHD instabilities and turbulence but also to challenge unresolved physics problems such as the nature of ELMs, ELM-crash dynamics and the role of the core

  9. First demonstration of cerebrospinal fluid and plasma A beta lowering with oral administration of a beta-site amyloid precursor protein-cleaving enzyme 1 inhibitor in nonhuman primates.

    Science.gov (United States)

    Sankaranarayanan, Sethu; Holahan, Marie A; Colussi, Dennis; Crouthamel, Ming-Chih; Devanarayan, Viswanath; Ellis, Joan; Espeseth, Amy; Gates, Adam T; Graham, Samuel L; Gregro, Allison R; Hazuda, Daria; Hochman, Jerome H; Holloway, Katharine; Jin, Lixia; Kahana, Jason; Lai, Ming-tain; Lineberger, Janet; McGaughey, Georgia; Moore, Keith P; Nantermet, Philippe; Pietrak, Beth; Price, Eric A; Rajapakse, Hemaka; Stauffer, Shaun; Steinbeiser, Melissa A; Seabrook, Guy; Selnick, Harold G; Shi, Xiao-Ping; Stanton, Matthew G; Swestock, John; Tugusheva, Katherine; Tyler, Keala X; Vacca, Joseph P; Wong, Jacky; Wu, Guoxin; Xu, Min; Cook, Jacquelynn J; Simon, Adam J

    2009-01-01

    beta-Site amyloid precursor protein (APP)-cleaving enzyme (BACE) 1 cleavage of amyloid precursor protein is an essential step in the generation of the potentially neurotoxic and amyloidogenic A beta 42 peptides in Alzheimer's disease. Although previous mouse studies have shown brain A beta lowering after BACE1 inhibition, extension of such studies to nonhuman primates or man was precluded by poor potency, brain penetration, and pharmacokinetics of available inhibitors. In this study, a novel tertiary carbinamine BACE1 inhibitor, tertiary carbinamine (TC)-1, was assessed in a unique cisterna magna ported rhesus monkey model, where the temporal dynamics of A beta in cerebrospinal fluid (CSF) and plasma could be evaluated. TC-1, a potent inhibitor (IC(50) approximately 0.4 nM), has excellent passive membrane permeability, low susceptibility to P-glycoprotein transport, and lowered brain A beta levels in a mouse model. Intravenous infusion of TC-1 led to a significant but transient lowering of CSF and plasma A beta levels in conscious rhesus monkeys because it underwent CYP3A4-mediated metabolism. Oral codosing of TC-1 with ritonavir, a potent CYP3A4 inhibitor, twice daily over 3.5 days in rhesus monkeys led to sustained plasma TC-1 exposure and a significant and sustained reduction in CSF sAPP beta, A beta 40, A beta 42, and plasma A beta 40 levels. CSF A beta 42 lowering showed an EC(50) of approximately 20 nM with respect to the CSF [TC-1] levels, demonstrating excellent concordance with its potency in a cell-based assay. These results demonstrate the first in vivo proof of concept of CSF A beta lowering after oral administration of a BACE1 inhibitor in a nonhuman primate.

  10. Electromagnetic effects on dynamics of high-beta filamentary structures

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Wonjae; Krasheninnikov, Sergei I., E-mail: skrash@mae.ucsd.edu [University of California, San Diego, 9500 Gilman Drive, La Jolla, California 92093 (United States); Umansky, Maxim V. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, California 94550 (United States); Angus, J. R. [Naval Research Laboratory, 4555 Overlook Avenue, Washington, DC 20375 (United States)

    2015-01-15

    The impacts of the electromagnetic effects on blob dynamics are considered. Electromagnetic BOUT++ simulations on seeded high-beta blobs demonstrate that inhomogeneity of magnetic curvature or plasma pressure along the filament leads to bending of the blob filaments and the magnetic field lines due to increased propagation time of plasma current (Alfvén time). The bending motion can enhance heat exchange between the plasma facing materials and the inner scrape-off layer (SOL) region. The effects of sheath boundary conditions on the part of the blob away from the boundary are also diminished by the increased Alfvén time. Using linear analysis and BOUT++ simulations, it is found that electromagnetic effects in high temperature and high density plasmas reduce the growth rate of resistive drift wave instability when resistivity drops below a certain value. The blobs temperature decreases in the course of its motion through the SOL and so the blob can switch from the electromagnetic to the electrostatic regime where resistive drift waves become important again.

  11. Equilibrium and stability of high-beta toroidal plasmas with toroidal and poloidal flow in reduced magnetohydrodynamic models

    International Nuclear Information System (INIS)

    Ito, A.; Nakajima, N.

    2010-11-01

    Effects of flow, finite ion temperature and pressure anisotropy on equilibrium and stability of a high-beta toroidal plasma are studied in the framework of reduced magnetohydrodynamics (MHD). A set of reduced equilibrium equations for high-beta tokamaks with toroidal and poloidal flow comparable to the poloidal sound velocity is derived in a unified form of single-fluid and Hall MHD models and a two-fluid MHD model with ion finite Larmor radius (FLR) terms. Pressure anisotropy is introduced with equations for the parallel heat flux which are closed by a fluid closure model. It is solved analytically for the single-fluid model and the solutions shows complicated characteristics in the region around the poloidal sound velocity due to pressure anisotropy and the parallel heat flux. Numerical solutions are found by using the finite element method for the two-fluid model with FLR effects in the case of isotropic, adiabatic pressure and indicate the following features of two-fluid equilibria: the isosurfaces of the magnetic flux, the pressure and the ion stream function do not coincide with each other, and the solutions depend on the sign of the radial electric field. Reduced single-fluid MHD equations with time evolution that are consistent with the above equilibria are also derived in order to study their stability. They conserve the energy up to the order required by the equilibria. (author)

  12. Experimental/theoretical comparisons of the turbulence in the scrape-off-layers of Alcator C-Mod, DIII-D, and NSTX

    International Nuclear Information System (INIS)

    Terry, J.L. . E-mail : terry@psfc.mit.edu; Zweben, S.J.; Rudakov, D.L.

    2003-01-01

    The intermittent turbulent transport in the scrape-off-layers of Alcator C-Mod, DIII-D, and NSTX is studied experimentally. On DIII-D the fluctuations of both density and temperature have strongly non-Gaussian statistics, and events with amplitudes above 10 times the mean level are responsible for large fractions of the net particle and heat transport, indicating the importance of turbulence on the transport. In C-Mod and NSTX the turbulence is imaged with a very high density of spatial measurements. The 2-D structure and dynamics of emission from a localized gas puff are observed, and intermittent features (also sometimes called 'blobs') are typically seen. On DIII-D the turbulence is imaged using BES and similar intermittent features are seen. The dynamics of these intermittent features are discussed. The experimental observations are compared with numerical simulations of edge turbulence. The electromagnetic turbulence in a 3-D geometry is computed using non-linear plasma fluid equations. The wavenumber spectra in the poloidal dimension of the simulations are in reasonable agreement with those of the C-Mod experimental images once the response of the optical system is accounted for. The resistive ballooning mode is the dominant linear instability in the simulations. (author)

  13. Plasma-assisted co-evaporation of {beta}-indium sulfide thin films

    Energy Technology Data Exchange (ETDEWEB)

    Kosaraju, Sreenivas; Marino, Joseph A.; Harvey, John A.; Wolden, Colin A. [Department of Chemical Engineering, Colorado School of Mines, Golden, CO 80401 (United States)

    2006-05-05

    This paper describes the development of plasma-assisted co-evaporation (PACE) for the formation of {beta}-In{sub 2}S{sub 3} thin films. Indium was supplied by conventional thermal evaporation, while the chalcogen gas precursor (H{sub 2}S) was activated using an inductively coupled plasma (ICP) source. Using a combination of optical emission spectroscopy and mass spectrometry it was shown that the ICP effectively dissociated H{sub 2}S, producing atomic sulfur. Transport modeling was used to quantify the flux distributions of the co-evaporated metal and the plasma-generated species impinging the substrate. Model predictions were validated by measurements of deposition rate and film properties. Substantial improvements in both materials utilization and substrate temperature reduction were realized with respect to conventional co-evaporation. {beta}-In{sub 2}S{sub 3} was formed as low as 100{sup o}C and it was observed that quality was a strong function of S/In ratio. The grain size decreased and the optical band gap increased as the substrate temperature was reduced. (author)

  14. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U.

    Science.gov (United States)

    Faust, I; Delgado-Aparicio, L; Bell, R E; Tritz, K; Diallo, A; Gerhardt, S P; LeBlanc, B; Kozub, T A; Parker, R R; Stratton, B C

    2014-11-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  15. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-Ua)

    Science.gov (United States)

    Faust, I.; Delgado-Aparicio, L.; Bell, R. E.; Tritz, K.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A.; Parker, R. R.; Stratton, B. C.

    2014-11-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  16. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Faust, I.; Parker, R. R. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Delgado-Aparicio, L.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Tritz, K. [The Johns Hopkins University, Baltimore, Maryland 21209 (United States); Stratton, B. C. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)

    2014-11-15

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

  17. Two-dimensional AXUV-based radiated power density diagnostics on NSTX-U

    International Nuclear Information System (INIS)

    Faust, I.; Parker, R. R.; Delgado-Aparicio, L.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B.; Kozub, T. A.; Tritz, K.; Stratton, B. C.

    2014-01-01

    A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed

  18. MHD Stability Calculations of High-Beta Quasi-Axisymmetric Stellarators

    International Nuclear Information System (INIS)

    Kessel, C.; Fu, G.Y.; Ku, L.P.; Redi, M.H.; Pomphrey, N.

    1999-01-01

    The MHD stability of quasi-axisymmetric compact stellarators is investigated. It is shown that bootstrap current driven external kink modes can be stabilized by a combination of edge magnetic shear and appropriate 3D plasma boundary shaping while maintaining good quasi-axisymmetry. The results demonstrate that there exists a new class of stellarators with quasi-axisymmetry, large bootstrap current, high MHD beta limit, and compact size

  19. Kinetic Modifications to MHD Phenomena in Toroidal Plasmas

    International Nuclear Information System (INIS)

    Cheng, C.Z.; Gorelenkov, N.N.; Kramer, G.J.; Fredrickson, E.

    2004-01-01

    Particle kinetic effects involving small spatial and fast temporal scales can strongly affect MHD phenomena and the long time behavior of plasmas. In particular, kinetic effects such as finite ion gyroradii, trapped particle dynamics, and wave-particle resonances have been shown to greatly modify the stability of MHD modes. Here, the kinetic effects of trapped electron dynamics and finite ion gyroradii are shown to have a large stabilizing effect on kinetic ballooning modes in low aspect ratio toroidal plasmas such as NSTX [National Spherical Torus Experiment]. We also present the analysis of Toroidicity-induced Alfven Eigenmodes (TAEs) destabilized by fast neutral-beam injected ions in NSTX experiments and TAE stability in ITER due to alpha-particles and MeV negatively charged neutral beam injected ions

  20. Studies on limiter confined toroidal plasma in BETA

    International Nuclear Information System (INIS)

    Bera, D.; Reddy, C.; Jayakumar, R.; Kaw, P.K.

    1984-01-01

    Plasma equilibrium and stability in the presence of a toroidal magnetic field and a poloidal limiter is being studied experimentally in the BETA experiment. In a simple toroidal magnetic field, plasma cannot be in equilibrium because of the effect of the magnetic field curvature, which tends to expand the plasma. The electric field, which causes this expansion, is short circuited if a poloidal conducting limiter is placed and this brings about a quasi-equilibrium. In the model the charge separation current flows on the surface of the plasma and closes the path by flowing parallel to the magnetic field away from the limiter and transverse to the field at the limiter. For such an equilibrium, the vertical pressure profile is expected to be uniform, while the radial pressure profile is determined by transport. Such a profile is unstable to Rayleigh-Taylor instabilities, if the magnetic field gradient and the pressure gradient have the same sense

  1. Precision metrology of NSTX surfaces using coherent laser radar ranging

    International Nuclear Information System (INIS)

    Kugel, H.W.; Loesser, D.; Roquemore, A. L.; Menon, M. M.; Barry, R. E.

    2000-01-01

    A frequency modulated Coherent Laser Radar ranging diagnostic is being used on the National Spherical Torus Experiment (NSTX) for precision metrology. The distance (range) between the 1.5 microm laser source and the target is measured by the shift in frequency of the linearly modulated beam reflected off the target. The range can be measured to a precision of < 100microm at distances of up to 22 meters. A description is given of the geometry and procedure for measuring NSTX interior and exterior surfaces during open vessel conditions, and the results of measurements are elaborated

  2. MHD Calculation of halo currents and vessel forces in NSTX VDEs

    Science.gov (United States)

    Breslau, J. A.; Strauss, H. R.; Paccagnella, R.

    2012-10-01

    Research tokamaks such as ITER must be designed to tolerate a limited number of disruptions without sustaining significant damage. It is therefore vital to have numerical tools that can accurately predict the effects of these events. The 3D nonlinear extended MHD code M3D [1] can be used to simulate disruptions and calculate the associated wall currents and forces. It has now been validated against halo current data from NSTX experiments in which vertical displacement events (VDEs) were deliberately induced by turning off vertical feedback control. The results of high-resolution numerical simulations at realistic Lundquist numbers show reasonable agreement with the data, supporting a model in which the most dangerously asymmetric currents and heat loads, and the largest horizontal forces, arise in situations where a fast-growing ideal 2,1 external kink mode is destabilized by the scraping-off of flux surfaces with safety factor q>2 during the course of the VDE. [4pt] [1] W. Park, et al., Phys. Plasmas 6 (1999) 1796.

  3. Three-dimensional Reconstruction of Dust Particle Trajectories in the NSTX

    International Nuclear Information System (INIS)

    Boeglin, W.U.; Roquemore, A.L.; Maqueda, R.

    2009-01-01

    Highly mobile incandescent dust particles are routinely observed on NSTX using two fast cameras operating in the visible region. An analysis method to reconstruct dust particle trajectories in space using two fast cameras is presented in this paper. Position accuracies of a few millimeters depending on the particle's location have been achieved and particle velocities between 10 and 200 m/s have been observed

  4. Alignment of the Thomson scattering diagnostic on NSTX

    International Nuclear Information System (INIS)

    LeBlanc, B P; Diallo, A

    2013-01-01

    The Thomson scattering diagnostic can provide profile measurement of the electron temperature, T e , and density, n e , in plasmas. Proper laser beam path and optics arrangement permits profiles T e (R) and n e (R) measurement along the major radius R. Keeping proper alignment between the laser beam path and the collection optics is necessary for an accurate determination of the electron density. As time progresses the relative position of the collection optics field of view with respect to the laser beam path will invariably shift. This can be kept to a minimum by proper attention to the physical arrangement of the collection and laser-beam delivery optics. A system has been in place to monitor the relative position between laser beam and collection optics. Variation of the alignment can be detected before it begins to affect the quality of the profile data. This paper discusses details of the instrumentation and techniques used to maintain alignment during NSTX multi-month experimental campaigns

  5. Dependence of {beta} {center_dot} {tau} on plasma shape in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Lazarus, E.A. [Oak Ridge National Lab., TN (United States)

    1993-12-31

    In this paper we discuss the observed variation in plasma performance with plasma shape, in particular, we shall compare single and double null diverted plasmas. The product {beta} {center_dot} {tau} has been used as a figure-of-merit for comparing different toroidal magnetic configurations. Here we shall use it as the figure-of-merit for comparing differing configurations within the DIII-D tokamak. (author) 5 refs., 5 figs.

  6. Impact of ELM filaments on divertor heat flux dynamics in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, J.-W., E-mail: jahn@pppl.gov [Oak Ridge National Laboratory, Oak Ridge (United States); Maingi, R. [Princeton Plasma Physics Laboratory, Princeton (United States); Canik, J.M. [Oak Ridge National Laboratory, Oak Ridge (United States); Gan, K.F. [Institute of Plasma Physics, Chinese Academy of Science, Hefei (China); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore (United States)

    2015-08-15

    The ELM induced change in wetted area (A{sub wet}) and peak heat flux (q{sub peak}) of divertor heat flux is investigated as a function of the number of striations, which represent ELM filaments, observed in the heat flux profile in NSTX. More striations are found to lead to larger A{sub wet} and lower q{sub peak}. The typical number of striations observed in NSTX is 0–9, while 10–15 striations are normally observed in other machines such as JET, and the ELM contracts heat flux profile when the number of striations is less than 3–4 but broadens it with more of them. The smaller number of striations in NSTX is attributed to the fact that NSTX ELMs are against kink/peeling boundary with lower toroidal mode number (n = 1–5), while typical peeling–ballooning ELMs have higher mode number of n = 10–20. For ELMs with smaller number of striations, relative A{sub wet} change is rather constant and q{sub peak} change rapidly increases with increasing ELM size, while A{sub wet} change slightly increases leading to a weaker increase of q{sub peak} change for ELMs with larger number of striations, both of which are unfavourable trend for the material integrity of divertor tiles.

  7. Continuum Gyrokinetic Simulations of Turbulence in a Helical Model SOL with NSTX-type parameters

    Science.gov (United States)

    Hammett, G. W.; Shi, E. L.; Hakim, A.; Stoltzfus-Dueck, T.

    2017-10-01

    We have developed the Gkeyll code to carry out 3D2V full- F gyrokinetic simulations of electrostatic plasma turbulence in open-field-line geometries, using special versions of discontinuous-Galerkin algorithms to help with the computational challenges of the edge region. (Higher-order algorithms can also be helpful for exascale computing as they reduce the ratio of communications to computations.) Our first simulations with straight field lines were done for LAPD-type cases. Here we extend this to a helical model of an SOL plasma and show results for NSTX-type parameters. These simulations include the basic elements of a scrape-off layer: bad-curvature/interchange drive of instabilities, narrow sources to model plasma leaking from the core, and parallel losses with model sheath boundary conditions (our model allows currents to flow in and out of the walls). The formation of blobs is observed. By reducing the strength of the poloidal magnetic field, the heat flux at the divertor plate is observed to broaden. Supported by the Max-Planck/Princeton Center for Plasma Physics, the SciDAC Center for the Study of Plasma Microturbulence, and DOE Contract DE-AC02-09CH11466.

  8. Beta-Suppression of Alfven Cascade Modes in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; N.A. Crocker; N.N. Gorelenkov; W.W. Heidbrink; S. Kubota; F.M. Levinton; H. Yuh; J.E. Menard; Bell, R.E.

    2007-01-01

    The coupling of Alfven Cascade (AC) modes or reversed-shear Alfven 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 β e was increased from shot to shot, the range of the AC frequency sweep was reduced, in agreement with this theoretical prediction.

  9. Beta limit of crescent and bean shaped tokamaks

    International Nuclear Information System (INIS)

    Naitou, H.; Yamazaki, K.

    1988-01-01

    The maximum attainable beta values which can be expected in tokamaks with crescent (BEAN 1) and rounded (BEAN 2) bean shaped cross-sections are obtained numerically by using the linear ideal MHD stability analysis code ERATO. The current profiles are optimized with a fixed pressure profile for high values of beta, keeping Mercier, high-n ballooning and n=1 kink modes stable. The poloidal plasma cross-sections are inscribed in a rectangle with an aspect ratio of three and an ellipticity of two. A confocal wall, the distance of which from the plasma surface is equal to the horizontal minor plasma radius, is present to stabilize against the kink mode. Depending on the shape and triangularity (indentation), a beta value of 10 to 17% is obtained. It is also shown that the coefficient of the Troyon-type beta scaling increases for an indented plasma. In the case of small indentation, the BEAN 1 type tokamaks show higher beta values than the BEAN 2 type. For strong indentation, the BEAN 2 type gives the highest beta value. (author). 29 refs, 15 figs

  10. Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

    Energy Technology Data Exchange (ETDEWEB)

    McLean, A.G., E-mail: mclean@fusion.gat.com [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Gan, K.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Ahn, J.-W.; Gray, T.K.; Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Abrams, T.; Jaworski, M.A.; Kaita, R.; Kugel, H.W. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

    2013-07-15

    Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of T{sub surface} near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q{sub ⊥,peak} = 5 MW/m{sup 2} inter-ELM and up to 10 MW/m{sup 2} during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

  11. Plasma pre beta-HDL formation is decreased by atorvastatin treatment in type 2 diabetes mellitus : Role of phospholipid transfer protein

    NARCIS (Netherlands)

    Dallinga-Thie, G. M.; van Tol, A.; Dullaart, R. P. F.

    Atorvastatin lowers plasma phospholipid transfer protein (PLTP) activity, which stimulates pre-beta-HDL, generation in vitro. We determined the effect of atorvastatin on pre-beta-HDL formation and its relation with PLTP activity in type 2 diabetes. Methods: Plasma pre-beta-HDL formation as well as

  12. Parametric instabilities of parallel propagating incoherent Alfven waves in a finite ion beta plasma

    International Nuclear Information System (INIS)

    Nariyuki, Y.; Hada, T.; Tsubouchi, K.

