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Sample records for torus experiment h-mode

  1. Gyrokinetic Stability Studies of the Microtearing Mode in the National Spherical Torus Experiment H-mode

    International Nuclear Information System (INIS)

    Baumgaertel J.A., Redi M.H., Budny R.V., Rewoldt G., Dorland W.

    2005-01-01

    Insight into plasma microturbulence and transport is being sought using linear simulations of drift waves on the National Spherical Torus Experiment (NSTX), following a study of drift wave modes on the Alcator C-Mod Tokamak. Microturbulence is likely generated by instabilities of drift waves, which cause transport of heat and particles. Understanding this transport is important because the containment of heat and particles is required for the achievement of practical nuclear fusion. Microtearing modes may cause high heat transport through high electron thermal conductivity. It is hoped that microtearing will be stable along with good electron transport in the proposed low collisionality International Thermonuclear Experimental Reactor (ITER). Stability of the microtearing mode is investigated for conditions at mid-radius in a high density NSTX high performance (H-mode) plasma, which is compared to the proposed ITER plasmas. The microtearing mode is driven by the electron temperature gradient, and believed to be mediated by ion collisions and magnetic shear. Calculations are based on input files produced by TRXPL following TRANSP (a time-dependent transport analysis code) analysis. The variability of unstable mode growth rates is examined as a function of ion and electron collisionalities using the parallel gyrokinetic computational code GS2. Results show the microtearing mode stability dependence for a range of plasma collisionalities. Computation verifies analytic predictions that higher collisionalities than in the NSTX experiment increase microtearing instability growth rates, but that the modes are stabilized at the highest values. There is a transition of the dominant mode in the collisionality scan to ion temperature gradient character at both high and low collisionalities. The calculations suggest that plasma electron thermal confinement may be greatly improved in the low-collisionality ITER

  2. Enhancement of mode-converted electron Bernstein wave emission during National Spherical Torus Experiment H-mode plasmas

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C.; Jones, B.; Le Blanc, B.P.; Maingi, R.

    2002-01-01

    A sudden, threefold increase in emission from fundamental electrostatic electron Bernstein waves (EBW) which mode convert and tunnel to the electromagnetic X-mode has been observed during high energy and particle confinement (H-mode) transitions in the National Spherical Torus Experiment (NSTX) plasma [M. Ono, S. Kaye, M. Peng et al., in Proceedings of the 17th IAEA Fusion Energy Conference (IAEA, Vienna, Austria, 1999), Vol. 3, p. 1135]. The mode-converted EBW emission viewed normal to the magnetic field on the plasma midplane increases when the density profile steepens in the vicinity of the mode conversion layer, which is located in the plasma scrape off. The measured conversion efficiency during the H-mode is consistent with the calculated EBW to X-mode conversion efficiency derived using edge density data. Calculations indicate that there may also be a small residual contribution to the measured X-mode electromagnetic radiation from polarization-scrambled, O-mode emission, converted from EBWs

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

    Maqueda, R.J.; Stotler, D.P.; Zweben, S.J.

    2010-01-01

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

  6. The Bumpy Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Cobble, James Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-06-09

    This document summarizes the Bumpy Torus Experiment as a viable fusion reactor concept. Conclusions reached include the following: In 30 years, order-of-magnitude technological advances have occurred in multiple areas of plasma heating and confinement. The ORNL bumpy torus of the 1970s was technology limited. Now that ITER is technology limited, an alternate concept is needed. A device built on such a concept should be current free, CW, modular, have a gentle shutdown, and demonstrable stability. The bumpy torus meets or has the potential to meet all of these criteria. Earlier, stability was not possible due to power limits; it has not been fully tested. It is time to revisit the bumpy-torus concept with a modest new machine.

  7. The Bumpy Torus Experiment

    International Nuclear Information System (INIS)

    Cobble, James Allen

    2016-01-01

    This document summarizes the Bumpy Torus Experiment as a viable fusion reactor concept. Conclusions reached include the following: In 30 years, order-of-magnitude technological advances have occurred in multiple areas of plasma heating and confinement. The ORNL bumpy torus of the 1970s was technology limited. Now that ITER is technology limited, an alternate concept is needed. A device built on such a concept should be current free, CW, modular, have a gentle shutdown, and demonstrable stability. The bumpy torus meets or has the potential to meet all of these criteria. Earlier, stability was not possible due to power limits; it has not been fully tested. It is time to revisit the bumpy-torus concept with a modest new machine.

  8. Limiter H-mode experiments on TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Bush, C [Oak Ridge National Lab., TN (USA); Bretz, N L; Fredrickson, E D; McGuire, K M; Nazikian, R; Park, H K; Schivell, J; Taylor, G; Bitter, B; Budny, R; Cohen, S A; Kilpatrick, S J; LeBlanc, B; Manos, D M; Meade, D; Paul, S F; Scott, S D; Stratton, B C; Synakowski, E J; Towner, H H; Weiland, R M; Arunasalam, V; Bateman, G; Bell, M G; Bell, R; Boivin, R; Cavallo, A; Cheng, C Z; Chu, T K; Cowl,

    1990-12-15

    Limiter H-modes with centrally peaked density profiles have been obtained in TFTR using a highly conditioned graphite limiter. The transition to these centrally peaked H-modes takes place from the supershot to the H-mode rather than the usual L- to H-mode transition observed on other tokamaks. Bi-directional beam heating is required to induce the transition. Density peaking factors, n{sub e}(0)/{l angle}n{sub e}{r angle}, >2.3 are obtained and at the same time the H-mode characteristics are similar to those of limiter H-modes on other tokamaks and the global confinement, {tau}{sub E}, can be >2.5 times L-mode scaling. The TRANSP analysis shows that transport in these H-modes is similar to that of supershots within the inner 60 cm of the plasma, but the stored electron energy (calculated using measured values of T{sub e} and n{sub e}) is higher for the H-mode at the plasma edge. Microwave scattering near the edge shows broad spectra at k = 5.5 cm{sup {minus}1} which begin at the drop in D{sub {alpha}} radiation and are strongly shifted in the electron diamagnetic drift direction. At the same time beam emission spectroscopy shows a coherent mode near the boundary with m = 15--20 at 20--30 kHz which is propagating in the ion direction. During an ELM event these apparent rotations cease and Mirnov fluctuations in the 50--500 kHz increase in intensity.

  9. 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.

  10. JET Radiative Mantle Experiments in ELMy H-Mode

    International Nuclear Information System (INIS)

    Budny, R.; Coffey, I.; Dumortier, P.; Grisolia, C.; Strachan, J.D.

    1999-01-01

    Radiative mantle experiments were performed on JET ELMy H-mode plasmas. The Septum configuration was used where the X-point is embedded into the top of the Septum. Argon radiated 50% of the input power from the bulk plasma while Z eff rose from an intrinsic level of 1.5 to about 1.7 due to the injected Argon. The total energy content and global energy confinement time decreased 15% when the impurities were introduced. In contrast, the effective thermal diffusivity in the core confinement region (r/a = .4--.8) decreased by 30%. Usually, JET ELMy H-mode plasmas have confinement that is correlated to the edge pedestal pressure. The radiation lowered the edge pedestal and consequently lowered the global confinement. Thus the confinement was changed by a competition between the edge pedestal reduction lowering the confinement and the weaker RI effect upon the core transport coefficients raising the confinement. The ELM frequency increased from 10 Hz Type I ELMs, to 200 Hz type III ELMs. The energy lost by each ELM reduced to 0.5% of the plasma energy content

  11. Recent results from the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Maingi, R; Bell, M G; Bell, R E; Bialek, J; Bourdelle, C; Bush, C E; Darrow, D S; Fredrickson, E D; Gates, D A; Gilmore, M; Gray, T; Jarboe, T R; Johnson, D W; Kaita, R; Kaye, S M; Kubota, S; Kugel, H W; LeBlanc, B P; Maqueda, R J; Mastrovito, D; Medley, S S; Menard, J E; Mueller, D; Nelson, B A; Ono, M; Paoletti, F; Park, H K; Paul, S F; Peebles, T; Peng, Y-K M; Phillips, C K; Raman, R; Rosenberg, A L; Roquemore, A L; Ryan, P M; Sabbagh, S A; Skinner, C H; Soukhanovskii, V A; Stutman, D; Swain, D W; Synakowski, E J; Taylor, G; Wilgen, J; Wilson, J R; Wurden, G A; Zweben, S J

    2003-01-01

    The National Spherical Torus Experiment (NSTX) is a low aspect-ratio fusion research facility whose research goal is to make a determination of the attractiveness of the spherical torus concept in the areas of high-β stability, confinement, current drive, and divertor physics. Remarkable progress was made in extending the operational regime of the device in FY 2002. In brief, β t of 34% and β N of 6.5 were achieved. H-mode became the main operational regime, and energy confinement exceeded conventional aspect-ratio tokamak scalings. Heating was demonstrated with the radiofrequency antenna, and signatures of current drive were observed. Current initiation with coaxial helicity injection produced discharges of 400 kA, and first measurements of divertor heat flux profiles in H-mode were made

  12. 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

  13. 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

  14. Feasibility study for the Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Lazarus, E.A.; Attenberger, S.E.; Baylor, L.R.

    1985-10-01

    The design of the Spherical Torus Experiment (STX) is discussed. The physics of the plasma are given in a magnetohydrodynamic model. The structural aspects and instrumentation of the device are described. 19 refs., 103 figs

  15. Results of the H-mode experiments with JT-60 outer and lower divertors

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Tsuji, Shunji; Nagami, Masayuki

    1989-08-01

    In JT-60, hydrogen H-mode experiments with outer and lower divertors were performed. In the outer divertor, H-mode were obtained, similar to the ones observed in the other lower/upper divertors. Its threshold absorbed power and electron density were 16 MW and 1.8 x 10 19 m -3 . In the two combined heatings with NB+ICRF and NB+LHRF, H-mode discharges are also obtained. Moreover, in new configuration of lower divertor, H-mode phases without and with ELM are obtained. Typical results of the lower divertor are shown to compare the H-mode characteristics between the two configurations. Improvement of the energy confinement time in the two divertors was limited to 10 %. Analyses on ballooning/interchange instabilities were carried out with precise equlibria of JT-60. These results showed that the both modes were enough stable. (author)

  16. Summary of US compact torus experiments

    International Nuclear Information System (INIS)

    Hartman, C.W.

    1981-01-01

    During the past several years a rapid increase has occurred in compact torus (CT) research in the United States, reflecting renewed interest in this simplified reactor consequences of this configuration. This paper reviews early approaches to CT formation and results and summarizes present experimental studies. Recent experiments have demonstrated a number of macroscopic aspects of the CT, including the conditions under which a macroscopically stable CT can be formed and maintained. Scaling experiments and more detailed studies of plasma transport in progress are discussed along with experiments under construction

  17. H-mode physics

    International Nuclear Information System (INIS)

    Itoh, Sanae.

    1991-06-01

    After the discovery of the H-mode in ASDEX ( a tokamak in Germany ) the transition between the L-mode ( Low confinement mode ) and H-mode ( High confinement mode ) has been observed in many tokamaks in the world. The H-mode has made a breakthrough in improving the plasma parameters and has been recognized to be a universal phenomena. Since its discovery, the extensive studies both in experiments and in theory have been made. The research on H-mode has been casting new problems of an anomalous transport across the magnetic surface. This series of lectures will provide a brief review of experiments for explaining H-mode and a model theory of H-mode transition based on the electric field bifurcation. If the time is available, a new theoretical model of the temporal evolution of the H-mode will be given. (author)

  18. 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

  19. Characterization and parametric dependencies of low wavenumber pedestal turbulence in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D. R.; Fonck, R. J.; McKee, G. R.; Thompson, D. S. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Bell, R. E.; Diallo, A.; Guttenfelder, W.; Kaye, S. M.; LeBlanc, B. P.; Podesta, M. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2013-05-15

    The spherical torus edge region is among the most challenging regimes for plasma turbulence simulations. Here, we measure the spatial and temporal properties of ion-scale turbulence in the steep gradient region of H-mode pedestals during edge localized mode-free, MHD quiescent periods in the National Spherical Torus Experiment. Poloidal correlation lengths are about 10 ρ{sub i}, and decorrelation times are about 5 a/c{sub s}. Next, we introduce a model aggregation technique to identify parametric dependencies among turbulence quantities and transport-relevant plasma parameters. The parametric dependencies show the most agreement with transport driven by trapped-electron mode, kinetic ballooning mode, and microtearing mode turbulence, and the least agreement with ion temperature gradient turbulence. In addition, the parametric dependencies are consistent with turbulence regulation by flow shear and the empirical relationship between wider pedestals and larger turbulent structures.

  20. Structure and motion of edge turbulence in the National Spherical Torus Experiment and Alcator C-Moda)

    Science.gov (United States)

    Zweben, S. J.; Maqueda, R. J.; Terry, J. L.; Munsat, T.; Myra, J. R.; D'Ippolito, D.; Russell, D. A.; Krommes, J. A.; LeBlanc, B.; Stoltzfus-Dueck, T.; Stotler, D. P.; Williams, K. M.; Bush, C. E.; Maingi, R.; Grulke, O.; Sabbagh, S. A.; White, A. E.

    2006-05-01

    In this paper we compare the structure and motion of edge turbulence observed in L-mode vs. H-mode plasmas in the National Spherical Torus Experiment (NSTX) [M. Ono, M. G. Bell, R. E. Bell et al., Plasma Phys. Controlled Fusion 45, A335 (2003)]. The radial and poloidal correlation lengths are not significantly different between the L-mode and the H-mode in the cases examined. The poloidal velocity fluctuations are lower and the radial profiles of the poloidal turbulence velocity are somewhat flatter in the H-mode compared with the L-mode plasmas. These results are compared with similar measurements Alcator C-Mod [E. Marmar, B. Bai, R. L. Boivin et al., Nucl. Fusion 43, 1610 (2003)], and with theoretical models.

  1. Tungsten Transport in the Core of JET H-mode Plasmas, Experiments and Modelling

    Science.gov (United States)

    Angioni, Clemente

    2014-10-01

    The physics of heavy impurity transport in tokamak plasmas plays an essential role towards the achievement of practical fusion energy. Reliable predictions of the behavior of these impurities require the development of realistic theoretical models and a complete understanding of present experiments, against which models can be validated. Recent experimental campaigns at JET with the ITER-like wall, with a W divertor, provide an extremely interesting and relevant opportunity to perform this combined experimental and theoretical research. Theoretical models of both neoclassical and turbulent transport must consistently include the impact of any poloidal asymmetry of the W density to enable quantitative predictions of the 2D W density distribution over the poloidal cross section. The agreement between theoretical predictions and experimentally reconstructed 2D W densities allows the identification of the main mechanisms which govern W transport in the core of JET H-mode plasmas. Neoclassical transport is largely enhanced by centrifugal effects and the neoclassical convection dominates, leading to central accumulation in the presence of central peaking of the density profiles and insufficiently peaked ion temperature profiles. The strength of the neoclassical temperature screening is affected by poloidal asymmetries. Only around mid-radius, turbulent diffusion offsets neoclassical transport. Consistently with observations in other devices, ion cyclotron resonance heating in the plasma center can flatten the electron density profile and peak the ion temperature profile and provide a means to reverse the neoclassical convection. MHD activity may hamper or speed up the accumulation process depending on mode number and plasma conditions. Finally, the relationship of JET results to a parallel modelling activity of the W behavior in the core of ASDEX Upgrade plasmas is presented. This project has received funding from the European Union's Horizon 2020 research and innovation

  2. Power requirements for superior H-mode confinement on Alcator C-Mod: experiments in support of ITER

    International Nuclear Information System (INIS)

    Hughes, J.W.; Reinke, M.L.; Terry, J.L.; Brunner, D.; Greenwald, M.; Hubbard, A.E.; LaBombard, B.; Lipschultz, B.; Ma, Y.; Wolfe, S.; Wukitch, S.J.; Loarte, A.

    2011-01-01

    Power requirements for maintaining sufficiently high confinement (i.e. normalized energy confinement time H 98 ≥ 1) in H-mode and its relation to H-mode threshold power scaling, P th , are of critical importance to ITER. In order to better characterize these power requirements, recent experiments on the Alcator C-Mod tokamak have investigated H-mode properties, including the edge pedestal and global confinement, over a range of input powers near and above P th . In addition, we have examined the compatibility of impurity seeding with high performance operation, and the influence of plasma radiation and its spatial distribution on performance. Experiments were performed at 5.4 T at ITER relevant densities, utilizing bulk metal plasma facing surfaces and an ion cyclotron range of frequency waves for auxiliary heating. Input power was scanned both in stationary enhanced D α (EDA) H-modes with no large edge localized modes (ELMs) and in ELMy H-modes in order to relate the resulting pedestal and confinement to the amount of power flowing into the scrape-off layer, P net , and also to the divertor targets. In both EDA and ELMy H-mode, energy confinement is generally good, with H 98 near unity. As P net is reduced to levels approaching that in L-mode, pedestal temperature diminishes significantly and normalized confinement time drops. By seeding with low-Z impurities, such as Ne and N 2 , high total radiated power fractions are possible, along with substantial reductions in divertor heat flux (>4x), all while maintaining H 98 ∼ 1. When the power radiated from the confined versus unconfined plasma is examined, pedestal and confinement properties are clearly seen to be an increasing function of P net , helping to unify the results with those from unseeded H-modes. This provides increased confidence that the power flow across the separatrix is the correct physics basis for ITER extrapolation. The experiments show that P net /P th of one or greater is likely to lead to H

  3. 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

  4. Beta II compact torus experiment plasma equilibrium and power balance

    International Nuclear Information System (INIS)

    Turner, W.C.; Goldenbaum, G.C.; Granneman, E.H.A.; Prono, D.S.; Hartman, C.W.; Taska, J.

    1982-01-01

    In this paper we follow up some of our earlier work that showed the compact torus (CT) plasma equilibrium produced by a magnetized coaxial plasma gun is nearly force free and that impurity radiation plays a dominant role in determining the decay time of plasma currents in present generation experiments

  5. 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

  6. 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

  7. Design innovations of the next-step spherical torus experiment and spherical torus development path

    International Nuclear Information System (INIS)

    Ono, M.; Kessel, C.; Peng, M.

    2003-01-01

    The spherical torus (ST) fusion energy development path is complementary to the tokamak burning plasma experiment such as ITER as it focuses toward the compact Component Test Facility (CTF) and higher toroidal beta regimes to improve the design of DEMO and a Power Plant. To support the ST development path, one option of a Next Step Spherical Torus (NSST) device is examined. NSST is a 'performance extension' (PE) stage ST with a plasma current of 5 - 10 MA, R = 1.5, B T ≤ 2.7 T with flexible physics capability to 1) Provide a sufficient physics basis for the design of the CTF, 2) Explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, 3) Contribute to the general plasma/fusion science of high β toroidal plasmas. The NSST facility is designed to utilize the TFTR site to minimize the cost and time required for the construction. (author)

  8. Next-Step Spherical Torus Experiment and Spherical Torus Strategy in the Fusion Energy Development Path

    International Nuclear Information System (INIS)

    Ono, M.; Peng, M.; Kessel, C.; Neumeyer, C.; Schmidt, J.; Chrzanowski, J.; Darrow, D.; Grisham, L.; Heitzenroeder, P.; Jarboe, T.; Jun, C.; Kaye, S.; Menard, J.; Raman, R.; Stevenson, T.; Viola, M.; Wilson, J.; Woolley, R.; Zatz, I.

    2003-01-01

    A spherical torus (ST) fusion energy development path which is complementary to proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact Component Test Facility (CTF) and higher performance advanced regimes leading to more attractive DEMO and Power Plant scale reactors. To provide the physics basis for the CTF an intermediate step needs to be taken which we refer to as the ''Next Step Spherical Torus'' (NSST) device and examine in some detail herein. NSST is a ''performance extension'' (PE) stage ST with the plasma current of 5-10 MA, R = 1.5 m, and Beta(sub)T less than or equal to 2.7 T with flexible physics capability. The mission of NSST is to: (1) provide a sufficient physics basis for the design of CTF, (2) explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, and (3) contribute to the general plasma/fusion science of high beta toroidal plasmas. The NSST facility is designed to utilize the Tokamak Fusion Test Reactor (or similar) site to minimize the cost and time required for the design and construction

  9. Electron Bernstein wave emission based diagnostic on National Spherical Torus Experiment (invited)

    International Nuclear Information System (INIS)

    Diem, S.; Taylor, G.; Caughman, John B.; Efthimion, P.C.; Kugel, H.; LeBlanc, B.; Preinhaelter, J.; Sabbagh, S.A.; Urban, J.

    2008-01-01

    National Spherical Torus Experiment (NSTX) is a spherical tokamak (ST) that operates with n(e) up to 10(20) m(-3) and B-T less than 0.6 T, cutting off low harmonic electron cyclotron (EC) emission widely used for T-e measurements on conventional aspect ratio tokamaks. The electron Bernstein wave (EBW) can propagate in ST plasmas and is emitted at EC harmonics. These properties suggest thermal EBW emission (EBE) may be used for local T-e measurements in the ST. Practically, a robust T-e(R,t) EBE diagnostic requires EBW transmission efficiencies of >90% for a wide range of plasma conditions. EBW emission and coupling physics were studied on NSTX with an obliquely viewing EBW to O-mode (B-X-O) diagnostic with two remotely steered antennas, coupled to absolutely calibrated radiometers. While T-e(R,t) measurements with EBW emission on NSTX were possible, they were challenged by several issues. Rapid fluctuations in edge n(e) scale length resulted in >20% changes in the low harmonic B-X-O transmission efficiency. Also, B-X-O transmission efficiency during H modes was observed to decay by a factor of 5-10 to less than a few percent. The B-X-O transmission behavior during H modes was reproduced by EBE simulations that predict that EBW collisional damping can significantly reduce emission when T-e < 30 eV inside the B-X-O mode conversion (MC) layer. Initial edge lithium conditioning experiments during H modes have shown that evaporated lithium can increase T-e inside the B-X-O MC layer, significantly increasing B-X-O transmission.

  10. Electron Bernstein Wave Emission Based Diagnostic on National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Diem, S.; Taylor, G.; Caughman, John B.; Efthimion, P.C.; Kugel, H.; LeBlanc, B.; Preinhaelter, J.; Sabbagh, S.A.; Urban, J.; Wilgen, John B.

    2008-01-01

    National Spherical Torus Experiment (NSTX) is a spherical tokamak (ST) that operates with n(e) up to 10(20) m(-3) and B(T) less than 0.6 T, cutting off low harmonic electron cyclotron (EC) emission widely used for T(e) measurements on conventional aspect ratio tokamaks. The electron Bernstein wave (EBW) can propagate in ST plasmas and is emitted at EC harmonics. These properties suggest thermal EBW emission (EBE) may be used for local T(e) measurements in the ST. Practically, a robust T(e)(R,t) EBE diagnostic requires EBW transmission efficiencies of >90% for a wide range of plasma conditions. EBW emission and coupling physics were studied on NSTX with an obliquely viewing EBW to O-mode (B-X-O) diagnostic with two remotely steered antennas, coupled to absolutely calibrated radiometers. While T(e)(R,t) measurements with EBW emission on NSTX were possible, they were challenged by several issues. Rapid fluctuations in edge n(e) scale length resulted in >20% changes in the low harmonic B-X-O transmission efficiency. Also, B-X-O transmission efficiency during H modes was observed to decay by a factor of 5-10 to less than a few percent. The B-X-O transmission behavior during H modes was reproduced by EBE simulations that predict that EBW collisional damping can significantly reduce emission when T(e)< 30 eV inside the B-X-O mode conversion (MC) layer. Initial edge lithium conditioning experiments during H modes have shown that evaporated lithium can increase T(e) inside the B-X-O MC layer, significantly increasing B-X-O transmission.

  11. 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

  12. The H-mode pedestal, ELMs and TF ripple effects in JT-60U/JET dimensionless identity experiments

    International Nuclear Information System (INIS)

    Saibene, G.; Oyama, N.; Loennroth, J.; Andrew, Y.; Luna, E. de la; Giroud, C.; Huysmans, G.T.A.; Kamada, Y.; Kempenaars, M.A.H.; Loarte, A.; Donald, D. Mc; Nave, M.M.F.; Meiggs, A.; Parail, V.; Sartori, R.; Sharapov, S.; Stober, J.; Suzuki, T.; Takechi, M.; Toi, K.; Urano, H.

    2007-01-01

    This paper summarizes results of dimensionless identity experiments in JT-60U and JET, aimed at the comparison of the H-mode pedestal and ELM behaviour in the two devices. Given their similar size, dimensionless matched plasmas are also similar in their dimensional parameters (in particular, the plasma minor radius a is the same in JET and JT-60U). Power and density scans were carried out at two values of I p , providing a q scan (q 95 = 3.1 and 5.1) with fixed (and matched) toroidal field. Contrary to initial expectations, a dimensionless match between the two devices was quite difficult to achieve. In general, p ped in JT-60U is lower than in JET and, at low q, the pedestal pressure of JT-60U with a Type I ELMy edge is matched in JET only in the Type III ELM regime. At q 95 = 5.1, a dimensionless match in ρ*, ν* and β p,ped is obtained with Type I ELMs, but only with low power JET H-modes. These results motivated a closer investigation of experimental conditions in the two devices, to identify possible 'hidden' physics that prevents obtaining a good match of pedestal values over a large range of plasmas parameters. Ripple-induced ion losses of the medium bore plasma used in JT-60U for the similarity experiments are identified as the main difference with JET. The magnitude of the JT-60U ripple losses is sufficient to induce counter-toroidal rotation in co-injected plasma. The influence of ripple losses was demonstrated at q 95 = 5.1: reducing ripple losses by ∼2 (from 4.3 to 1.9 MW) by replacing positive with negative neutral beam injection at approximately constant P in resulted in an increased p ped in JT-60U, providing a good match to full power JET H-modes. At the same time, the counter-toroidal rotation decreased. Physics mechanisms relating ripple losses to pedestal performance are not yet identified, and the possible role of velocity shear in the pedestal stability, as well as the possible influence of ripple on thermal ion transport are briefly

  13. 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.

  14. Divertor Heat Flux Reduction and Detachment in the National Spherical Torus eXperiment.

    Science.gov (United States)

    Soukhanovskii, Vsevolod

    2007-11-01

    Steady-state handling of the heat flux is a critical divertor issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) devices. Because of an inherently compact divertor, it was thought that ST-based devices might not be able to fully utilize radiative and dissipative divertor techniques based on induced power and momentum loss. However, initial experiments conducted in the National Spherical Torus Experiment in an open geometry horizontal carbon plate divertor using 0.8 MA 2-6 MW NBI-heated lower single null H-mode plasmas at the lower end of elongations κ=1.8-2.4 and triangularities δ=0.45-0.75 demonstrated that high divertor peak heat fluxes, up to 6-10 MW/ m^2, could be reduced by 50-75% using a high-recycling radiative divertor regime with D2 injection. Furthermore, similar reduction was obtained with a partially detached divertor (PDD) at high D2 injection rates, however, it was accompanied by an X-point MARFE that quickly led to confinement degradation. Another approach takes advantage of the ST relation between strong shaping and high performance, and utilizes the poloidal magnetic flux expansion in the divertor region. Up to 60 % reduction in divertor peak heat flux was achieved at similar levels of scrape-off layer power by varying plasma shaping and thereby increasing the outer strike point (OSP) poloidal flux expansion from 4-6 to 18-22. In recent experiments conducted in highly-shaped 1.0-1.2 MA 6 MW NBI heated H-mode plasmas with divertor D2 injection at rates up to 10^22 s-1, a PDD regime with OSP peak heat flux 0.5-1.5 MW/m^2 was obtained without noticeable confinement degradation. Calculations based on a two point scrape-off layer model with parameterized power and momentum losses show that the short parallel connection length at the OSP sets the upper limit on the radiative exhaust channel, and both the impurity radiation and large momentum sink achievable only at high divertor neutral pressures are required

  15. Physics results from the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Kaye, S.; Bell, M.

    2000-01-01

    The mission of the National Spherical Torus Experiment (NSTX) is to extend the understanding of toroidal physics to low aspect ratio (R/a ∼ 1.25) in low collisionality regimes. NSTX is designed to operate with up to 6 MW of High Harmonic Fast Wave (HHFW) heating and current drive, 5 MW of Neutral Beam Injection (NBI), and Co-Axial Helicity Injection (CHI) for non-inductive startup. Initial experiments focused on establishing conditions that will allow NSTX to achieve its aims of simultaneous high-β t and high-bootstrap current fraction, and to develop methods for non-inductive operation, which will be necessary for Spherical Torus power plants. Ohmic discharges with plasma currents up to 1 MA, stored energies up to 55 kJ, β t ∼ 10%, and a range of shapes and configurations were produced. Density limits in deuterium and helium reached 80% and 120% of the Greenwald limit respectively. Significant electron heating was observed with up to 2.3 MW of HHFW. Up to 270 kA of toroidal current for up to 200 msec was produced noninductively using CHI. Initial NBI experiments were carried out with up to two beam sources (3.2 MW). Plasmas with stored energies of up to 140 kJ and β t =21% were produced

  16. Recent H-mode density limit experiments on DIII-D

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Maingi, R.; Hyatt, A.W.

    1997-06-01

    A vast body of tokamak data is in good agreement with the empirical density limit scalings proposed by Hugill and Greenwald. These scalings have common puzzling features of showing no dependence on either impurity concentration or heating power, since the density limit is frequently correlated with a rapid rise of the edge radiation. Despite the resiliency of these scalings, several machines under restrictive conditions have operated at densities well above the predictions of these scalings, albeit with pellet injection and at varying degrees of confinement degradation. Furthermore, data from several machines display a weak dependence on heating power. These results cast doubt on the universal validity of both of these scalings. Nevertheless the fact remains that access to densities above Hugill-Greenwald scaling is very difficult. A number of theories based on radiative power balance in the plasma boundary have explained some but not all features of tokamak density limit behavior, and as ITER design studies recently brought to focus, a satisfactory understanding of this phenomenon is lacking. Motivated by a need for better understanding of effects of density and fueling on tokamak plasmas in general, the authors have conducted a series of experiments designed to identify and isolate physical effects suspected to be directly or indirectly responsible for the density limit. The physical effects being considered include: divertor power balance, MARFE, poloidally symmetric radiative instabilities, MHD instabilities, and transport. In this paper they first present a brief summary of the experimental results up to the writing of this paper. The remainder of the paper is devoted to a comparison of this data at the onset of the MARFE instability with predictions of theory and the implications of the results on access to densities beyond the Hugill-Greenwald limit

  17. Control System Development Plan for the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Neumeyer, C.; Mueller, D.; Gates, D.A.; Ferron, J.R.

    1999-01-01

    The National Spherical Torus Experiment (NSTX) has as one of its primary goals the demonstration of the attractiveness of the spherical torus concept as a fusion power plant. Central to this goal is the achievement of high plasma β ( = 2 micro 0 /B 2 a measure of the efficiency of a magnetic plasma confinement system). It has been demonstrated both theoretically and experimentally that the maximum achievable β is a strong function of both local and global plasma parameters. It is therefore important to optimize control of the plasma. To this end a phased development plan for digital plasma control on NSTX is presented. The relative level of sophistication of the control system software and hardware will be increased according to the demands of the experimental program in a three phase plan. During Day 0 (first plasma), a simple coil current control algorithm will initiate plasma operations. During the second phase (Day 1) of plasma operations the control system will continue to use the preprogrammed algorithm to initiate plasma breakdown but will then change over to a rudimentary plasma control scheme based on linear combinations of measured plasma fields and fluxes. The third phase of NSTX plasma control system development will utilize the rtEFIT code, first used on DIII-D, to determine, in real-time, the full plasma equilibrium by inverting the Grad-Shafranov equation. The details of the development plan, including a description of the proposed hardware will be presented

  18. Measurement of Poloidal Velocity on the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ronald E. Bell and Russell Feder

    2010-06-04

    A diagnostic suite has been developed to measure impurity poloidal flow using charge exchange recombination spectroscopy on the National Spherical Torus Experiment. Toroidal and poloidal viewing systems measure all quantities required to determine the radial electric field. Two sets of up/down symmetric poloidal views are used to measure both active emission in the plane of the neutral heating beams and background emission in a radial plane away from the neutral beams. Differential velocity measurements isolate the line-integrated poloidal velocity from apparent flows due to the energy-dependent chargeexchange cross section. Six f/1.8 spectrometers measure 276 spectra to obtain 75 active and 63 background channels every 10 ms. Local measurements from a similar midplane toroidal viewing system are mapped into two dimensions to allow the inversion of poloidal line-integrated measurements to obtain local poloidal velocity profiles. Radial resolution after inversion is 0.6-1.8 cm from the plasma edge to the center.

  19. Measurement of Poloidal Velocity on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Bell, Ronald E.; Feder, Russell

    2010-01-01

    A diagnostic suite has been developed to measure impurity poloidal flow using charge exchange recombination spectroscopy on the National Spherical Torus Experiment. Toroidal and poloidal viewing systems measure all quantities required to determine the radial electric field. Two sets of up/down symmetric poloidal views are used to measure both active emission in the plane of the neutral heating beams and background emission in a radial plane away from the neutral beams. Differential velocity measurements isolate the line-integrated poloidal velocity from apparent flows due to the energy-dependent chargeexchange cross section. Six f/1.8 spectrometers measure 276 spectra to obtain 75 active and 63 background channels every 10 ms. Local measurements from a similar midplane toroidal viewing system are mapped into two dimensions to allow the inversion of poloidal line-integrated measurements to obtain local poloidal velocity profiles. Radial resolution after inversion is 0.6-1.8 cm from the plasma edge to the center.

  20. Physics design of a 28 GHz electron heating system for the National Spherical Torus experiment upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, G.; Bertelli, N.; Ellis, R. A.; Gerhardt, S. P.; Hosea, J. C.; Poli, F. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Harvey, R. W. [CompX, Del Mar, California 92014 (United States); Raman, R. [University of Washington, Seattle, Washington 98195 (United States); Smirnov, A. P. [M.V. Lomonosov Moscow State University, Moscow (Russian Federation)

    2014-02-12

    A megawatt-level, 28 GHz electron heating system is being designed to support non-inductive (NI) plasma current (I{sub p}) start-up and local heating and current drive (CD) in H-mode discharges in the National Spherical Torus Experiment Upgrade (NSTX-U). The development of fully NI I{sub p} start-up and ramp-up is an important goal of the NSTXU research program. 28 GHz electron cyclotron (EC) heating is predicted to rapidly increase the central electron temperature (T{sub e}(0)) of low density NI plasmas generated by Coaxial Helicity Injection (CHI). The increased T{sub e}(0) will significantly reduce the I{sub p} decay rate of CHI plasmas, allowing the coupling of fast wave heating and neutral beam injection. Also 28 GHz electron Bernstein wave (EBW) heating and CD can be used during the I{sub p} flat top in NSTX-U discharges when the plasma is overdense. Ray tracing and Fokker-Planck numerical simulation codes have been used to model EC and EBW heating and CD in NSTX-U. This paper presents a pre-conceptual design for the 28 GHz heating system and some of the results from the numerical simulations.

  1. 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

  2. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    Science.gov (United States)

    Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

    2009-11-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  3. Current drive experiments on the HIT-II spherical torus

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Raman, R.; Nelson, B.A.; Holcomb, C.T.; McCollam, K.J.; Sieck, P.E.

    1999-01-01

    This paper describes the following new achievements from the Helicity Injected Torus (HIT) program: a) formation and sustainment of a toroidal magnetic equilibrium using coaxial helicity injection (CHI) in a conducting shell that has an L/R time much shorter than the pulse length; b) static formation of a spherical torus with plasma current over 180 kA using a transformer and feedback controlled equilibrium coils; and c) production of a current increase in a transformer produced spherical torus using CHI. (author)

  4. Current drive experiments on the HIT-II spherical torus

    International Nuclear Information System (INIS)

    Jarboe, T.; Raman, R.; Nelson, B.; Holcomb, C.T.; McCollam, K.J.; Sieck, P.E.

    2001-01-01

    This paper describes the following new achievements from the Helicity Injected Torus (HIT) program: a) formation and sustainment of a toroidal magnetic equilibrium using coaxial helicity injection (CHI) in a conducting shell that has an L/R time much shorter than the pulse length; b) static formation of a spherical torus with plasma current over 180 kA using a transformer and feedback controlled equilibrium coils; and c) production of a current increase in a transformer produced spherical torus using CHI. (author)

  5. Benchmark experiments on neutron streaming through JET Torus Hall penetrations

    Science.gov (United States)

    Batistoni, P.; Conroy, S.; Lilley, S.; Naish, J.; Obryk, B.; Popovichev, S.; Stamatelatos, I.; Syme, B.; Vasilopoulou, T.; contributors, JET

    2015-05-01

    Neutronics experiments are performed at JET for validating in a real fusion environment the neutronics codes and nuclear data applied in ITER nuclear analyses. In particular, the neutron fluence through the penetrations of the JET torus hall is measured and compared with calculations to assess the capability of state-of-art numerical tools to correctly predict the radiation streaming in the ITER biological shield penetrations up to large distances from the neutron source, in large and complex geometries. Neutron streaming experiments started in 2012 when several hundreds of very sensitive thermo-luminescence detectors (TLDs), enriched to different levels in 6LiF/7LiF, were used to measure the neutron and gamma dose separately. Lessons learnt from this first experiment led to significant improvements in the experimental arrangements to reduce the effects due to directional neutron source and self-shielding of TLDs. Here we report the results of measurements performed during the 2013-2014 JET campaign. Data from new positions, at further locations in the South West labyrinth and down to the Torus Hall basement through the air duct chimney, were obtained up to about a 40 m distance from the plasma neutron source. In order to avoid interference between TLDs due to self-shielding effects, only TLDs containing natural Lithium and 99.97% 7Li were used. All TLDs were located in the centre of large polyethylene (PE) moderators, with natLi and 7Li crystals evenly arranged within two PE containers, one in horizontal and the other in vertical orientation, to investigate the shadowing effect in the directional neutron field. All TLDs were calibrated in the quantities of air kerma and neutron fluence. This improved experimental arrangement led to reduced statistical spread in the experimental data. The Monte Carlo N-Particle (MCNP) code was used to calculate the air kerma due to neutrons and the neutron fluence at detector positions, using a JET model validated up to the

  6. 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

  7. New Capabilities and Results for the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    M.G. Bell, R.E. Bell, D.A. Gates, S.M. Kaye, H. Kugel, B.P. LeBlanc, F.M. Levinton, R. Maingi, J.E. Menard, R. Raman, S.A. Sabbagh, D. Stutman and the NSTX Research Team

    2008-02-29

    The National Spherical Torus Experiment (NSTX) produces plasmas with toroidal aspect ratio as low as 1.25, which can be heated by up to 6 MW High-Harmonic Fast Waves and up to 7 MW of deuterium Neutral Beam Injection. Using new poloidal fields coils, plasmas with cross-section elongation up to 2.7, triangularity 0.8, plasma currents Ip up to 1.5 MA and normalized currents Ip/a·BT up to 7.5 MA/m·T have been achieved. A significant extension of the plasma pulse length, to 1.5 s at a plasma current of 0.7 MA, has been achieved by exploiting the bootstrap and NBI-driven currents to reduce the dissipation of poloidal flux. Inductive plasma startup has been supplemented by Coaxial Helicity Injection (CHI) and the production of persistent current on closed flux surfaces by CHI has now been demonstrated in NSTX. The plasma response to magnetic field perturbations with toroidal mode numbers n = 1 or 3 and the effects on the plasma rotation have been investigated using three pairs of coils outside the vacuum vessel. Recent studies of both MHD stability and of transport benefitted from improved diagnostics, including measurements of the internal poloidal field using the motional Stark effect (MSE). In plasmas with a region of reversed magnetic shear in the core, now confirmed by the MSE data, improved electron confinement has been observed.

  8. Nonlocal neoclassical transport in tokamak and spherical torus experiments

    International Nuclear Information System (INIS)

    Wang, W. X.; Rewoldt, G.; Tang, W. M.; Hinton, F. L.; Manickam, J.; Zakharov, L. E.; White, R. B.; Kaye, S.

    2006-01-01

    Large ion orbits can produce nonlocal neoclassical effects on ion heat transport, the ambipolar radial electric field, and the bootstrap current in realistic toroidal plasmas. Using a global δf particle simulation, it is found that the conventional local, linear gradient-flux relation is broken for the ion thermal transport near the magnetic axis. With regard to the transport level, it is found that details of the ion temperature profile determine whether the transport is higher or lower when compared with the predictions of standard neoclassical theory. Particularly, this nonlocal feature is suggested to exist in the National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, Y.-K. M. Peng et al., Nucl. Fusion 40, 557 (2000)], being consistent with NSTX experimental evidence. It is also shown that a large ion temperature gradient can increase the bootstrap current. When the plasma rotation is taken into account, the toroidal rotation gradient can drive an additional parallel flow for the ions and then additional bootstrap current, either positive or negative, depending on the gradient direction. Compared with the carbon radial force balance estimate for the neoclassical poloidal flow, our nonlocal simulation predicts a significantly deeper radial electric field well at the location of an internal transport barrier of an NSTX discharge

  9. Transient Transport Experiments in the CDX-U Spherical Torus

    International Nuclear Information System (INIS)

    T. Munsat; P.C. Efthimion; B. Jones; R. Kaita; R. Majeski; D. Stutman; G. Taylor

    2001-01-01

    Electron transport has been measured in the Current Drive Experiment-Upgrade (CDX-U) using two separate perturbative techniques. Gas modulation at the plasma edge was used to introduce cold-pulses which propagate towards the plasma center, providing time-of-flight information leading to a determination of chi(subscript e) as a function of radius. Sawteeth at the q=1 radius (r/a ∼ 0.15) induced heat-pulses which propagated outward towards the plasma edge, providing a complementary time-of-flight based chi(subscript e) profile measurement. This work represents the first localized measurement of chi(subscript e) in a spherical torus. It is found that chi(subscript e) = 1-2 meters squared per second in the plasma core (r/a < 1/3), increasing by an order of magnitude or more outside of this region. Furthermore, the chi(subscript e) profile exhibits a sharp transition near r/a = 1/3. Spectral and profile analyses of the soft X-rays, scanning interferometer, and edge probe data show no evidence of a significant magnetic island causing the high chi(subscript e) region

  10. 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

  11. Simulation of microtearing turbulence in national spherical torus experiment

    Energy Technology Data Exchange (ETDEWEB)

    Guttenfelder, W.; Kaye, S. M.; Bell, R. E.; Hammett, G. W.; LeBlanc, B. P.; Mikkelsen, D. R.; Ren, Y. [Princeton Plasma Physics Laboratory, Princeton New Jersey 08543 (United States); Candy, J. [General Atomics, San Diego, California 92186 (United States); Nevins, W. M.; Wang, E. [Lawrence Livermore National Laboratory, Livermore, California 04551 (United States); Zhang, J.; Crocker, N. A. [University of California Los Angeles, California 90095 (United States); Yuh, H. [Nova Photonics Inc., Princeton, New Jersey 08540 (United States)

    2012-05-15

    Thermal energy confinement times in National Spherical Torus Experiment (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 spherical tokamak (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 ({approx}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 Multiplication-Sign B flows as experimental values of E Multiplication-Sign 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.