    2007-01-01

    Large amplitude, low-frequency Alfven waves constitute one of the most essential elements of magnetohydrodynamic (MHD) turbulence in the fast solar wind. Due to small collisionless dissipation rates, the waves can propagate long distances and efficiently convey such macroscopic quantities as momentum, energy, and helicity. Since loading of such quantities is completed when the waves damp away, it is important to examine how the waves can dissipate in the solar wind. Among various possible dissipation processes of the Alfven waves, parametric instabilities have been believed to be important. In this paper, we numerically discuss the parametric instabilities of coherent/incoherent Alfven waves in a finite ion beta plasma using a one-dimensional hybrid (superparticle ions plus an electron massless fluid) simulation, in order to explain local production of sunward propagating Alfven waves, as suggested by Helios/Ulysses observation results. Parameter studies clarify the dependence of parametric instabilities of coherent/incoherent Alfven waves on the ion and electron beta ratio. Parametric instabilities of coherent Alfven waves in a finite ion beta plasma are vastly different from those in the cold ions (i.e., MHD and/or Hall-MHD systems), even if the collisionless damping of the Alfven waves are neglected. Further, ''nonlinearly driven'' modulational instability is important for the dissipation of incoherent Alfven waves in a finite ion beta plasma regardless of their polarization, since the ion kinetic effects let both the right-hand and left-hand polarized waves become unstable to the modulational instability. The present results suggest that, although the antisunward propagating dispersive Alfven waves are efficiently dissipated through the parametric instabilities in a finite ion beta plasma, these instabilities hardly produce the sunward propagating waves

  13. Spectroscopic study of turbulent heating in the high beta tokamak - Torus II

    International Nuclear Information System (INIS)

    Georgiou, G.E.

    1979-01-01

    Visible spectroscopy, involving line profile and line intensity measurements, was used to study the turbulent heating of the rectangular cross-section high-beta tokamak Torus II. The spectroscopy was done in the visible wave-length region using a six channel polychrometer having 0.2 A resolution, which is capable of radial scans of the plasma. The plasma, obtained by ionizing helium, is heated by poloidal skin currents, induced by a rapid (tau/sub R/ approx. = 1.7 μsec) change of the toroidal magnetic field either parallel or anti-parallel to the initial toroidal bias magnetic field, which converts a cold toroidal Z-pinch plasma into a hot tokamak plasma

  14. Effects of q and high beta on tokamak stability

    International Nuclear Information System (INIS)

    Brickhouse, N.S.; Callen, J.D.; Dexter, R.N.

    1984-08-01

    In the Columbia University Torus II tokamak plasmas have been studied with volume averaged toroidal beta values as high as 15%. Experimental equilibria have been compared with a 2D free boundary MHD equilibrium code PSEC. The stability of these equilibria has been computed using PEST, the predictions of which are compatible with an observed instability in Torus II which may be characterized as a high toroidal mode number ballooning fluctuation. In the University of Wisconsin Tokapole II tokamak disruptive instability behavior is investigated, with plasma able to be confined on closed magnetic surfaces in the scrape-off region, as the cylindrical edge safety factor is varied from q approx. 3 to q approx. 0.5. It is observed that at q/sub a/ approx. 3 major disruption activity occurs without current terminations, at q/sub a/ less than or equal to 2 well-confined plasmas are obtained without major disruption, and at q/sub a/ approx. 0.5 only partial reconnection accompanies minor disruptions

  15. Stability of high-beta tokamak equilibria and transport in Belt-Pinch IIa

    Energy Technology Data Exchange (ETDEWEB)

    Becker, G; Gruber, O; Krause, H; Mast, F; Wilhelm, R [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.)

    1978-01-01

    In Belt-Pinch IIa, highly elongated equilibria with poloidal beta values up to the aspect ratio have been achieved. In these tokamak-like configurations, no fast-growing MHD instabilities such as external kink and ballooning modes have been observed. Rigid displacement instabilities have been stabilized by an appropriate poloidal magnetic field configuration and by a conducting shell. By comparing simulation experiments using the Garching high-beta transport code with measurements, it has been found that in the collision-dominated plasma no anomalously enhanced transport occurs. Transport theory in the Pfirsch-Schlueter regime, which includes elongation and high-beta effects, has been confirmed by the experiment. In particular, it has been shown that the perpendicular electrical conductivity is also classical. Detailed investigations of oxygen and carbon impurity losses demonstrated that the impurity subprograms commonly used for tokamaks underestimate the radiation losses in the range Tsub(e)=10 to 30 eV.

  16. Vessel Eddy Current Measurement for the National Spherical Torus Experiment (NSTX)

    International Nuclear Information System (INIS)

    Gates, D.A.; Menard, J.; Marsala, R.

    2004-01-01

    A simple analog circuit that measures the NSTX axisymmetric eddy current distribution has been designed and constructed. It is based on simple circuit model of the NSTX vacuum vessel that was calibrated using a special axisymmetric eddy current code which was written so that accuracy was maintained in the vicinity of the current filaments. The measurement and the model have been benchmarked against data from numerous vacuum shots and they are in excellent agreement. This is an important measurement that helps give more accurate equilibrium reconstructions

  17. Modelling of NSTX hot vertical displacement events using M 3 D -C 1

    Science.gov (United States)

    Pfefferlé, D.; Ferraro, N.; Jardin, S. C.; Krebs, I.; Bhattacharjee, A.

    2018-05-01

    The main results of an intense vertical displacement event (VDE) modelling activity using the implicit 3D extended MHD code M3D-C1 are presented. A pair of nonlinear 3D simulations are performed using realistic transport coefficients based on the reconstruction of a so-called NSTX frozen VDE where the feedback control was purposely switched off to trigger a vertical instability. The vertical drift phase is solved assuming axisymmetry until the plasma contacts the first wall, at which point the intricate evolution of the plasma, decaying to large extent in force-balance with induced halo/wall currents, is carefully resolved via 3D nonlinear simulations. The faster 2D nonlinear runs allow to assess the sensitivity of the simulations to parameter changes. In the limit of perfectly conducting wall, the expected linear relation between vertical growth rate and wall resistivity is recovered. For intermediate wall resistivities, the halo region contributes to slowing the plasma down, and the characteristic VDE time depends on the choice of halo temperature. The evolution of the current quench and the onset of 3D halo/eddy currents are diagnosed in detail. The 3D simulations highlight a rich structure of toroidal modes, penetrating inwards from edge to core and cascading from high-n to low-n mode numbers. The break-up of flux-surfaces results in a progressive stochastisation of field-lines precipitating the thermalisation of the plasma with the wall. The plasma current then decays rapidly, inducing large currents in the halo region and the wall. Analysis of normal currents flowing in and out of the divertor plate reveals rich time-varying patterns.

  18. The National Spherical Tokamak Experiment at the Princeton Plasma Physics Laboratory

    International Nuclear Information System (INIS)

    1995-12-01

    The Department of Energy (DOE) has prepared an Environmental Assessment (EA), DOE/EA-1108, evaluating the environmental effects of the proposed construction and operation of the National Spherical Tokamak Experiment (NSTX) within the existing C-Stellarator (CS) Building at the Princeton Plasma Physics Laboratory, Princeton, New Jersey. The purpose of the NSTX is to investigate the physics of spherically shaped plasmas as an alternative path to conventional tokamaks for development of fusion energy. Fusion energy has the potential to help compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Construction of the NSTX in the CS Building would require the dismantling and removal of the existing unused Princeton Large Torus (PLT) device, part of which would be reused to construct the NSTX. Based on the analyses in the EA, the DOE has determined that the proposed action does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, 42 U.S.C. 4,321 et seq. The preparation of an Environmental Impact Statement is not required. Thus, the DOE is issuing a FONSI pursuant to the Council on Environmental Quality regulations implementing NEPA (40 CFR Parts 1500--1508) and the DOE NEPA implementing regulations (10 CFR Part 1021)

  19. Magnetic field structure of experimental high beta tokamak equilibria

    International Nuclear Information System (INIS)

    Deniz, A.V.

    1986-01-01

    The magnetic field structure of several low and high β tokamaks in the Columbia High Beta Tokamak (HBT) was determined by high-impedance internal magnetic probes. From the measurement of the magnetic field, the poloidal flux, toroidal flux, toroidal current, and safety factor are calculated. In addition, the plasma position and cross-sectional shape are determined. The extent of the perturbation of the plasma by the probe was investigated and was found to be acceptably small. The tokamaks have major radii of approx.0.24 m, minor radii of approx.0.05 m, toroidal plasma current densities of approx.10 6 A/m 2 , and line-integrated electron densities of approx.10 20 m -2 . The major difference between the low and high β tokamaks is that the high β tokamak was observed to have an outward shift in major radius of both the magnetic center and peak of the toroidal current density. The magnetic center moves inward in major radius after 20 to 30 μsec, presumably because the plasma maintains major radial equilibrium as its pressure decreases from radiation due to impurity atoms. Both the equilibrium and the production of these tokamaks from a toroidal field stabilized z-pinch are modeled computationally. One tokamak evolves from a state with low β features, through a possibly unstable state, to a state with high β features

  20. Kinetic parameters for plasma. beta. -endorphin in lean and obese Zucker rats

    Energy Technology Data Exchange (ETDEWEB)

    Rodd, D.; Farrell, P.A.; Caston, A.L.; Green, M.H. (Department of Exercise and Sport Science, Pennsylvania State University, University Park (USA))

    1991-03-01

    To determine plasma clearance kinetics for {beta}-endorphin (BE) by empirical compartmental analysis, a bolus of radioactive labeled 125I-BE was rapidly injected into a carotid artery catheter of unanesthetized lean (L) and obese (O) Zucker rats. The plasma disappearance of 125I was followed over a 3-h period. A 3-component exponential equation provided the best fit for plasma data. Plasma transit times were very short (10 s); however, plasma fractional catabolic rate was much slower. Plasma mean residence time was similar for both groups (50 min) as was recycle time (1.3 min). These data suggest that BE plasma disappearance kinetics are similar in L and O rats.

  1. High performance deuterium-tritium plasmas in TFTR

    International Nuclear Information System (INIS)

    Sabbagh, S.A.; Bell, M.G.

    1995-03-01

    Plasmas composed of nominally equal concentrations of deuterium and tritium (DT) have been created in TFTR with the goals of producing significant levels of fusion power and of examining the effects of DT fusion alpha particles. Conditioning of the limiter by the injection of lithium pellets has led to an approximate doubling of the energy confinement time, τ E , in supershot plasmas at high plasma current (I p ≤ 2.5 MA) and high heating power (P b ≤ 33 MW). Operation with DT typically results in an additional 20% increase in τ E . In the high poloidal beta, advanced tokamak regime in TFTR, confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.5 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasmas, exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Confinement of alpha particles appears to be classical and losses due to collective effects have not been observed. While small fluctuations in fusion product loss were observed during ELMs, no large loss was detected in DT plasmas

  2. A study of X-divertor in NSTX-U with SOLPS simulations

    Science.gov (United States)

    Chen, Zhong-Ping; Kotschenreuther, Mike; Mahajan, Swadesh; Gerhardt, Stefan

    2018-03-01

    The X-divertor (XD) geometry in NSTX-U is demonstrated, via SOLPS simulations, to perform better than the standard divertor (SD); in particular, it allows detachment at a lower upstream density and stabilizes the detachment front near the target, away from the main X-point. Consequently a stable detached operation becomes possible—the localization near the plate allows a vast reduction of heat fluxes without degrading the core plasma. Indeed, it is confirmed by our simulation that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures, resulting in scrape-off layers that are more favorable for advanced tokamak operation. These advantages are attributed to the unique geometric characteristics of XD—poloidal flaring near the target.

  3. Transport of carbon ion test particles and hydrogen recycling in the plasma of the Columbia tokamak ''HBT'' [High Beta Tokamak

    International Nuclear Information System (INIS)

    Wang, Jian-Hua.

    1990-01-01

    Carbon impurity ion transport is studied in the Columbia High Beta Tokamak (HBT), using a carbon tipped probe which is inserted into the plasma (n e ∼ 1 - 5 x 10 14 (cm -3 ), T e ∼ 4 - 10 (eV), B t ∼ 0.2 - 0.4(T)). Carbon impurity light, mainly the strong lines of C II (4267A, emitted by the C + ions) and C III (4647A, emitted by the C ++ ions), is formed by the ablation or sputtering of plasma ions and by the discharge of the carbon probe itself. The diffusion transport of the carbon ions is modeled by measuring the space-and-time dependent spectral light emission of the carbon ions with a collimated optical beam and photomultiplier. The point of emission can be observed in such a way as to sample regions along and transverse to the toroidal magnetic field. The carbon ion diffusion coefficients are obtained by fitting the data to a diffusion transport model. It is found that the diffusion of the carbon ions is ''classical'' and is controlled by the high collisionality of the HBT plasma; the diffusion is a two-dimensional problem and the expected dependence on the charge of the impurity ion is observed. The measurement of the spatial distribution of the H α emissivity was obtained by inverting the light signals from a 4-channel polychromator, the data were used to calculate the minor-radial influx, the density, and the recycling time of neutral hydrogen atoms or molecules. The calculation shows that the particle recycling time τ p is comparable with the plasma energy confinement time τ E ; therefore, the recycling of the hot plasma ions with the cold neutrals from the walls is one of the main mechanisms for loss of plasma energy

  4. Characteristics of the First H-mode Discharges in NSTX

    International Nuclear Information System (INIS)

    Maingi, R.; Bell, M.G.; Bell, R.E.; Bush, C.E.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; Kugel, H.W.; LeBlanc, B.P.; Menard, J.E.; Mueller, D.; Sabbagh, S.A.; Stutman, D.; Taylor, G.; Johnson, D.W.; Kaita, R.; Maqueda, R.J.; Ono, M.; Paoletti, F.; Peng, Y-K.M.; Roquemore, A.L.; Skinner, C.H.; Soukhanovskii, V.A.; Synakowski, E.J.

    2001-01-01

    We report observations of the first low-to-high (L-H) confinement mode transitions in the National Spherical Torus Experiment (NSTX). The H-mode energy confinement time increased over reference L-mode discharges transiently by 100-300%, as high as ∼150 ms. This confinement time is ∼1.8-2.3 times higher than predicted by a multi-machine ELM-free H-mode scaling. This achievement extends the H-mode window of fusion devices down to a record low aspect ratio (R/a) ∼ 1.3, challenging both confinement and L-H power thresholds scalings based on conventional aspect ratio tokamaks

  5. Diagnostic method for measuring plasma-induced voltages on the PBX-M [Princeton Beta Experiment-Modified] stabilizing shell

    International Nuclear Information System (INIS)

    Kugel, H.W.; Okabayashi, M.; Schweitzer, S.

    1990-07-01

    The Princeton Beta Experiment-Modified (PBX-M) has a close-fitting conducting, passive plate, stabilizing shell which nearly surrounds highly indented, bean-shaped plasmas. The proximity of this electrically isolated shell to a large fraction of the plasma surface allows measurements similar to previous work on other tokamaks using floating probes and limiters. Measurements were performed to characterize the plasma-induced voltages on the PBX-M passive plate stabilizing shell during highplasmas. Voltage differences were measured between the respective passive plate toroidal and poloidal gaps, the respective passive plates and the vessel, and an outer poloidal graphite limiter and its passive plate. The calibration and qualification testing procedures are discussed. The initial measurements found that the largest voltages were observed at plasma start-up and at the plasma current disruption and exhibited characteristics depending on operating conditions. The highest voltages observed have been at disruption and were less than 2 kV. 9 refs., 5 figs

  6. Experimental studies of lithium-based surface chemistry for fusion plasma-facing materials applications

    International Nuclear Information System (INIS)

    Allain, J.P.; Rokusek, D.L.; Harilal, S.S.; Nieto-Perez, M.; Skinner, C.H.; Kugel, H.W.; Heim, B.; Kaita, R.; Majeski, R.

    2009-01-01

    Lithium has enhanced the operational performance of fusion devices such as: TFTR, CDX-U, FTU, T-11 M, and NSTX. Lithium in the solid and liquid state has been studied extensively in laboratory experiments including its erosion and hydrogen-retaining properties. Reductions in physical sputtering up to 40-60% have been measured for deuterated solid and liquid lithium surfaces. Computational modeling indicates that up to a 1:1 deuterium volumetric retention in lithium is possible. This paper presents the results of systematic in situ laboratory experimental studies on the surface chemistry evolution of ATJ graphite under lithium deposition. Results are compared to post-mortem analysis of similar lithium surface coatings on graphite exposed to deuterium discharge plasmas in NSTX. Lithium coatings on plasma-facing components in NSTX have shown substantial reduction of hydrogenic recycling. Questions remain on the role lithium surface chemistry on a graphite substrate has on particle sputtering (physical and chemical) as well as hydrogen isotope recycling. This is particularly due to the lack of in situ measurements of plasma-surface interactions in tokamaks such as NSTX. Results suggest that the lithium bonding state on ATJ graphite is lithium peroxide and with sufficient exposure to ambient air conditions, lithium carbonate is generated. Correlation between both results is used to assess the role of lithium chemistry on the state of lithium bonding and implications on hydrogen pumping and lithium sputtering. In addition, reduction of factors between 10 and 30 reduction in physical sputtering from lithiated graphite compared to pure lithium or carbon is also measured.

  7. Experimental observations of MHD instabilities in the high-beta tokamak Torus-II

    International Nuclear Information System (INIS)

    Machida, M.

    1982-01-01

    The CO 2 laser scattering and interferometry diagnostics have been used to study the MHD instabilities in the high-beta tokamak Torus-II. Detailed measurements of the density and density fluctuation profiles have been performed. In order to measure density fluctuations with wavelengths longer than 2 cm, an interferometric like, phase matching technique has been developed. The toroidal and poloidal mode numbers have been measured using a double-beam, two-position technique. Working at high-beta values, average β greater than or equal to 10%, we have found parameters where the growing instabilities are created or suppressed. The plasma lifetime for both cases is seen to be about the same and the loss of the plasma appears to be caused by the decay in the external fields. The growing instability parameters are within the MHD regime, and it only grows at the outer edge of the plasma. This is in agreement with the theoretical Ballooning mode instability. The frequency and mode number measurements also agree with the Kinetic theory description of Ballooning modes. The comparison with possible other modes, such as Tearing and Drift instabilities, is performed and the Ballooning growth rate is shown to be the best fit to the experimental values

  8. Safety and deterministic failure analyses in high-beta D-D tokamak reactors

    International Nuclear Information System (INIS)

    Selcow, E.C.

    1984-01-01

    Safety and deterministic failure analyses were performed to compare major component failure characteristics for different high-beta D-D tokamak reactors. The primary focus was on evaluating damage to the reactor facility. The analyses also considered potential hazards to the general public and operational personnel. Parametric designs of high-beta D-D tokamak reactors were developed, using WILDCAT as the reference. The size, and toroidal field strength were reduced, and the fusion power increased in an independent manner. These changes were expected to improve the economics of D-D tokamaks. Issues examined using these designs were radiation induced failurs, radiation safety, first wall failure from plasma disruptions, and toroidal field magnet coil failure

  9. Fast wave power flow along SOL field lines in NSTX

    Science.gov (United States)

    Perkins, R. J.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; Leblanc, B. P.; Kramer, G. J.; Phillips, C. K.; Roquemore, L.; Taylor, G.; Wilson, J. R.; Ahn, J.-W.; Gray, T. K.; Green, D. L.; McLean, A.; Maingi, R.; Ryan, P. M.; Jaeger, E. F.; Sabbagh, S.

    2012-10-01

    On NSTX, a major loss of high-harmonic fast wave (HHFW) power can occur along open field lines passing in front of the antenna over the width of the scrape-off layer (SOL). Up to 60% of the RF power can be lost and at least partially deposited in bright spirals on the divertor floor and ceiling [1,2]. The flow of HHFW power from the antenna region to the divertor is mostly aligned along the SOL magnetic field [3], which explains the pattern of heat deposition as measured with infrared (IR) cameras. By tracing field lines from the divertor back to the midplane, the IR data can be used to estimate the profile of HHFW power coupled to SOL field lines. We hypothesize that surface waves are being excited in the SOL, and these results should benchmark advanced simulations of the RF power deposition in the SOL (e.g., [4]). Minimizing this loss is critical optimal high-power long-pulse ICRF heating on ITER while guarding against excessive divertor erosion.[4pt] [1] J.C. Hosea et al., AIP Conf Proceedings 1187 (2009) 105. [0pt] [2] G. Taylor et al., Phys. Plasmas 17 (2010) 056114. [0pt] [3] R.J. Perkins et al., to appear in Phys. Rev. Lett. [0pt] [4] D.L. Green et al., Phys. Rev. Lett. 107 (2011) 145001.

  10. Solid State Neutral Particle Analyzer Array on NSTX

    International Nuclear Information System (INIS)

    Shinohara, K.; Darrow, D.S.; Roquemore, A.L.; Medley, S.S.; Cecil, F.E.