  12. Comparative study of poloidal field systems for the torus II experiment

    International Nuclear Information System (INIS)

    Farvaque, L.; Ghazal, S.; Leloup, C.; Pariente, M.; CEA Centre d'Etudes Nucleaires de Fontenay-aux-Roses, 92

    1976-11-01

    Three types of transformer for the TORUS II experiment are compared: a saturated iron core transformer with an entire magnetic circuit, an air core transformer and a saturated iron core transformer restricted to the central limb [fr

  13. 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

  14. 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

  15. Snowflake divertor configuration studies in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); and others

    2012-08-15

    Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width {lambda}{sub q} was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m{sup 2} (from 4-7 MW/m{sup 2} in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P{sub rad} in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

  16. Electromagnetic diagnostic system for the Keda Torus eXperiment

    Science.gov (United States)

    Tu, Cui; Liu, Adi; Li, Zichao; Tan, Mingsheng; Luo, Bing; You, Wei; Li, Chenguang; Bai, Wei; Fu, Chenshuo; Huang, Fangcheng; Xiao, Bingjia; Shen, Biao; Shi, Tonghui; Chen, Dalong; Mao, Wenzhe; Li, Hong; Xie, Jinglin; Lan, Tao; Ding, Weixing; Xiao, Chijin; Liu, Wandong

    2017-09-01

    A system for electromagnetic measurements was designed and installed on the Keda Torus eXperiment (KTX) reversed field pinch device last year. Although the unique double-C structure of the KTX, which allows the machine to be opened easily without disassembling the poloidal field windings, makes the convenient replacement and modification of the internal inductive coils possible, it can present difficulties in the design of flux coils and magnetic probes at the two vertical gaps. Moreover, the KTX has a composite shell consisting of a 6 mm stainless steel vacuum chamber and a 1.5 mm copper shell, which results in limited space for the installation of saddle sensors. Therefore, the double-C structure and composite shell should be considered, especially during the design and installation of the electromagnetic diagnostic system (EDS). The inner surface of the vacuum vessel includes two types of probes. One type is for the measurement of the global plasma parameters, and the other type is for studying the local behavior of the plasma and operating the new saddle coils. In addition, the probes on the outer surface of the composite shell are used for measurements of eddy currents. Finally, saddle sensors for radial field measurements for feedback control were installed between the conducting shell and the vacuum vessel. The entire system includes approximately 1100 magnetic probes, 14 flux coils, 4 ×26 ×2 saddle sensors, and 16 Rogowski coils. Considering the large number of probes and limited space available in the vacuum vessel, the miniaturization of the probes and optimization of the probe distribution are necessary. In addition, accurate calibration and careful mounting of the probes are also required. The frequency response of the designed magnetic probes is up to 200 kHz, and the resolution is 1 G. The EDS, being spherical and of high precision, is one of the most basic and effective diagnostic tools of the KTX and meets the demands imposed by requirements on

  17. 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

  18. 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

  19. Space potential profiles in ELMO Bumpy Torus (EBT) experiment

    International Nuclear Information System (INIS)

    Bieniosek, F.M.; Connor, K.A.

    1983-01-01

    Spatially resolved measurements of the electric space potential in the ELMO Bumpy Torus (EBT) have been made by a heavy ion beam probe. The EBT-I device is characterized by positive potentials in the surface plasma the order of 100 V and by a nearly symmetric potential well in the core plasma of up to 300 V with respect to the surface potential. The EBT-S device has a similar potential structure with well depth and peak potential similar to or greater than that of EBT-I. Peak potential and well depth increase as the edge gas pressure is lowered and as the microwave power is increased. The potential structure is strongly linked to the specific heating geometry. The ambipolar electric field is large enough generally to dominate the core electron neoclassical diffusion. The potential profile is approximately parabolic in the core, which is shown to be a natural consequence of the spatially uniform plasma source function

  20. Comparison of beam emission spectroscopy and gas puff imaging edge fluctuation measurements in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Sechrest, Y.; Munsat, T. [Department of Physics, University of Colorado, Boulder, Colorado 80309 (United States); Smith, D. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Stotler, D. P.; Zweben, S. J. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)

    2015-05-15

    In this study, the close physical proximity of the Gas Puff Imaging (GPI) and Beam Emission Spectroscopy (BES) diagnostics on the National Spherical torus Experiment (NSTX) is leveraged to directly compare fluctuation measurements, and to study the local effects of the GPI neutral deuterium puff during H-mode plasmas without large Edge Localized Modes. The GPI and BES views on NSTX provide partially overlapping coverage of the edge and scrape-off layer (SOL) regions above the outboard midplane. The separation in the toroidal direction is 16°, and field lines passing through diagnostic views are separated by ∼20 cm in the direction perpendicular to the magnetic field. Strong cross-correlation is observed, and strong cross-coherence is seen for frequencies between 5 and 15 kHz. Also, probability distribution functions of fluctuations measured ∼3 cm inside the separatrix exhibit only minor deviations from a normal distribution for both diagnostics, and good agreement between correlation length estimates, decorrelation times, and structure velocities is found at the ±40% level. While the two instruments agree closely in many respects, some discrepancies are observed. Most notably, GPI normalized fluctuation levels exceed BES fluctuations by a factor of ∼9. BES mean intensity is found to be sensitive to the GPI neutral gas puff, and BES normalized fluctuation levels for frequencies between 1 and 10 kHz are observed to increase during the GPI puff.

  1. Reconciliation of Measured and TRANSP-calculated Neutron Emission Rates in the National Spherical Torus Experiment: Circa 2002-2005

    International Nuclear Information System (INIS)

    Medley, S.S.; Darrow, D.S.; Roquemore, A.L.

    2005-01-01

    A change in the response of the neutron detectors on the National Spherical Torus Experiment occurred between the 2002-2003 and 2004 experimental run periods. An analysis of this behavior by investigating the neutron diagnostic operating conditions and comparing measured and TRANSP-calculated neutron rates is presented. Also a revised procedure for cross calibration of the neutron scintillator detectors with the fission chamber detectors was implemented that delivers good agreement amongst the measured neutron rates for all neutron detectors and all run periods. For L-mode discharges, the measured and TRANSP-calculated neutron rates now match closely for all run years. For H-mode discharges over the entire 2002-2004 period, the 2FG scintillator and fission chamber measurements match each other but imply a neutron deficit of 11.5% relative to the TRANSP-calculated neutron. The results of this report impose a modification on all of the previously used calibration factors for the entire neutron detector suite over the 2002-2004 period. A tabular summary of the new calibration factors is provided including certified calibration factors for the 2005 run

  2. Comparison of beam emission spectroscopy and gas puff imaging edge fluctuation measurements in National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Sechrest, Y.; Munsat, T.; Smith, D.; Stotler, D. P.; Zweben, S. J.

    2015-01-01

    In this study, the close physical proximity of the Gas Puff Imaging (GPI) and Beam Emission Spectroscopy (BES) diagnostics on the National Spherical torus Experiment (NSTX) is leveraged to directly compare fluctuation measurements, and to study the local effects of the GPI neutral deuterium puff during H-mode plasmas without large Edge Localized Modes. The GPI and BES views on NSTX provide partially overlapping coverage of the edge and scrape-off layer (SOL) regions above the outboard midplane. The separation in the toroidal direction is 16°, and field lines passing through diagnostic views are separated by ∼20 cm in the direction perpendicular to the magnetic field. Strong cross-correlation is observed, and strong cross-coherence is seen for frequencies between 5 and 15 kHz. Also, probability distribution functions of fluctuations measured ∼3 cm inside the separatrix exhibit only minor deviations from a normal distribution for both diagnostics, and good agreement between correlation length estimates, decorrelation times, and structure velocities is found at the ±40% level. While the two instruments agree closely in many respects, some discrepancies are observed. Most notably, GPI normalized fluctuation levels exceed BES fluctuations by a factor of ∼9. BES mean intensity is found to be sensitive to the GPI neutral gas puff, and BES normalized fluctuation levels for frequencies between 1 and 10 kHz are observed to increase during the GPI puff

  3. Results from the RACE [Ring ACceleration Experiment] Compact Torus Acceleration Experiment

    International Nuclear Information System (INIS)

    Hammer, J.H.; Hartman, C.W.; Eddleman, J.L.; Kusse, B.

    1987-06-01

    RACE (Ring ACceleration Experiment) is a proof-of-principle experiment aimed at demonstrating acceleration of magnetically confined compact torus plasma rings to directed kinetic energies well in excess of their magnetic and thermal energies. In the course of the first year of operation the following have been observed: successful formation of rings in the RACE geometry; acceleration of rings with large forces, F/sub accelerate/ ∼F/sub equilibrium/ without apparent degradation of the ring structure; peak velocities of ≅2.5 x 10 8 cm/sec; acceleration efficiency of >30% at speeds of 1.5 x 10 8 cm/sec inferred from trajectory and capacitor bank data; kinetic to magnetic energy ratios ∼10 were observed. Experiments in the near future will be aimed at confirmation of the mass/energy measurements by calorimetry and direct density measurements

  4. The H-mode of ASDEX

    International Nuclear Information System (INIS)

    1989-01-01

    The paper is a review of investigations of the H-mode on ASDEX performed since its discovery in 1982. The topics discussed are: (1) the development of the plasma profiles, with steep gradients in the edge region and flat profiles in the bulk plasma, (2) the MHD properties resulting from the profile changes, including an extensive stability analysis, (3) the impurity development, with special emphasis on the MHD aspects and on neoclassical impurity transport effects in quiescent H-phases, and (4) the properties of the edge plasma, including the evidence of three-dimensional distortions at the edge. The part on confinement includes scaling studies and the results of transport analysis. The power threshold of the H-mode is found to depend weakly on the density, but there is probably no dependence on the toroidal field or the current. For the operational range of the H-mode, new results for the limiter H-mode on ASDEX and the development of the H-mode under beam current drive conditions are included. A number of experiments are described which demonstrate the crucial role of the edge electron temperature in the L-H transition. New results of magnetic and density fluctuation studies at the plasma edge within the edge transport barrier are presented. Finally, the findings on ASDEX are compared with results obtained on other machines and are used to test various H-mode theories. (author). 131 refs, 103 figs, 1 tab

  5. ELMs and constraints on the H-mode pedestal: peeling-ballooning stability calculation and comparison with experiment

    International Nuclear Information System (INIS)

    Snyder, P.B.; Ferron, J.R.; Wilson, H.R.

    2004-01-01

    We review and test the peeling-ballooning model for edge localized modes (ELMs) and pedestal constraints, a model based upon theoretical analysis of magnetohydrodynamic (MHD) instabilities that can limit the pedestal height and drive ELMs. A highly efficient MHD stability code, ELITE, is used to calculate quantitative stability constraints on the pedestal, including constraints on the pedestal height. Because of the impact of collisionality on the bootstrap current, these pedestal constraints are dependent on the density and temperature separately, rather than simply on the pressure. ELITE stability calculations are directly compared with experimental data for a series of plasmas in which the density is varied and ELM characteristics change. In addition, a technique is developed whereby peeling-ballooning pedestal constraints are calculated as a function of key equilibrium parameters via ELITE calculations using series of model equilibria. This technique is used to successfully compare the expected pedestal height as a function of density, triangularity and plasma current with experimental data. Furthermore, the technique can be applied for parameter ranges beyond the purview of present experiments, and we present a brief projection of peeling-ballooning pedestal constraints for burning plasma tokamak designs. (author)

  6. Modification Of The Electron Energy Distribution Function During Lithium Experiments On The National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Jaworski, M A; Gray, T K; Kaita, R; Kallman, J; Kugel, H; LeBlanc, B; McLean, A; Sabbagh, S A; Soukanovskii, V; Stotler, D P

    2011-06-03

    The National Spherical Torus Experiment (NSTX) has recently studied the use of a liquid lithium divertor (LLD). Divertor Langmuir probes have also been installed for making measurements of the local plasma conditions. A non-local probe interpretation method is used to supplement the classical probe interpretation and obtain measurements of the electron energy distribution function (EEDF) which show the occurrence of a hot-electron component. Analysis is made of two discharges within a sequence that exhibited changes in plasma fueling efficiency. It is found that the local electron temperature increases and that this increase is most strongly correlated with the energy contained within the hot-electron population. Preliminary interpretative modeling indicates that kinetic effects are likely in the NSTX.

  7. 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.

  8. Confinement of Neutral Beam Ions in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Darrow, D.S.; Medley, S.S.; Roquemore, A.L.; Rosenberg, A.

    2001-01-01

    The loss of neutral-beam ions to the wall has been measured in the National Spherical Torus Experiment (NSTX) by means of thermocouples, an infrared (IR) camera, and a Faraday cup probe. The losses tend to exhibit the expected dependences on plasma current, tangency radius of the injector, and plasma outer gap. However, the thermocouples and the Faraday cups indicate substantially different levels of loss and this difference has yet to be understood

  9. Remote handling of JET in-torus components. A practical experience

    International Nuclear Information System (INIS)

    Mills, S.; Brade, R.; Edwards, P.

    2000-01-01

    This paper summarises the experiences gained from the extensive handling of JET components inside the torus. The problems involved with handling components not designed to be remotely handled and the methods used to overcome them are described and discussed with specific examples from recent JET remote operations. The method employed for remotely producing structural TIG welds is explained. The problems of dextrous manipulation in an inverted attitude are discussed and the methods of amelioration are described

  10. Overview of physics results from the conclusive operation of the National Spherical Torus Experiment

    Science.gov (United States)

    Sabbagh, S. A.; Ahn, J.-W.; Allain, J.; Andre, R.; Balbaky, A.; Bastasz, R.; Battaglia, D.; Bell, M.; Bell, R.; Beiersdorfer, P.; Belova, E.; Berkery, J.; Betti, R.; Bialek, J.; Bigelow, T.; Bitter, M.; Boedo, J.; Bonoli, P.; Boozer, A.; Bortolon, A.; Boyle, D.; Brennan, D.; Breslau, J.; Buttery, R.; Canik, J.; Caravelli, G.; Chang, C.; Crocker, N.; Darrow, D.; Davis, B.; 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.; 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.; Kramer, G.; Krasheninnikov, S.; Kubota, S.; Kugel, H.; La Haye, R. J.; Lao, L.; LeBlanc, B.; Lee, W.; Lee, K.; Leuer, J.; Levinton, F.; Liang, Y.; Liu, D.; Lore, J.; Luhmann, N., Jr.; Maingi, R.; Majeski, R.; Manickam, J.; Mansfield, D.; Maqueda, R.; Mazzucato, E.; McLean, A.; McCune, D.; McGeehan, B.; McKee, G.; Medley, S.; Meier, E.; 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.; Park, Y. S.; Paul, S.; Peebles, W.; Penaflor, B.; Perkins, R. J.; Phillips, C.; Pigarov, A.; Podesta, M.; Preinhaelter, J.; Raman, R.; Ren, Y.; Rewoldt, G.; Rognlien, T.; Ross, P.; Rowley, C.; Ruskov, E.; Russell, D.; Ruzic, D.; Ryan, P.; 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, G.; Taylor, C.; Tritz, K.; Tsarouhas, D.; Umansky, M.; Urban, J.; Untergberg, E.; Walker, M.; Wampler, W.; Wang, W.; Whaley, J.; White, R.; Wilgen, J.; Wilson, R.; Wong, K. L.; Wright, J.; Xia, Z.; Youchison, D.; Yu, G.; Yuh, H.; Zakharov, L.; Zemlyanov, D.; Zimmer, G.; Zweben, S. J.

    2013-10-01

    Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique ST operational space is leveraged to test physics theories for next-step tokamak operation, including ITER. Present research also examines implications for the coming device upgrade, NSTX-U. An energy confinement time, τE, scaling unified for varied wall conditions exhibits a strong improvement of BTτE with decreased electron collisionality, accentuated by lithium (Li) wall conditioning. This result is consistent with nonlinear microtearing simulations that match the experimental electron diffusivity quantitatively and predict reduced electron heat transport at lower collisionality. Beam-emission spectroscopy measurements in the steep gradient region of the pedestal indicate the poloidal correlation length of turbulence of about ten ion gyroradii increases at higher electron density gradient and lower Ti gradient, consistent with turbulence caused by trapped electron instabilities. Density fluctuations in the pedestal top region indicate ion-scale microturbulence compatible with ion temperature gradient and/or kinetic ballooning mode instabilities. Plasma characteristics change nearly continuously with increasing Li evaporation and edge localized modes (ELMs) stabilize due to edge density gradient alteration. Global mode stability studies show stabilizing resonant kinetic effects are enhanced at lower collisionality, but in stark contrast have almost no dependence on collisionality when the plasma is off-resonance. Combined resistive wall mode radial and poloidal field sensor feedback was used to control n = 1 perturbations and improve stability. The disruption probability due to unstable resistive wall modes (RWMs) was surprisingly reduced at very high βN/li > 10 consistent with low frequency magnetohydrodynamic spectroscopy measurements of mode stability. Greater

  11. Torus theory

    International Nuclear Information System (INIS)

    Namsrai, Kh.

    2001-11-01

    Geometrical structure and physical characteristics of a torus are investigated in detail. Newtonian and electromagnetic potentials of the torus are defined at short and long distances. It is shown that torus potential at small distances has attractive oscillator behaviour. Motion of a particle in the torus potential is studied. The inertia tensor of the torus and its dynamics are obtained. Rotating torus whose tip is held fixed by two massless rigid threads and moves in a gravitational field is considered. (author)

  12. L to H-mode Power Threshold and Confinement Characteristics of H-modes in KSTAR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H. S.; Na, Y.S., E-mail: ftwalker.hyuns@gmail.com [Seoul National University, Seoul (Korea, Republic of); Ahn, J. W. [Oak Ridge National Laboratory, Oak Ridge (United States); Jeon, Y. M.; Yoon, S. W.; Lee, K. D.; Ko, W. H.; Bae, Y. S.; Kim, W. C.; Kwak, J. G. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2012-09-15

    Full text: Since KSTAR has obtained the H-mode in 2010 campaign, H-mode plasmas were routinely obtained with combined heating of NBI with maximum power of 1.5 MW and ECRH with maximum power of {approx} 0.3 MW and {approx} 0.6 MW for 110 GHz and 170 GHz, respectively. The L- to H-mode power threshold and confinement properties of KSTAR H-modes are investigated in this work. Firstly, the L- to H-mode power threshold and the power loss to the seperatrix are calculated by power balance analysis for about collected 400 shots. As a result, a trend of roll-over is observed in the power threshold of KSTAR H-mode compared with the multi-machine power threshold scaling in the low density regime. Dependence of the power threshold on other parameters are also investigated such as the X-point position and shaping parameters like as triangularity and elongation. In addition, the reason of reduction of power threshold in 2011 campaign compared with that in 2010 is addressed. Secondly, the confinement enhancement factors are calculated to evaluate the performance of KSTAR H-modes. The calculated H{sub 89-p} and H{sub 98} (y, 2) represent that the confinement is enhanced in most KSTAR H-mode discharges. Interestingly, even in L-mode phases, confinement is observed to be enhanced against the multi-machine scalings. H{sub exp} factor is newly introduced to evaluate the amount of confinement improvement in the H-mode phase compared with the L-mode phase in a single discharge. H{sub exp} exhibits that the global energy confinement time of the H-mode phase is improved about 1.3 - 2.0 times compared with that of the L-mode phase. Finally, interpretive and predictive numerical simulations are carried out using the ASTRA code for typical KSTAR H-mode discharges. The Weiland model and the GLF23 model are employed for calculating the anomalous contributions of both electron and ion heat transport in predictive simulations. For the H-mode phase, the Weiland model reproduces the experiment

  13. Reduced model prediction of electron temperature profiles in microtearing-dominated National Spherical Torus eXperiment plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Kaye, S. M., E-mail: skaye@pppl.gov; Guttenfelder, W.; Bell, R. E.; Gerhardt, S. P.; LeBlanc, B. P.; Maingi, R. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States)

    2014-08-15

    A representative H-mode discharge from the National Spherical Torus eXperiment is studied in detail to utilize it as a basis for a time-evolving prediction of the electron temperature profile using an appropriate reduced transport model. The time evolution of characteristic plasma variables such as β{sub e}, ν{sub e}{sup ∗}, the MHD α parameter, and the gradient scale lengths of T{sub e}, T{sub i}, and n{sub e} were examined as a prelude to performing linear gyrokinetic calculations to determine the fastest growing micro instability at various times and locations throughout the discharge. The inferences from the parameter evolutions and the linear stability calculations were consistent. Early in the discharge, when β{sub e} and ν{sub e}{sup ∗} were relatively low, ballooning parity modes were dominant. As time progressed and both β{sub e} and ν{sub e}{sup ∗} increased, microtearing became the dominant low-k{sub θ} mode, especially in the outer half of the plasma. There are instances in time and radius, however, where other modes, at higher-k{sub θ}, may, in addition to microtearing, be important for driving electron transport. Given these results, the Rebut-Lallia-Watkins (RLW) electron thermal diffusivity model, which is based on microtearing-induced transport, was used to predict the time-evolving electron temperature across most of the profile. The results indicate that RLW does a good job of predicting T{sub e} for times and locations where microtearing was determined to be important, but not as well when microtearing was predicted to be stable or subdominant.

  14. 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

  15. 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

  16. X-ray production experiments on the RACE Compact Torus Accelerator

    International Nuclear Information System (INIS)

    Hammer, J.H.; Eddleman, J.L.; Hartman, C.W.; McLean, H.S.; Molvik, A.W.; Gee, M.

    1989-12-01

    The Purpose of the Compact Torus Accelerator (CTA) program at LLNL is to prove the principle of a unique accelerator concept based on magnetically confined compact torus (CT) plasma rings and to study applications. Successful achievement of these goals could lead to a high power-density driver for many applications including an intense x-ray source for nuclear weapons effects simulation and an inertial fusion driver. Fusion applications and a description of the CTA concept are included in a companion paper at this conference. This paper will describe the initial experiments on soft x-ray production conducted on the plasma Ring ACcelerator Experiment (RACE) and compare the results to modeling studies. The experiments on CT stagnation and soft x-ray production were conducted with unfocused rings as a first of CT dynamics and the physics of x-ray production. The x-ray fluences observed are consistent with expectations based on calculations employing a radiation-hydrodynamics code. We conclude with a diffusion of future x-ray production studies that can be conducted on RACE and a possible multi-megajoule upgrade

  17. Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

    International Nuclear Information System (INIS)

    Kugel, H.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.P.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.K.; Menard, J.; Stotler, D.; Zakharov, L.E.; Maingi, Rajesh; Nygren, R.E.; Soukhanovskii, V.; Wakeland, P.

    2009-01-01

    Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15 25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW 1), to enable ne scan capability (2) in the H-mode, to test the ability to operate at significantly lower density (e.g., ne/nGW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  18. Ultrasoft x-ray imaging system for the National Spherical Torus Experiment

    Science.gov (United States)

    Stutman, D.; Finkenthal, M.; Soukhanovskii, V.; May, M. J.; Moos, H. W.; Kaita, R.

    1999-01-01

    A spectrally resolved ultrasoft x-ray imaging system, consisting of arrays of high resolution (the National Spherical Torus Experiment. Initially, three poloidal arrays of diodes filtered for C 1s-np emission will be implemented for fast tomographic imaging of the colder start-up plasmas. Later on, mirrors tuned to the C Lyα emission will be added in order to enable the arrays to "see" the periphery through the hot core and to study magnetohydrodynamic activity and impurity transport in this region. We also discuss possible core diagnostics, based on tomographic imaging of the Lyα emission from the plume of recombined, low Z impurity ions left by neutral beams or fueling pellets. The arrays can also be used for radiated power measurements and to map the distribution of high Z impurities injected for transport studies. The performance of the proposed system is illustrated with results from test channels on the CDX-U spherical torus at Princeton Plasma Physics Laboratory.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, W.; Bell, M.; Berzak,L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren,R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov L. E.

    2008-09-26

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

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

    International Nuclear Information System (INIS)

    Kugel, W.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren, R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov, L. E.

    2008-01-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW∼1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization

  1. The effect of lithium surface coatings on plasma performance in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Kugel, H.; Bell, M.; Ahn, J.W.; Bush, C.E.; Maingi, R.

    2008-01-01

    National Spherical Torus Experiment (which M. Ono, Nucl. Fusion 40, 557 (2000)) high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to plasma facing components found previously in limited plasmas can occur also in high-power diverted configurations. Lithium coatings were applied with pellets injected into helium discharges, and also with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium oven depositions from a few milligrams to 1 g have been applied between discharges. Benefits from the lithium coatings were sometimes, but not always, seen. These benefits sometimes included decreases in plasma density, inductive flux consumption, and edge-localized mode occurrence, and increases in electron temperature, ion temperature, energy confinement, and periods of edge and magnetohydrodynamic quiescence. In addition, reductions in lower divertor D, C, and O luminosity were measured.

  2. 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.

  3. Scintillator Based Energetic Ion Loss Diagnostic for the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Darrow, D.S.

    2007-01-01

    A scintillator based energetic ion loss detector has been built and installed on the National Spherical Torus Experiment (NSTX) to measure the loss of neutral beam ions. The detector is able to resolve the pitch angle and gyroradius of the lost energetic ions. It has a wide acceptance range in pitch angle and energy, and is able to resolve the full, one-half, and one-third energy components of the 80 keV D neutral beams up to the maximum toroidal magnetic field of NSTX. Multiple Faraday cups have been embedded behind the scintillator to allow easy absolute calibration of the diagnostic and to measure the energetic ion loss to several ranges of pitch angle with good time resolution. Several small, vacuum compatible lamps allow simple calibration of the scintillator position within the field of view of the diagnostic's video camera

  4. Scintillator Based Energetic Ion Loss Diagnostic for the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    D.S. Darrow

    2007-07-02

    A scintillator based energetic ion loss detector has been built and installed on the National Spherical Torus Experiment (NSTX) to measure the loss of neutral beam ions. The detector is able to resolve the pitch angle and gyroradius of the lost energetic ions. It has a wide acceptance range in pitch angle and energy, and is able to resolve the full, one-half, and one-third energy components of the 80 keV D neutral beams up to the maximum toroidal magnetic field of NSTX. Multiple Faraday cups have been embedded behind the scintillator to allow easy absolute calibration of the diagnostic and to measure the energetic ion loss to several ranges of pitch angle with good time resolution. Several small, vacuum compatible lamps allow simple calibration of the scintillator position within the field of view of the diagnostic's video camera.

  5. 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

  6. Prompt Loss of Energetic Ions during Early Neutral Beam Injection in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Medley, S.S.; Darrow, D.S.; Liu, D.; Roquemore, A.L.

    2005-01-01

    Early neutral-beam injection is used in the National Spherical Torus Experiment (NSTX) to heat the electrons and slow current penetration which keeps q(0) elevated to avoid deleterious MHD activity and at the same time reduces Ohmic flux consumption, all of which aids long-pulse operation. However, the low plasma current (I p ∼ 0.5 MA) and electron density (n e ∼ 1 x 10 13 cm -3 ) attending early injection lead to elevated orbit and shine through losses. The inherent orbit losses are aggravated by large excursions in the outer gap width during current ramp-up. An investigation of this behavior using various energetic particle diagnostics on NSTX and TRANSP code analysis is presented

  7. Simulation and design of feedback control on resistive wall modes in Keda Torus eXperiment

    International Nuclear Information System (INIS)

    Li, Chenguang; Liu, Wandong; Li, Hong

    2014-01-01

    The feedback control of resistive wall modes (RWMs) in Keda Torus eXperiment (KTX) (Liu et al., Plasma Phys. Controlled Fusion 56, 094009 (2014)) is investigated by simulation. A linear model is built to describe the growth of the unstable modes in the absence of feedback and the resulting mode suppression due to feedback, given the typical reversed field pinch plasma equilibrium. The layout of KTX with two shell structures (the vacuum vessel and the stabilizing shell) is taken into account. The feedback performance is explored both in the scheme of “clean mode control” (Zanca et al., Nucl. Fusion 47, 1425 (2007)) and “raw mode control.” The discrete time control model with specific characteristic times will mimic the real feedback control action and lead to the favored control cycle. Moreover, the conceptual design of feedback control system is also presented, targeting on both RWMs and tearing modes

  8. 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

  9. Studies of improved electron confinement in low density L-mode National Spherical Torus Experiment discharges

    International Nuclear Information System (INIS)

    Stutman, D.; Finkenthal, M.; Tritz, K.; Redi, M. H.; Kaye, S. M.; Bell, M. G.; Bell, R. E.; LeBlanc, B. P.; Hill, K. W.; Medley, S. S.; Menard, J. E.; Rewoldt, G.; Wang, W. X.; Synakowski, E. J.; Levinton, F.; Kubota, S.; Bourdelle, C.; Dorland, W.; The NSTX Team

    2006-01-01

    Electron transport is rapid in most National Spherical Torus Experiment, M. Ono et al., Nucl. Fusion 40, 557 (2000) beam heated plasmas. A regime of improved electron confinement is nevertheless observed in low density L-mode (''low-confinement'') discharges heated by early beam injection. Experiments were performed in this regime to study the role of the current profile on thermal transport. Variations in the magnetic shear profile were produced by changing the current ramp rate and onset of neutral beam heating. An increased electron temperature gradient and local minimum in the electron thermal diffusivity were observed at early times in plasmas with the fastest current ramp and earliest beam injection. In addition, an increased ion temperature gradient associated with a region of reduced ion transport is observed at slightly larger radii. Ultrasoft x-ray measurements of double-tearing magnetohydrodynamic activity, together with current diffusion calculations, point to the existence of negative magnetic shear in the core of these plasmas. Discharges with slower current ramp and delayed beam onset, which are estimated to have more monotonic q-profiles, do not exhibit regions of reduced transport. The results are discussed in the light of the initial linear microstability assessment of these plasmas, which suggests that the growth rate of all instabilities, including microtearing modes, can be reduced by negative or low magnetic shear in the temperature gradient region. Several puzzles arising from the present experiments are also highlighted

  10. Surface Treatment of a Lithium Limiter for Spherical Torus Plasma Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R.; Majeski, R.; Doerner, R.; Antar, G.; Timberlake, J.; Spaleta, J.; Hoffman, D.; Jones, B.; Munsat, T.; Kugel, H.; Taylor, G.; Stutman, D.; Soukhanovskii, V.; Maingi, R.; Molesa, S.; Efthimion, P.; Menard, J.; Finkenthal, M.; Luckhardt, S.

    2001-03-20

    The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. As part of investigations to determine the feasibility of this approach, plasma interaction questions in a toroidal plasma geometry are being addressed in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The first experiments involved a toroidally local lithium limiter (L3). Measurements of pumpout rates indicated that deuterium pumping was greater for the L3 compared to conventional boron carbide limiters. The difference in the pumpout rates between the two limiter types decreased with plasma exposure, but argon glow discharge cleaning was able to restore the pumping effectiveness of the L3. At no point, however, was the extremely low recycling regime reported in previous lithium experiments achieved. This may be due to the much larger lithium surfaces that were exposed to the plasma in the earlier work. The possibility will be studied in the next set of CDX-U experiments, which are to be conducted with a large area, fully toroidal lithium limiter.

  11. Surface Treatment of a Lithium Limiter for Spherical Torus Plasma Experiments

    International Nuclear Information System (INIS)

    Kaita, R.; Majeski, R.; Doerner, R.; Antar, G.; Timberlake, J.; Spaleta, J.; Hoffman, D.; Jones, B.; Munsat, T.; Kugel, H.; Taylor, G.; Stutman, D.; Soukhanovskii, V.; Maingi, R.; Molesa, S.; Efthimion, P.; Menard, J.; Finkenthal, M.; Luckhardt, S.

    2001-01-01

    The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. As part of investigations to determine the feasibility of this approach, plasma interaction questions in a toroidal plasma geometry are being addressed in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The first experiments involved a toroidally local lithium limiter (L3). Measurements of pumpout rates indicated that deuterium pumping was greater for the L3 compared to conventional boron carbide limiters. The difference in the pumpout rates between the two limiter types decreased with plasma exposure, but argon glow discharge cleaning was able to restore the pumping effectiveness of the L3. At no point, however, was the extremely low recycling regime reported in previous lithium experiments achieved. This may be due to the much larger lithium surfaces that were exposed to the plasma in the earlier work. The possibility will be studied in the next set of CDX-U experiments, which are to be conducted with a large area, fully toroidal lithium limiter

  12. 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

  13. 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

  14. Control and data acquisition system for versatile experiment spherical torus at SNU

    Energy Technology Data Exchange (ETDEWEB)

    An, YoungHwa [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of); Chung, Kyoung-Jae, E-mail: jkjlsh1@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of); Na, DongHyeon; Hwang, Y.S. [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of)

    2013-10-15

    A control and data acquisition system for VEST (Versatile Experiment Spherical Torus) at Seoul National University (SNU) has been developed to enable remote operation from a central control room. The control and data acquisition system consists of three subsystems; a main control and data acquisition system that triggers each device at the preprogrammed timing and collects various diagnostic signals during discharges, a monitoring system that watches and logs the device status continuously, and a data storage and distribution system that stores collected data and provides data access layer via Ethernet. The system is designed to be cost-effective, extensible and easy to develop by using well-established standard technologies and solutions. Combining broad accessibility with modern information technology, alarm signal can be sent immediately to the registered cell phones when the abnormal status of devices is found, and the web data distribution system enables data access from almost everywhere using smart phones or tablet computers. Since December 2011, VEST is operational and the control and data acquisition system has been successfully used for remote operation of VEST.

  15. A megawatt-level 28 GHz heating system for the National Spherical Torus Experiment Upgrade

    Directory of Open Access Journals (Sweden)

    Taylor G.

    2015-01-01

    Full Text Available The National Spherical Torus Experiment Upgrade (NSTX-U will operate at axial toroidal fields of ≤ 1 T and plasma currents, Ip ≤ 2 MA. The development of non-inductive (NI plasmas is a major long-term research goal for NSTX-U. Time dependent numerical simulations of 28 GHz electron cyclotron (EC heating of low density NI start-up plasmas generated by Coaxial Helicity Injection (CHI in NSTX-U predict a significant and rapid increase of the central electron temperature (Te(0 before the plasma becomes overdense. The increased Te(0 will significantly reduce the Ip decay rate of CHI plasmas, allowing the coupling of fast wave heating and neutral beam injection. A megawatt-level, 28 GHz electron heating system is planned for heating NI start-up plasmas in NSTX-U. In addition to EC heating of CHI start-up discharges, this system will be used for electron Bernstein wave (EBW plasma start-up, and eventually for EBW heating and current drive during the Ip flattop.

  16. Hybrid simulation of toroidal Alfvén eigenmode on the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Liu, D., E-mail: deyongl@uci.edu [Department of Physics and Astronomy, University of California, Irvine, California 92697 (United States); Fu, G. Y.; Podestà, M.; Breslau, J. A.; Fredrickson, E. D. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Crocker, N. A.; Kubota, S. [Department of Physics and Astronomy, University of California, Los Angles, California 90095 (United States)

    2015-04-15

    Energetic particle modes and Alfvén eigenmodes driven by super-Alfvénic fast ions are routinely observed in neutral beam heated plasmas on the National Spherical Torus eXperiment (NSTX). These modes can significantly impact fast ion transport and thus cause fast ion redistribution or loss. Self-consistent linear simulations of Toroidal Alfvén Eigenmodes (TAEs) in NSTX plasmas have been carried out with the kinetic/magnetohydrodynamic hybrid code M3D-K using experimental plasma parameters and profiles including plasma toroidal rotation. The simulations show that unstable TAEs with n=3,4, or 5 can be excited by the fast ions from neutral beam injection. The simulated mode frequency, mode radial structure, and phase shift are consistent with measurements from a multi-channel microwave reflectometer diagnostic. A sensitivity study on plasma toroidal rotation, safety factor q profile, and initial fast ion distribution is performed. The simulations show that rotation can have a significant destabilizing effect when the rotation is comparable or larger than the experimental level. The mode growth rate is sensitive to q profile and fast ion distribution. Although mode structure and peak position depend somewhat on q profile and plasma rotation, the variation of synthetic reflectometer response is within experimental uncertainty and it is not sensitive enough to see the difference clearly.

  17. Non-linear dynamics of toroidicity-induced Alfven eigenmodes on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Podesta, M.; Bell, R.E.; Fredrickson, E.D.; Gorelenkov, N.N.; LeBlanc, B.P.; Crocker, N.A.; Kubota, S.; Heidbrink, W.W.; Yuh, H.

    2011-01-01

    The National Spherical Torus Experiment (NSTX, (Ono et al 2000 Nucl. Fusion 40 557)) routinely operates with neutral beam injection as the primary system for heating and current drive. The resulting fast ion population is super-Alfvenic, with velocities 1 fast /v Alfven < 5. This provides a strong drive for toroidicity-induced Alfven eigenmodes (TAEs). As the discharge evolves, the fast ion population builds up and TAEs exhibit increasing bursts in amplitude and down-chirps in frequency, which eventually lead to a so-called TAE avalanche. Avalanches cause large (∼<30%) fast ion losses over ∼1 ms, as inferred from the neutron rate. The increased fast ion losses correlate with a stronger activity in the TAE band. In addition, it is shown that a n = 1 mode with frequency well below the TAE gap appears in the Fourier spectrum of magnetic fluctuations as a result of non-linear mode coupling between TAEs during avalanche events. The non-linear coupling between modes, which leads to enhanced fast ion transport during avalanches, is investigated.

  18. Non-linear Dynamics Of Toroidicity-induced Alfven Eigenmodes On The National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Podesta, M.; Bell, R.E.; Crocker, N.A.; Fredrickson, E.D.; Gorelenkov, N.N.; Heidbrink, W.W.; Kubota, S.; LeBlanc, B.P.; Yu, H.

    2011-01-01

    The National Spherical Torus Experiment (NSTX, (M. Ono et al., Nucl. Fusion 40, 557 (2000))) routinely operates with neutral beam injection as the primary system for heating and current drive. The resulting fast ion population is super-Alfvenic, with velocities 1 fast /v Alfven < 5. This provides a strong drive for toroidicity-induced Alfven eigenmodes (TAEs). As the discharge evolves, the fast ion population builds up and TAEs exhibit increasing bursts in amplitude and down-chirps in frequency, which eventually lead to a so-called TAE avalanche. Avalanches cause large (∼<30%) fast ion losses over ∼ 1 ms, as inferred from the neutron rate. The increased fast ion losses correlate with a stronger activity in the TAE band. In addition, it is shown that a n = 1 mode with frequency well below the TAE gap appears in the Fourier spectrum of magnetic fluctuations as a result of non-linear mode coupling between TAEs during avalanche events. The non-linear coupling between modes, which leads to enhanced fast ion transport during avalanches, is investigated.

  19. Non-linear Dynamics Of Toroidicity-induced Alfven Eigenmodes On The National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Podesta, M; Crocker, N A; Fredrickson, E D; Gorelenkov, N N; Heidbrink, W W; Kubota, S; LeBlanc, B P

    2011-04-26

    The National Spherical Torus Experiment (NSTX, [M. Ono et al., Nucl. Fusion 40, 557 (2000)]) routinely operates with neutral beam injection as the primary system for heating and current drive. The resulting fast ion population is super-Alfv enic, with velocities 1 < vfast=vAlfven < 5. This provides a strong drive for toroidicity-induced Alfv en eigenmodes (TAEs). As the discharge evolves, the fast ion population builds up and TAEs exhibit increasing bursts in amplitude and down-chirps in frequency, which eventually lead to a so-called TAE avalanche. Avalanches cause large (≤ 30%) fast ion losses over ~ 1 ms, as inferred from the neutron rate. The increased fast ion losses correlate with a stronger activity in the TAE band. In addition, it is shown that a n = 1 mode with frequency well below the TAE gap appears in the Fourier spectrum of magnetic fluctuations as a result of non-linear mode coupling between TAEs during avalanche events. The non-linear coupling between modes, which leads to enhanced fast ion transport during avalanches, is investigated.

  20. Spectroscopic diagnostics for liquid lithium divertor studies on National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Soukhanovskii, V. A.; Roquemore, A. L.; Bell, R. E.; Kaita, R.; Kugel, H. W.

    2010-01-01

    The use of lithium-coated plasma facing components for plasma density control is studied in the National Spherical Torus Experiment (NSTX). A recently installed liquid lithium divertor (LLD) module has a porous molybdenum surface, separated by a stainless steel liner from a heated copper substrate. Lithium is deposited on the LLD from two evaporators. Two new spectroscopic diagnostics are installed to study the plasma surface interactions on the LLD: (1) A 20-element absolute extreme ultraviolet (AXUV) diode array with a 6 nm bandpass filter centered at 121.6 nm (the Lyman-α transition) for spatially resolved divertor recycling rate measurements in the highly reflective LLD environment, and (2) an ultraviolet-visible-near infrared R=0.67 m imaging Czerny-Turner spectrometer for spatially resolved divertor D I, Li I-II, C I-IV, Mo I, D 2 , LiD, CD emission and ion temperature on and around the LLD module. The use of photometrically calibrated measurements together with atomic physics factors enables studies of recycling and impurity particle fluxes as functions of LLD temperature, ion flux, and divertor geometry.

  1. Intermittent Divertor Filaments in the National Spherical Torus Experiment and Their Relation to Midplane Blobs

    International Nuclear Information System (INIS)

    Maqueda, R.J.; Stotler, D.P.

    2010-01-01

    While intermittent filamentary structures, also known as blobs, are routinely seen in the low-field-side scrape-off layer of the National Spherical Torus Experiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557), fine structured filaments are also seen on the lower divertor target plates of NSTX. These filaments, not associated with edge localized modes, correspond to the interaction of the turbulent blobs seen near the midplane with the divertor plasma facing components. The fluctuation level of the neutral lithium light observed at the divertor, and the skewness and kurtosis of its probability distribution function, is similar to that of midplane blobs seen in D α ; e.g. increasing with increasing radii outside the outer strike point (OSP) (separatrix). In addition, their toroidal and radial movement agrees with the typical movement of midplane blobs. Furthermore, with the appropriate magnetic topology, i.e. mapping between the portion of the target plates being observed into the field of view of the midplane gas puff imaging diagnostic, very good correlation is observed between the blobs and the divertor filaments. The correlation between divertor plate filaments and midplane blobs is lost close to the OSP. This latter observation is consistent with the existence of 'magnetic shear disconnection' due to the lower X-point, as proposed by Cohen and Ryutov (1997 Nucl. Fusion 37 621).

  2. The confinement of dilute populations of beam ions in the national spherical torus experiment

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Miah, M.; Darrow, D.; Le Blanc, B.; Medley, S.; Roquemore, A.L.; Cecil, F.E.