    2004-01-01

    A Solid State Neutral Particle Analyzer (SSNPA) array has been installed on the National Spherical Torus Experiment (NSTX). The array consists of four chords viewing through a common vacuum flange. The tangency radii of the viewing chords are 60, 90, 100, and 120 cm. They view across the three co-injection neutral beam lines (deuterium, 80 keV (typ.) with tangency radii 48.7, 59.2, and 69.4 cm) on NSTX and detect co-going energetic ions. A silicon photodiode used was calibrated by using a mono-energetic deuteron beam source. Deuterons with energy above 40 keV can be detected with the present setup. The degradation of the performance was also investigated. Lead shots and epoxy are used for neutron shielding to reduce handling any hazardous heavy metal. This method also enables us to make an arbitrary shape to be fit into the complex flight tube

  11. [Associations of insulin resistance and pancreatic beta-cell function with plasma glucose level in type 2 diabetes].

    Science.gov (United States)

    Nian, Xiaoping; Sun, Gaisheng; Dou, Chunmei; Hou, Hongbo; Fan, Xiuping; Yu, Hongmei; Ma, Ling; He, Bingxian

    2002-06-10

    To investigate the influence of insulin resistance and pancreatic beta-cell function on plasma glucose level in type 2 diabetes so as to provide theoretical basis for reasonable selection of hypoglycemic agents. The plasma non-specific insulin (NSINS), true insulin (TI) and glucose in eight-one type 2 diabetics, 38 males and 43 females, with a mean age of 53 years, were examined 0, 30, 60 and 120 minutes after they had 75 grams of instant noodles. The patients were divided into two groups according to their fasting plasma glucose (FPG): group A (FPG = 8.89 mmol/L). The insulin resistance was evaluated by HOMA-IR, the beta-cell function was evaluated by HOMA-beta formula and the formula deltaI(30)/deltaG(30) = (deltaI(30)-deltaI(0))/(deltaG(30)-deltaG(0)). The insulin area under curve (INSAUC) was evaluated by the formula INSAUC=FINS/2+INS(30)+INS(60)+INS(120)/2. The mean FPG was 6.23 mmol/L in group A and 12.6 mmol/L in group B. PG2H was 11.7 mmol/L in group A and 19.2 mmol/L in group B. The TI levels in group B at 0, 30, 60, 120 min during standard meal test were significantly higher than those in group A: 6.15 +/- 1.06 vs 4.77 +/- 1.06, 9.76 +/- 1.1 vs 5.88 +/- 1.1,14.68 +/- 1.11 vs 6.87 +/- 1.1 and 17.13 +/- 1.12 vs 8.0 +/- 1.1 microU/dl (all P< 0.01). The NSINS showed the same trend. The insulin resistance in group B was 1.5 times that in group A. With the insulin resistance adjusted, the beta cell function in group A was 5 to 6 times that in group B. The INSAUC in group A was 1.66 times larger than that in group B, especially the INSAUC for true insulin (2 times larger). The contribution of insulin resistance and beta cell function to PG2H was half by half in group A and 1:8 in group B. beta cell function calculated by insulin (Homa-beta) explained 41% of the plasma glucose changes in group A and 54% of the plasma glucose changes in group B. The contribution of insulin deficiency to plasma glocose was 3.3.times that of insulin resistance in group A and was 9

  12. Fueling by coaxial plasma guns

    International Nuclear Information System (INIS)

    Marshall, J.

    1977-01-01

    A review of the operational characteristics of ''snowplow'' and ''deflagration'' coaxial plasma guns is given. The injection of these plasmas into containment fields is discussed. The effect of a background plasma on low-beta injection is mentioned. The use of high-beta injection for reactor plasmas is described

  13. 100% noninductive operation at high beta using off-axis ECCD

    International Nuclear Information System (INIS)

    Murakami, M.; Greenfield, C.M.; Wade, M.R.

    2005-01-01

    The Advanced Tokamak (AT) program on DIII-D is developing the scientific basis for steady-state, high-performance operation in future devices. The key element of the program is to demonstrate sustainment of 100% noninductive current for several seconds at high beta. Guided by integrated modeling, recent experiments using up to 2.5 MW of off-axis electron cyclotron current drive (ECCD) and up to 15 MW neutral beam injection (NBI) with q 95 ∼ 5 have sustained ∼ 100% of the plasma current noninductively for 1 s at high beta (β ∼ 3.6%, β N ∼ 3.4, above the no-wall limit) with q min ≥ 1.5 and good confinement (H 89 ∼ 2.3). Integrated modeling using both empirical and theory-based models is used to design experiments and to interpret their results. These experiments have achieved the parameters required for the ITER Q=5 steady-state scenario, and the same modeling tools are applied to ITER AT scenario development. (author)

  14. 100% NONINDUCTIVE OPERATION AT HIGH BETA USING OFF-AXIS ECCD

    International Nuclear Information System (INIS)

    MURAKAMI, M.; GREENFIELD, C.M.; WADE, M.R.; LUCE, T.C.; FERRON, J.R.; ST JOHN, H.E.; MAKOWSKI, M.A.; AUSTIN, M.E.; ALLEN, S.L.; BRENNAN, D.P.; BURRELL, K.H.; CASPER, T.A.; DeBOO, J.C.; DOYLE, E.J.; GAROFALO, A.M.; GOHIL, P.; GORELOV, I.A.; GOEBNER, R.J.; HOBIRK, J.; HYATT, A.W.; JAYAKUMAR, R.J.; KAJIWARA, K.; KESSEL, C.E.; KINSEY, J.E.; LA HAYE, R.J.; KIM, J.Y.; LAO, L.L.; LOHR, J.; MENARD, J.E.; PETTY, C.C.; PETRIE, T.W.; PINSKER, R.I.; POLITZER, P.A.; PRATER, R.; RHODES, T.L.; SIPS, A.C.C.; STAEBLER, G.M.; TAYLOR, T.S.; WANG, G.; WEST, W.P.; ZENG, L.

    2004-01-01

    The Advanced Tokamak (AT) program on DIII-D is developing the scientific basis for steady-state, high-performance operation in future devices. The key element of the program is to demonstrate sustainment of 100% noninductive current for several seconds at high beta. Guided by integrated modeling, recent experiments using up to 2.5 MW of off-axis electron cyclotron current drive (ECCD) and up to 15 MW neutral beam injection (NBI) with q 95 ∼ 5 have sustained ∼ 100% of the plasma current noninductively for 1 s at high beta (β ∼ 3.6%, β N ∼ 3.4, above the no-wall limit) with q min (ge) 1.5 and good confinement (H 89 ∼ 2.3). Integrated modeling using both empirical and theory-based models is used to design experiments and to interpret their results. These experiments have achieved the parameters required for the ITER Q=5 steady-state scenario, and the same modeling tools are applied to ITER AT scenario development

  15. Effect of ion cyclotron acceleration on frequency chirping beam-driven instabilities in NSTX

    International Nuclear Information System (INIS)

    Ruskov, E.; Heidbrink, W.W.; Fredrickson, E.D.; Darrow, D.; Medley, S.; Gorelenkov, N.

    2006-01-01

    The fast-ion distribution function in the National Spherical Torus Experiment (NSTX) is modified from shot to shot while keeping the total injected power at ∼2 MW. Deuterium beams of different energy and tangency radius are injected into helium L-mode plasmas, producing a rich set of instabilities, including TAE modes, 50-100∼kHz instabilities with rapid frequency sweeps or chirps, and strong, low frequency (10-20 kHz) fishbones. The experiment was motivated by a theory that attributes frequency chirping to the formation of holes and clumps in phase space. In the theory, increasing the effective collision frequency of the fast ions that drive the instability can suppress frequency chirping. In the experiment, high-power (∼3 MW) harmonic fast wave (HHFW) heating accelerates the fast ions in an attempt to alter the effective collision frequency. Steady-frequency TAE modes excited early in the discharge are affected by the HHFW heating but there is no evidence that the chirping of 20-100 kHz modes is suppressed. (author)

  16. Effect of Ion Cyclotron Acceleration on Frequency Chirping Beam-Driven Instabilities in NSTX

    International Nuclear Information System (INIS)

    Ruskov, E.; Heidbrink, W.W.; Fredrickson, E.D.; Darrow, D.; Medley, S.; Gorelenkov, N.

    2006-01-01

    The fast-ion distribution function in the National Spherical Torus Experiment (NSTX) is modified from shot to shot while keeping the total injected power at ∼2 MW. Deuterium beams of different energy and tangency radius are injected into helium L-mode plasmas, producing a rich set of instabilities, including TAE modes, 50-100∼kHz instabilities with rapid frequency sweeps or chirps, and strong, low frequency (10-20 kHz) fishbones. The experiment was motivated by a theory that attributes frequency chirping to the formation of holes and clumps in phase space. In the theory, increasing the effective collision frequency of the fast ions that drive the instability can suppress frequency chirping. In the experiment, high-power (∼3 MW) harmonic fast wave (HHFW) heating accelerates the fast ions in an attempt to alter the effective collision frequency. Steady-frequency TAE modes excited early in the discharge are affected by the HHFW heating but there is no evidence that the chirping of 20-100 kHz modes is suppressed. (author)

  17. Proposal for the construction of a High-Beta Tokamak at LASL

    International Nuclear Information System (INIS)

    Van der Laan, P.C.T.; Freidberg, J.P.; Thomas, K.S.

    1976-06-01

    The large heating rate inherent to implosion heating allows the rapid generation of high-beta tokamak plasmas. A study of these plasmas in the proposed HBT machine can give information on how MHD equilibrium and stability limit β and q. Both a wide current profile and a moderate elongation of the minor cross section should help to raise the permissible peak β values in HBT to at least 20 percent. The longer term loss processes occurring in MHD-stable plasmas are to be investigated. The main parameters of HBT are: R = 0.30 m, minor cross section a racetrack of width and height 0.24 m and 0.48 m, B/sub phi/ = 2 T, I/sub phi/ approximately 750 kA

  18. Fast ion loss diagnostic plans for NSTX

    International Nuclear Information System (INIS)

    Darrow, D. S.; Bell, R.; Johnson, R.; Kugel, H.; Wilson, J. R.; Cecil, F. E.; Maingi, R.; Krasilnikov, A.; Alekseyev, A.

    2000-01-01

    The prompt loss of neutral beam ions from the National Spherical Torus Experiment (NSTX) is expected to be between 12% and 42% of the total 5 MW of beam power. There may, in addition, be losses of fast ions arising from high harmonic fast wave (HHFW) heating. Most of the lost ions will strike the HHFW antenna or the neutral beam dump. To measure these losses in the 2000 experimental campaign, thermocouples in the antenna, several infrared camera views, and a Faraday cup lost ion probe will be employed. The probe will measure loss of fast ions with E > 1 keV at three radial locations, giving the scrape-off length of the fast ions

  19. Plasma characteristics of a high power helicon discharge

    International Nuclear Information System (INIS)

    Ziemba, T; Euripides, P; Slough, J; Winglee, R; Giersch, L; Carscadden, J; Schnackenberg, T; Isley, S

    2006-01-01

    A new high power helicon (HPH) plasma system has been designed to provide input powers of several tens of kilowatts to produce a large area (0.5 m 2 ) of uniform high-density, of at least 5 x 10 17 m -3 , plasma downstream from the helicon coil. Axial and radial plasma characteristics show that the plasma is to a lesser extent created in and near the helicon coil and then is accelerated into the axial and equatorial regions. The bulk acceleration of the plasma is believed to be due to a coupling of the bulk of the electrons to the helicon field, which in turn transfers energy to the ions via ambipolar diffusion. The plasma beta is near unity a few centimetres away from the HPH system and Bdot measurements show ΔB perturbations in the order of the vacuum magnetic field magnitude. In the equatorial region, a magnetic separatrix is seen to develop roughly at the mid-point between the helicon and chamber wall. The magnetic perturbation develops on the time scale of the plasma flow speed and upon the plasma reaching the chamber wall decays to the vacuum magnetic field configuration within 200 μs

  20. Importance of Plasma Response to Non-axisymmetric Perturbations in Tokamaks

    International Nuclear Information System (INIS)

    Park, Jong-kyu; Boozer, Allen H.; Menard, Jonathan E.; Garofalo, Andrea M.; Schaffer, Michael J.; Hawryluk, Richard J.; Kaye, Stanley M.; Gerhardt, Stefan P.; Sabbagh, Steve A. and the NSTX Team

    2009-01-01

    Tokamaks are sensitive to deviations from axisymmetry as small as (delta)B/B 0 ∼ 10 -4 . These non-axisymmetric perturbations greatly modify plasma confinement and performance by either destroying magnetic surfaces with subsequent locking or deforming magnetic surfaces with associated non-ambipolar transport. The Ideal Perturbed Equilibrium Code (IPEC) calculates ideal perturbed equilibria and provides important basis for understanding the sensitivity of tokamak plasmas to perturbations. IPEC calculations indicate that the ideal plasma response, or equivalently the effect by ideally perturbed plasma currents, is essential to explain locking experiments on National Spherical Torus eXperiment (NSTX) and DIII-D. The ideal plasma response is also important for Neoclassical Toroidal Viscosity (NTV) in non-ambipolar transport. The consistency between NTV theory and magnetic braking experiments on NSTX and DIII-D can be improved when the variation in the field strength in IPEC is coupled with generalized NTV theory. These plasma response effects will be compared with the previous vacuum superpositions to illustrate the importance. However, plasma response based on ideal perturbed equilibria is still not sufficiently accurate to predict the details of NTV transport, and can be inconsistent when currents associated with a toroidal torque become comparable to ideal perturbed currents

  1. Production and analysis of thermonuclear plasmas in high beta devices. Progress report

    International Nuclear Information System (INIS)

    1976-01-01

    During the October 1975--July 1976 reporting period, significant progress was reported in all ongoing projects funded by ERDA. Construction of the major new experiment THOR was delayed due to difficulties in bringing the Maxwell Laboratory swinging LC-pulse generators up to specifications. These technical difficulties have now been overcome and the pulsers were accepted late in April. THOR is almost back on schedule and physics results on plasma heating are expected by September of this year. TERP, originally funded as an Exploratory Concept, has been operating successfully for over a year and given confinement physics results which are important for the development of Maximum Beta Tokamaks. The Measurement and Instrumentation efforts have resulted in a number of instrument developments that have been successfully tested on FTP or STP, our fast or small theta pinches. Both FTP and STP were studied intensively to obtain ion and electron heating rates and anomalous post-implosion resistivities. These results are supported by the University of Maryland theory and simulation programs and SAI

  2. Response of NSTX liquid lithium divertor to high heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Abrams, T., E-mail: tabrams@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaworski, M.A. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Kallman, J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Foley, E.L. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kugel, H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Levinton, F. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2013-07-15

    Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ∼1.5 MW/m{sup 2} for 1–3 s. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the “bare” sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface.

  3. Modeling detachment physics in the NSTX snowflake divertor

    Energy Technology Data Exchange (ETDEWEB)

    Meier, E.T., E-mail: emeier@wm.edu [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Bell, R.E.; Diallo, A.; Kaita, R.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Podestà, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); Rognlien, T.D.; Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

    2015-08-15

    The snowflake divertor is a proposed technique for coping with the tokamak power exhaust problem in next-step experiments and eventually reactors, where extreme power fluxes to material surfaces represent a leading technological and physics challenge. In lithium-conditioned National Spherical Torus Experiment (NSTX) discharges, application of the snowflake divertor typically induced partial outer divertor detachment and severalfold heat flux reduction. UEDGE is used to analyze and compare conventional and snowflake divertor configurations in NSTX. Matching experimental upstream profiles and divertor measurements in the snowflake requires target recycling of 0.97 vs. 0.91 in the conventional case, implying partial saturation of the lithium-based pumping mechanism. Density scans are performed to analyze the mechanisms that facilitate detachment in the snowflake, revealing that increased divertor volume provides most of the parallel heat flux reduction. Also, neutral gas power loss is magnified by the increased wetted area in the snowflake, and plays a key role in generating volumetric recombination.

  4. A stable route to high-{beta}{sub p} plasmas with non-monotonic q-profiles

    Energy Technology Data Exchange (ETDEWEB)

    Soeldner, F X; Baranov, Y; Bhatnagar, V P; Bickley, A J; Challis, C D; Fischer, B; Gormezano, C; Huysmans, G T.A.; Kerner, W; Rimini, F; Sips, A C.C.; Springmann, R; Taroni, A [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Goedbloed, J P; Holties, H A [Institute for Plasmas Physics, Nieuwegein (Netherlands); Parail, V V; Pereverzev, G V [Kurchatov Institute of Atomic Energy, Moscow (Russian Federation)

    1994-07-01

    Steady-state operation of tokamak reactors seems feasible in so-called Advanced Scenarios with high bootstrap current in high-beta{sub p} operation. The stabilization of such discharges with noninductive profile control will be attempted on JET in pursuit of previous high bootstrap current studies. Results of modelling studies of full noninductive current drive scenarios in JET and ITER are presented. Fast Waves (FW), Lower Hybrid (LH) Waves and Neutral Beam Injection (NBI) are used for heating and current drive, alternatively or in combination. A stable route to nonmonotonic q-profiles has been found with a specific ramp-up scenario which combines LH-current drive (LHCD) and a fast Ohmic ramp-up. A hollow current profile with deep shear reversal over the whole central region is thereby formed in an early low-beta phase and frozen in by additional heating. (authors). 5 refs., 4 figs.

  5. Temperature gradient driven electron transport in NSTX and Tore Supra

    International Nuclear Information System (INIS)

    Horton, W.; Wong, H.V.; Morrison, P.J.; Wurm, A.; Kim, J.H.; Perez, J.C.; Pratt, J.; Hoang, G.T.; LeBlanc, B.P.; Ball, R.

    2005-01-01

    Electron thermal fluxes are derived from the power balance for Tore Supra (TS) and NSTX discharges with centrally deposited fast wave electron heating. Measurements of the electron temperature and density profiles, combined with ray tracing computations of the power absorption profiles, allow detailed interpretation of the thermal flux versus temperature gradient. Evidence supporting the occurrence of electron temperature gradient turbulent transport in the two confinement devices is found. With control of the magnetic rotational transform profile and the heating power, internal transport barriers are created in TS and NSTX discharges. These partial transport barriers are argued to be a universal feature of transport equations in the presence of invariant tori that are intrinsic to non-monotonic rotational transforms in dynamical systems

  6. Measurement of high-beta tokamak pressure profiles with multipoint Thomson scattering

    International Nuclear Information System (INIS)

    Levinton, F.M.

    1983-01-01

    A multipoint Thomson-scattering system has been developed to obtain pressure profiles along the major radius of Torus II, a high-beta tokamak. The profiles obtained during the 20 to 25 μs lifetime of the discharge indicates that the plasma has a peak temperature of 80 eV and density of 1.0 x 10 15 cm - 3 . The profiles remain fairly constant during this time until the equilibrium is lost, after which the temperature and density decays to 10 eV and 10 14 cm - 3 very quickly (approx. 1 μs). Experimental results show Torus II has a high-beta ( approx. 10%) equilibrium, with a strong shift of the peak of the pressure profile towards the outside. Numerical results from a 2-D free boundary MHD equilibrium code have obtained equilibria which closely approximate the experimentally measured profiles

  7. High normal fasting glucose level in obese youth: a marker for insulin resistance and beta cell dysregulation.

    LENUS (Irish Health Repository)

    O'Malley, G

    2010-06-01

    A high but normal fasting plasma glucose level in adults is a risk factor for future development of type 2 diabetes mellitus and cardiovascular disease. We investigated whether normal fasting plasma glucose levels (<5.60 mmol\\/l) are associated with decreases in insulin sensitivity and beta cell function, as well as an adverse cardiovascular profile in obese youth.

  8. Effect of the scrape-off layer in AORSA full wave simulations of fast wave minority, mid/high harmonic, and helicon heating regimes

    Energy Technology Data Exchange (ETDEWEB)

    Bertelli, N., E-mail: nbertell@pppl.gov; Gerhardt, S.; Hosea, J. C.; LeBlanc, B.; Perkins, R. J.; Phillips, C. K.; Taylor, G.; Valeo, E. J.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaeger, E. F. [XCEL Engineering Inc., Oak Ridge, TN 37830 (United States); Lau, C.; Blazevski, D.; Green, D. L.; Berry, L.; Ryan, P. M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Bonoli, P. T.; Wright, J. C. [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Pinsker, R. I.; Prater, R. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Qin, C. M. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); and others

    2015-12-10

    Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves, have found strong interactions between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 3D AORSA results for the National Spherical Torus eXperiment (NSTX), where a full antenna spectrum is reconstructed, are shown, confirming the same behavior found for a single toroidal mode results in Bertelli et al, Nucl. Fusion, 54 083004, 2014, namely, a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is moved away from in front of the antenna by increasing the edge density. Additionally, full wave simulations have been extended to “conventional” tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for Alcator C-Mod and EAST, which operate in the minority heating regime unlike NSTX/NSTX-U and DIII-D, which operate in the mid/high harmonic regime. A substantial discussion of some of the main aspects, such as (i) the pitch angle of the magnetic field; (ii) minority heating vs. mid/high harmonic regimes is presented showing the different behavior of the RF field in the SOL region for NSTX-U scenarios with different plasma current. Finally, the preliminary results of the impact of the SOL region on the evaluation of the helicon current drive efficiency in DIII-D is presented for the first time and briefly compared with the different regimes

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

    International Nuclear Information System (INIS)

    Maingi, R.; Bell, M.; Bell, R.; Biewer, T.; Bush, C.; Chang, C.S.; Gates, D.; Kaye, S.; Kugel, H.; LeBlanc, B.; Maqueda, R.; Menard, J.; Mueller, D.; Raman, R.; Sabbagh, S.; Soukhanovskii, V.