    2003-01-01

    Short ∼3 ms pulses of 80 keV deuterium neutrals are injected at three different tangency radii into the national spherical torus experiment. The confinement is studied as a function of tangency radius, plasma current (between 0.4 and 1.0 MA), and toroidal field (between 2.5 and 5.0 kG). The jump in neutron emission during the pulse is used to infer prompt losses of beam ions. In the absence of MHD, the neutron data show the expected dependences on beam angle and plasma current; the average jump in the neutron signal is 88±39% of the expected jump. The decay of the neutron and neutral particle signals following the blip are compared to the expected classical deceleration to detect losses on a 10 ms timescale. The temporal evolution of these signals are consistent with Coulomb scattering rates, implying an effective beam-ion confinement time > or ∼ 100 ms. The confinement is insensitive to the toroidal field despite large values of ρ∇B/B < or ∼(0.25), so any effects of non-conservation of the adiabatic invariant μ are smaller than the experimental error. (author)

  3. Development of wall conditioning and impurity monitoring systems in Versatile Experiment Spherical Torus (VEST)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.Y., E-mail: brbbebbero@snu.ac.kr [Seoul National University, Seoul (Korea, Republic of); Yang, J.; Kim, Y.G.; Yang, S.M.; Kim, Y.S.; Lee, K.H. [Seoul National University, Seoul (Korea, Republic of); An, Y.H. [National Fusion Research Institute, Daejon (Korea, Republic of); Chung, K.J.; Na, Y.S. [Seoul National University, Seoul (Korea, Republic of); Hwang, Y.S., E-mail: yhwang@snu.ac.kr [Seoul National University, Seoul (Korea, Republic of)

    2016-11-01

    Highlights: • The baking for partial wall heating and H{sub 2}/He GDC systems are developed in VEST. • The RGA and OES systems for monitoring impurities are constructed in VEST. • The partial baking and He GDC show limited effects on plasma characteristics. • H{sub 2} GDC above 4 h enables the longer plasma current duration up to ∼15 ms. • After H{sub 2} GDC, the discharge should be conducted within 3 h from treatment. - Abstract: Wall conditioning and impurity monitoring systems are developed in Versatile Experiment Spherical Torus (VEST). As a wall conditioning system, a baking system covering the vacuum vessel wall partially and a glow discharge cleaning (GDC) system using two electrodes with dc and 50 kHz power supplies are installed. The GDC system operates with hydrogen and helium gases for both chemical and physical desorption. The impurity monitoring system with residual gas analyzer (RGA), operating at <10{sup −5} Torr with a differential pumping system, is installed along with the optical emission spectroscopy (OES) system to monitor the hydrogen and impurity radiation lines. Effects of these wall conditioning techniques are investigated with the impurity monitoring system for ohmic discharges of VEST. The partial baking and He GDC show limited effects on plasma characteristics but sufficient H{sub 2} GDC above 4 h enables the longer plasma current duration up to ∼15 ms within 3 h from the end of treatment.

  4. Development of wall conditioning and impurity monitoring systems in Versatile Experiment Spherical Torus (VEST)

    International Nuclear Information System (INIS)

    Lee, H.Y.; Yang, J.; Kim, Y.G.; Yang, S.M.; Kim, Y.S.; Lee, K.H.; An, Y.H.; Chung, K.J.; Na, Y.S.; Hwang, Y.S.

    2016-01-01

    Highlights: • The baking for partial wall heating and H_2/He GDC systems are developed in VEST. • The RGA and OES systems for monitoring impurities are constructed in VEST. • The partial baking and He GDC show limited effects on plasma characteristics. • H_2 GDC above 4 h enables the longer plasma current duration up to ∼15 ms. • After H_2 GDC, the discharge should be conducted within 3 h from treatment. - Abstract: Wall conditioning and impurity monitoring systems are developed in Versatile Experiment Spherical Torus (VEST). As a wall conditioning system, a baking system covering the vacuum vessel wall partially and a glow discharge cleaning (GDC) system using two electrodes with dc and 50 kHz power supplies are installed. The GDC system operates with hydrogen and helium gases for both chemical and physical desorption. The impurity monitoring system with residual gas analyzer (RGA), operating at <10"−"5 Torr with a differential pumping system, is installed along with the optical emission spectroscopy (OES) system to monitor the hydrogen and impurity radiation lines. Effects of these wall conditioning techniques are investigated with the impurity monitoring system for ohmic discharges of VEST. The partial baking and He GDC show limited effects on plasma characteristics but sufficient H_2 GDC above 4 h enables the longer plasma current duration up to ∼15 ms within 3 h from the end of treatment.

  5. Fast response of electron-scale turbulence to auxiliary heating cessation in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Y.; Wang, W. X.; LeBlanc, B. P.; Guttenfelder, W.; Kaye, S. M.; Ethier, S.; Mazzucato, E.; Bell, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Lee, K. C. [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Domier, C. W. [University of California at Davis, Davis, California 95616 (United States); Smith, D. R. [University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Yuh, H. [Nova Photonics, Inc., Princeton, New Jersey 08540 (United States)

    2015-11-15

    In this letter, we report the first observation of the fast response of electron-scale turbulence to auxiliary heating cessation in National Spherical Torus eXperiment [Ono et al., Nucl. Fusion 40, 557 (2000)]. The observation was made in a set of RF-heated L-mode plasmas with toroidal magnetic field of 0.55 T and plasma current of 300 kA. It is observed that electron-scale turbulence spectral power (measured with a high-k collective microwave scattering system) decreases significantly following fast cessation of RF heating that occurs in less than 200 μs. The large drop in the turbulence spectral power has a short time delay of about 1–2 ms relative to the RF cessation and happens on a time scale of 0.5–1 ms, much smaller than the energy confinement time of about 10 ms. Power balance analysis shows a factor of about 2 decrease in electron thermal diffusivity after the sudden drop of turbulence spectral power. Measured small changes in equilibrium profiles across the RF cessation are unlikely able to explain this sudden reduction in the measured turbulence and decrease in electron thermal transport, supported by local linear stability analysis and both local and global nonlinear gyrokinetic simulations. The observations imply that nonlocal flux-driven mechanism may be important for the observed turbulence and electron thermal transport.

  6. Control and data acquisition system for versatile experiment spherical torus at SNU

    International Nuclear Information System (INIS)

    An, YoungHwa; Chung, Kyoung-Jae; Na, DongHyeon; Hwang, Y.S.

    2013-01-01

    A control and data acquisition system for VEST (Versatile Experiment Spherical Torus) at Seoul National University (SNU) has been developed to enable remote operation from a central control room. The control and data acquisition system consists of three subsystems; a main control and data acquisition system that triggers each device at the preprogrammed timing and collects various diagnostic signals during discharges, a monitoring system that watches and logs the device status continuously, and a data storage and distribution system that stores collected data and provides data access layer via Ethernet. The system is designed to be cost-effective, extensible and easy to develop by using well-established standard technologies and solutions. Combining broad accessibility with modern information technology, alarm signal can be sent immediately to the registered cell phones when the abnormal status of devices is found, and the web data distribution system enables data access from almost everywhere using smart phones or tablet computers. Since December 2011, VEST is operational and the control and data acquisition system has been successfully used for remote operation of VEST

  7. Observation of Beam Driven Modes during Neutral Beam Heating on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Fredrickson, E.D.; Gorelenkov, E.D.; Cheng, C.Z.; Bell, R.; Darrow, D.; Johnson, D.; Kaye, S.; LeBlanc, B.; Menard, J.; Kubota, S.; Peebles, W.

    2001-01-01

    With the first injection of neutral beams on the National Spherical Torus Experiment (NSTX), a broad and complicated spectrum of coherent modes was seen between approximately 0.4 MHz and 2.5 MHz [where f(subscript ''ci'')] for deuterium is approximately 2.2 MHz. The modes have been observed with high bandwidth magnetic pick-up coils and with a reflectometer. The parametric scaling of the mode frequency with density and magnetic field is consistent with Alfvenic modes (linear in B, inversely with the square root of density). These modes have been identified as magnetosonic waves or compressional Alfven eigenmodes (CAE) excited by a cyclotron resonance with the neutral-beam ions. Modes have also been observed in the frequency range 50-150 kHz with toroidal mode numbers n = 1-5. These lower frequency modes are thought to be related to the TAE [Toroidal Alfven Eigenmode] seen commonly in tokamaks and driven by energetic fast ion populations resulting from ICRF [ion cyclotron range of frequency] and NBI [neutral-beam injection] heating. There is no clear indication of enhanced fast ion losses associated with the modes

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

    International Nuclear Information System (INIS)

    Gerhardt, S.P.; Menard, J.E.

    2008-01-01

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

  9. Liquid-metal plasma-facing component research on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Jaworski, M A; Khodak, A; Kaita, R

    2013-01-01

    Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation of economically viable fusion power reactors. Liquid metals face critical issues in three key areas: free-surface stability, material migration and demonstration of integrated scenarios. To date, few demonstrations exist of this approach in a diverted tokamak and we here provide an overview of such work on the National Spherical Torus Experiment (NSTX). The liquid lithium divertor (LLD) was installed and operated for the 2010 run campaign using evaporated coatings as the filling method. Despite a nominal liquid level exceeding the capillary structure and peak current densities into the PFCs exceeding 100 kA m −2 , no macroscopic ejection events were observed. The stability can be understood from a Rayleigh–Taylor instability analysis. Capillary restraint and thermal-hydraulic considerations lead to a proposed liquid-metal PFCs scheme of actively-supplied, capillary-restrained systems. Even with state-of-the-art cooling techniques, design studies indicate that the surface temperature with divertor-relevant heat fluxes will still reach temperatures above 700 °C. At this point, one would expect significant vapor production from a liquid leading to a continuously vapor-shielded regime. Such high-temperature liquid lithium PFCs may be possible on the basis of momentum-balance arguments. (paper)

  10. Liquid-metal plasma-facing component research on the National Spherical Torus Experiment

    Science.gov (United States)

    Jaworski, M. A.; Khodak, A.; Kaita, R.

    2013-12-01

    Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation of economically viable fusion power reactors. Liquid metals face critical issues in three key areas: free-surface stability, material migration and demonstration of integrated scenarios. To date, few demonstrations exist of this approach in a diverted tokamak and we here provide an overview of such work on the National Spherical Torus Experiment (NSTX). The liquid lithium divertor (LLD) was installed and operated for the 2010 run campaign using evaporated coatings as the filling method. Despite a nominal liquid level exceeding the capillary structure and peak current densities into the PFCs exceeding 100 kA m-2, no macroscopic ejection events were observed. The stability can be understood from a Rayleigh-Taylor instability analysis. Capillary restraint and thermal-hydraulic considerations lead to a proposed liquid-metal PFCs scheme of actively-supplied, capillary-restrained systems. Even with state-of-the-art cooling techniques, design studies indicate that the surface temperature with divertor-relevant heat fluxes will still reach temperatures above 700 °C. At this point, one would expect significant vapor production from a liquid leading to a continuously vapor-shielded regime. Such high-temperature liquid lithium PFCs may be possible on the basis of momentum-balance arguments.

  11. Design and development of the helicity injection system in Versatile Experiment Spherical Torus

    International Nuclear Information System (INIS)

    Park, JongYoon; An, Younghwa; Jung, Bongki; Lee, Jeongwon; Lee, HyunYoung; Chung, Kyoung-Jae; Na, Yong-Su; Hwang, Y.S.

    2015-01-01

    Graphical abstract: - Highlights: • A high current electron gun with single pulse power for both arc and extraction is developed. • The optimal gun operation is confirmed by impedance matching between the PFN and plasma. • The gun injected currents of 0.95 kA with the voltage of ∼410 V for 5 ms with a 1.2 kV PFN. • The helicity injection system using the gun has been developed and tested successfully in VEST. • Toroidal currents of up to 3.8 kA confirm possible relaxation into tokamak-like plasma. - Abstract: A helicity injection system for the Versatile Experiment Spherical Torus (VEST) has been successfully developed and commissioned. A high current electron gun utilizing hollow cathode and washer stacks has been designed and constructed with a single pulse power system that can provide voltages for both arc discharge and extraction sequentially. Tests for electron gun operation with the single pulse power system have been conducted under various toroidal and poloidal field strengths. The estimated plasma impedance, depending on the injection magnetic field structure, can be utilized for the optimal gun operation by impedance matching between the pulse power system and plasma. With the charging voltage of 1.2 kV, injection current of 0.95 kA has been obtained with the injection voltage of 410 V for about 5 ms. Initial helicity injection experiments have been conducted under various toroidal and poloidal field strengths and a toroidal plasma current of up to 3.8 kA is observed with the current multiplication larger than the geometric stacking ratio, confirming the possibility of relaxation into tokamak-like plasma with closed flux formation.

  12. Design and development of the helicity injection system in Versatile Experiment Spherical Torus

    Energy Technology Data Exchange (ETDEWEB)

    Park, JongYoon; An, Younghwa; Jung, Bongki; Lee, Jeongwon; Lee, HyunYoung; Chung, Kyoung-Jae; Na, Yong-Su; Hwang, Y.S., E-mail: yhwang@snu.ac.kr

    2015-10-15

    Graphical abstract: - Highlights: • A high current electron gun with single pulse power for both arc and extraction is developed. • The optimal gun operation is confirmed by impedance matching between the PFN and plasma. • The gun injected currents of 0.95 kA with the voltage of ∼410 V for 5 ms with a 1.2 kV PFN. • The helicity injection system using the gun has been developed and tested successfully in VEST. • Toroidal currents of up to 3.8 kA confirm possible relaxation into tokamak-like plasma. - Abstract: A helicity injection system for the Versatile Experiment Spherical Torus (VEST) has been successfully developed and commissioned. A high current electron gun utilizing hollow cathode and washer stacks has been designed and constructed with a single pulse power system that can provide voltages for both arc discharge and extraction sequentially. Tests for electron gun operation with the single pulse power system have been conducted under various toroidal and poloidal field strengths. The estimated plasma impedance, depending on the injection magnetic field structure, can be utilized for the optimal gun operation by impedance matching between the pulse power system and plasma. With the charging voltage of 1.2 kV, injection current of 0.95 kA has been obtained with the injection voltage of 410 V for about 5 ms. Initial helicity injection experiments have been conducted under various toroidal and poloidal field strengths and a toroidal plasma current of up to 3.8 kA is observed with the current multiplication larger than the geometric stacking ratio, confirming the possibility of relaxation into tokamak-like plasma with closed flux formation.

  13. Ohmic H-mode studies in TUMAN-3

    International Nuclear Information System (INIS)

    Lebedev, S.V.; Andrejko, M.V.; Askinazi, L.G.; Golant, V.E.; Kornev, V.A.; Levin, L.S.; Tukachinsky, A.S.; Tendler, M.

    1994-01-01

    The spontaneous transition from Ohmically heated limiter discharges into the regime with improved confinement termed as ''Ohmic H-mode'' has been investigated in ''TUMAN-3''. The typical signatures of H-mode in tokamaks with powerful auxiliary heating have been observed: sharp drop of D α radiation with simultaneous increase in the electron density and stored energy, suppression of the density fluctuations and establishing the steep gradient near the periphery. The crucial role of the radial electric field in the L-H transition was found in the experiments with boundary biasing. The possibility of initiating the H-mode using single pellet injection was demonstrated. In Ohmic H-mode strong dependencies of τ E on plasma current and on input power and weak dependence on density were found. Thermal energy confinement time enhanced by a factor of 10 compared to predictions of Neo-Alcator scaling. Longest energy confinement time (30 ms) was obtained in the small tokamak TUMAN-3. Absolute values of the energy confinement time are in agreement with scaling proposed for description of the ELM-free H-modes in devices with powerful auxiliary heating (''DIII-D/JET H-mode'' scaling). (author)

  14. 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

  15. 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

  16. Stabilization of electron-scale turbulence by electron density gradient in national spherical torus experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz Ruiz, J.; White, A. E. [MIT-Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Ren, Y.; Guttenfelder, W.; Kaye, S. M.; Leblanc, B. P.; Mazzucato, E. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Lee, K. C. [National Fusion Research Institute, Daejeon (Korea, Republic of); Domier, C. W. [University of California at Davis, Davis, California 95616 (United States); Smith, D. R. [University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Yuh, H. [Nova Photonics, Inc., Princeton, New Jersey 08540 (United States)

    2015-12-15

    Theory and experiments have shown that electron temperature gradient (ETG) turbulence on the electron gyro-scale, k{sub ⊥}ρ{sub e} ≲ 1, can be responsible for anomalous electron thermal transport in NSTX. Electron scale (high-k) turbulence is diagnosed in NSTX with a high-k microwave scattering system [D. R. Smith et al., Rev. Sci. Instrum. 79, 123501 (2008)]. Here we report on stabilization effects of the electron density gradient on electron-scale density fluctuations in a set of neutral beam injection heated H-mode plasmas. We found that the absence of high-k density fluctuations from measurements is correlated with large equilibrium density gradient, which is shown to be consistent with linear stabilization of ETG modes due to the density gradient using the analytical ETG linear threshold in F. Jenko et al. [Phys. Plasmas 8, 4096 (2001)] and linear gyrokinetic simulations with GS2 [M. Kotschenreuther et al., Comput. Phys. Commun. 88, 128 (1995)]. We also found that the observed power of electron-scale turbulence (when it exists) is anti-correlated with the equilibrium density gradient, suggesting density gradient as a nonlinear stabilizing mechanism. Higher density gradients give rise to lower values of the plasma frame frequency, calculated based on the Doppler shift of the measured density fluctuations. Linear gyrokinetic simulations show that higher values of the electron density gradient reduce the value of the real frequency, in agreement with experimental observation. Nonlinear electron-scale gyrokinetic simulations show that high electron density gradient reduces electron heat flux and stiffness, and increases the ETG nonlinear threshold, consistent with experimental observations.

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

    International Nuclear Information System (INIS)

    Maingi, R.; Tritz, K.; Fredrickson, E.D.; Menard, J.E.; Sabbagh, S.A.; Stutman, D.; Bell, M.G.; Bell, R.E.; Bush, C.E.; 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.

    2004-01-01

    We report observation of a high performance scenario in the National Spherical Torus Experiment with very small edge-localized modes (ELMs). These ELMs have no measurable impact on stored energy and are consistent with high bootstrap current operation with line average density approaching Greenwald scaling. The ELM perturbation is observed to typically originate near the lower divertor region, as opposed to the outer midplane for ELMs described in the literature. If extrapolable, this scenario would provide an attractive operating regime for next step fusion experiments

  18. Spherical Torus Center Stack Design

    International Nuclear Information System (INIS)

    C. Neumeyer; P. Heitzenroeder; C. Kessel; M. Ono; M. Peng; J. Schmidt; R. Woolley; I. Zatz

    2002-01-01

    The low aspect ratio spherical torus (ST) configuration requires that the center stack design be optimized within a limited available space, using materials within their established allowables. This paper presents center stack design methods developed by the National Spherical Torus Experiment (NSTX) Project Team during the initial design of NSTX, and more recently for studies of a possible next-step ST (NSST) device

  19. H-mode study in CHS

    International Nuclear Information System (INIS)

    Toi, K.; Morisaki, T.; Sakakibara, S.

    1995-02-01

    In CHS rapid H-mode transition is observed in NBI heated deuterium and hydrogen plasmas without obvious isotope effect, when a net plasma current is ramped up to increase the external rotational transform. The H-mode of CHS has many similarities with those in tokamaks. Recent measurement with fast response Langmuir probes has revealed that the rapid change in floating potential occurs at the transition, but the change follows the formation of edge transport barrier. The presence of ι/2π = 1 surface near the edge and sawtooth crash triggered by internal modes may play an important role for determining the H-mode transition in CHS. (author)

  20. Pellet fuelling and ELMy H-mode physics at JET

    International Nuclear Information System (INIS)

    Horton, L.D.

    2001-01-01

    As the reference operating regime for ITER, investigations of the ELMy H-mode have received high priority in the JET experimental programme. Recent experiments have concentrated in particular on operation simultaneously at high density and high confinement using high field side (HFS) pellet launch. The enhanced fuelling efficiency of HFS pellet fuelling is found to scale favourably to a large machine such as JET. The achievable density of ELMy H-mode plasmas in JET has been significantly increased using HFS fuelling although at the expense of confinement degradation back to L-mode levels. Initial experiments using control of the pellet injection frequency have shown that density and confinement can simultaneously be increased close to the values necessary for ITER. The boundaries of the available ELMy H-mode operational space have also been extensively explored. The power necessary to maintain the high confinement normally associated with ELMy H-mode operation is found to be substantially higher than the H-mode threshold power. The compatibility of ELMy H-modes with divertor operation acceptable for a fusion device has been studied. Narrow energy scrape-off widths are measured which place stringent limits on divertor power handling. Deuterium and tritium codeposition profiles are measured to be strongly in/out asymmetric. Successful modelling of these profiles requires the inclusion of the (measured) scrape-off layer flows and of the production in the divertor of hydrocarbon molecules with sticking coefficients below unity. Helium exhaust and compression are found to be within the limits sufficient for a reactor. (author)

  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. The H-mode operational window as determined from the ITER H-mode database

    International Nuclear Information System (INIS)

    Ryter, F.; Kardaun, O.J.W.F.; Stroth, U.

    1994-01-01

    The H-mode is a promising regime for fusion reactors and it is essential to be able to predict its operational window in future devices. The 'H-Mode Database Working Group' started in 1992 to gather, analyze and compare H-mode threshold data from several divertor tokamaks so that predictions could be made. The database and first results were presented and the threshold database has been improved and extended since. The work has two objectives: 1) to predict the minimum heating power necessary to reach the H-mode in future devices, 2) to contribute to physics studies of the L-H transition. (author) 11 refs., 2 figs

  3. Effects of Toroidal Rotation Sshear on Toroidicity-induced Alfven Eigenmodes in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Podesta, M; Fredrickson, E D; Gorelenkov, N N; LeBlanc, B P; Heidbrink, W W; Crocker, N A; Kubota, S

    2010-08-19

    The effects of a sheared toroidal rotation on the dynamics of bursting Toroidicity-induced Alfven eigenmodes are investigated in neutral beam heated plasmas on the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40 557 (2000)]. The modes have a global character, extending over most of the minor radius. A toroidal rotation shear layer is measured at the location of maximum drive for the modes. Contrary to results from other devices, no clear evidence of increased damping is found. Instead, experiments with simultaneous neutral beam and radio-frequency auxiliary heating show a strong correlation between the dynamics of the modes and the instability drive. It is argued that kinetic effects involving changes in the mode drive and damping mechanisms other than rotation shear, such as continuum damping, are mostly responsible for the bursting dynamics of the modes.

  4. Effects of Toroidal Rotation Shear on Toroidicity-induced Alfven Eigenmodes in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Podesta, M.; Bell, R.E.; Fredrickson, E.D.; Gorelenkov, N.N.; LeBlanc, B.P.; Heidbrink, W.W.; Crocker, N.A.; Kubota, S.; Yuh, H.

    2010-01-01

    The effects of a sheared toroidal rotation on the dynamics of bursting Toroidicity-induced Alfven eigenmodes are investigated in neutral beam heated plasmas on the National Spherical Torus Experiment (NSTX) (M. Ono et al., Nucl. Fusion 40 557 (2000)). The modes have a global character, extending over most of the minor radius. A toroidal rotation shear layer is measured at the location of maximum drive for the modes. Contrary to results from other devices, no clear evidence of increased damping is found. Instead, experiments with simultaneous neutral beam and radio-frequency auxiliary heating show a strong correlation between the dynamics of the modes and the instability drive. It is argued that kinetic effects involving changes in the mode drive and damping mechanisms other than rotation shear, such as continuum damping, are mostly responsible for the bursting dynamics of the modes.

  5. Physics of the H-mode

    International Nuclear Information System (INIS)

    Hinton, F.L.; Chu, M.S.; Dominguez, R.R.

    1985-01-01

    A theoretical picture of the H-mode is proposed which explains some of the most important features of this good confinement mode in neutral beam heated plasmas with divertors. From consideration of the transport through the separatrix and along the open field lines outside the separatrix, as well as the stability of the plasma inside the separatrix, we show that a bifurcation in the operating parameters is possible. At high edge temperatures, very large particle confinement times are possible because of the Ware pinch. The transport of particles and heat along the open field lines to the divertor region depends on temperature in a non-monotonic way, and the bifurcation of the thermal equilibrium which is implied may correspond to the L- to H-mode transition. The improvement of the interior confinement in the H-mode, when the edge temperature is higher, is shown to follow from the tearing mode stability properties of current profiles with pedestals. (author)

  6. 'Snowflake' H Mode in a Tokamak Plasma

    International Nuclear Information System (INIS)

    Piras, F.; Coda, S.; Duval, B. P.; Labit, B.; Marki, J.; Moret, J.-M.; Pitzschke, A.; Sauter, O.; Medvedev, S. Yu.

    2010-01-01

    An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a 'snowflake' (second-order null) divertor configuration for the first time in the TCV tokamak. This regime exhibits 2 to 3 times lower ELM frequency and 20%-30% increased normalized ELM energy (ΔW ELM /W p ) compared to an identically shaped, conventional single-null diverted H mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. The capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally.

  7. Offset linear scaling for H-mode confinement

    International Nuclear Information System (INIS)

    Miura, Yukitoshi; Tamai, Hiroshi; Suzuki, Norio; Mori, Masahiro; Matsuda, Toshiaki; Maeda, Hikosuke; Takizuka, Tomonori; Itoh, Sanae; Itoh, Kimitaka.

    1992-01-01

    An offset linear scaling for the H-mode confinement time is examined based on single parameter scans on the JFT-2M experiment. Regression study is done for various devices with open divertor configuration such as JET, DIII-D, JFT-2M. The scaling law of the thermal energy is given in the MKSA unit as W th =0.0046R 1.9 I P 1.1 B T 0.91 √A+2.9x10 -8 I P 1.0 R 0.87 P√AP, where R is the major radius, I P is the plasma current, B T is the toroidal magnetic field, A is the average mass number of plasma and neutral beam particles, and P is the heating power. This fitting has a similar root mean square error (RMSE) compared to the power law scaling. The result is also compared with the H-mode in other configurations. The W th of closed divertor H-mode on ASDEX shows a little better values than that of open divertor H-mode. (author)

  8. Measurements with magnetic field in the National Spherical Torus Experiment using the motional Stark effect with laser induced fluorescence diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Foley, E. L.; Levinton, F. M. [Nova Photonics, Inc., Princeton, New Jersey 08540 (United States)

    2013-04-15

    The motional Stark effect with laser-induced fluorescence diagnostic (MSE-LIF) has been installed and tested on the National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Lab. The MSE-LIF diagnostic will be capable of measuring radially resolved profiles of magnetic field magnitude or pitch angle in NSTX plasmas. The system includes a diagnostic neutral hydrogen beam and a laser which excites the n = 2 to n = 3 transition. A viewing system has been implemented which will support up to 38 channels from the plasma edge to past the magnetic axis. First measurements of MSE-LIF signals in the presence of small applied magnetic fields in neutral gas are reported.

  9. Measurements with magnetic field in the National Spherical Torus Experiment using the motional Stark effect with laser induced fluorescence diagnostic

    Science.gov (United States)

    Foley, E. L.; Levinton, F. M.

    2013-04-01

    The motional Stark effect with laser-induced fluorescence diagnostic (MSE-LIF) has been installed and tested on the National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Lab. The MSE-LIF diagnostic will be capable of measuring radially resolved profiles of magnetic field magnitude or pitch angle in NSTX plasmas. The system includes a diagnostic neutral hydrogen beam and a laser which excites the n = 2 to n = 3 transition. A viewing system has been implemented which will support up to 38 channels from the plasma edge to past the magnetic axis. First measurements of MSE-LIF signals in the presence of small applied magnetic fields in neutral gas are reported.

  10. Design of interferometer system for Keda Torus eXperiment using terahertz solid-state diode sources

    International Nuclear Information System (INIS)

    Xie, Jinlin; Wang, Haibo; Li, Hong; Lan, Tao; Liu, Adi; Liu, Wandong; Yu, Changxuan; Ding, Weixing

    2014-01-01

    A solid-state source based terahertz (THz) interferometer diagnostic system has been designed and characterized for the Keda Torus eXperiment (KTX). The THz interferometer utilizes the planar diodes based frequency multiplier (X48) to provide the probing beam at fixed frequency 0.650 THz, and local oscillator is provided by an independent solid-state diode source with tunable frequency (0.650 THz +/− 10 MHz). Both solid-state sources have approximately 1 mW power. The planar-diode mixers optimized for high sensitivity, ∼750 mV/mW, are used in the heterodyne detection system, which permits multichannel interferometer on KTX with a low phase noise. A sensitivity of e l> min = 4.5 × 10 16 m −2 and a temporal resolution of 0.2 μs have been achieved during the initial bench test

  11. Compact and lightweight support platform with electromagnetic disturbance elimination for interferometer on reversed field pinch Keda Torus eXperiment

    Science.gov (United States)

    Mao, Wenzhe; Yuan, Peng; Zheng, Jian; Ding, Weixing; Li, Hong; Lan, Tao; Liu, Adi; Liu, Wandong; Xie, Jinlin

    2016-11-01

    A compact and lightweight support platform has been used as a holder for the interferometer system on the Keda Torus eXperiment (KTX), which is a reversed field pinch device. The vibration caused by the interaction between the time-varying magnetic field and the induced current driven in the metal optical components has been measured and, following comparison with the mechanical vibration of the KTX device and the refraction effect of the ambient turbulent air flow, has been identified as the primary vibration source in this case. To eliminate this electromagnetic disturbance, nonmetallic epoxy resin has been selected as the material for the support platform and the commercially available metal optical mounts are replaced. Following these optimization steps and mechanical reinforcements, the stability of the interferometer platform has improved significantly. The phase shift caused by the vibration has been reduced to the level of background noise.

  12. Experimental demonstration of tokamak inductive flux saving by transient coaxial helicity injection on national spherical torus experiment

    Energy Technology Data Exchange (ETDEWEB)

    Raman, R.; Jarboe, T. R.; Nelson, B. A. [University of Washington, Seattle, Washington 98195 (United States); Mueller, D.; Bell, M. G.; Gerhardt, S.; LeBlanc, B.; Menard, J.; Ono, M.; Roquemore, L. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2011-09-15

    Discharges initiated by transient coaxial helicity injection in National Spherical Torus Experiment have attained peak toroidal plasma currents up to 300 kA. When induction from the central solenoid is then applied, these discharges develop up to 300 kA additional current compared to discharges initiated by induction only. CHI initiated discharges in NSTX have achieved 1 MA of plasma current using only 258 mWb of solenoid flux whereas standard induction-only discharges require about 50% more solenoid flux to reach 1 MA. In addition, the CHI-initiated discharge has lower plasma density and a low normalized internal plasma inductance of 0.35, as needed for achieving advanced scenarios in NSTX.

  13. Suppressing electron turbulence and triggering internal transport barriers with reversed magnetic shear in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, J. L. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Bell, R.; Guttenfelder, W.; Hammett, G. W.; Kaye, S. M.; LeBlanc, B.; Mikkelsen, D. R. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Candy, J. [General Atomics, San Diego, California 92186 (United States); Smith, D. R. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Yuh, H. Y. [Nova Photonics Inc., Princeton, New Jersey 08540 (United States)

    2012-05-15

    The National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)] can achieve high electron plasma confinement regimes that are super-critically unstable to the electron temperature gradient driven (ETG) instability. These plasmas, dubbed electron internal transport barriers (e-ITBs), occur when the magnetic shear becomes strongly negative. Using the gyrokinetic code GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)], the first nonlinear ETG simulations of NSTX e-ITB plasmas reinforce this observation. Local simulations identify a strongly upshifted nonlinear critical gradient for thermal transport that depends on magnetic shear. Global simulations show e-ITB formation can occur when the magnetic shear becomes strongly negative. While the ETG-driven thermal flux at the outer edge of the barrier is large enough to be experimentally relevant, the turbulence cannot propagate past the barrier into the plasma interior.

  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. Comparison of Poloidal Velocity Meassurements to Neoclassical Theory on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Bell, R.E.; Andre, R.; Kaye, S.M.; Kolesnikov, R.A.; LeBlance, B.P.; Rewolldt, G.; Wang, W.X.; Sabbagh, S.A.

    2010-01-01

    Knowledge of poloidal velocity is necessary for the determination of the radial electric field, Er, which along with its gradient is linked to turbulence suppression and transport barrier formation. Recent measurements of poloidal flow on conventional tokamaks have been reported to be an order of magnitude larger than expected from neoclassical theory. In contrast, recent poloidal velocity measurements on the NSTX spherical torus (S. M. Kaye et al., Phys. Plasmas 8, 1977 (2001)) are near or below neoclassical estimates. A novel charge exchange recombination spectroscopy diagnostic is used, which features active and passive sets of up/down symmetric views to produce line-integrated poloidal velocity measurements that do not need atomic physics corrections. Local profiles are obtained with an inversion. Poloidal velocity measurements are compared with neoclassical values computed with the codes NCLASS (W. A. Houlberg et al., Phys. Plasmas 4, 3230 (1997)) and GTC-Neo (W. X. Wang, et al., Phys. Plasmas 13, 082501 (2006)), which has been updated to handle impurities.

  16. Comparative study of the electron density profiles in the compact torus plasma merging experiments

    International Nuclear Information System (INIS)

    Hayashiya, Hitoshi; Asaka, Takeo; Katsurai, Makoto

    2003-01-01

    Following two previous papers on the comparative studies of the electron density distributions for a single compact torus (CT) and a spherical tokamak (ST), and for the a single ST and a merged ST, a comparative study on the dynamics of the electron density profile and after the CT and ST plasma merging process was performed. The sharpness of the peak in the electron density profile around the mid-plane just after the merging of CT with a low safety factor (q value) such as RFP or spheromak is found to be related to the speed of the magnetic axis during the plasma merging process. It is also found that the electron density gradient near the plasma edge in a high q ST is larger than that of a low q CT. High q ST is found to be provided with the magnetic structure which is able to sustain a large thermal pressure by a strong j x B force. Despite these differences in the electron density profile between CT and ST during merging, the confinement characteristics evaluated from the number of electrons confined within the magnetic separatrix after the completion of the merging is almost similar between in the merging CT and in the merging ST. For all configurations, the electron density profiles after the completion of the merging are analogous to those of the corresponding single configuration produced without the merging process. (author)

  17. 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

  18. H-modes studies in PDX

    International Nuclear Information System (INIS)

    Fonck, R.J.; Beirsdorfer, P.; Bell, M.

    1984-07-01

    A regime of enhanced energy confinement during neutral beam heating has been obtained routinely in the PDX tokamak after modifications to form a closed divertor geometry. Plasma density profiles were broad and the electron temperature at the plasma edge reached values of approx. 400 eV in the H-mode phase of a discharge. A comparison of closed divertor discharges with moderate and intense gas puffing indicates that a requirement for obtaining high confinement times is the localization of the plasma fueling source in the divertor throat region. While high confinement was attained at moderate injected powers (P/sub INJ/ less than or equal to 3 MW), confinement was degraded at higher powers due to both increased edge instabilities and, especially, the intense gas puffing needed to prevent disruptions. Initial results with a particle scoop limiter indicate high particle confinement times and energy confinement times approaching those of diverted H-mode plasmas

  19. ELM Dynamics in TCV H-modes

    Science.gov (United States)

    Degeling, A. W.; Martin, Y. R.; Lister, J. B.; Llobet, X.; Bak, P. E.

    2003-06-01

    TCV (Tokamak à Configuration Variable, R = 0.88 m, a limited and diverted plasmas, with the primary aim of investigating the effects of plasma shape and current profile on tokamak physics and performance. L-mode to H-mode transitions are regularly obtained in TCV over a wide range of configurations. Under most conditions, the H-mode is ELM-free and terminates in a high density disruption. The conditions required for a transition to an ELMy H-mode were investigated in detail, and a reliable gateway in parameter space for the transition was identified. Once established, the ELMy H-mode is robust to changes in plasma current, elongation, divertor geometry and plasma density over ranges that are much wider than the size of the gateway in these parameters. There exists marked irregularity in the time interval between consecutive ELMs. Transient signatures in the time-series revealing the existence of an underlying chaotic dynamical system are repeatedly observed in a sizable group of discharges [1]. The properties of these signatures (called unstable periodic orbits, or UPOs) are found to vary systematically with parameters such as the plasma current, density and inner plasma — wall gap. A link has also been established between the dynamics of ELMs and sawteeth in TCV: under certain conditions a clear preference is observed in the phase between ELMs and sawtooth crashes, and the ratio of the ELM frequency (felm) to sawtooth frequency (fst) is found to prefer simple rational values (e.g. 1/1, 2/1 or 1/2). An attempt to control the ELM dynamics was made by applying a perturbation signal to the radial field coils used for vertical stabilisation. Phase synchronisation was found with the external perturbation, and felm was found to track limited scans in the driver frequency about the unperturbed value, albeit with intermittent losses in phase lock.

  20. ELM Dynamics in TCV H-modes

    International Nuclear Information System (INIS)

    Degeling, A.W.; Martin, Y.R.; Lister, J.B.; Llobet, X.; Bak, P.E.

    2003-01-01

    TCV (Tokamak a Configuration Variable, R = 0.88 m, a < 0.25 m, BT < 1.54 T) is a highly elongated tokamak, capable of producing limited and diverted plasmas, with the primary aim of investigating the effects of plasma shape and current profile on tokamak physics and performance. L-mode to H-mode transitions are regularly obtained in TCV over a wide range of configurations. Under most conditions, the H-mode is ELM-free and terminates in a high density disruption. The conditions required for a transition to an ELMy H-mode were investigated in detail, and a reliable gateway in parameter space for the transition was identified. Once established, the ELMy H-mode is robust to changes in plasma current, elongation, divertor geometry and plasma density over ranges that are much wider than the size of the gateway in these parameters. There exists marked irregularity in the time interval between consecutive ELMs. Transient signatures in the time-series revealing the existence of an underlying chaotic dynamical system are repeatedly observed in a sizable group of discharges [1]. The properties of these signatures (called unstable periodic orbits, or UPOs) are found to vary systematically with parameters such as the plasma current, density and inner plasma -- wall gap. A link has also been established between the dynamics of ELMs and sawteeth in TCV: under certain conditions a clear preference is observed in the phase between ELMs and sawtooth crashes, and the ratio of the ELM frequency (felm) to sawtooth frequency (fst) is found to prefer simple rational values (e.g. 1/1, 2/1 or 1/2). An attempt to control the ELM dynamics was made by applying a perturbation signal to the radial field coils used for vertical stabilisation. Phase synchronisation was found with the external perturbation, and felm was found to track limited scans in the driver frequency about the unperturbed value, albeit with intermittent losses in phase lock

  1. Weak effect of ion cyclotron acceleration on rapidly chirping beam-driven instabilities in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Heidbrink, W W; Ruskov, E; Fredrickson, E D; Gorelenkov, N; Medley, S S; Berk, H L; Harvey, R W

    2006-01-01

    The fast-ion distribution function in the National Spherical Torus Experiment 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 compressional Alfven eigenmodes, toroidicity-induced Alfven eigenmodes (TAE), 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) high harmonic fast wave (HHFW) heating accelerates the fast ions in an attempt to alter the nonlinear dynamics. Steady-frequency TAE modes diminish during the HHFW heating but there is little evidence that frequency chirping is suppressed

  2. Weak effect of ion cyclotron acceleration on rapidly chirping beam-driven instabilities in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Heidbrink, W W [University of California, Irvine, California (United States); Ruskov, E [University of California, Irvine, California (United States); Fredrickson, E D [Princeton Plasma Physics Laboratory, Princeton, New Jersey (United States); Gorelenkov, N [Princeton Plasma Physics Laboratory, Princeton, New Jersey (United States); Medley, S S [Princeton Plasma Physics Laboratory, Princeton, New Jersey (United States); Berk, H L [University of Texas, Austin, Texas (United States); Harvey, R W [CompX, Del Mar, California (United States)

    2006-09-15

    The fast-ion distribution function in the National Spherical Torus Experiment is modified from shot to shot while keeping the total injected power at {approx}2 MW. Deuterium beams of different energy and tangency radius are injected into helium L-mode plasmas, producing a rich set of instabilities, including compressional Alfven eigenmodes, toroidicity-induced Alfven eigenmodes (TAE), 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 ({approx}<3 MW) high harmonic fast wave (HHFW) heating accelerates the fast ions in an attempt to alter the nonlinear dynamics. Steady-frequency TAE modes diminish during the HHFW heating but there is little evidence that frequency chirping is suppressed.

  3. Weak effect of ion cyclotron acceleration on rapidly chirping beam-driven instabilities in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    W W,Heidbrink; E,Ruskov; E D,Fredrickson; N,Gorelenkov; S S,Medley; H L,Berk; R W,Harvey

    2006-09-01

    The fast-ion distribution function in the National Spherical Torus Experiment 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 compressional Alfven eigenmodes, toroidicity-induced Alfven eigenmodes (TAE), 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 (≤3MW) high harmonic fast wave (HHFW) heating accelerates the fast ions in an attempt to alter the nonlinear dynamics. Steady-frequency TAE modes diminish during the HHFW heating but there is little evidence that frequency chirping is suppressed.

  4. Investigation of self-organized criticality behavior of edge plasma transport in Torus experiment of technology oriented research

    International Nuclear Information System (INIS)

    Xu, Y.H.; Jachmich, S.; Weynants, R.R.; Huber, A.; Unterberg, B.; Samm, U.

    2004-01-01

    The self-organized criticality (SOC) behavior of the edge plasma transport has been studied using fluctuation data measured in the plasma edge and the scrape-off layer of Torus experiment of technology oriented research tokamak [H. Soltwisch et al., Plasma Phys. Controlled Fusion 26, 23 (1984)] before and during the edge biasing experiments. In the 'nonshear' discharge phase before biasing, the fluctuation data clearly show some of the characteristics associated with SOC, including similar frequency spectra to those obtained in 'sandpile' transport and other SOC systems, slowly decaying long tails in the autocorrelation function, values of Hurst parameters larger than 0.5 at all the detected radial locations, and a radial propagation of avalanchelike events in the edge plasma area. During the edge biasing phase, with the generation of an edge radial electric field E r and thus of E r xB flow shear, contrary to theoretical expectation, the Hurst parameters are substantially enhanced in the negative flow shear region and in the scrape-off layer as well. Concomitantly, it is found that the local turbulence is well decorrelated by the E r xB velocity shear, consistent with theoretical predictions

  5. Divertor heat flux mitigation in the National Spherical Torus Experimenta)

    Science.gov (United States)

    Soukhanovskii, V. A.; Maingi, R.; Gates, D. A.; Menard, J. E.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Bell, M. G.; Bell, R. E.; Boedo, J. A.; Bush, C. E.; Kaita, R.; Kugel, H. W.; Leblanc, B. P.; Mueller, D.; NSTX Team

    2009-02-01

    Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

  6. ELMO Bumpy Torus

    International Nuclear Information System (INIS)

    Berry, L.A.; Hedrick, C.L.; Uckan, N.A.