    2003-01-01

    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

  10. Response to “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh

    2014-01-01

    Relying on coil positions relative to the plasma, the “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the “proximity condition,” used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors

  11. Response to “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh [Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States)

    2014-05-15

    Relying on coil positions relative to the plasma, the “Comment on ‘Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake’ ” [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the “proximity condition,” used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors.

  12. Vortices, Reconnection and Turbulence in High Electron-Beta Plasmas

    International Nuclear Information System (INIS)

    Stenzel, R. L.

    2004-01-01

    Plasmas in which the kinetic energy exceeds the magnetic energy by a significant factor are common in space and in the laboratory. Such plasmas can convect magnetic fields and create null points in whose vicinity first the ions become unmagnetized, then the electrons. This project focuses on the detailed study of the transition regime of these plasmas

  13. Thomson scattering measurements on the high beta pinch Extrap-T1

    International Nuclear Information System (INIS)

    Karlsson, P.

    1989-11-01

    Electron temperature and density measurement on a high beta discharge in the Extrap-T1 device have been performed with Thomson scattering. It was found that the signal levels were low and the plasma background radiation high. The spread of the measured temperatures and densities was large. A computer code was developed to investigate whether this spread in measured temperatures was due to shot to shot variations or to photon statistics. The code showed that the scattered data could be explained by photon statistics

  14. High-beta experiments with neutral-beam injection on PDX

    International Nuclear Information System (INIS)

    Johnson, D.; Bell, M.; Bitter, M.

    1983-01-01

    Experimental investigations of high-beta plasmas produced in PDX with near-perpendicular neutral-beam injection are reported. Systematic power scans have been performed over a wide range of toroidal fields (νsub(T)q.7 T< Bsub(T)<2.2 T) and plasma currents (200 kA< Isub(p)<500 kA). At high toroidal fields, the change in total stored energy due to beam injection increases linearly with input power and also increases with plasma current. At lower toroidal fields and low injection power levels, the stored energy also increases with power and plasma current. However, at high power and low toroidal fields, a saturation in heating is observed. This result suggests the onset of a νsub(T) limit for circular cross-section tokamaks with near-perpendicular injection. Scaling experiments indicate that this νsub(T) limit increases with rising 1/q. Values of νsub(T)approx.=3% at qsub(PSI)=1.8 have been achieved. At high values of νsub(T)q, short bursts of MHD activity are observed, synchronized with sharply increased fluxes of perpendicular charge-exchange neutrals and rapid decreases in the rate of beam-driven neutron production. When strong bursts occur, there is a significant depletion of the fast-ion population. Estimates of the fast-ion loss indicate that it could explain the observed decrease in heating, although an additional reduction in thermal-plasma confinement cannot be ruled out. Numerical studies using measured pressure profiles predict that the equilibria obtained become unstable to the ideal n=1 internal mode, at about the same value of 0 where the new fluctuations are observed. (author)

  15. Edge Turbulence Imaging on NSTX and Alcator C-Mod

    International Nuclear Information System (INIS)

    S.J. Zweben; R.A. Maqueda; J.L. Terry; B. Bai; C.J. Boswell; C.E. Bush; D. D'Ippolito; E.D. Fredrickson; M. Greenwald; K. Hallatschek; S. Kaye; B. LaBombard; R. Maingi; J. Myra; W.M. Nevins; B.N. Rogers; D.P. Stotler; J. Wilgen; and X.Q. Xu

    2002-01-01

    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

  16. Measurements of Prompt and MHD-Induced Fast Ion Loss from National Spherical Torus Experiment Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    D.S. Darrow; S.S. Medley; A.L. Roquemore; W.W. Heidbrink; A. Alekseyev; F.E. Cecil; J. Egedal; V.Ya. Goloborod' ko; N.N. Gorelenkov; M. Isobe; S. Kaye; M. Miah; F. Paoletti; M.H. Redi; S.N. Reznik; A. Rosenberg; R. White; D. Wyatt; V.A. Yavorskij

    2002-10-15

    A range of effects may make fast ion confinement in spherical tokamaks worse than in conventional aspect ratio tokamaks. Data from neutron detectors, a neutral particle analyzer, and a fast ion loss diagnostic on the National Spherical Torus Experiment (NSTX) indicate that neutral beam ion confinement is consistent with classical expectations in quiescent plasmas, within the {approx}25% errors of measurement. However, fast ion confinement in NSTX is frequently affected by magnetohydrodynamic (MHD) activity, and the effect of MHD can be quite strong.

  17. Multi-energy soft-x-ray technique for impurity transport measurements in the fusion plasma edge

    International Nuclear Information System (INIS)

    Clayton, D J; Tritz, K; Stutman, D; Finkenthal, M; Kumar, D; Kaye, S M; LeBlanc, B P; Paul, S; Sabbagh, S A

    2012-01-01

    A new diagnostic technique was developed to produce high-resolution impurity transport measurements of the steep-gradient edge of fusion plasmas. Perturbative impurity transport measurements were performed for the first time in the NSTX plasma edge (r/a ∼ 0.6 to the SOL) with short neon gas puffs, and the resulting line and continuum emission was measured with the new edge multi-energy soft-x-ray (ME-SXR) diagnostic. Neon transport is modeled with the radial impurity transport code STRAHL and the resulting x-ray emission is computed using the ADAS atomic database. The radial transport coefficient profiles D(r) and v(r), and the particle flux from the gas puff Φ(t), are the free parameters in this model and are varied to find the best fit to experimental x-ray emissivity measurements, with bolometry used to constrain the impurity source. Initial experiments were successful and results were consistent with previous measurements of core impurity transport and neoclassical transport calculations. New diagnostic tools will be implemented on NSTX-U to further improve these transport measurements. (paper)

  18. THz Backward-wave oscillators for plasma diagnostic in nuclear fusion

    OpenAIRE

    Paoloni, Claudio; Yue, Lingna; Tang, Xiaopin; Zhang, Fuzhi; Popovic, Branko; Himes, Logan; Barchfeld, Robert; Gamzina, Diana; Mineo, Mauro; Letizia, Rosa; Luhmann Jr., Neville C.

    2015-01-01

    Summary form only given. The understanding of plasma turbulence in nuclear fusion is related to the availability of powerful THz sources and the possibility to map wider plasma regions. A novel approach to realize compact THz sources to be implemented in the plasma diagnostic at NSTX experiment (Princeton Plasma Physics Laboratory, USA) is reported.Two novel 0.346 THz Backward-Wave Oscillators (BWOs) have been designed and are presently in the fabrication phase. One BWO is based on the Double...

  19. Plasma characteristics of a high power helicon discharge

    Energy Technology Data Exchange (ETDEWEB)

    Ziemba, T; Euripides, P; Slough, J; Winglee, R; Giersch, L; Carscadden, J; Schnackenberg, T; Isley, S [Box 351310, University of Washington, Seattle WA, 98195 (United States)

    2006-08-01

    A new high power helicon (HPH) plasma system has been designed to provide input powers of several tens of kilowatts to produce a large area (0.5 m{sup 2}) of uniform high-density, of at least 5 x 10{sup 17} m{sup -3}, plasma downstream from the helicon coil. Axial and radial plasma characteristics show that the plasma is to a lesser extent created in and near the helicon coil and then is accelerated into the axial and equatorial regions. The bulk acceleration of the plasma is believed to be due to a coupling of the bulk of the electrons to the helicon field, which in turn transfers energy to the ions via ambipolar diffusion. The plasma beta is near unity a few centimetres away from the HPH system and Bdot measurements show {delta}B perturbations in the order of the vacuum magnetic field magnitude. In the equatorial region, a magnetic separatrix is seen to develop roughly at the mid-point between the helicon and chamber wall. The magnetic perturbation develops on the time scale of the plasma flow speed and upon the plasma reaching the chamber wall decays to the vacuum magnetic field configuration within 200 {mu}s.

  20. Neutral Particle Analyzer Diagnostic on NSTX

    International Nuclear Information System (INIS)

    Medley, S.S.; Roquemore, A.L.

    2004-01-01

    The Neutral Particle Analyzer (NPA) diagnostic on the National Spherical Torus Experiment (NSTX) utilizes a PPPL-designed E||B spectrometer that measures the energy spectra of minority hydrogen and bulk deuterium species simultaneously with 39 energy channels per mass specie and a time resolution of 1 ms. The calibrated energy range is E = 0.5-150 keV and the energy resolution varies from AE/E = 3-7% over the surface of the microchannel plate detector

  1. Neutral Particle Analyzer Diagnostic on NSTX

    Energy Technology Data Exchange (ETDEWEB)

    S.S. Medley; A.L. Roquemore

    2004-03-16

    The Neutral Particle Analyzer (NPA) diagnostic on the National Spherical Torus Experiment (NSTX) utilizes a PPPL-designed E||B spectrometer that measures the energy spectra of minority hydrogen and bulk deuterium species simultaneously with 39 energy channels per mass specie and a time resolution of 1 ms. The calibrated energy range is E = 0.5-150 keV and the energy resolution varies from AE/E = 3-7% over the surface of the microchannel plate detector.

  2. 2D full-wave simulation of waves in space and tokamak plasmas

    Directory of Open Access Journals (Sweden)

    Kim Eun-Hwa

    2017-01-01

    Full Text Available Simulation results using a 2D full-wave code (FW2D for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF waves in the scape-off layer (SOL of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

  3. 2D full-wave simulation of waves in space and tokamak plasmas

    Science.gov (United States)

    Kim, Eun-Hwa; Bertelli, Nicola; Johnson, Jay; Valeo, Ernest; Hosea, Joel

    2017-10-01

    Simulation results using a 2D full-wave code (FW2D) for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry) and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC) waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF) waves in the scape-off layer (SOL) of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

  4. Investigation of MHD instabilities and control in KSTAR preparing for high beta operation

    Science.gov (United States)

    Park, Y. S.; Sabbagh, S. A.; Bialek, J. M.; Berkery, J. W.; Lee, S. G.; Ko, W. H.; Bak, J. G.; Jeon, Y. M.; Park, J. K.; Kim, J.; Hahn, S. H.; Ahn, J.-W.; Yoon, S. W.; Lee, K. D.; Choi, M. J.; Yun, G. S.; Park, H. K.; You, K.-I.; Bae, Y. S.; Oh, Y. K.; Kim, W.-C.; Kwak, J. G.

    2013-08-01

    Initial H-mode operation of the Korea Superconducting Tokamak Advanced Research (KSTAR) is expanded to higher normalized beta and lower plasma internal inductance moving towards design target operation. As a key supporting device for ITER, an important goal for KSTAR is to produce physics understanding of MHD instabilities at long pulse with steady-state profiles, at high normalized beta, and over a wide range of plasma rotation profiles. An advance from initial plasma operation is a significant increase in plasma stored energy and normalized beta, with Wtot = 340 kJ, βN = 1.9, which is 75% of the level required to reach the computed ideal n = 1 no-wall stability limit. The internal inductance was lowered to 0.9 at sustained H-mode duration up to 5 s. In ohmically heated plasmas, the plasma current reached 1 MA with prolonged pulse length up to 12 s. Rotating MHD modes are observed in the device with perturbations having tearing rather than ideal parity. Modes with m/n = 3/2 are triggered during the H-mode phase but are relatively weak and do not substantially reduce Wtot. In contrast, 2/1 modes to date only appear when the plasma rotation profiles are lowered after H-L back-transition. Subsequent 2/1 mode locking creates a repetitive collapse of βN by more than 50%. Onset behaviour suggests the 3/2 mode is close to being neoclassically unstable. A correlation between the 2/1 mode amplitude and local rotation shear from an x-ray imaging crystal spectrometer suggests that the rotation shear at the mode rational surface is stabilizing. As a method to access the ITER-relevant low plasma rotation regime, plasma rotation alteration by n = 1, 2 applied fields and associated neoclassical toroidal viscosity (NTV) induced torque is presently investigated. The net rotation profile change measured by a charge exchange recombination diagnostic with proper compensation of plasma boundary movement shows initial evidence of non-resonant rotation damping by the n = 1, 2 applied

  5. Investigation of MHD instabilities and control in KSTAR preparing for high beta operation

    International Nuclear Information System (INIS)

    Park, Y.S.; Sabbagh, S.A.; Bialek, J.M.; Berkery, J.W.; Lee, S.G.; Ko, W.H.; Bak, J.G.; Jeon, Y.M.; Kim, J.; Hahn, S.H.; Yoon, S.W.; Lee, K.D.; You, K.-I.; Bae, Y.S.; Oh, Y.K.; Park, J.K.; Ahn, J.-W.; Choi, M.J.; Yun, G.S.; Park, H.K.

    2013-01-01

    Initial H-mode operation of the Korea Superconducting Tokamak Advanced Research (KSTAR) is expanded to higher normalized beta and lower plasma internal inductance moving towards design target operation. As a key supporting device for ITER, an important goal for KSTAR is to produce physics understanding of MHD instabilities at long pulse with steady-state profiles, at high normalized beta, and over a wide range of plasma rotation profiles. An advance from initial plasma operation is a significant increase in plasma stored energy and normalized beta, with W tot = 340 kJ, β N = 1.9, which is 75% of the level required to reach the computed ideal n = 1 no-wall stability limit. The internal inductance was lowered to 0.9 at sustained H-mode duration up to 5 s. In ohmically heated plasmas, the plasma current reached 1 MA with prolonged pulse length up to 12 s. Rotating MHD modes are observed in the device with perturbations having tearing rather than ideal parity. Modes with m/n = 3/2 are triggered during the H-mode phase but are relatively weak and do not substantially reduce W tot . In contrast, 2/1 modes to date only appear when the plasma rotation profiles are lowered after H–L back-transition. Subsequent 2/1 mode locking creates a repetitive collapse of β N by more than 50%. Onset behaviour suggests the 3/2 mode is close to being neoclassically unstable. A correlation between the 2/1 mode amplitude and local rotation shear from an x-ray imaging crystal spectrometer suggests that the rotation shear at the mode rational surface is stabilizing. As a method to access the ITER-relevant low plasma rotation regime, plasma rotation alteration by n = 1, 2 applied fields and associated neoclassical toroidal viscosity (NTV) induced torque is presently investigated. The net rotation profile change measured by a charge exchange recombination diagnostic with proper compensation of plasma boundary movement shows initial evidence of non-resonant rotation damping by the n = 1, 2

  6. Easy web interfaces to IDL code for NSTX Data Analysis

    International Nuclear Information System (INIS)

    Davis, W.M.

    2012-01-01

    Highlights: ► Web interfaces to IDL code can be developed quickly. ► Dozens of Web Tools are used effectively on NSTX for Data Analysis. ► Web interfaces are easier to use than X-window applications. - Abstract: Reusing code is a well-known Software Engineering practice to substantially increase the efficiency of code production, as well as to reduce errors and debugging time. A variety of “Web Tools” for the analysis and display of raw and analyzed physics data are in use on NSTX [1], and new ones can be produced quickly from existing IDL [2] code. A Web Tool with only a few inputs, and which calls an IDL routine written in the proper style, can be created in less than an hour; more typical Web Tools with dozens of inputs, and the need for some adaptation of existing IDL code, can be working in a day or so. Efficiency is also increased for users of Web Tools because of the familiar interface of the web browser, and not needing X-windows, or accounts and passwords, when used within our firewall. Web Tools were adapted for use by PPPL physicists accessing EAST data stored in MDSplus with only a few man-weeks of effort; adapting to additional sites should now be even easier. An overview of Web Tools in use on NSTX, and a list of the most useful features, is also presented.

  7. A Model for Straight and Helical Solar Jets: II. Parametric Study of the Plasma Beta

    Science.gov (United States)

    Pariat, E.; Dalmasse, K.; DeVore, C. R.; Antiochos, S. K.; Karpen, J. T.

    2016-01-01

    Context. Jets are dynamic, impulsive, well-collimated plasma events that develop at many different scales and in different layers of the solar atmosphere. Aims. Jets are believed to be induced by magnetic reconnection, a process central to many astrophysical phenomena. Within the solar atmosphere, jet-like events develop in many different environments, e.g. in the vicinity of active regions as well as in coronal holes, and at various scales, from small photospheric spicules to large coronal jets. In all these events, signatures of helical structure and/or twisting/rotating motions are regularly observed. The present study aims to establish that a single model can generally reproduce the observed properties of these jet-like events. Methods. In this study, using our state-of-the-art numerical solver ARMS, we present a parametric study of a numerical tridimensional magnetohydrodynamic (MHD) model of solar jet-like events. Within the MHD paradigm, we study the impact of varying the atmospheric plasma beta on the generation and properties of solar-like jets. Results. The parametric study validates our model of jets for plasma beta ranging from 10(sup 3) to 1, typical of the different layers and magnetic environments of the solar atmosphere. Our model of jets can robustly explain the generation of helical solar jet-like events at various beta less than or equal to 1. We show that the plasma beta modifies the morphology of the helical jet, explaining the different observed shapes of jets at different scales and in different layers of the solar atmosphere. Conclusions. Our results allow us to understand the energisation, triggering, and driving processes of jet-like events. Our model allows us to make predictions of the impulsiveness and energetics of jets as determined by the surrounding environment, as well as the morphological properties of the resulting jets.

  8. Princeton Plasma Physics Laboratory Annual Site Environmental Report for Calendar Year 1999

    International Nuclear Information System (INIS)

    Finley, Virginia

    2001-01-01

    The results of the 1999 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and non-radioactive pollutants (if any) that are added to the environment as a result of PPPL's operations. The report also summarizes environmental initiatives, assessments, and programs that were undertaken in 1999. The Princeton Plasma Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program is to create innovations to make fusion power a practical reality--an alternative energy source. 1999 marked the first year of National Spherical Torus Experiment (NSTX) operations and Tokamak Fusion Test Reactor (TFTR) dismantlement and deconstruction activities. A collaboration among fourteen national laboratories, universities, and research institutions, the NSTX is a major element in the U.S. Fusion Energy Sciences Program. It has been designed to test the physics principles of spherical torus (ST) plasmas. The ST concept could play an important role in the development of smaller, more economical fusion reactors. With its completion within budget and ahead of its target schedule, NSTX first plasma occurred on February 12, 1999. The 1999 performance of the Princeton Plasma Physics Laboratory was rated ''outstanding'' by the U.S. Department of Energy in the Laboratory Appraisal report issued early in 2000. The report cited the Laboratory's consistently excellent scientific and technological achievements, its successful management practices, and included high marks in a host of other areas including environmental management, employee health and safety, human resources administration, science education, and communications. Groundwater investigations continued under a voluntary agreement with the New Jersey

  9. Princeton Plasma Physics Laboratory Annual Site Environmental Report for Calendar Year 1999

    Energy Technology Data Exchange (ETDEWEB)

    Virginia Finley

    2001-04-20

    The results of the 1999 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and non-radioactive pollutants (if any) that are added to the environment as a result of PPPL's operations. The report also summarizes environmental initiatives, assessments, and programs that were undertaken in 1999. The Princeton Plasma Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program is to create innovations to make fusion power a practical reality--an alternative energy source. 1999 marked the first year of National Spherical Torus Experiment (NSTX) operations and Tokamak Fusion Test Reactor (TFTR) dismantlement and deconstruction activities. A collaboration among fourteen national laboratories, universities, and research institutions, the NSTX is a major element in the U.S. Fusion Energy Sciences Program. It has been designed to test the physics principles of spherical torus (ST) plasmas. The ST concept could play an important role in the development of smaller, more economical fusion reactors. With its completion within budget and ahead of its target schedule, NSTX first plasma occurred on February 12, 1999. The 1999 performance of the Princeton Plasma Physics Laboratory was rated ''outstanding'' by the U.S. Department of Energy in the Laboratory Appraisal report issued early in 2000. The report cited the Laboratory's consistently excellent scientific and technological achievements, its successful management practices, and included high marks in a host of other areas including environmental management, employee health and safety, human resources administration, science education, and communications. Groundwater investigations continued under a voluntary

  10. Time-dependent analysis of visible helium line-ratios for electron temperature and density diagnostic using synthetic simulations on NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Muñoz Burgos, J. M., E-mail: jmunozbu@pppl.gov; Stutman, D.; Tritz, K. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, Maryland 21218 (United States); Barbui, T.; Schmitz, O. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States)

    2016-11-15

    Helium line-ratios for electron temperature (T{sub e}) and density (n{sub e}) plasma diagnostic in the Scrape-Off-Layer (SOL) and edge regions of tokamaks are widely used. Due to their intensities and proximity of wavelengths, the singlet, 667.8 and 728.1 nm, and triplet, 706.5 nm, visible lines have been typically preferred. Time-dependency of the triplet line (706.5 nm) has been previously analyzed in detail by including transient effects on line-ratios during gas-puff diagnostic applications. In this work, several line-ratio combinations within each of the two spin systems are analyzed with the purpose of eliminating transient effects to extend the application of this powerful diagnostic to high temporal resolution characterization of plasmas. The analysis is done using synthetic emission modeling and diagnostic for low electron density NSTX SOL plasma conditions by several visible lines. Quasi-static equilibrium and time-dependent models are employed to evaluate transient effects of the atomic population levels that may affect the derived electron temperatures and densities as the helium gas-puff penetrates the plasma. The analysis of a wider range of spectral lines will help to extend this powerful diagnostic to experiments where the wavelength range of the measured spectra may be constrained either by limitations of the spectrometer or by other conflicting lines from different ions.