    1979-03-01

    The ELMO Bumpy Torus (EBT) program of experiment, theory, and reactor studies has been a remarkably successful one. In the five years since EBT-I began operating, work has progressed from a demonstration of macrostability to an increasingly detailed understanding of transport properties. Collisionless scaling (tau/sub E/ increases with temperature) has been observed, and the magnitude of the energy confinement time is consistent with neoclassical theory. Experiments on EBT-S (for scale) are now being conducted at the increased magnetic field levels and higher microwave power and frequency made possible by a 28-GHz gyrotron development program. A review of the program is given

  7. Compact and multi-view solid state neutral particle analyzer arrays on National Spherical Torus Experiment-Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Liu, D., E-mail: deyongl@uci.edu; Heidbrink, W. W.; Hao, G. Z.; Zhu, Y. B. [Departments of Physics and Astronomy, University of California, Irvine, California 92697 (United States); Tritz, K. [Departments of Physics and Astronomy, Johns Hopkins University, Baltimore, Maryland 21218 (United States); Fredrickson, E. D. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2016-11-15

    A compact and multi-view solid state neutral particle analyzer (SSNPA) diagnostic based on silicon photodiode arrays has been successfully tested on the National Spherical Torus Experiment-Upgrade. The SSNPA diagnostic provides spatially, temporally, and pitch-angle resolved measurements of fast-ion distribution by detecting fast neutral flux resulting from the charge exchange (CX) reactions. The system consists of three 16-channel subsystems: t-SSNPA viewing the plasma mid-radius and neutral beam (NB) line #2 tangentially, r-SSNPA viewing the plasma core and NB line #1 radially, and p-SSNPA with no intersection with any NB lines. Due to the setup geometry, the active CX signals of t-SSNPA and r-SSNPA are mainly sensitive to passing and trapped particles, respectively. In addition, both t-SSNPA and r-SSNPA utilize three vertically stacked arrays with different filter thicknesses to obtain coarse energy information. The experimental data show that all channels are operational. The signal to noise ratio is typically larger than 10, and the main noise is x-ray induced signal. The active and passive CX signals are clearly observed on t-SSNPA and r-SSNPA during NB modulation. The SSNPA data also indicate significant losses of passing particles during sawteeth, while trapped particles are weakly affected. Fluctuations up to 120 kHz have been observed on SSNPA, and they are strongly correlated with magnetohydrodynamics instabilities.

  8. Charge exchange measurements of MHD activity during neutral beam injection in the Princeton Large Torus and the Poloidal Divertor Experiment

    International Nuclear Information System (INIS)

    Goldston, R.J.; Kaita, R.; Beiersdorfer, P.; Gammel, G.; Herndon, D.L.; McCune, D.C.; Meyerhofer, D.D.

    1987-01-01

    The horizontally scanning, multi-angle charge exchange analysers on the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) were used to study the effects of MHD activity on the background ion distribution function and on the beam ion slowing-down process during high power neutral injection. Sawtooth oscillations were observed in the fast ion flux on PLT and PDX, and measurements with neutral beams providing local neutral density enhancement indicated that the ions were transported radially when these events occurred. With near-perpendicular injection in PDX, at the lower toroidal fields necessary to maximize the plasma beta, repetitive bursts of greatly enhanced charge exchange flux were observed. These were associated with the 'fishbone' MHD instability, and a substantial depletion of the perpendicular slowing-down spectrum below the injection energy was seen. A simple phenomenological model for this loss mechanism was developed, and its use in simulation codes has been successful in providing good agreement with the experimental data. The behaviour and characteristics of this model are well matched by direct theoretical calculations. (author)

  9. High power millimeter wave experiment of torus diamond window prototype for ITER EC H and CD system

    International Nuclear Information System (INIS)

    Takahashi, K.; Kajiwara, K.; Oda, Y.; Sakamoto, K.; Omori, T.; Henderson, M.

    2013-01-01

    Highlights: ► The diamond window prototype was fabricated based on the reliable and manufacturable design. ► Transmission of 740 kW-100 s on the window prototype was successfully demonstrated. ► tan δ = 7.8 × 10 −6 , which was the lowest value that we had ever obtained at JAEA, was evaluated. ► The window structure promising the high power transmission more than 1 MW was obtained. -- Abstract: The design of the torus diamond window for the ITER electron cyclotron heating and current drive (EC H and CD) system has advanced considering a reliable and manufacturable structure. The diamond window prototype was fabricated based on the design and the high power experiment was carried out to verify the millimeter wave transmission capability. Transmission of 740 kW-100 s was demonstrated and no significant temperature increase of the window structure and no damage on the diamond disk were obtained. The temperature saturation of the cooling water for the window was observed and loss tangent of 7.8 × 10 −6 , which was the lowest value that we had ever obtained at JAEA, was evaluated. This result indicates that the diamond window design is feasible and promising the high power more than 1 MW transmission

  10. Profiles of fast ions that are accelerated by high harmonic fast waves in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Liu, D; Heidbrink, W W; Podesta, M; Ruskov, E; Bell, R E; Fredrickson, E D; Medley, S S; Harvey, R W

    2010-01-01

    Combined neutral beam injection and high-harmonic fast-wave (HHFW) heating accelerate deuterium fast ions in the National Spherical Torus Experiment (NSTX). With 1.1 MW of HHFW power, the neutron emission rate is about three times larger than in the comparison discharge without HHFW heating. Acceleration of fast ions above the beam injection energy is evident on an E||B type neutral particle analyzer (NPA), a 4-chord solid state neutral particle analyzer (SSNPA) array and a 16-channel fast-ion D-alpha (FIDA) diagnostic. The accelerated fast ions observed by the NPA and SSNPA diagnostics mainly come from passive charge exchange reactions at the edge due to the NPA/SSNPA localization in phase space. The spatial profile of accelerated fast ions that is measured by the FIDA diagnostic is much broader than in conventional tokamaks because of the multiple resonance layers and large orbits in NSTX. The fast-ion distribution function calculated by the CQL3D Fokker-Planck code differs from the measured spatial profile, presumably because the current version of CQL3D uses a zero-banana-width model. In addition, compressional Alfven eigenmode activity is stronger during the HHFW heating and it may affect the fast-ion spatial profile.

  11. Suppressing Electron Turbulence and Triggering Internal Transport Barriers with Reversed Magnetic Shear in the National Spherical Torus Experiment

    Science.gov (United States)

    Peterson, Jayson Luc

    2011-10-01

    Observations in the National Spherical Torus Experiment (NSTX) have found electron temperature gradients that greatly exceed the linear threshold for the onset for electron temperature gradient-driven (ETG) turbulence. These discharges, deemed electron internal transport barriers (e-ITBs), coincide with a reversal in the shear of the magnetic field and with a reduction in electron-scale density fluctuations, qualitatively consistent with earlier gyrokinetic predictions. To investigate this phenomenon further, we numerically model electron turbulence in NSTX reversed-shear plasmas using the gyrokinetic turbulence code GYRO. These first-of-a-kind nonlinear gyrokinetic simulations of NSTX e-ITBs confirm that reversing the magnetic shear can allow the plasma to reach electron temperature gradients well beyond the critical gradient for the linear onset of instability. This effect is very strong, with the nonlinear threshold for significant transport approaching three times the linear critical gradient in some cases, in contrast with moderate shear cases, which can drive significant ETG turbulence at much lower gradients. In addition to the experimental implications of this upshifted nonlinear critical gradient, we explore the behavior of ETG turbulence during reversed shear discharges. This work is supported by the SciDAC Center for the Study of Plasma Microturbulence, DOE Contract DE-AC02-09CH11466, and used the resources of NCCS at ORNL and NERSC at LBNL. M. Ono et al., Nucl. Fusion 40, 557 (2000).

  12. H-mode pedestal characteristics on MAST

    International Nuclear Information System (INIS)

    Kirk, A; Counsell, G F; Arends, E; Meyer, H; Taylor, D; Valovic, M; Walsh, M; Wilson, H

    2004-01-01

    The H-mode pedestal characteristics on the mega ampere spherical tokamak (MAST) are measured in a variety of disconnected double null discharges and the effect of edge localized modes (ELMs) on the pedestal is presented. The edge density pedestal width in spatial co-ordinates is similar on both the inboard and outboard sides. Neutral penetration may be able to explain the density pedestal width but it alone cannot explain the characteristics of the temperature pedestal. The data from MAST can be used to improve temperature pedestal width scalings by extending the ranges in pedestal collisionality, magnetic field, elongation and aspect ratio studied by other machines. Convective transport is found to dominate energy losses during ELMs and the fractional loss of pedestal energy during an ELM on MAST correlates better with SOL ion transit time than with pedestal collisionality

  13. Application of divertor cryopumping to H-mode density control in DIII-D

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Ferron, J.R.; Hyatt, A.W.

    1993-11-01

    In this paper we describe the method and the results of experiments where a unique in-vessel cryopump-baffle system was used to control density of H-mode plasmas. We were able to independently regulate current and density of ELMing H-mode plasmas, each over a range of factor two, and measure the H-mode confinement scaling with plasma density and current. With a modest pumping speed of ∼40 kl/s, particle exhaust rates as high as 2 x 10 22 atom/s -1 have been observed

  14. 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.

  15. Bumpy torus annulus startup

    International Nuclear Information System (INIS)

    Sperling, J.L.; Hamasaki, S.; Krall, N.A.

    1982-01-01

    In order that a stable bumpy torus plasma configuration can be attained, it is first necessary to irradiate the plasma with sufficient external power to cause annulus formation. To estimate the power required to initiate annuli, it is assumed that quasilinear electron-cyclotron heating by microwaves is the dominant electron heating mechanism. A scaling law for required microwave power is derived which shows that annulus formation is assisted by smaller cross-section areas, lower density, lower microwave frequency, and higher C-mode temperature. The scaling law is quantitatively evaluated for NBT, EBT-1, EBT-S, EBT-P, and EBT-R parameters. The resulting power estimates are consistent with the available microwave power in previous and present experiments. In larger projected bumpy tori, like EBT-P and EBT-R, it may be necessary to initiate annulus formation at densities which are lower than in the stable T-mode so that the necessary microwave power can be reduced to reasonably modest levels. It is suggested that instabilities as well as rf heating can aid the formation of bumpy torus electron annuli. Rf experiments on NBT and EBT-S would be beneficial in determining the capability of rf power to assist annulus startup

  16. Ion orbit loss and pedestal width of H-mode tokamak plasmas in limiter geometry

    International Nuclear Information System (INIS)

    Xiao Xiaotao; Liu Lei; Zhang Xiaodong; Wang Shaojie

    2011-01-01

    A simple analytical model is proposed to analyze the effects of ion orbit loss on the edge radial electric field in a tokamak with limiter configuration. The analytically predicted edge radial electric field is consistent with the H-mode experiments, including the width, the magnitude, and the well-like shape. This model provides an explanation to the H-mode pedestal structure. Scaling of the pedestal width based on this model is proposed.

  17. Phenomenological model for H-mode

    International Nuclear Information System (INIS)

    Ohyabu, N.

    1985-08-01

    A phenomenological model has been developed to clarify the role of the boundary configuration in the heat transport of the H-mode regime. We assume that the dominant mechanism of heat loss at the edge of the plasma is convection and that the diffusion coefficient (D/sub edge/) at the edge of the plasma increases rapidly with plasma pressure, but drops to a low value when the temperature exceeds a certain threshold value. When particle refueling takes place without time delay, as in the case of a limiter discharge, the unfavorable temperature dependence of the D/sub edge/ prohibits even a modest rise of the edge temperature. In a divertor discharge, the particles lost from the closed surface are kept away from the edge region for a time comparable to or longer than the energy transport time in the edge region. Thus, rapid increase in the heat flux allows an excursion of the edge temperature to a higher value thereby reaching the threshold value of the H-transition

  18. H-mode development in TEXT-U limiter plasmas

    International Nuclear Information System (INIS)

    Roberts, D.R.; Bravenec, R.V.; Bengtson, R.D.

    1996-01-01

    H-mode transitions in TEXT-U limiter plasmas have been observed at q a ∼ 3 and I p ∼ 250 kA (P OH ∼ 300 kW) with at least 300 kW of central electron-cyclotron heating (ECH). These are dithering transitions which are induced by sawtooth crashes and display the typical signatures of H-modes (D α drop, spontaneous density increase, evidence of a transport barrier). However, they show only a slight improvement over L-mode energy confinement. The vessel walls are boronized and conditioned prior to experiments to achieve low-impurity influx and particle recycling. Discharges which undergo transitions are fuelled almost entirely on residual recycling. Transitions are observed when limited on a toroidally localized top or bottom limiter and, more often, when the limiter surface is 'fresh', which is achieved by alternating between top and bottom limiters on successive shots. No strong dependence upon the distance from the low-field-side limiter has been found. Transitions are not yet observed when limited on the high-field-side wall tiles or in the case of TEXT-U diverted configurations. Preliminary measurements with the 2 MeV heavy-ion beam probe (HIBP) (in the core) and Langmuir probes (in the edge) indicate that the plasma potential drops outside the q = 1 radius while only small changes are observed in the density fluctuations level. (author)

  19. H-mode and confinement studies in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Suttrop, W.; Ryter, F.; Mertens, V.; Gruber, O.; Murmann, H.; Salzmann, H.; Schweinzer, J.

    2001-01-01

    H-mode operational boundaries and H-mode confinement are investigated on ASDEX Upgrade. The local edge parameter threshold for H-mode holds independent of divertor geometry and changes little with ion mass. The deviation of the H-mode power threshold at densities near the Greenwald limit can be understood as a consequence of a confinement deterioration, caused by 'stiff' temperature profiles and lack of core density gradients in gas puff fuelled discharges. Ion and electron temperature profiles can be described by a lower limit of gradient length L T =T/T'. (author)

  20. ORIGINAL ARTICLE Torus Palatinus and Torus Mandibularis in a ...

    African Journals Online (AJOL)

    Ogunbodede

    ; 28:105-111. 4. Seah, Y. H. Torus Palatinus and. Torus Mandibularis: a Review of the Literature. Aust. Dent. J. 1995;. 40:318-321. 5. Bernal, B. A.; Moreira, D. E.;. Rodriguez, P., I [Prevalence of. Torus Palatinus and Torus. Mandibularis in the ...

  1. Energy confinement in Ohmic H-mode in TUMAN-3M

    International Nuclear Information System (INIS)

    Andrejko, M.V.; Askinazi, L.G.; Golant, V.E.; Kornev, V.A.; Lebedev, S.V.; Levin, L.S.; Tukachinsky, A.S.

    1997-01-01

    The spontaneous transition from Ohmically heated limiter discharges into the regime with improved confinement termed as ''Ohmic H-mode'' has been investigated in ''TUMAN-3''. The typical signatures of H-mode in tokamaks with powerful auxiliary heating have been observed: sharp drop of D α radiation with simultaneous increase in the electron density and stored energy, suppression of the density fluctuations and establishing the steep gradient near the periphery. In 1994 new vacuum vessel had been installed in TUMAN-3 tokamak. The vessel has the same sizes as old one (R 0 =0.55 m, a 1 =0.24 m). New vessel was designed to reduce mechanical stresses in the walls during B T ramp phase of a shot. Therefore modified device - TUMAN-3M is able to produce higher B T and I p , up to 2 T and 0.2 MA respectively. During first experimental run device was operated in Ohmic Regime. In these experiments the possibility to achieve Ohmic H-mode was studied. The study of the parametric dependencies of the energy confinement time in both OH and Ohmic H-mode was performed. In Ohmic H-mode strong dependencies of τ E on plasma current and on input power and weak dependence on density were found. Energy confinement time in TUMAN-3/TUMAN-3M Ohmic H-mode has revealed good agreement with JET/DIII-D/ASDEX scaling for ELM-free H-mode, resulting in very long τ E at the high plasma current discharges. (author)

  2. Origin of the various beta dependences of ELMy H-mode confinement properties

    International Nuclear Information System (INIS)

    Takizuka, T; Urano, H; Takenaga, H; Oyama, N

    2006-01-01

    Dependence of the energy confinement in ELMy H-mode tokamak on the beta has been investigated for a long time, but a common conclusion has not been obtained so far. Recent non-dimensional transport experiments in JT-60U demonstrated clearly the beta degradation. A database for JT-60U ELMy H-mode confinement was assembled. Analysis of this database is carried out, and the strong beta degradation consistent with the above experiments is confirmed. Two subsets of ASDEX Upgrade and JET data in the ITPA H-mode confinement database are analysed to find the origin of the various beta dependences. The shaping of the plasma cross section, as well as the fuelling condition, affects the confinement performance. The beta dependence is not identical for different devices and conditions. The shaping effect, as well as the fuelling effect, is a possible candidate for causing the variation of beta dependence

  3. Phase coherence of parametric-decay modes during high-harmonic fast-wave heating in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, J. A., E-mail: carlsson@pppl.gov [Crow Radio and Plasma Science, Princeton, New Jersey 08540 (United States); Wilson, J. R.; Hosea, J. C.; Greenough, N. L.; Perkins, R. J. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States)

    2016-06-15

    Third-order spectral analysis, in particular, the auto bicoherence, was applied to probe signals from high-harmonic fast-wave heating experiments in the National Spherical Torus Experiment. Strong evidence was found for parametric decay of the 30 MHz radio-frequency (RF) pump wave, with a low-frequency daughter wave at 2.7 MHz, the local majority-ion cyclotron frequency. The primary decay modes have auto bicoherence values around 0.85, very close to the theoretical value of one, which corresponds to total phase coherence with the pump wave. The threshold RF pump power for onset of parametric decay was found to be between 200 kW and 400 kW.

  4. Scaling of the H-mode power threshold for ITER

    International Nuclear Information System (INIS)

    1998-01-01

    Analysis of the latest ITER H-mode threshold database is presented. The power necessary for the transition to H-mode is estimated for ITER, with or without the inclusion of radiation losses from the bulk plasma, in terms of the main engineering variables. The main geometrical variables (aspect ratio ε, elongation κ and average triangularity δ) are also included in the analysis. The H-mode transition is also considered from the point of view of the local edge variables, and the electron temperature at 90% of the poloidal flux is expressed in terms of both local and global variables. (author)

  5. Essential elements of the high density H-mode on W7-AS

    International Nuclear Information System (INIS)

    McCormick, K.; Burhenn, R.; Grigull, P.

    2003-01-01

    The High Density H-Mode (HDH), discovered during the run-in phase of W7-AS divertor operation/1-3/, rapidly became the workhorse of the divertor program, combining optimal core behavior along with edge parameters necessary for successful operation of an Island Divertor. Its unique properties of high energy confinement along with low impurity retention and radiation localized at the edge under ELM-free steady-state conditions at high densities (to 4 x 10 20 m -3 ) and heating powers (to 1.7 MWm -3 ) make the HDH H-mode ideal for a reactor scenario, given it can be extended to higher temperatures in a larger machine. Hence, considerable effort has been invested to understand the nature of the HDH-mode in order to be able to extrapolate to next generation devices. To this end the present paper reports on experiments where two globally-similar ELM-free H-modes are compared: the classic quiescent H-mode H* where both impurity and density control are a severe problem and the HDH-mode with its contrasting steady-state behavior. Through modeling of the temporal behavior of laser-ablated aluminum spectral lines, as well as that of background impurities, it is concluded that a principle difference between the two H-modes is that of enhanced impurity diffusion in the edge gradient region of the HDH-mode. However, no direct indicators of enhanced diffusion have yet been identified. (orig.)

  6. 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)

  7. Torus type thermonuclear device

    International Nuclear Information System (INIS)

    Imura, Yasuya.

    1979-01-01

    Purpose: To attain supporting effect against electromagnetic force and moderate the inner stress applied to toroidal coils due to thermal expansion by intervening a stress relaxation member between the outer circumferential side of a torus and a support device in toroidal coils. Constitution: Toroidal coils for confining a plasma within a torus vacuum container is supported on a support secured to upper and lower bases. A thermoplastic stress relaxation material of a low young's modulus is put between the outer circumferential side of the torus container and the torus outer circumferential side of the support in the toroidal coil. Thermoplastic resin is best suited to the stress relaxation substance, although tetrafluoro resin may be used as the stress relaxation substance while packing non-woven tetron fabric or non-woven glass fabric impregnated with varnish in a gap between the stress relaxation substance and the support or the toroidal coils. (Seki, T.)

  8. Neutral Particle Analyzer Vertically Scanning Measurements of MHD-induced Energetic Ion Redistribution or Loss in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    S.S. Medley, R. Andre, R.E. Bell, D.S. Darrow, C.W. Domier, E.D. Fredrickson, N.N. Gorelenkov, S.M. Kaye, B.P. LeBlanc, K.C. Lee, F.M. Levinton, D. Liu, N.C. Luhmann, Jr., J.E. Menard, H. Park, D. Stutman, A.L. Roquemore, K. Tritz, H. Yuh and the NSTX Team

    2007-11-15

    Observations of magneto-hydro-dynamic (MHD) induced redistribution or loss of energetic ions measured using the vertically scanning capability of the Neutral Particle Analyzer diagnostic on the National Spherical Torus Experiment (NSTX) are presented along with TRANSP and ORBIT code analysis of the results. Although redistribution or loss of energetic ions due to bursting fishbone-like and low-frequency (f ~ 10 kHz) kinktype MHD activity has been reported previously, the primary goal of this work is to study redistribution or loss due to continuous Alfvénic (f ~ 20 – 150 kHz) modes, a topic that heretofore has not been investigated in detail for NSTX plasmas. Initial indications are that the former drive energetic ion loss whereas the continuous Alfvénic modes only cause redistribution and the energetic ions remain confined.

  9. Neutral Particle Analyzer Vertically Scanning Measurements of MHD-induced Energetic Ion Redistribution or Loss in the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Medley, S.S.; Andre, R.; Bell, R.E.; Darrow, D.S.; Domier, C.W.; Fredrickson, E.D.; Gorelenkov, N.N.; Kaye, S.M.; LeBlanc, B.P.; Lee, K.C.; Levinton, F.M.; Liu, D.; Luhmann, N.C. Jr.; Menard, J.E.; Park, H.; Stutman, D.; Roquemore, A.L.; Tritz, K.; Yuh, H

    2007-01-01

    Observations of magneto-hydro-dynamic (MHD) induced redistribution or loss of energetic ions measured using the vertically scanning capability of the Neutral Particle Analyzer diagnostic on the National Spherical Torus Experiment (NSTX) are presented along with TRANSP and ORBIT code analysis of the results. Although redistribution or loss of energetic ions due to bursting fishbone-like and low-frequency (f ∼ 10 kHz) kinktype MHD activity has been reported previously, the primary goal of this work is to study redistribution or loss due to continuous Alfvenic (f ∼ 20-150 kHz) modes, a topic that heretofore has not been investigated in detail for NSTX plasmas. Initial indications are that the former drive energetic ion loss whereas the continuous Alfvenic modes only cause redistribution and the energetic ions remain confined.