  11. Initial Results from Coaxial Helicity Injection Experiments in NSTX

    International Nuclear Information System (INIS)

    Raman, R.; Jarboe, T.R.; Mueller, D.; Schaffer, M.J.; Maqueda, R.; Nelson, B.A.; Sabbagh, S.; Bell, M.; Ewig, R.; Fredrickson, E.; Gates, D.; Hosea, J.; Ji, H.; Kaita, R.; Kaye, S.M.; Kugel, H.; Maingi, R.; Menard, J.; Ono, M.; Orvis, D.; Paolette, F.; Paul, S.; Peng, M.; Skinner, C.H.; Wilgen, W.; Zweben, S.

    2001-01-01

    Coaxial Helicity Injection (CHI) has been investigated on the National Spherical Torus Experiment (NSTX). Initial experiments produced 130 kA of toroidal current without the use of the central solenoid. The corresponding injector current was 20 kA. Discharges with pulse lengths up to 130 ms have been produced

  12. Toroidal confinement of non-neutral plasma - A new approach to high-beta equilibrium

    International Nuclear Information System (INIS)

    Yoshida, Z.; Ogawa, Y.; Morikawa, J.

    2001-01-01

    Departure from the quasi-neutral condition allows us to apply significant two-fluid effects that impart a new freedom to the design of high-performance fusion plasma. The self-electric field in a non-neutralized plasma induces a strong ExB-drift flow. A fast flow produces a large hydrodynamic pressure that can balance with the thermal pressure of the plasma. Basic concepts to produce a toroidal non-neutral plasma have been examined on the internal-conductor toroidal confinement device Proto-RT. A magnetic separatrix determines the boundary of the confinement region. Electrons describe chaotic orbits in the neighborhood of the magnetic null point on the separatrix. The chaos yields collisionless diffusion of electrons from the particle source (electron gun) towards the confinement region. Collisionless heating also occurs in the magnetic null region, which can be applied to produce a plasma. (author)

  13. Non-Inductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta in the Pegasus Toroidal Experiment

    Science.gov (United States)

    Reusch, Joshua

    2017-10-01

    A major goal of the spherical tokamak research program is accessing a state of low internal inductance li, high elongation κ, high toroidal and normalized beta (βt and βN) , and low collisionality without solenoidal current drive. A new local helicity injection (LHI) system in the lower divertor region of the ultra-low aspect ratio Pegasus ST provides non-solenoidally driven plasmas that exhibit most of these characteristics. LHI utilizes compact, edge-localized current sources (Ainj 4 cm2, Iinj 8 kA, Vinj 1.5 kV) for plasma startup and sustainment, and can sustain more than 200 kA of plasma current. Plasma growth via LHI is enhanced by a transition from a regime of high kink-like MHD activity to one of reduced MHD activity at higher frequencies and presumably shorter wavelengths. The strong edge current drive provided by LHI results in a hollow current density profile with low li. The low aspect ratio (R0 / a 1.2) of Pegasus allows ready access to high κ and MHD stable operation at very high normalized plasma currents (IN =Ip /aBT> 15). Thomson scattering measurements indicate Te 100 eV and ne 1 ×19 m-3. The impurity Ti evolution is correlated in time with high frequency magnetic fluctuations, implying substantial reconnection ion heating is driven by the applied helicity injection. Doppler spectroscopy indicates Ti >=Te and that the anomalous ion heating scales consistently with two fluid reconnection theory. Taken together, these features provide access to very high βt plasmas. Equilibrium analyses indicate βt up to 100% and βN 6.5 is achieved. At increasingly low BT, the discharge disrupts at the no-wall ideal stability limit. In these high βt discharges, a minimum |B| well forms over 50% of the plasma volume. This unique magnetic configuration may be of interest for testing predictions of stabilizing drift wave turbulence and/or improving energetic particle confinement. This work supported by US DOE Grants DE-FG02-96ER54375 and DE-SC0006928.

  14. Experiments close to the beta-limit in W7-AS

    International Nuclear Information System (INIS)

    Weller, A; Geiger, J; Werner, A; Zarnstorff, M C; Nuehrenberg, C; Sallander, E; Baldzuhn, J; Brakel, R; Burhenn, R; Dinklage, A; Fredrickson, E; Gadelmeier, F; Giannone, L; Grigull, P; Hartmann, D; Jaenicke, R; Klose, S; Knauer, J P; Koenies, A; Kolesnichenko, Ya I; Laqua, H P; Lutsenko, V V; McCormick, K; Monticello, D; Osakabe, M; Pasch, E; Reiman, A; Rust, N; Spong, D A; Wagner, F; Yakovenko, Yu V

    2003-01-01

    A major objective of the experimental program in the last phase of the W7-AS stellarator was to explore and demonstrate the high-beta performance of advanced stellarators. MHD-quiescent discharges at low impurity radiation levels with volume averaged beta-values of up to beta=3.4% have been achieved. A very important prerequisite was the attainment of the high density H-Mode (HDH) regime. This was made possible by the installation of extensive graphite plasma facing components designed for island divertor operation. The co-directed neutral beam injection provided increased absorbed heating power of up to 3.2 MW in high-beta plasmas with B<=1.25 T. The anticipated improved features concerning equilibrium and stability at high plasma beta could be verified experimentally by the comparison of x-ray data with free boundary equilibrium calculations. The maximum beta found in configurations with a rotational transform around t=0.5 is determined by the available heating power. No evidence of a stability limit has been found in the accessible configuration space, and the discharges are remarkably quiescent at maximum beta, most likely due the increase of the magnetic well depth. An increase in low m/n MHD activity is typically observed during the transition towards high beta. The beneficial stability properties of net-current-free configurations could be demonstrated by comparison with configurations where a significant inductive current drive was involved. Current driven instabilities such as tearing modes and soft disruptions can prevent access to beta-values as high as in the currentless case. The experimental results indicate that optimized stellarators such as W7-X can be considered as a viable option for an attractive stellarator fusion reactor

  15. Interaction with beta-arrestin determines the difference in internalization behavor between beta1- and beta2-adrenergic receptors.

    Science.gov (United States)

    Shiina, T; Kawasaki, A; Nagao, T; Kurose, H

    2000-09-15

    The beta(1)-adrenergic receptor (beta(1)AR) shows the resistance to agonist-induced internalization. As beta-arrestin is important for internalization, we examine the interaction of beta-arrestin with beta(1)AR with three different methods: intracellular trafficking of beta-arrestin, binding of in vitro translated beta-arrestin to intracellular domains of beta(1)- and beta(2)ARs, and inhibition of betaAR-stimulated adenylyl cyclase activities by beta-arrestin. The green fluorescent protein-tagged beta-arrestin 2 translocates to and stays at the plasma membrane by beta(2)AR stimulation. Although green fluorescent protein-tagged beta-arrestin 2 also translocates to the plasma membrane, it returns to the cytoplasm 10-30 min after beta(1)AR stimulation. The binding of in vitro translated beta-arrestin 1 and beta-arrestin 2 to the third intracellular loop and the carboxyl tail of beta(1)AR is lower than that of beta(2)AR. The fusion protein of beta-arrestin 1 with glutathione S-transferase inhibits the beta(1)- and beta(2)AR-stimulated adenylyl cyclase activities, although inhibition of the beta(1)AR-stimulated activity requires a higher concentration of the fusion protein than that of the beta(2)AR-stimulated activity. These results suggest that weak interaction of beta(1)AR with beta-arrestins explains the resistance to agonist-induced internalization. This is further supported by the finding that beta-arrestin can induce internalization of beta(1)AR when beta-arrestin 1 does not dissociate from beta(1)AR by fusing to the carboxyl tail of beta(1)AR.

  16. Approach to high stability beta limit and its control by fast wave current drive in reversed field pinch plasma

    International Nuclear Information System (INIS)

    Kusano, K.; Kondoh, Y.; Gesso, H.; Osanai, Y.; Saito, K.N.; Ukai, R.; Nanba, T.; Nagamine, Y.; Shiina, S.

    2001-01-01

    Before the generation of steady state, dynamo-free RFP configuration by rf current driving scheme, it is necessary to find an optimum configuration into high stability beta limit against m=1 resonant resistive MHD modes and reducing nonlinearly turbulent level with less rf power. As first step to the optimization study, we are interested in partially relaxed state model (PRSM) RFP configuration, which is considered to be closer to a relaxed state at finite beta since it has force-free fields for poloidal direction with a relatively shorter characteristic length of relaxation and a relatively higher stability beta limit to m=1 resonant ideal MHD modes. The stability beta limit to m=1 resonant resistive MHD modes can be predicted to be relatively high among other RFP models and to be enhanced by the current density profile control using fast magnetosonic waves (FMW), which are accessible to high density region with strong absorption rate. (author)

  17. Long pulse operation of high performance plasmas in JT-60U

    International Nuclear Information System (INIS)

    Ide, Shunsuke

    2005-01-01

    Recent experimental progress in JT-60U advanced tokamak research is presented; sustainment of the normalized beta (β N ) - 3 in a normal magnetic shear plasma, the bootstrap current fraction (f BS ) - 45% in a weak shear plasma and ∼75% in a reversed magnetic shear plasma in a nearly full non-inductive current drive condition for longer than the current relaxation time. Achievement of high-density high-radiation fraction together with high-confinement in advanced plasmas was demonstrated. Achievement and foundings in long pulse operations after system modification are presented as well. A 65 s discharge of I p =0.7 MA was successfully obtained. As a result, high-β N of 2.3 was successfully sustained for a very long period of 22.3 s. In addition, a 30 s standard ELMy H-mode plasma of I p up to 1.4 MA has also been obtained. Effectiveness of divertor pumping to control particle recycling and the electron density under the wall retention was saturated was demonstrated. These achievement and issues in the development will be discussed. (author)

  18. Ideal MHD stability of high poloidal beta equilibria in TFTR

    International Nuclear Information System (INIS)

    Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.; Bell, M.G.; Budny, R.V.; Chance, M.S.; Fredrickson, E.D.; Jardin, S.C.; Manickam, J.; McCune, D.C.; McGuire, K.M.; Wieland, R.M.; Zarnstorff, M.C.; Phillips, M.W.; Hughes, M.H.; Kesner, J.

    1991-01-01

    Recent experiments in TFTR have expanded the operating space of the device to include plasmas with values of var-epsilon β p dia ≡ 2μ 0 var-epsilon perpendicular >/ p >> 2 as large as 1.6, and Troyon normalized diamagnetic beta β N dia ≡ β t perpendicular aB t /10 -8 I p as large as 4.7. At values of var-epsilon β p dia ≥ 1.3, a separatrix was observed to enter the vacuum vessel, producing a naturally diverted discharge. Plasmas with large values of var-epsilon β p dia were created with both the plasma current, I p , held constant and with I p decreased, or ramped down, before the start of neutral beam injection. A convenient characterization of the change in I p using experimental parameters can be defined by the ratio of I p before the ramp down, to I p during the neutral beam heating phase, F I p . The ideal MHD stability of these equilibria is investigated to determine their location in stability space, and to study the role of plasma current and pressure profile modification in the creation of these high var-epsilon β p and β N plasmas. The evolution of these plasmas is modelled from experimental data using the TRANSP code. Two-dimensional equilibria are computed from the TRANSP results and used as input to both high and low-n stability codes including PEST. The high var-epsilon β p equilibria, which generally have an oblate cross-sectional shape, are in the first stability region to high-n ballooning modes. At constant I p , these equilibria generally have maximum pressure gradients near the magnetic axis and are stable to n=1 modes without a stabilizing conducting wall. The effect of the current profile shape on the stability of low-n kink/ballooning modes and the requirements for these plasmas to access the second stability region are examined. 6 refs

  19. Fluctuations in high βp plasmas in DIII-D

    International Nuclear Information System (INIS)

    Casper, T.A.; Chu, M.S.; Gohil, P.

    1994-07-01

    In our investigation of improved confinement in high poloidal beta (β p = 2 to 4) advanced tokamak experiments, we observe that the internal MHD activity evolves from an m/n = 2/1 to a 3/1 structure coincident with q o rising above 2, and consistent with the GATO code stability analysis. The plasma eventually evolves to a quiescent state at which time the stored energy increases, mostly as a result of improved particle confinement. The bootstrap fraction rises to 80%. The measured plasma pressure profiles during this time are calculated to be stable to high-n ballooning modes consistent with operation of the core in the second stable regime. The sustained improvement in confinement is ultimately limited by our ability to control the toroidal current profile

  20. Simultaneous extraction of. beta. -endorphin and leu- and met-enkephalins from human and rat plasma

    Energy Technology Data Exchange (ETDEWEB)

    Bhathena, S.J.; Smith, P.M.; Kennedy, B.W. (Dept. of Agriculture, Beltsville, MD (USA)); Voyles, N.R.; Recant, L. (Diabetes Research Laboratory, Washington, DC (USA))

    1989-01-01

    A simple, rapid and reliable procedure is described to simultaneously concentrated and purify {beta}-endorphin, leu-and met-enkephalins from small volumes of human and rat plasma before radioimmunoassay is performed. It uses C{sub 18} Sep-Pak reverse phase cartridges. The effectiveness of different protease inhibitors in preventing degradation of opiates by plasma and different solvent systems for eluting opiates is also evaluated.

  1. USXR Based MHD, Transport, Equilibria and Current Profile Diagnostics for NSTX. Final Report

    International Nuclear Information System (INIS)

    Finkenthal, Michael

    2009-01-01

    The present report resumes the research activities of the Plasma Spectroscopy/Diagnostics Group at Johns Hopkins University performed on the NSTX tokamak at PPPL during the period 1999-2009. During this period we have designed and implemented XUV based diagnostics for a large number of tasks: study of impurity content and particle transport, MHD activity, time-resolved electron temperature measeurements, ELM research, etc. Both line emission and continuum were used in the XUV range. New technics and novel methods have been devised within the framework of the present research. Graduate and post-graduate students have been involved at all times in addition to the senior research personnel. Several tens of papers have been published and lectures have been given based on the obtained results at conferences and various research institutions (lists of these activities were attached both in each proposal and in the annual reports submitted to our supervisors at OFES)

  2. Mass changes in NSTX Surface Layers with Li Conditioning as Measured by Quartz Microbalances

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.W.; Roquemore, A.L.; Krstic, P.S.; Beste, A.

    2008-01-01

    Dynamic retention, lithium deposition, and the stability of thick deposited layers were measured by three quartz crystal microbalances (QMB) deployed in plasma shadowed areas at the upper and lower divertor and outboard midplane in the National Spherical Torus Experiment (NSTX). Deposition of 185 (micro)/g/cm 2 over 3 months in 2007 was measured by a QMB at the lower divertor while a QMB on the upper divertor, that was shadowed from the evaporator, received an order of magnitude less deposition. During helium glow discharge conditioning both neutral gas collisions and the ionization and subsequent drift of Li + interrupted the lithium deposition on the lower divertor. We present calculations of the relevant mean free paths. Occasionally strong variations in the QMB frequency were observed of thick lithium films suggesting relaxation of mechanical stress and/or flaking or peeling of the deposited layers.

  3. Proceedings of 1999 U.S./Japan Workshop (99FT-05) On High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

    International Nuclear Information System (INIS)

    NYGREN, RICHARD E.; STAVROS, DIANA T.

    2000-01-01

    The 1999 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions in Next Step Fusion Devices was held at the St. Francis Hotel in Santa Fe, New Mexico, on November 1-4, 1999. There were 42 presentations as well as discussion on technical issues and planning for future collaborations. The participants included 22 researchers from Japan and the United States as well as seven researchers from Europe and Russia. There have been important changes in the programs in both the US and Japan in the areas of plasma surface interactions and plasma facing components. The US has moved away from a strong focus on the ITER Project and has introduced new programs on use of liquid surfaces for plasma facing components, and operation of NSTX has begun. In Japan, the Large Helical Device began operation. This is the first large world-class confinement device operating in a magnetic configuration different than a tokamak. In selecting the presentations for this workshop, the organizers sought a balance between research in laboratory facilities or confinement devices related to plasma surface interactions and experimental research in the development of plasma facing components. In discussions about the workshop itself, the participants affirmed their preference for a setting where ''work-in-progress'' could be informally presented and discussed

  4. Different antihypertensive effect of beta-blocking drugs in low and normal-high renin hypertension.

    Science.gov (United States)

    Kralberg, B E; Tolagen, K

    1976-05-31

    The treatment response to beta-adrenoceptor blocking drugs was compared in two groups of patients with primary (essential) hypertension and different renin levels. Each group consisted of 25 patients and was equally distributed regarding age, severity and stage of hypertension. In the first group (group 1), the mean upright plasma renin activity was 0.8 ng ml-1h-1 (range 0.3 to 1.5) and the patients were considered to have low renin hypertension. In the other group (group 2) the patients had a mean plasma renin activity of 2.1 ng ml-1h-1 (range 1.1 to 5.1) and were considered to have normal to high renin hypertension. In both groups the patients were initially treated with beta-blocking drugs; in group 1 with a beta-blocker corresponding to an average dose of 311 mg propranolol a day for at least eight weeks and in group 2 with propranolol 320 mg a day in a fixed dose for eight weeks. The hypotensive response differed significantly between the two groups (p less than 0.001). In group 1 the pretreatment blood pressure was 197/117 mm Hg supine and 198/120 mm Hg standing. During treatment blood pressure decreased only 5/3 mm Hg supine and 9/5 mm Hg standing. The pretreatment blood pressure in group 2 was 187/114 mm Hg supine and 186/117 mm Hg standing. Beta-blocking therapy reduced blood pressure 36/23 and 34/18 mm Hg, respectively (both p less than 0.001). Pulse rates fell significantly in the two groups, both in the lying and standing positions. In 17 patients with low renin hypertension (group 1), a volume-depleting drug was added (spironolactone, 14 patients; thiazides, 3 patients) and this achieved a marked fall in blood pressure levels of 38/16 mm Hg supine and 37/19 mm Hg standing (both p less than 0.001). These results suggest the following: (1) Most patients with normal to high plasma renin activity respond well to moderate doses of propranolol. (2) Propranolol given in the same doses is almost without antihypertensive effect in patients with low renin

  5. Confinement and heating of high beta plasma with emphasis on compact toroids. Compact toroid research

    International Nuclear Information System (INIS)

    Vlases, G.C.; Pietrzyk, Z.A.

    1984-11-01

    Two older projects associated with very high energy density plasmas, specifically the High Density Field Reversed Configuration and the Liner Plasma Compression Experiment, have been completed. Attention has been turned to compact toroid experiments of more conventional density, and three experiments have been initiated. These include the Coaxial Slow Source Experiment, the Variable Length FRC Experiment, and Variable Angle CthetaP Experiment. In each case, the project was begun in order to provide basic plasma physics information on specific unresolved issues of progammatic importance to the national CT Program

  6. A mechanism for large divertor plasma energy loss via lithium radiation in tokamaks

    Science.gov (United States)

    Rognlien, T. D.; Meier, E. T.; Soukhanovskii, V. A.

    2012-10-01

    Lithium has been used as a wall-conditioning element in a number of tokamaks over the years, including TFTR, FTU, and NSTX, where core plasma energy confinement and particle control are often found to improve following such conditioning. Here the possible role of Li in providing substantial energy loss for divertor plasmas via line radiation is reported. A multi-charge-state 2D UEDGE fluid model is used where the hydrogenic and Li ions and neutrals are each evolved as separate species and separate equations are solved for the electron and ion temperatures. It is shown that a sufficient level of Li neutrals evolving from the divertor surface via sputtering or evaporation can induce energy detachment of the divertor plasma, yielding a strongly radiating zone near the divertor where ionization and recombination from/to neutral Li can radiate most of the power flowing into the scrape-off layer while maintaining low core contamination. A local peaking of Li emissivity for electron temperatures near 1 eV appears to play an important role in the detachment of the mixed deuterium/Li plasma. Evidence of such behavior from NSTX discharges will be discussed.

  7. l=1,2 high-beta stellarator

    International Nuclear Information System (INIS)

    Bartsch, R.R.; Cantrell, E.L.; Gribble, R.F.; Klare, K.A.; Kutac, K.J.; Miller, G.; Siemon, R.E.

    1978-01-01

    The final scyllac experiments are described. These experiments utilized a feedback-stabilized, l=1,2 high-beta stellarator configuration and like the previous feedback-stabilization experiments were carried out in a toroidal sector, rather than a complete torus. The energy confinement time, obtained from excluded flux measurements, agrees with a two-dimensional calculation of particle end loss from a straight theta pinch. Because simple end loss was dominant, the energy confinement time was independent of whether equilibrium adjustment or feedback stabilization fields were applied. The dynamical characteristics of the toroidal equilibrium were improved by elimination of the l=0 field used previously, as expected from theory. A modal rather than local feedback control algorithm was used. Although feedback clearly decreased the m=1 motion of the plasma, the experimental test of modal feedback, which is expected from theory to be superior to local feedback, is considered inconclusive because of the limitations imposed by the sector configuration

  8. Bakery products enriched with phytosterol esters, alpha-tocopherol and beta-carotene decrease plasma LDL-cholesterol and maintain plasma beta-carotene concentrations in normocholesterolemic men and women.

    Science.gov (United States)

    Quílez, Joan; Rafecas, Magda; Brufau, Gemma; García-Lorda, Pilar; Megías, Isabel; Bulló, Mònica; Ruiz, Joan A; Salas-Salvadó, Jordi

    2003-10-01

    The hypocholesterolemic effects of phytosterols have not been evaluated in bakery products, and the addition of liposoluble antioxidants to the carrier has never been tested. We investigated the effects of consuming croissants and magdalenas (Spanish muffins) enriched with sterol esters, alpha-tocopherol and beta-carotene on plasma lipid and fat-soluble antioxidant concentrations in normocholesterolemic, habitual consumers of bakery products following their usual diet and lifestyle. Using a randomized, double-blind, placebo-controlled design, the control (C) group (n = 29) received two pieces daily (standard croissant and muffin) and the sterol ester (SE) group (n = 28), the same products with sterol esters added (3.2 g/d) for 8 wk. Total and LDL cholesterol (LDL-C) decreased in the SE group by 0.24 mmol/L (P bakery products are excellent carriers for phytosterols, and their consumption is associated with a decrease in total and LDL-C concentrations, with no changes in alpha-tocopherol and beta-carotene. The ability of bakery products to include sufficient quantities of beta-carotene to compensate for a potential deficiency, and the fact that their efficacy was not associated with the time of day at which they were consumed, are interesting findings.

  9. Hypervelocity Dust Injection for Plasma Diagnostic Applications

    Science.gov (United States)

    Ticos, Catalin

    2005-10-01

    Hypervelocity micron-size dust grain injection was proposed for high-temperature magnetized plasma diagnosis. Multiple dust grains are launched simultaneously into high temperature plasmas at several km/s or more. The hypervelocity dust grains are ablated by the electron and ion fluxes. Fast imaging of the resulting luminous plumes attached to each grain is expected to yield local magnetic field vectors. Combination of multiple local magnetic field vectors reproduces 2D or even 3D maps of the internal magnetic field topology. Key features of HDI are: (1) a high spatial resolution, due to a relatively small transverse size of the elongated tail, and (2) a small perturbation level, as the dust grains introduce negligible number of particles compared to the plasma particle inventory. The latter advantage, however, could be seriously compromised if the gas load from the accelerator has an unobstructed access to the diagnosed plasma. Construction of a HDI diagnostic for National Spherical Torus Experiment (NSTX), which includes a coaxial plasma gun for dust grain acceleration, is underway. Hydrogen and deuterium gas discharges inside accelerator are created by a ˜ 1 mF capacitor bank pre-charged up to 10 kV. The diagnostic apparatus also comprises a dust dispenser for pre-loading the accelerator with dust grains, and an imaging system that has a high spatial and temporal resolution.

  10. Plasma Interactions with Mixed Materials and Impurity Transport

    Energy Technology Data Exchange (ETDEWEB)

    Rognlien, T. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beiersdorfer, Peter [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chernov, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frolov, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Magee, E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rudd, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Umansky, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-28

    The project brings together three discipline areas at LLNL to develop advanced capability to predict the impact of plasma/material interactions (PMI) on metallic surfaces in magnetic fusion energy (MFE) devices. These areas are (1) modeling transport of wall impurity ions through the edge plasma to the core plasma, (2) construction of a laser blow-off (LBO) system for injecting precise amounts of metallic atoms into a tokamak plasma, and (3) material science analysis of fundamental processes that modify metallic surfaces during plasma bombardment. The focus is on tungsten (W), which is being used for the ITER divertor and in designs of future MFE devices. In area (1), we have worked with the University of California, San Diego (UCSD) on applications of the UEDGE/DUSTT coupled codes to predict the influx of impurity ions from W dust through the edge plasma, including periodic edge-plasma oscillations, and revived a parallel version of UEDGE to speed up these simulations. In addition, the impurity transport model in the 2D UEDGE code has been implemented into the 3D BOUT++ turbulence/transport code to allow fundamental analysis of the impact of strong plasma turbulence on the impurity transport. In area (2), construction and testing of the LBO injection system has been completed. The original plan to install the LBO on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton and its use to validate the impurity transport simulations is delayed owing to NSTX-U being offline for substantial magnetic coil repair period. In area (3), an analytic model has been developed to explain the growth of W tendrils (or fuzz) observed for helium-containing plasmas. Molecular dynamics calculations of W sputtering by W and deuterium (D) ions shows that a spatial blending of interatomic potentials is needed to describe the near-surface and deeper regions of the material.

  11. Plasma Interactions with Mixed Materials and Impurity Transport

    International Nuclear Information System (INIS)

    Rognlien, T. D.; Beiersdorfer, Peter; Chernov, A.; Frolov, T.; Magee, E.; Rudd, R.; Umansky, M.

    2016-01-01

    The project brings together three discipline areas at LLNL to develop advanced capability to predict the impact of plasma/material interactions (PMI) on metallic surfaces in magnetic fusion energy (MFE) devices. These areas are (1) modeling transport of wall impurity ions through the edge plasma to the core plasma, (2) construction of a laser blow-off (LBO) system for injecting precise amounts of metallic atoms into a tokamak plasma, and (3) material science analysis of fundamental processes that modify metallic surfaces during plasma bombardment. The focus is on tungsten (W), which is being used for the ITER divertor and in designs of future MFE devices. In area (1), we have worked with the University of California, San Diego (UCSD) on applications of the UEDGE/DUSTT coupled codes to predict the influx of impurity ions from W dust through the edge plasma, including periodic edge-plasma oscillations, and revived a parallel version of UEDGE to speed up these simulations. In addition, the impurity transport model in the 2D UEDGE code has been implemented into the 3D BOUT++ turbulence/transport code to allow fundamental analysis of the impact of strong plasma turbulence on the impurity transport. In area (2), construction and testing of the LBO injection system has been completed. The original plan to install the LBO on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton and its use to validate the impurity transport simulations is delayed owing to NSTX-U being offline for substantial magnetic coil repair period. In area (3), an analytic model has been developed to explain the growth of W tendrils (or fuzz) observed for helium-containing plasmas. Molecular dynamics calculations of W sputtering by W and deuterium (D) ions shows that a spatial blending of interatomic potentials is needed to describe the near-surface and deeper regions of the material.

  12. Waves and Instabilities in Steady-State High-Beta Plasmas

    Science.gov (United States)

    1976-07-01

    us working on magnetospheric related problems. Several groups are now constructing identical devices including Y. Nishida of Utsunomiya University...and other satellites operate in the magnetospheric plasma environment at the geosynchronous orbit (%6.6 earth radii). Arc- related deterioration of the...carefully 16 - 3diagnosed device produces a plasma of density n 3 x 10 cm and temperature Te = Ti W 1.6eV. (3) Heat Flow Measurements in a Laser-Heated

  13. BETA (Bitter Electromagnet Testing Apparatus)

    Science.gov (United States)

    Bates, Evan M.; Birmingham, William J.; Rivera, William F.; Romero-Talamas, Carlos A.

    2017-10-01

    The Bitter Electromagnet Testing Apparatus (BETA) is a 1-Tesla (T) prototype of the 10-T Adjustable Long Pulse High-Field Apparatus (ALPHA). These water-cooled resistive magnets use high DC currents to produce strong uniform magnetic fields. Presented here is the successful completion of the BETA project and experimental results validating analytical magnet designing methods developed at the Dusty Plasma Laboratory (DPL). BETA's final design specifications will be highlighted which include electromagnetic, thermal and stress analyses. The magnet core design will be explained which include: Bitter Arcs, helix starters, and clamping annuli. The final version of the magnet's vessel and cooling system are also presented, as well as the electrical system of BETA, which is composed of a unique solid-state breaker circuit. Experimental results presented will show the operation of BETA at 1 T. The results are compared to both analytical design methods and finite element analysis calculations. We also explore the steady state maximums and theoretical limits of BETA's design. The completion of BETA validates the design and manufacturing techniques that will be used in the succeeding magnet, ALPHA.

  14. Stability and control of resistive wall modes in high beta, low rotation DIII-D plasmas

    International Nuclear Information System (INIS)

    Garofalo, A.M.; Jackson, G.L.; Haye, R.J. La; Okabayashi, M.; Reimerdes, H.; Strait, E.J.; Ferron, J.R.; Groebner, R.J.; In, Y.; Lanctot, M.J.; Matsunaga, G.; Navratil, G.A.; Solomon, W.M.; Takahashi, H.; Takechi, M.; Turnbull, A.D.

    2007-01-01

    Recent high-β DIII-D (Luxon J.L. 2002 Nucl. Fusion 42 64) experiments with the new capability of balanced neutral beam injection show that the resistive wall mode (RWM) remains stable when the plasma rotation is lowered to a fraction of a per cent of the Alfven frequency by reducing the injection of angular momentum in discharges with minimized magnetic field errors. Previous DIII-D experiments yielded a high plasma rotation threshold (of order a few per cent of the Alfven frequency) for RWM stabilization when resonant magnetic braking was applied to lower the plasma rotation. We propose that the previously observed rotation threshold can be explained as the entrance into a forbidden band of rotation that results from torque balance including the resonant field amplification by the stable RWM. Resonant braking can also occur naturally in a plasma subject to magnetic instabilities with a zero frequency component, such as edge localized modes. In DIII-D, robust RWM stabilization can be achieved using simultaneous feedback control of the two sets of non-axisymmetric coils. Slow feedback control of the external coils is used for dynamic error field correction; fast feedback control of the internal non-axisymmetric coils provides RWM stabilization during transient periods of low rotation. This method of active control of the n = 1 RWM has opened access to new regimes of high performance in DIII-D. Very high plasma pressure combined with elevated q min for high bootstrap current fraction, and internal transport barriers for high energy confinement, are sustained for almost 2 s, or 10 energy confinement times, suggesting a possible path to high fusion performance, steady-state tokamak scenarios

  15. NSTX Disruption Simulations of Detailed Divertor and Passive Plate Models by Vector Potential Transfer from OPERA Global Analysis Results

    International Nuclear Information System (INIS)

    Titus, P.H.; Avasaralla, S.; Brooks, A.; Hatcher, R.

    2010-01-01

    The National Spherical Torus Experiment (NSTX) project is planning upgrades to the toroidal field, plasma current and pulse length. This involves the replacement of the center-stack, including the inner legs of the TF, OH, and inner PF coils. A second neutral beam will also be added. The increased performance of the upgrade requires qualification of the remaining components including the vessel, passive plates, and divertor for higher disruption loads. The hardware needing qualification is more complex than is typically accessible by large scale electromagnetic (EM) simulations of the plasma disruptions. The usual method is to include simplified representations of components in the large EM models and attempt to extract forces to apply to more detailed models. This paper describes a more efficient approach of combining comprehensive modeling of the plasma and tokamak conducting structures, using the 2D OPERA code, with much more detailed treatment of individual components using ANSYS electromagnetic (EM) and mechanical analysis. This capture local eddy currents and resulting loads in complex details, and allows efficient non-linear, and dynamic structural analyses.

  16. NSTX Disruption Simulations of Detailed Divertor and Passive Plate Models by Vector Potential Transfer from OPERA Global Analysis Results

    Energy Technology Data Exchange (ETDEWEB)

    P. H. Titus, S. Avasaralla, A.Brooks, R. Hatcher

    2010-09-22

    The National Spherical Torus Experiment (NSTX) project is planning upgrades to the toroidal field, plasma current and pulse length. This involves the replacement of the center-stack, including the inner legs of the TF, OH, and inner PF coils. A second neutral beam will also be added. The increased performance of the upgrade requires qualification of the remaining components including the vessel, passive plates, and divertor for higher disruption loads. The hardware needing qualification is more complex than is typically accessible by large scale electromagnetic (EM) simulations of the plasma disruptions. The usual method is to include simplified representations of components in the large EM models and attempt to extract forces to apply to more detailed models. This paper describes a more efficient approach of combining comprehensive modeling of the plasma and tokamak conducting structures, using the 2D OPERA code, with much more detailed treatment of individual components using ANSYS electromagnetic (EM) and mechanical analysis. This capture local eddy currents and resulting loads in complex details, and allows efficient non-linear, and dynamic structural analyses.

  17. High-beta characteristics of first and second-stable spherical tokamaks in reconnection heating experiments of TS-3

    International Nuclear Information System (INIS)

    Ono, Y.

    2002-01-01

    Novel formations of ultra-high-beta Spherical Tokamak (ST) have been developed in the TS-3 device using high power heating of merging/ reconnection. In Type-A merging, two STs were merged together to build up the plasma beta. In Type-B merging, an oblate FRC was initially formed by merging of two spheromaks with opposing toroidal field B t and was transformed into an ultra-high-beta ST by applying external B t . Ballooning stability analyses confirmed formations of the first-stable STs by Type- A merging and the second-stable STs by Type-B merging and also the unstable STs by both mergings, revealing the ballooning stability window consistent with measured high-n instabilities. We made (1) those model analyses of the produced STs for the first time using the BALLOO stability code, revealing that hollowness/ broadness of current/pressure profiles widen significantly the window to the second-stable regime. This paper also addresses (2) normalized betas of the second-stable STs as large as 6-17 for comparison with the Troyon scaling and (3) a promising scaling of the reconnection heating energy. (author)

  18. Progress with helicity injection current drive

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Raman, R.; Nelson, B.A.

    2003-01-01

    Coaxial Helicity Injection (CHI) experiments in the NSTX and HIT-II devices are reported. NSTX has produced toroidal currents of 0.4 MA and pulse lengths of up to 0.33 s. These discharges nearly fill the NSTX main chamber, and show the n=1 rotating distortion characteristic of high-performance CHI plasmas. CHI has been used in HIT-II to provide a closed flux startup plasma for inductive drive. The CHI startup method saves transformer volt-seconds and greatly improves reproducibility and reliability of inductively driven discharges, even in the presence of diminishing wall conditions. (author)

  19. A validated liquid chromatography-tandem mass spectrometry method for the quantitative determination of 4 beta-hydroxycholesterol in human plasma

    NARCIS (Netherlands)

    van de Merbel, Nico C.; Bronsema, Kees J.; van Hout, Mischa W. J.; Nilsson, Ralf; Sillen, Henrik

    2011-01-01

    A novel liquid chromatography-tandem mass spectrometry method is described for the quantitative determination of the endogenous CYP 3A4/5 marker 4 beta-hydroxycholesterol in human K(2)-EDTA plasma. It is based on alkaline hydrolysis to convert esterified to free 4 beta-hydroxycholesterol, followed

  20. Electron acceleration during the decay of nonlinear Whistler waves in low-beta electron-ion plasma

    International Nuclear Information System (INIS)

    Umeda, Takayuki; Saito, Shinji; Nariyuki, Yasuhiro

    2014-01-01

    Relativistic electron acceleration through dissipation of a nonlinear, short-wavelength, and monochromatic electromagnetic whistler wave in low-beta plasma is investigated by utilizing a one-dimensional fully relativistic electromagnetic particle-in-cell code. The nonlinear (large-amplitude) parent whistler wave decays through the parametric instability which enhances electrostatic ion acoustic waves and electromagnetic whistler waves. These waves satisfy the condition of three-wave coupling. Through the decay instability, the energy of electron bulk velocity supporting the parent wave is converted to the thermal energy perpendicular to the background magnetic field. Increase of the perpendicular temperature triggers the electron temperature anisotropy instability which generates broadband whistler waves and heats electrons in the parallel direction. The broadband whistler waves are inverse-cascaded during the relaxation of the electron temperature anisotropy. In lower-beta conditions, electrons with a pitch angle of about 90° are successively accelerated by inverse-cascaded whistler waves, and selected electrons are accelerated to over a Lorentz factor of 10. The result implies that the nonlinear dissipation of a finite-amplitude and short-wavelength whistler wave plays an important role in producing relativistic nonthermal electrons over a few MeV especially at lower beta plasmas.

  1. Metabolic clearance rate and urinary clearance of purified beta-core

    International Nuclear Information System (INIS)

    Wehmann, R.E.; Blithe, D.L.; Flack, M.R.; Nisula, B.C.

    1989-01-01

    We injected a highly purified preparation of the beta-core molecule, a fragment of hCG beta excreted in pregnancy urine, into five men and three women to determine its kinetic parameters, MCR, and urinary clearance. The beta-core molecule was distributed in an initial volume [1950 +/- 156 (mean +/- SEM) mL/m2 body surface area] approximately equal to the estimated plasma volume. Its disappearance was multiexponential on a semilogarithmic plot, with a rapid phase t1/2 of 3.5 +/- 0.7 min and a slow phase t1/2 of 22.4 +/- 4.2 min. The transit time (the mean time spent by a molecule of beta-core in transit) was 20.6 +/- 2.1 min. The MCR was 192.0 +/- 8.0 mL/min.m2 body surface area. About 5% of the injected dose of beta-core was excreted into the urine in the first 30 min after injection, and low levels of excretion persisted for up to 7 days. The urinary clearance rate of beta-core was 13.7 +/- 1.4 mL/min.m2, accounting for about 8% of the elimination of beta-core from the plasma. The beta-core immunoreactivity in serum and urine was characterized by gel filtration and three independent RIA systems to show that its properties were indistinguishable from those of the injected beta-core. Serum levels of beta-core in pregnant women were less than 0.2 ng/mL, while the amounts excreted in their urine were as much as 5 mg/day. Based on these clearance parameters of beta-core in normal subjects, less than 0.2% of the beta-core excreted in pregnancy urine is derived by urinary clearance of plasma beta-core. Therefore, more than 99% of the beta-core excreted in pregnancy urine is derived from beta-core in a compartment separate from plasma. In particular, these data indicate that there is relatively little placental secretion of beta-core into plasma and that placental secretion does not account for the vast majority of beta-core in pregnancy urine

  2. Fluctuations in high βp plasmas in DIII-D

    International Nuclear Information System (INIS)

    Casper, T.A.; Chu, M.S.; Gohil, C.P.

    1994-06-01

    In our investigation of improved confinement in high poloidal beta (β p = 2 to 4) advanced tokamak experiments, coincident with q 0 rising above 2, we observe the internal MHD activity to evolve from an m/n = 2/1 to a 3/1 structure consistent with the GATO code stability analysis. The plasma eventually evolves to a quiescent state at which time the stored energy increases, mostly as a result of improved particle confinement. The measured plasma pressure profiles during this time are also calculated to be stable to high-n ballooning modes consistent with operation of the core in the second stable regime. The sustained improvement in confinement is ultimately limited by our ability to control the toroidal current profile of which the bootstrap current contributes a large fraction (up to 80%)

  3. Fluctuations in high βp plasmas in DIII-D

    International Nuclear Information System (INIS)

    Casper, T.A.; Chu, M.S.; Gohil, P.

    1994-01-01

    In our investigation of improved confinement in high poloidal beta (β p = 2 to 4) advanced tokamak experiments, we observe that the internal MHD activity evolves from an m/n = 2/1 to a 3/1 structure coincident with q 0 rising above 2, and consistent with the GATO code stability analysis. The plasma eventually evolves to a quiescent state at which time the stored energy increases, mostly as a result of improved particle confinement. The bootstrap fraction rises to 80%. The measured plasma pressure profiles during this time are calculated to be stable to high-n ballooning modes consistent with operation of the core in the second stable regime. The sustained improvement in confinement is ultimately limited by our ability to control the toroidal current profile. (author) 4 refs., 6 figs

  4. Axisymmetric stability of vertically asymmetric Tokamaks at large beta poloidal

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, K.; Fishman, H.; Okabayashi, M. (Princeton Univ., NJ (USA). Plasma Physics Lab.); Todd, A.M.M. (Grumman Aerospace Corp., Princeton, NJ (USA))

    1983-11-01

    The rigid-mode stability of high-..beta.. vertically asymmetric Tokamak equilibria with quasi-uniform current profile is investigated analytically using toroidicity-shaping double expansion method. It is found that vertical stability at large beta poloidal is mainly determined by a coupling between the shape of the plasma surface and the Shafranov shift of the magnetic axis. To the lowest order, symmetric components of the plasma surface shape are found to be the critical destabilizing elements. Asymmetric components have little effect. The inclusion of higher order terms in the high-..beta.. Tokamak expansion leads to further destabilization. These analytic insights are qualitatively confirmed by numerical stability calculations using the PEST code with parabolic safety-factor profile.

  5. Diet, plasma levels of beta-carotene and alpha-tocopherol, and risk of malignant melanoma.

    Science.gov (United States)

    Stryker, W S; Stampfer, M J; Stein, E A; Kaplan, L; Louis, T A; Sober, A; Willett, W C

    1990-04-01

    Dietary intake and the plasma levels of retinol, alpha-tocopherol, lycopene, alpha-carotene, and beta-carotene for 204 cases with malignant melanoma were compared with those of 248 controls. Cases and controls were patients 18 years of age or older making their first visit to a dermatology subspecialty clinic for pigmented lesions from July 1, 1982 to September 1, 1985. Intakes of nutrients were estimated using a semiquantitative food frequency questionnaire. No significant associations with malignant melanoma were observed for higher plasma levels of lycopene, retinol, or alpha-carotene in logistic regression analyses after controlling for age, sex, plasma lipids, and known constitutional risk factors (hair color and ability to tan). In similar models, the odds ratio comparing the highest with the lowest quintile was 0.9 (95% confidence interval (CI) 0.5-1.5) for plasma beta-carotene, 0.7 (95% CI 0.5-1.3) for plasma alpha-tocopherol, 0.7 (95% CI 0.4-1.2) for carotene intake, and 0.7 (95% CI 0.4-1.3) for total vitamin E intake. A trend toward reduced risk of melanoma was observed for increasing intake of iron (not including supplements); this was related to the more frequent consumption of baked goods, such as cake, among controls. Alcohol consumption was positively associated with risk of melanoma (chi for trend = 2.1, p = 0.03); the odds ratio for consumption of over 10 g/day compared with persons with no alcohol intake was 1.8 (95% CI 1.0-3.3).

  6. Ion temperature anisotropy limitation in high beta plasmas

    International Nuclear Information System (INIS)

    Scime, Earl E.; Keiter, Paul A.; Balkey, Matthew M.; Boivin, Robert F.; Kline, John L.; Blackburn, Melanie; Gary, S. Peter

    2000-01-01

    Measurements of parallel and perpendicular ion temperatures in the Large Experiment on Instabilities and Anisotropies (LEIA) space simulation chamber display an inverse correlation between the upper bound on the ion temperature anisotropy and the parallel ion beta (β=8πnkT/B 2 ). Fluctuation measurements indicate the presence of low frequency, transverse, electromagnetic waves with wave numbers and frequencies that are consistent with predictions for Alfven Ion Cyclotron instabilities. These observations are also consistent with in situ spacecraft measurements in the Earth's magnetosheath and with a theoretical/computational model that predicts that such an upper bound on the ion temperature anisotropy is imposed by scattering from enhanced fluctuations due to growth of the Alfven ion cyclotron instability. (c) 2000 American Institute of Physics

  7. Progress Toward Long Pulse, High Performance Plasmas in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    P.A. Politzer; T.C. Luce; M.E. Austin; J.R. Ferron, A.M. Garofalo; C.M. Greenfield; A.W. Hyatt; R.J. La Haye; L.L. Lao; E.A. Lazarus; M.A. Makowski; M. Murakami; C.C. Petty; R.I. Pinsker; B.W. Rice; E.J. Strait, M.R. Wade; J.G. Watkins

    2000-01-01

    A major portion of the research program of the DIII-D tokamak collaboration is devoted to the development and demonstration of high performance advanced tokamak plasmas, with profiles as close as possible to those anticipated for steady-state operation. The work during the 1999 campaign has resulted in significant progress toward this goal. High normalized performance ((beta)(sub N)(approx) 4 and(beta)(sub N) H(sub 89)(approx) 9) discharges have been sustained for up to 2 s. These plasmas are in H-mode with rapid ELMs. The most common limiting phenomena are resistive wall modes (RWMs) rather than neoclassical tearing modes (NTMs). NTMs do occur, apparently triggered by the RWMs. The observed pressure is well above the calculated beta limit without a wall, and(beta)(sub N) and gt; 4(ell)(sub i) throughout the high performance phase. The bootstrap current is estimated to be and gt;50% of the total, and measurements of the internal loop voltage show that only about 25% of the current is inductively driven. The central q profile is flat, as is the calculated bootstrap current profile, due to the absence of any localized pressure gradients. The residual inductive current is localized around r/a(approx) 0.5. To demonstrate quasi-stationary operation, it will be necessary to replace the residual inductive current with ECCD at the same minor radius. To effectively apply ECH and ECCD to these discharges, density control will be needed. Preliminary experiments using the DIII-D cryopump have reduced the density by(approx)20%. A new EC power system and a new private flux cryopump will be available for the 2000 campaign

  8. Turbulence Scattering of High Harmonic Fast Waves

    International Nuclear Information System (INIS)

    M. Ono; J. Hosea; B. LeBlanc; J. Menard; C.K. Phillips; R. Wilson; P. Ryan; D. Swain; J. Wilgen; S. Kubota; and T.K. Mau

    2001-01-01

    Effect of scattering of high-harmonic fast-magnetosonic waves (HHFW) by low-frequency plasma turbulence is investigated. Due to the similarity of the wavelength of HHFW to that of the expected low-frequency turbulence in the plasma edge region, the scattering of HHFW can become significant under some conditions. The scattering probability increases with the launched wave parallel-phase-velocity as the location of the wave cut-off layer shifts toward the lower density edge. The scattering probability can be reduced significantly with higher edge plasma temperature, steeper edge density gradient, and magnetic field. The theoretical model could explain some of the HHFW heating observations on the National Spherical Torus Experiment (NSTX)

  9. Soft x-ray measurements of resistive wall mode behavior in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Delgado-Aparicio, L; Bell, R E; Gerhardt, S P; LeBlanc, B; Menard, J; Paul, S; Roquemore, L [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Stutman, D; Tritz, K; Finkenthal, M [Johns Hopkins University, Baltimore, MD 21218 (United States); Sabbagh, S A; Berkery, J W; Levesque, J P [Columbia University, New York, NY 10027 (United States); Lee, K C, E-mail: ldelgado@pppl.gov [University of California at Davis, Davis, CA 95616 (United States)

    2011-03-15

    A multi-energy soft x-ray (ME-SXR) array is used for the characterization of resistive wall modes (RWMs) in the National Spherical Torus Experiment (NSTX). Modulations in the time history of the ME-SXR emissivity profiles indicate the existence of edge density and core temperature fluctuations in good agreement with the slow evolution of the n = 1 magnetic perturbation measured by the poloidal and radial RWM coils. The characteristic 20-25 Hz frequency in the SXR diagnostics is approximately that of the n = 1 stable RWM, which is also near the measured peak of the resonant field amplification (RFA) and inversely proportional to the wall time. Together with the magnetics, the ME-SXR measurements suggest that in NSTX the RWM is not restricted exclusively to the reactor wall and edge, and that acting with the stabilizing coils on its global structure may result in density and temperature fluctuations that can be taken into account when designing the feedback process.

  10. Beta decay of highly charged ions

    International Nuclear Information System (INIS)

    Litvinov, Yuri A; Bosch, Fritz

    2011-01-01

    Beta decay of highly charged ions has attracted much attention in recent years. An obvious motivation for this research is that stellar nucleosynthesis proceeds at high temperatures where the involved atoms are highly ionized. Another important reason is addressing decays of well-defined quantum-mechanical systems, such as one-electron ions where all interactions with other electrons are excluded. The largest modifications of nuclear half-lives with respect to neutral atoms have been observed in beta decay of highly charged ions. These studies can be performed solely at ion storage rings and ion traps, because there high atomic charge states can be preserved for extended periods of time (up to several hours). Currently, all experimental results available in this field originate from experiments at the heavy-ion complex GSI in Darmstadt. There, the fragment separator facility FRS allows the production and separation of exotic, highly charged nuclides, which can then be stored and investigated in the storage ring facility ESR. In this review, we present and discuss in particular two-body beta decays, namely bound-state beta decay and orbital electron capture. Although we focus on experiments conducted at GSI, we will also attempt to provide general requirements common to any other experiment in this context. Finally, we address challenging but not yet performed experiments and we give prospects for the new radioactive beam facilities, such as FAIR in Darmstadt, IMP in Lanzhou and RIKEN in Wako.

  11. Effect of beta-endorphin imprinting during late pregnancy on the brain serotonin and plasma nocistatin levels of adult male rats.

    Science.gov (United States)

    Tekes, K; Gyenge, M; Hantos, M; Csaba, G

    2007-07-01

    Female rats were treated with 10 microg of beta-endorphin on the 19th day of pregnancy. Offspring were studied when five months old. Serotonin (5-HT) and 5-hydroxyindoleacetic acid (5-HIAA) content in four brain regions were determined by HPLC-EC and the nocistatin levels of blood plasma using RIA methods. In each brain region studied, the 5-HT levels were highly significantly reduced and that of 5-HIAA in three regions was highly significantly increased. When 5HIAA/5HT ratios, as a measure of serotonin turnover, were calculated, imprinted animals showed extremely high values. Plasma nocistatin level was also significantly elevated. The results call attention to the effect of perinatal endorphin imprinting and its long-term consequences (e.g., setting of aggressiveness, pain tolerance).

  12. PARTICLE-IN-CELL SIMULATIONS OF CONTINUOUSLY DRIVEN MIRROR AND ION CYCLOTRON INSTABILITIES IN HIGH BETA ASTROPHYSICAL AND HELIOSPHERIC PLASMAS

    International Nuclear Information System (INIS)

    Riquelme, Mario A.; Quataert, Eliot; Verscharen, Daniel

    2015-01-01

    We use particle-in-cell simulations to study the nonlinear evolution of ion velocity space instabilities in an idealized problem in which a background velocity shear continuously amplifies the magnetic field. We simulate the astrophysically relevant regime where the shear timescale is long compared to the ion cyclotron period, and the plasma beta is β ∼ 1-100. The background field amplification in our calculation is meant to mimic processes such as turbulent fluctuations or MHD-scale instabilities. The field amplification continuously drives a pressure anisotropy with p > p ∥ and the plasma becomes unstable to the mirror and ion cyclotron instabilities. In all cases, the nonlinear state is dominated by the mirror instability, not the ion cyclotron instability, and the plasma pressure anisotropy saturates near the threshold for the linear mirror instability. The magnetic field fluctuations initially undergo exponential growth but saturate in a secular phase in which the fluctuations grow on the same timescale as the background magnetic field (with δB ∼ 0.3 (B) in the secular phase). At early times, the ion magnetic moment is well-conserved but once the fluctuation amplitudes exceed δB ∼ 0.1 (B), the magnetic moment is no longer conserved but instead changes on a timescale comparable to that of the mean magnetic field. We discuss the implications of our results for low-collisionality astrophysical plasmas, including the near-Earth solar wind and low-luminosity accretion disks around black holes

  13. Plasma facing surface composition during NSTX Li experiments

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H., E-mail: cskinner@pppl.gov [Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States); Sullenberger, R. [Department of Mechanical and Aerospace Engineering, Princeton University, NJ 08540 (United States); Koel, B.E. [Department of Chemical and Biological Engineering, Princeton University, NJ 08540 (United States); Jaworski, M.A.; Kugel, H.W. [Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States)

    2013-07-15

    Lithium conditioned plasma facing surfaces have lowered recycling and enhanced plasma performance on many fusion devices. However, the nature of the plasma–lithium surface interaction has been obscured by the difficulty of in-tokamak surface analysis. We report laboratory studies of the chemical composition of lithium surfaces exposed to typical residual gases found in tokamaks. Solid lithium and a molybdenum alloy (TZM) coated with lithium have been examined using X-ray photoelectron spectroscopy, temperature programmed desorption, and Auger electron spectroscopy both in ultrahigh vacuum conditions and after exposure to trace gases. Lithium surfaces near room temperature were oxidized after exposure to 1–2 Langmuirs of oxygen or water vapor. The oxidation rate by carbon monoxide was four times less. Lithiated PFC surfaces in tokamaks will be oxidized in about 100 s depending on the tokamak vacuum conditions.

  14. Proceedings of 1999 U.S./Japan Workshop (99FT-05) On High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

    Energy Technology Data Exchange (ETDEWEB)

    NYGREN,RICHARD E.; STAVROS,DIANA T.

    2000-06-01

    The 1999 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions in Next Step Fusion Devices was held at the St. Francis Hotel in Santa Fe, New Mexico, on November 1-4, 1999. There were 42 presentations as well as discussion on technical issues and planning for future collaborations. The participants included 22 researchers from Japan and the United States as well as seven researchers from Europe and Russia. There have been important changes in the programs in both the US and Japan in the areas of plasma surface interactions and plasma facing components. The US has moved away from a strong focus on the ITER Project and has introduced new programs on use of liquid surfaces for plasma facing components, and operation of NSTX has begun. In Japan, the Large Helical Device began operation. This is the first large world-class confinement device operating in a magnetic configuration different than a tokamak. In selecting the presentations for this workshop, the organizers sought a balance between research in laboratory facilities or confinement devices related to plasma surface interactions and experimental research in the development of plasma facing components. In discussions about the workshop itself, the participants affirmed their preference for a setting where ''work-in-progress'' could be informally presented and discussed.

  15. Sustained high βN plasmas on EAST tokamak

    Science.gov (United States)

    Gao, Xiang; the EAST team

    2018-05-01

    Sustained high normalized beta (βN ∼ 1.9) plasmas with an ITER-like tungsten divertor have been achieved on EAST tokamak recently. The high power NBI heating system of 4.8 MW and the 4.6 GHz lower hybrid wave of 1 MW were developed and applied to produce edge and internal transport barriers in high βN discharges. The central flat q profile with q (ρ) ∼ 1 at ρ safety factor q95 = 4.7 is identified by the multi-channel far-infrared laser polarimeter and the EFIT code. The fraction of non-inductive current is about 40%. The relation between fishbone activity and ITB formation is observed and discussed.

  16. The high beta tokamak-extended pulse magnetohydrodynamic mode control research program

    International Nuclear Information System (INIS)

    Maurer, D A; Bialek, J; Byrne, P J; De Bono, B; Levesque, J P; Li, B Q; Mauel, M E; Navratil, G A; Pedersen, T S; Rath, N; Shiraki, D

    2011-01-01

    The high beta tokamak-extended pulse (HBT-EP) magnetohydrodynamic (MHD) mode control research program is studying ITER relevant internal modular feedback control coil configurations and their impact on kink mode rigidity, advanced digital control algorithms and the effects of plasma rotation and three-dimensional magnetic fields on MHD mode stability. A new segmented adjustable conducting wall has been installed on the HBT-EP and is made up of 20 independent, movable, wall shell segments instrumented with three distinct sets of 40 saddle coils, totaling 120 in-vessel modular feedback control coils. Each internal coil set has been designed with varying toroidal angular coil coverage of 5, 10 and 15 0 , spanning the toroidal angle range of an ITER port plug based internal coil to test resistive wall mode (RWM) interaction and multimode MHD plasma response to such highly localized control fields. In addition, we have implemented 336 new poloidal and radial magnetic sensors to quantify the applied three-dimensional fields of our control coils along with the observed plasma response. This paper describes the design and implementation of the new control shell incorporating these control and sensor coils on the HBT-EP, and the research program plan on the upgraded HBT-EP to understand how best to optimize the use of modular feedback coils to control instability growth near the ideal wall stabilization limit, answer critical questions about the role of plasma rotation in active control of the RWM and the ferritic resistive wall mode, and to improve the performance of MHD control systems used in fusion experiments and future burning plasma systems.

  17. On the conditions for the onset of nonlinear chirping structures in NSTX

    Science.gov (United States)

    Duarte, Vinicius; Podesta, Mario; Berk, Herbert; Gorelenkov, Nikolai

    2015-11-01

    The nonlinear dynamics of phase space structures is a topic of interest in tokamak physics in connection with fast ion loss mechanisms. The onset of phase-space holes and clumps has been theoretically shown to be associated with an explosive solution of an integro-differential, nonlocal cubic equation that governs the early mode amplitude evolution in the weakly nonlinear regime. The existence and stability of the solutions of the cubic equation have been theoretically studied as a function of Fokker-Planck coefficients for the idealized case of a single resonant point of a localized mode. From realistic computations of NSTX mode structures and resonant surfaces, we calculate effective pitch angle scattering and slowing-down (drag) collisional coefficients and analyze NSTX discharges for different cases with respect to chirping experimental observation. Those results are confronted to the theory that predicts the parameters region that allow for chirping to take place.

  18. Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta

    International Nuclear Information System (INIS)

    Schlossberg, David J.; Bodner, Grant M.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.

    2017-01-01

    Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β t ), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β t up to ~100% with a minimum |B| well spanning up to ~50% of the plasma volume.

  19. Confinement and heating of high beta plasmas with emphasis on compact toroids: Task 2, Stellarator and heliac research: Annual report, October 1, 1985-September 30, 1986

    International Nuclear Information System (INIS)

    Ribe, F.L.

    1986-01-01

    Progress over the past year has enabled us to complete the comprehensive experimental investigation of the equilibrium and stability of the linear heliac and the linear l = 1 stellarator configration. In the case of the heliac work, we discoverd axisymmetric hot plasma near the axial conductor (''hardcore'') in amounts which are comparable to the helically symmetric hot plasma centered on the magnetic axis. From this result came the motivation for an extended investigation which concentrates on the details of high-beta heliac formation. Important first results of the formation study have been obtained. In the case of the stellarator work, we have observed the flute-like m = 1 instability foe a specific set of experimental parameters and, for a different set, we have observed the mode stabilized by the combined effects of a finite ion Larmor radius and a nearby conducting wall. The single-discharge CV ion-temperature diagnostic system has been debugged and has yielded a heliac temperature measurement of (90 +- 10)eV. The plasma density diagnostic system based on cross-tube interferometry has been modified from its previous design for use with the somewhat wrinkly helical discharge tube

  20. Development of a High Resolution X-Ray Imaging Crystal Spectrometer for Measurement of Ion-Temperature and Rotation-Velocity Profiles in Fusion Energy Research Plasmas

    International Nuclear Information System (INIS)

    Hill, K.W.; Bitter, M.L.; Broennimann, Ch.; Eikenberry, E.F.; Ince-Cushman, A.; Lee, S.G.; Rice, J.E.; Scott, S.; Barnsley, R.

    2008-01-01

    A new imaging high resolution x-ray crystal spectrometer (XCS) has been developed to measure continuous profiles of ion temperature and rotation velocity in fusion plasmas. Following proof-of-principle tests on the Alcator C-Mod tokamak and the NSTX spherical tokamak, and successful testing of a new silicon, pixilated detector with 1MHz count rate capability per pixel, an imaging XCS is being designed to measure full profiles of T i and ν φ on C-Mod. The imaging XCS design has also been adopted for ITER. Ion-temperature uncertainty and minimum measurable rotation velocity are calculated for the C-Mod spectrometer. The affects of x-ray and nuclear-radiation background on the measurement uncertainties are calculated to predict performance on ITER

  1. Drift-kinetic Alfven modes in high performance tokamaks

    International Nuclear Information System (INIS)

    Jaun, A.; Fasoli, A.F.; Testa, D.; Vaclavik, J.; Villard, L.

    2001-01-01

    The stability of fast-particle driven Alfven eigenmodes is modeled in high performance tokamaks, successively with a conventional shear, an optimized shear and a tight aspect ratio plasma. A large bulk pressure yields global kinetic Alfven eigenmodes that are stabilized by mode conversion in the presence of a divertor. This suggests how conventional reactor scenarii could withstand significant pressure gradients from the fusion products. A large safety factor in the core q 0 >2.5 in deeply shear reversed configurations and a relatively large bulk ion Larmor radius in a low magnetic field can trigger global drift-kinetic Alfven eigenmodes that are unstable in high performance JET, NSTX and ITER plasmas. (author)

  2. Glycosaminoglycan and transforming growth factor beta1 changes in human plasma and urine during the menstrual cycle, in vitro fertilization treatment, and pregnancy.

    Science.gov (United States)

    De Muro, Pierina; Capobianco, Giampiero; Formato, Marilena; Lepedda, Antonio Junior; Cherchi, Gian Mario; Gordini, Laila; Dessole, Salvatore

    2009-07-01

    To evaluate transforming growth factor beta1 (TGF-beta1) and glycosaminoglycans (GAG) changes in human plasma and urine during the menstrual cycle, IVF-ET, and pregnancy. Prospective clinical study. University hospital. Thirteen women with apparently normal menstrual cycle (group 1); 18 women undergoing IVF-ET (group 2); and 14 low-risk pregnant women (group 3). We assayed plasma and urine concentrations of TGF-beta1, urine content, and distribution of GAG. Blood and urine samples were collected during days 2 to 3, 12 to 13, and 23 to 24 in group 1; in group 2, samples were obtained at menstrual phase, oocyte pick-up day, and 15 days after ET; in group 3, samples were obtained during gestational weeks 10-12, 22-24, and 30-32 and 1 month after delivery. Changes in TGF-beta1 and GAG content. The mean value of total urinary trypsin inhibitor/chondroitin sulfate (UTI/CS) showed a distinct peak at day 12 of the menstrual cycle in the fertile women in whom we monitored the ovulatory period. In the IVF-ET group, GAG distribution and TGF-beta1 levels showed significant differences during the cycle. We observed increased levels of plasma TGF-beta1 15 days after ET. A significant increase of total UTI/CS value with increasing gestation was detected. Transforming growth factor beta1 and GAG levels could represent an additional tool to monitor reproductive events and could be useful, noninvasive markers of ovulation and ongoing pregnancy.

  3. Energy Exchange Dynamics across L-H transitions in NSTX

    Science.gov (United States)

    Diallo, Ahmed

    2017-10-01

    H-mode is planned for future devices such as ITER, and is preceded by a low (L) to high (H) transition. A key question remains. What is the mechanism behind the L-H transition? Most theoretical descriptions of the L-H transition are based on the shear of the radial electric field and coincident ExB poloidal flow shear, which is thought to be responsible for the onset of the anomalous transport suppression that leads to the L-H transition. This talk will focus on the analysis of the flow dynamics across the L-H transition in NSTX. We analyze the L-H transition dynamics using the velocimetry of 2D edge turbulence data from gas-puff imaging (GPI). We determine the velocity components at the edge across the L-H transition for 17 discharges with three types of heating power (NBI, ohmic, and RF). Using a reduced model equation of edge flows and turbulence, the energy transfer dynamics is compared with the turbulence depletion hypothesis of the predator-prey model. In order for Reynolds work to suppress the turbulence, it must deplete the total turbulent free energy, including the thermal free-energy term. For this to occur, the increase in kinetic energy in the mean flow over the L-H transition must be comparable to the pre-transition thermal free energy. However, this ratio was found to be of order 10-2. Although there are significant simplifications in the theoretical model, they are unlikely to cause inaccuracy by two orders of magnitude, suggesting that direct turbulence depletion by the Reynolds work may not be large enough to explain the L-H transition on NSTX, contrary to the predator-prey model. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

  4. Energy exchange dynamics across L-H transitions in NSTX

    Science.gov (United States)

    Diallo, A.; Banerjee, S.; Zweben, S. J.; Stoltzfus-Dueck, T.

    2017-06-01

    We studied the energy exchange dynamics across the low-to-high-confinement (L-H) transition in NSTX discharges using the gas-puff imaging (GPI) diagnostic. The investigation focused on the energy exchange between flows and turbulence to help clarify the mechanism of the L-H transition. We applied this study to three types of heating schemes, including a total of 17 shots from the NSTX 2010 campaign run. Results show that the edge fluctuation characteristics (fluctuation levels, radial and poloidal correlation lengths) measured using GPI do not vary just prior to the H-mode transition, but change after the transition. Using a velocimetry approach (orthogonal-dynamics programming), velocity fields of a 24× 30 cm GPI view during the L-H transition were obtained with good spatial (˜1 cm) and temporal (˜2.5 μs) resolutions. Analysis using these velocity fields shows that the production term is systematically negative just prior to the L-H transition, indicating a transfer from mean flows to turbulence, which is inconsistent with the predator-prey paradigm. Moreover, the inferred absolute value of the production term is two orders of magnitude too small to explain the observed rapid L-H transition. These discrepancies are further reinforced by consideration of the ratio between the kinetic energy in the mean flow to the thermal free energy, which is estimated to be much less than 1, suggesting again that the turbulence depletion mechanism may not play an important role in the transition to the H-mode. Although the Reynolds work therefore appears to be too small to directly deplete the turbulent free energy reservoir, order-of-magnitude analysis shows that the Reynolds stress may still make a non-negligible contribution to the observed poloidal flows.

  5. Origins of Beta Tantalum in Sputtered Coatings

    National Research Council Canada - National Science Library

    Mulligan, C

    2001-01-01

    .... Some of the most recent work has attempted to relate the energetics (i.e., atom/ion energy) of the plasma to the alpha right arrow beta transition. It has been shown that the energetics of the plasma can relate to the most crucial sputtering parameters. The most significant feature of the use of plasma energy to explain the alpha right arrow beta transition is that it relates the formation of beta-tantalum to a quantifiable measure.

  6. DbAccess: Interactive Statistics and Graphics for Plasma Physics Databases

    International Nuclear Information System (INIS)

    Davis, W.; Mastrovito, D.

    2003-01-01

    DbAccess is an X-windows application, written in IDL(reg s ign), meeting many specialized statistical and graphical needs of NSTX [National Spherical Torus Experiment] plasma physicists, such as regression statistics and the analysis of variance. Flexible ''views'' and ''joins,'' which include options for complex SQL expressions, facilitate mixing data from different database tables. General Atomics Plot Objects add extensive graphical and interactive capabilities. An example is included for plasma confinement-time scaling analysis using a multiple linear regression least-squares power fit

  7. Dissipation of Alfven Waves at Fluid Scale through Parametric Decay Instabilities in Low-beta Turbulent Plasma

    Science.gov (United States)

    Fu, X.; Li, H.; Guo, F.; Li, X.; Roytershteyn, V.

    2017-12-01

    The solar wind is a turbulent magnetized plasma extending from the upper atmosphere of the sun to the edge of the heliosphere. It carries charged particles and magnetic fields originated from the Sun, which have great impact on the geomagnetic environment and human activities in space. In such a magnetized plasma, Alfven waves play a crucial role in carrying energy from the surface of the Sun, injecting into the solar wind and establishing power-law spectra through turbulent energy cascades. On the other hand, in compressible plasmas large amplitude Alfven waves are subject to a parametric decay instability (PDI) which converts an Alfven wave to another counter-propagating Alfven wave and an ion acoustic wave (slow mode). The counter-propagating Alfven wave provides an important ingredient for turbulent cascade, and the slow-mode wave provides a channel for solar wind heating in a spatial scale much larger than ion kinetic scales. Growth and saturation of PDI in quiet plasma have been intensively studied using linear theory and nonlinear simulations in the past. Here using 3D hybrid simulations, we show that PDI is still effective in turbulent low-beta plasmas, generating slow modes and causing ion heating. Selected events in WIND data are analyzed to identify slow modes in the solar wind and the role of PDI, and compared with our simulation results. We also investigate the validity of linear Vlasov theory regarding PDI growth and slow mode damping in turbulent plasmas. Since PDI favors low plasma beta, we expect to see more evidence of PDI in the solar wind close to the Sun, especially from the upcoming NASA's Parker Solar Probe mission which will provide unprecedented wave and plasma data as close as 8.5 solar radii from the Sun.

  8. Princeton Plasma Physics Laboratory Annual Site Environmental Report for Calendar Year 2000

    Energy Technology Data Exchange (ETDEWEB)

    Virginia L. Finley

    2002-04-22

    The results of the 2000 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and nonradioactive pollutants (if any) that are added to the environment as a result of PPPL's operations. The report also summarizes environmental initiatives, assessments, and programs that were undertaken in 2000. The Princeton Plasma Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program is to create innovations to make fusion power a practical reality -- an alternative energy source. The year 2000 marked the second year of National Spherical Torus Experiment (NSTX) operations and Tokamak Fusion Test Reactor (TFTR) dismantlement and deconstruction activities. A collaboration among fourteen national laboratories, universities, and research institutions, the NSTX is a major element in the U.S. Fusion Energy Sciences Program. It has been designed to test the physics principles of spherical torus (ST) plasmas. The ST concept could play an important role in the development of smaller, more economical fusion power plants. With its completion within budget and ahead of its target schedule, NSTX first plasma occurred on February 12, 1999. In 2000, PPPL's radiological environmental monitoring program measured tritium in the air at on-site and off-site sampling stations. PPPL is capable of detecting small changes in the ambient levels of tritium by using highly sensitive monitors. The operation of an in-stack monitor located on D-site is a requirement of the National Emission Standard for Hazardous Air Pollutants (NESHAPs) regulations with limits set by the Environmental Protection Agency (EPA). Also included in PPPL's radiological environmental monitoring program, are precipitation, surface

  9. Princeton Plasma Physics Laboratory Annual Site Environmental Report for Calendar Year 2000

    International Nuclear Information System (INIS)

    Virginia L. Finley

    2002-04-01

    The results of the 2000 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and nonradioactive pollutants (if any) that are added to the environment as a result of PPPL's operations. The report also summarizes environmental initiatives, assessments, and programs that were undertaken in 2000. The Princeton Plasma Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program is to create innovations to make fusion power a practical reality -- an alternative energy source. The year 2000 marked the second year of National Spherical Torus Experiment (NSTX) operations and Tokamak Fusion Test Reactor (TFTR) dismantlement and deconstruction activities. A collaboration among fourteen national laboratories, universities, and research institutions, the NSTX is a major element in the U.S. Fusion Energy Sciences Program. It has been designed to test the physics principles of spherical torus (ST) plasmas. The ST concept could play an important role in the development of smaller, more economical fusion power plants. With its completion within budget and ahead of its target schedule, NSTX first plasma occurred on February 12, 1999. In 2000, PPPL's radiological environmental monitoring program measured tritium in the air at on-site and off-site sampling stations. PPPL is capable of detecting small changes in the ambient levels of tritium by using highly sensitive monitors. The operation of an in-stack monitor located on D-site is a requirement of the National Emission Standard for Hazardous Air Pollutants (NESHAPs) regulations with limits set by the Environmental Protection Agency (EPA). Also included in PPPL's radiological environmental monitoring program, are precipitation, surface, ground, a nd

  10. Plasma Membrane Protein Profiling in Beta-Amyloid-Treated Microglia Cell Line.

    Science.gov (United States)

    Correani, Virginia; Di Francesco, Laura; Mignogna, Giuseppina; Fabrizi, Cinzia; Leone, Stefano; Giorgi, Alessandra; Passeri, Alessia; Casata, Roberto; Fumagalli, Lorenzo; Maras, Bruno; Schininà, M Eugenia

    2017-09-01

    In the responsiveness of microglia to toxic stimuli, plasma membrane proteins play a key role. In this study we treated with a synthetic beta amyloid peptide murine microglial cells metabolically differently labelled with stable isotope amino acids (SILAC). The plasma membrane was selectively enriched by a multi-stage aqueous two-phase partition system. We were able to identify by 1D-LC-MS/MS analyses 1577 proteins, most of them are plasma membrane proteins according to the Gene Ontology annotation. An unchanged level of amyloid receptors in this data set suggests that microglia preserve their responsiveness capability to the environment even after 24-h challenge with amyloid peptides. On the other hand, 14 proteins were observed to change their plasma membrane abundance to a statistically significant extent. Among these, we proposed as reliable biomarkers of the inflammatory microglia phenotype in AD damaged tissues MAP/microtubule affinity-regulating kinase 3 (MARK3), Interferon-induced transmembrane protein 3 (IFITM3), Annexins A5 and A7 (ANXA5, ANXA7) and Neuropilin-1 (NRP1), all proteins known to be involved in the inflammation processes and in microtubule network assembly rate. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  11. Operation of the NSTX Thomson Scattering System

    International Nuclear Information System (INIS)

    LeBlanc, B.P.; Bell, R.E.; Johnson, D.W.; Hoffman, D.E.; Long, D.C.; Palladino, R.W.

    2002-01-01

    The NSTX multi-point Thomson scattering system has been in operation for nearly two years and provides routine Te(R,t) and ne(R,t) measurements. The laser beams from two 30-Hz Nd:YAG lasers are imaged by a spherical mirror onto 36 fiber-optics bundles. In the present configuration, the output ends of 20 of these bundles are instrumented with filter polychromators and avalanche photodiode detectors. In this paper, we discuss the laser implementation and the installed collection optics. We follow with examples of raw and analyzed data. We close with some comments about calibration

  12. Plasma tomographic reconstruction from tangentially viewing camera with background subtraction

    Czech Academy of Sciences Publication Activity Database

    Odstrčil, M.; Mlynář, Jan; Weinzettl, Vladimír; Háček, Pavel; Odstrčil, T.; Verdoolaege, G.; Berta, M.; Szabolics, T.; Bencze, A.

    2014-01-01

    Roč. 85, č. 1 (2014), 013509-013509 ISSN 0034-6748 R&D Projects: GA ČR GAP205/10/2055 Institutional support: RVO:61389021 Keywords : X-ray tomography * edge turbulence * tokamak * NSTX * TCV * COMPASS Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.614, year: 2014 http://scitation.aip.org/content/aip/journal/rsi/85/1/10.1063/1.4862652

  13. Beta normal control of TFTR using fuzzy logic

    International Nuclear Information System (INIS)

    Lawson, J.E.; Bell, M.G.; Marsala, R.J.; Mueller, D.

    1995-01-01

    In TFTR plasmas heated by neutral beam injection, the fusion power yield increases rapidly with the plasma pressure. However, the pressure is limited by the onset of instabilities which may result in plasma disruptions that would have had an adverse effect on the performance of subsequent discharges and increase the risk of damage to internal components. The likelihood of disruption has been found to correlate with the normalized beta, defined as βN = 2 x 10 8 μ circle left angle p perpendicular to right angle a / BTIp where left angle p perpendicular to right angle is the volume-average plasma perpendicular pressure, a the mid-plane minor radius of the plasma, BT the toroidal magnetic field and Ip the plasma current. Other variables, such as the peaking of the plasma pressure and current profiles, have been found to influence the threshold of βN at which the probability of disruption begins to increase significantly. For TFTR plasmas with high fusion performance (TFTR ''supershots'') the probability of disruption has been found to increase rapidly for βN > 1.8. Since confinement in this regime is affected by plasma-wall interaction, which can vary from shot to shot, operation at high βN with preprogrammed heating power pulses can produce an unacceptably high risk of disruption. To reduce the risk of producing beta-limit disruptions during neutral beam heating experiments, a control system, the Neutral Beam Power Feedback System (NBPFS), has been developed to modulate the total heating power by switching individual neutral beam sources on and off in response to the evolution of the normalized beta so that the limit will not be exceeded. The value of βN is calculated in real time and transmitted to the NBPFS. The value of βN and its calculated time derivative are input to a fuzzy logic controller which implements a proportional-derivative control based on the difference between βN and a programmed reference level βNREF which can be programmed as a function

  14. Controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    1994-07-01

    40 papers are presented at this 21. conference on controlled fusion and plasma physics (JET). Titles are: effects of sawtooth crashes on beams ions and fusion product tritons; beta limits in H-modes and VH-modes; impurity induced neutralization of MeV energy protons in JET plasmas; lost α particle diagnostic for high-yield D-T fusion plasmas; 15-MeV proton emission from ICRF-heated plasmas; pulse compression radar reflectometry for density measurements; gamma-ray emission profile measurements during ICRH discharges; the new JET phase ICRH array; simulation of triton burn-up; parametric dependencies of JET electron temperature profiles; detached divertor plasmas; excitation of global Alfven Eigenmodes by RF heating; mechanisms of toroidal rotation; effect of shear in the radial electric field on confinement; plasma transport properties at the L-H transition; numerical study of plasma detachment conditions in JET divertor plasmas; the SOL width and the MHD interchange instability; non linear magnetic reconnection in low collisionality plasmas; topology and slowing down of high energy ion orbits; sawtooth crashes at high beta; fusion performances and alpha heating in future JET D-T plasmas; a stable route to high-beta plasmas with non-monotonic q-profiles; theory of propagation of changes to confinement; spatial distribution of gamma emissivity and fast ions during ICRF heating; multi-camera soft X-ray diagnostic; radiation phenomena and particle fluxes in the X-event; local measurement of transport parameters for laser injected trace impurities; impurity transport of high performance discharges; negative snakes and negative shear; neural-network charge exchange analysis; ion temperature anisotropy in helium neutral beam fuelling; impurity line emission due to thermal charge exchange in edge plasmas; control of convection by fuelling and pumping; VH mode accessibility and global H-mode properties; ion cyclotron emission by spontaneous emission; LHCD/ICRH synergy

  15. Ion Heating by Fast Particle Induced Alfven Turbulence

    International Nuclear Information System (INIS)

    Gates, D.; Gorelenkov, N.; White, R.B.

    2001-01-01

    A novel mechanism that directly transfers energy from Super-Alfvenic energetic ions to thermal ions in high-beta plasmas is described. The mechanism involves the excitation of compressional Alfvin eigenmodes (CAEs) in the frequency range with omega less than or approximately equal to omega(subscript ci). The broadband turbulence resulting from the large number of excited modes causes stochastic diffusion in velocity space, which transfers wave energy to thermal ions. This effect may be important on the National Spherical Torus Experiment (NSTX), and may scale up to reactor scenarios. This has important implications for low aspect ratio reactor concepts, since it potentially allows for the modification of the ignition criterion

  16. High-beta linac structures

    International Nuclear Information System (INIS)

    Schriber, S.O.

    1979-01-01

    Accelerating structures for high-beta linacs that have been and are in use are reviewed in terms of their performance. Particular emphasis is given to room-temperature structures and the disk-and-washer structure. The disk-and-washer structure has many attractive features that are discussed for pulsed high-gradient linacs, for 100% duty-cycle medium-gradient linacs and for high-current linacs requiring maximal amounts of stored energy in the electric fields available to the beam

  17. Equilibrium and stability studies for high beta plasmas in torsatron/heliotron devices

    International Nuclear Information System (INIS)

    Carreras, B.A.; Cooper, W.A.; Charlton, L.A.

    1983-01-01

    The equilibrium and stability properties of high β plasmas in torsatron/heliotron devices have been investigated. Three numerical approaches have been used to study plasma equilibria for a range of coil configurations. The method of averaging permits fast equilibrium and stability calculations. Two fully 3-D codes, namely the Chodura-Schluter code, and the NEAR code recently developed at ORNL, are used to explore selected regions of parameter space. The resulting equilibria calculated with different methods are in good agreement. This validates the average method approach and enhances its usefulness. Results are presented for configurations with different aspect ratios and number of field periods. The role of the vertical field has also been studied in detail. The main conclusion is that for moderate aspect ratios (Asub(p) <= 8), the self-stabilizing effect of the magnetic axis shift is large enough to open a direct path to the second stability regime. (author)

  18. Theory and observation of compressional Alfven eigenmodes in low aspect ratio plasma

    International Nuclear Information System (INIS)

    Gorelenkov, N.N.

    2002-01-01

    A new theory of radially and poloidally localized Compressional Alfven Eigenmodes (CAE) in low aspect ratio plasma is reported. The theory is applied to identify recently observed instabilities in the MHz frequency range in National Spherical Torus experiments (NSTX). The frequency of observed CAEs is correlated with the characteristic Alfven velocity of the plasma. The observed high frequency modes are explained as CAEs driven by energetic beam ions. The CAE frequency is determined by the Alfven frequency at the mode location on the low field side of the plasma and is given approximately by ω CAE v A m=r, where m is the poloidal mode number, and r is the local minor radius. CAEs are destabilized by free energy in the energetic ion velocity space gradient via Doppler shifted cyclotron resonance with beam ions. Properties of the CAE instability driven by different NBI ion distributions are presented. (author)

  19. Elimination of inter-discharge helium glow discharge cleaning with lithium evaporation in NSTX

    Directory of Open Access Journals (Sweden)

    R. Maingi

    2017-08-01

    Full Text Available Operation in the National Spherical Torus Experiment (NSTX typically used either periodic boronization and inter-shot helium glow discharge cleaning (HeGDC, or inter-shot lithium evaporation without boronization, and initially with inter-shot HeGDC. To assess the viability of operation without HeGDC, dedicated experiments were conducted in which Li evaporation was used while systematically shrinking the HeGDC between shots from the standard 10min to zero (10→6.5→4→0. Good shot reproducibility without HeGDC was achieved with lithium evaporations of 100mg or higher; evaporations of 200–300mg typically resulted in very low ELM frequency or ELM-free operation, reduced wall fueling, and improved energy confinement. The use of HeGDC before lithium evaporation modestly reduced Dα in the outer scrape-off layer, but not at the strike point. Pedestal electron and ion temperature also improved modestly, suggesting that HeGDC prior to lithium evaporation is a useful tool for experiments that seek to maximize plasma performance.

  20. Equilibrium and stability studies for high-beta plasmas in torsatron/heliotron devices

    International Nuclear Information System (INIS)

    Carreras, B.A.; Charlton, L.A.; Cooper, W.A.

    1983-01-01

    The equilibrium and stability properties of high-#betta# plasmas in torsatron/heliotron devices have been investigated. Three numerical approaches have been used to study plasma equilibria for a range of coil configurations. The method of averaging permits fast equilibrium and stability calculations. Two fully 3-D codes, namely the Chodura-Schluter code, and the NEAR code recently developed at ORNL, are used to explore selected regions of parameter space. The resulting equilibria calculated with different methods are in good agreement. This validates the average method approach and enhances its usefulness. Results are presented for configurations with different aspect ratios and number of field periods. The role of the vertical field has also been studied in detail. The main conclusion is that for moderate aspect ratios (A/sub p/ less than or equal to 8), the self-stabilizing effect of the magnetic-axis shift is large enough to open a direct path to the second-stability regime