  10. ELMO Bumpy Torus

    International Nuclear Information System (INIS)

    1978-01-01

    The ELMO Bumpy Torus (EBT) program of experiment, theory, and reactor studies has been a remarkably successful one. In the five years since EBT-I began operating, work has progressed from demonstrating macrostability to an increasingly detailed understanding of transport properties. Collisionless scaling (tau/sub E/ increases with temperature) has been observed and the magnitude of the energy confinement time is consistent with neoclassical theory. Experiments on EBT-S are now being conducted at the increased magnetic field levels and higher microwave power and frequency made possible by a 28-GHz gyrotron development program. Initial results confirm our assumptions of neoclassical scaling. In conjunction with the experimental advances, EBT theory now has a well-developed transport theory which models the physics which we now think to be important: for example, it yields negative ambipolar electric fields which are consistent with those measured. Stability calculations continue to predict stable equilibrium with β/sub ring/ approx. β/sub core/ approx. 20 to 40%

  11. Comparison between dominant NB and dominant IC heated ELMy H-mode discharges in JET

    NARCIS (Netherlands)

    Versloot, T.W.; Sartori, R.; de Vries, P.C.; et al, [No Value

    2011-01-01

    Abstract The experiment described in this paper is aimed at characterization of ELMy H-mode discharges with varying momentum input, rotation, power deposition profiles and ion to electron heating ratio obtained by varying the proportion between ion cyclotron (IC) and neutral beam (NB) heating. The

  12. Comparison between dominant NB and dominant IC heated ELMy H-mode discharges in JET

    NARCIS (Netherlands)

    Versloot, T. W.; Sartori, R.; Rimini, F.; de Vries, P. C.; Saibene, G.; Parail, V.; Beurskens, M. N. A.; Boboc, A.; Budny, R.; Crombe, K.; de la Luna, E.; Durodie, F.; Eich, T.; Giroud, C.; Kiptily, V.; Johnson, T.; Mantica, P.; Mayoral, M. L.; McDonald, D. C.; Monakhov, I.; Nave, M. F. F.; Voitsekhovitch, I.; Zastrow, K. D.

    2011-01-01

    The experiment described in this paper is aimed at characterization of ELMy H-mode discharges with varying momentum input, rotation, power deposition profiles and ion to electron heating ratio obtained by varying the proportion between ion cyclotron (IC) and neutral beam (NB) heating. The motivation

  13. Electron confinement studies on the EBT-S Bumpy Torus Experiment using soft x-ray techniques

    International Nuclear Information System (INIS)

    Hillis, D.L.; Haste, G.R.; Berry, L.A.

    1983-01-01

    Soft x-ray bremsstrahlung measurements have been performed on the ELMO Bumpy Torus (EBT-S) plasma to determine the electron temperature T/sub e/ and electron density density n/sub e/ using a calibrated Si(Li) detector over a wide range of operating conditions. The purpose of this paper is to outline the necessary assumptions and essential x-ray techniques that are inherent in soft x-ray measurements in order to investigate the electron heating and confinement properties of EBT-S. In addition, by utilizing the electron density as determined by the soft x-ray measurements, the previous EBT-S confinement analyses have been extended. The steady-state plasma of EBT-S is heated by microwaves using a continuous wave (cw) gyrotron that can operate up to power levels of 200 kW. From the soft x-ray measurements, both the electron temperature and density are found to increase at higher microwave power levels. For operation at microwave power levels of 200 kW, T/sub e/ approaches 1 keV while n/sub e/ approaches 1.2 x 10 12 cm - 3 . In general, confinement properties are found to improve with increased microwave power. The data are compared with neoclassical transport scaling and the electron transport is found to be collisionless (#betta#/Ω<1) as well as neoclassical

  14. Improved H-mode access in connected DND in MAST

    International Nuclear Information System (INIS)

    Meyer, H; Carolan, P G; Conway, N J; Counsell, G F; Cunningham, G; Field, A R; Kirk, A; McClements, K G; Price, M; Taylor, D

    2005-01-01

    In the Mega-Amp Spherical Tokamak, MAST, the formation of the edge transport barrier leading to the high-confinement (H-mode) regime is greatly facilitated by operating in a double null diverted (DND) configuration where both X-points are practically on the same flux surface. Ohmic H-modes are presently only obtained in these connected double null diverted (CDND) configurations. The ease of H-mode access is lost if the two flux surfaces passing through the X-points are radially separated by more than one ion Larmor radius (ρ i ∼ 6 mm) at the low-field-side mid-plane. The change of the magnetic configuration from disconnected to CDND is accompanied by a change in the radial electric field of about ΔE ψ ∼ -1 kV m -1 and a reduction of the electron temperature decay length in the high-field-side scrape-off-layer. Other parameters at the plasma edge, in particular those affecting the H-mode access criteria of common L/H transition theories, are not affected by the slight changes to the magnetic configuration. It is believed that the observed change in E ψ , which may result from differences in ion orbit losses, leads to a higher initial E x B flow shear in CDND configurations which could lead to the easier H-mode access

  15. On global H-mode scaling laws for JET

    International Nuclear Information System (INIS)

    Kardaun, O.; Lackner, K.; Thomsen, K.; Christiansen, J.; Cordey, J.; Gottardi, N.; Keilhacker, M.; Smeulders, P.

    1989-01-01

    Investigation of the scaling of the energy confinement time τ E with various plasma parameters has since long been an interesting, albeit not uncontroversial topic in plasma physics. Various global scaling laws have been derived for ohmic as well as (NBI and/or RF heated) L-mode discharges. Due to the scarce availability of computerised, extensive and validated H-mode datasets, systematic statistical analysis of H-mode scaling behaviour has hitherto been limited. A common approach is to fit the available H-mode data by an L-mode scaling law (e.g., Kaye-Goldston, Rebut-Lallia) with one or two adjustable constant terms. In this contribution we will consider the alternative approach of fitting all free parameters of various simple scaling models to two recently compiled datasets consisting of about 140 ELM-free and 40 ELMy H-mode discharges, measured at JET in the period 1986-1988. From this period, approximately all known H-mode shots have been included that satisfy the following criteria: D-injected D + discharges with no RF heating, a sufficiently long (≥300 ms) and regular P NBI flat-top, and validated main diagnostics. (author) 13 refs., 1 tab

  16. Edge transport studies in the edge and scrape-off layer of the National Spherical Torus Experiment with Langmuir probes

    Energy Technology Data Exchange (ETDEWEB)

    Boedo, J. A., E-mail: jboedo@ucsd.edu; Rudakov, D. L. [University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093 (United States); Myra, J. R.; D' Ippolito, D. A. [Lodestar Research Corp, 2400 Central Ave., Boulder, Colorado 80301 (United States); Zweben, S.; Maingi, R.; Maqueda, R. J.; Bell, R.; Kugel, H.; Leblanc, B.; Roquemore, L. A. [Princeton University, PO Box 451, Princeton, New Jersey 08543 (United States); Soukhanovskii, V. A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Ahn, J. W.; Canik, J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, Tennessee 37830 (United States); Crocker, N. [University of California Los Angeles, PO Box 957099, Los Angeles, California 90095 (United States)

    2014-04-15

    Transport and turbulence profiles were directly evaluated using probes for the first time in the edge and scrape-off layer (SOL) of NSTX [Ono et al., Nucl. Fusion 40, 557 (2000)] in low (L) and high (H) confinement, low power (P{sub in}∼ 1.3 MW), beam-heated, lower single-null discharges. Radial turbulent particle fluxes peak near the last closed flux surface (LCFS) at ≈4×10{sup 21} s{sup −1} in L-mode and are suppressed to ≈0.2×10{sup 21} s{sup −1} in H mode (80%–90% lower) mostly due to a reduction in density fluctuation amplitude and of the phase between density and radial velocity fluctuations. The radial particle fluxes are consistent with particle inventory based on SOLPS fluid modeling. A strong intermittent component is identified. Hot, dense plasma filaments 4–10 cm in diameter, appear first ∼2 cm inside the LCFS at a rate of ∼1×10{sup 21} s{sup −1} and leave that region with radial speeds of ∼3–5 km/s, decaying as they travel through the SOL, while voids travel inward toward the core. Profiles of normalized fluctuations feature levels of 10% inside LCFS to ∼150% at the LCFS and SOL. Once properly normalized, the intermittency in NSTX falls in similar electrostatic instability regimes as seen in other devices. The L-H transition causes a drop in the intermittent filaments velocity, amplitude and number in the SOL, resulting in reduced outward transport away from the edge and a less dense SOL.

  17. Fabrication of an alumina torus for thermonuclear fusion containment

    International Nuclear Information System (INIS)

    Hauth, W.E.; Blake, R.D.; Dickinson, J.M.; Rutz, H.L.; Stoddard, S.D.

    1978-05-01

    A 235-cm-diam torus has been fabricated for plasma containment during thermonuclear fusion experiments. This 30-cm-diam torus consists of sixty 99.5%-alumina segments, 80% of which are assembled by forming vacuum-tight ceramic-to-ceramic seals. Selection of sealing materials and techniques are discussed

  18. LASL Compact Torus Program

    International Nuclear Information System (INIS)

    Linford, R.K.; Armstrong, W.T.; Bartsch, R.R.

    1981-01-01

    The Compact Torus (CT) concept includes any axisymmetric toroidal plasma configuration, which does not require the linking of any material through the hole in the torus. Thus, the magnet coils, vacuum vessel, etc., have a simple cylindrical or spherical geometry instead of the toroidal geometry required for Tokamaks and RFP's. This simplified geometry results in substantial engineering advantages in CT reactor embodiments while retaining the good confinement properties afforded by an axisymmetric toroidal plasma-field geometry. CT's can be classified into three major types by using the ion gyro radius rho/sub i/ and the magnitude of the maximum toroidal field B/sub tm/

  19. Suppression of Alfven Modes on the National Spherical Torus Experiment Upgrade with Outboard Beam Injection [Suppression of Alfven Modes on the NSTX-U with Outboard Beam Injection

    International Nuclear Information System (INIS)

    Fredrickson, E. D.; Belova, E. V.; Battaglia, D. J.

    2017-01-01

    In this paper we present data from experiments on the National Spherical Torus Experiment Upgrade, where it is shown for the first time that small amounts of high pitch-angle beam ions can strongly suppress the counterpropagating global Alfven eigenmodes (GAE). GAE have been implicated in the redistribution of fast ions and modification of the electron power balance in previous experiments on NSTX. The ability to predict the stability of Alfven modes, and developing methods to control them, is important for fusion reactors like the International Tokamak Experimental Reactor, which are heated by a large population of nonthermal, super-Alfvenic ions consisting of fusion generated alpha's and beam ions injected for current profile control. We present a qualitative interpretation of these observations using an analytic model of the Doppler-shifted ion-cyclotron resonance drive responsible for GAE instability which has an important dependence on k(perpendicular to rho L). A quantitative analysis of this data with the HYM stability code predicts both the frequencies and instability of the GAE prior to, and suppression of the GAE after the injection of high pitch-angle beam ions.

  20. Transition to H-mode by energetic electrons

    International Nuclear Information System (INIS)

    Itoh, Kimitaka; Itoh, Sanae.

    1992-07-01

    Effect of the electron loss due to the toroidal ripple on an H-mode transition is studied. When energetic electrons exist in tokamaks, e.g., in the case of the current drive by lower hybrid (LH) waves, the edge electric field can show the bifurcation to the more positive value. In this state, both the electron loss and ion loss (such as loss cone loss) are reduced. The criterion for the transition is derived. Comparison with H-mode in JT-60 LH plasma shows a qualitative agreement. (author)

  1. Change of transport at L- and H-mode transition

    International Nuclear Information System (INIS)

    Itoh, Sanae-I; Itoh, Kimitaka.

    1990-01-01

    A new refined model of the L-mode and H-mode transition in tokamaks is presented based on the bifurcation of the radial electric field, E r , near edge. The radial gradient of E r is newly introduced to explain the sudden change of fluctuations as well as plasma fluxes at the onset of transition. This model predicts that the L-to H-mode transition is associated with the decrease of dE r /dr causing reduction of particle and energy fluxes at critical gradient. (author)

  2. Collisional drift waves in the H-mode edge

    International Nuclear Information System (INIS)

    Sen, S.

    1994-01-01

    The stability of the collisional drift wave in a sheared slab geometry is found to be severely restricted at the H-mode edge plasma due to the very steep density gradient. However, a radially varying transverse velocity field is found to play the key role in stability. Velocity profiles usually found in the H-mode plasma stabilize drift waves. On the other hand, velocity profiles corresponding to the L-mode render collisional drift waves unstable even though the magnetic shear continues to play its stabilizing role. (author). 24 refs

  3. Plasma current dependence of the edge pedestal height in JET ELM-free H-modes

    International Nuclear Information System (INIS)

    Nave, M.F.F; Lomas, P.; Gowers, C.; Guo, H.; Hawkes, N.; Huysmans, G.T.A.; Jones, T.; Parail, V.V.; Rimini, F.; Schunke, B.

    2000-01-01

    Some models for the suppression of turbulence in the L to H transition, suggest that the width of the H-mode edge barrier is either proportional or is of the order of the thermal or the fast-ion poloidal Larmor radius. This would require that the width of the edge barrier should depend on the plasma current. This dependence has been clearly verified at JET in experiments designed to control the edge MHD stability of ELM-free hot-ion H-mode plasmas. The effects of isotopic mass and the applicability of several edge barrier models to the hot-ion H-mode plasmas were analysed in (Guo H Y et al 2000 Edge transport barrier in JET hot-ion H-modes Nucl. Fusion 40 69) using a large database containing both deuterium-only and deuterium-tritium plasmas. This database has now been enlarged to include discharges from a plasma shape scan, allowing one to study the dependence of the pedestal height on the edge shear. In addition, the range of plasma currents was extended up to 6 MA. It is shown that the edge data are best described by a model where the edge barrier width is determined by the fast ions weighted towards the components with largest poloidal Larmor radii. However, it is not possible to conclusively eliminate the thermal ion model. (author)

  4. Integration of Microsoft Windows applications with MDSplus data acquisition on the National Spherical Torus Experiment at the Princeton Plasma Physics Laboratory

    International Nuclear Information System (INIS)

    Mastrovito, Dana M.

    2002-01-01

    Data acquisition on the National Spherical Torus Experiment at the Princeton Plasma Physics Laboratory (PPPL) has increasingly involved the use of Personal Computers and specially developed 'turn-key' hardware and software systems to control diagnostics. Interaction with these proprietary software packages is accomplished through use of Visual Basic, or Visual C++ and Component Object Model (COM) technology. COM is a software architecture that allows the components made by different software vendors to be combined into a variety of applications. This technology is particularly well suited to these systems because of its programming language independence, standards for function calling between components, and ability to transparently reference remote processes. COM objects make possible the creation of acquisition software that can control the experimental parameters of both the hardware and software. Synchronization of these applications for diagnostics, such as charged couple device cameras and residual gas analyzers, with the rest of the experiment event cycle at PPPL has been made possible by utilization of the MDSplus libraries for Windows. Instead of transferring large data files to remote disk space, Windows MDSplus events and I/O functions allow us to put raw data into MDSplus directly from interactive data language for Windows and Visual Basic. The combination of COM technology and the MDSplus libraries for Windows provide the tools for many new possibilities in versatile acquisition applications and future diagnostics

  5. Integration of Microsoft Windows Applications with MDSplus Data Acquisition on the National Spherical Torus Experiment at the Princeton Plasma Physics Laboratory; TOPICAL

    International Nuclear Information System (INIS)

    Dana M. Mastrovito

    2002-01-01

    Data acquisition on the National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Laboratory (PPPL) has increasingly involved the use of Personal Computers (PC's) and specially developed ''turn-key'' hardware and software systems to control diagnostics. Interaction with these proprietary software packages is accomplished through use of Visual Basic, or Visual C++ and COM (Component Object Model) technology. COM is a software architecture that allows the components made by different software vendors to be combined into a variety of applications. This technology is particularly well suited to these systems because of its programming language independence, standards for function calling between components, and ability to transparently reference remote processes. COM objects make possible the creation of acquisition software that can control the experimental parameters of both the hardware and software. Synchronization of these applications for diagnostics, such as CCD camer as and residual gas analyzers, with the rest of the experiment event cycle at PPPL has been made possible by utilization of the MDSplus libraries for Windows. Instead of transferring large data files to remote disk space, Windows MDSplus events and I/O functions allow us to put raw data into MDSplus directly from IDL for Windows and Visual Basic. The combination of COM technology and the MDSplus libraries for Windows provide the tools for many new possibilities in versatile acquisition applications and future diagnostics

  6. Integration of Microsoft Windows Applications with MDSplus Data Acquisition on the National Spherical Torus Experiment at the Princeton Plasma Physics Laboratory

    International Nuclear Information System (INIS)

    Dana M. Mastrovito

    2002-03-01

    Data acquisition on the National Spherical Torus Experiment (NSTX) at the Princeton Plasma Physics Laboratory (PPPL) has increasingly involved the use of Personal Computers (PC's) and specially developed ''turn-key'' hardware and software systems to control diagnostics. Interaction with these proprietary software packages is accomplished through use of Visual Basic, or Visual C++ and COM (Component Object Model) technology. COM is a software architecture that allows the components made by different software vendors to be combined into a variety of applications. This technology is particularly well suited to these systems because of its programming language independence, standards for function calling between components, and ability to transparently reference remote processes. COM objects make possible the creation of acquisition software that can control the experimental parameters of both the hardware and software. Synchronization of these applications for diagnostics, such as CCD camer as and residual gas analyzers, with the rest of the experiment event cycle at PPPL has been made possible by utilization of the MDSplus libraries for Windows. Instead of transferring large data files to remote disk space, Windows MDSplus events and I/O functions allow us to put raw data into MDSplus directly from IDL for Windows and Visual Basic. The combination of COM technology and the MDSplus libraries for Windows provide the tools for many new possibilities in versatile acquisition applications and future diagnostics

  7. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the group...

  8. Torus sector handling system

    International Nuclear Information System (INIS)

    Grisham, D.L.

    1981-01-01

    A remote handling system is proposed for moving a torus sector of the accelerator from under the cryostat to a point where it can be handled by a crane and for the reverse process for a new sector. Equipment recommendations are presented, as well as possible alignment schemes. Some general comments about future remote-handling methods and the present capabilities of existing systems will also be included. The specific task to be addressed is the removal and replacement of a 425 to 450 ton torus sector. This requires a horizontal movement of approx. 10 m from a normal operating position to a point where its further transport can be accomplished by more conventional means (crane or floor transporter). The same horizontal movement is required for reinstallation, but a positional tolerance of 2 cm is required to allow reasonable fit-up for the vacuum seal from the radial frames to the torus sector. Since the sectors are not only heavy but rather tall and narrow, the transport system must provide a safe, stable, and repeatable method fo sector movement. This limited study indicates that the LAMPF-based method of transporting torus sectors offers a proven method of moving heavy items. In addition, the present state of the art in remote equipment is adequate for FED maintenance

  9. LH transition theories and theory of H-mode

    International Nuclear Information System (INIS)

    Ward, D.J.

    1996-01-01

    Recent developments in H-mode theory are discussed with earlier work described to put new theories in context. Much of the recent work concerns the development of the radial electric field near the plasma edge and its impact on transport driven by fluctuations, and is the main topic discussed. (author)

  10. Particle transport in JET and TCV-H mode plasmas

    International Nuclear Information System (INIS)

    Maslov, M.

    2009-10-01

    Understanding particle transport physics is of great importance for magnetically confined plasma devices and for the development of thermonuclear fusion power for energy production. From the beginnings of fusion research, more than half a century ago, the problem of heat transport in tokamaks attracted the attention of researchers, but the particle transport phenomena were largely neglected until fairly recently. As tokamak physics advanced to its present level, the physics community realized that there are many hurdles to the development of fusion power beyond the energy confinement. Particle transport is one of the outstanding issues. The aim of this thesis work is to study the anomalous (turbulence driven) particle transport in tokamaks on the basis of experiments on two different devices: JET (Joint European Torus) and TCV (Tokamak à Configuration Variable). In particular the physics of particle inward convection (pinch), which causes formation of peaked density profiles, is addressed in this work. Density profile peaking has a direct, favorable effect on fusion power in a reactor, we therefore also propose an extrapolation to the international experimental reactor ITER, which is currently under construction. To complete the thesis research, a comprehensive experimental database was created on the basis of data collected on JET and TCV during the duration of the thesis. Improvements of the density profile measurements techniques and careful analysis of the experimental data allowed us to derive the dependencies of density profile shape on the relevant plasma parameters. These improved techniques also allowed us to dispel any doubts that had been voiced about previous results. The major conclusions from previous work on JET and other tokamaks were generally confirmed, with some minor supplements. The main novelty of the thesis resides in systematic tests of the predictions of linear gyrokinetic simulations of the ITG (Ion Temperature Gradient) mode against the

  11. Charge-exchange measurements of MHD activity during neutral beam injection in the Princeton Large Torus and the Poloidal Divertor Experiment

    International Nuclear Information System (INIS)

    Goldston, R.J.; Kaita, R.; Beiersdorfer, P.; Gammel, G.; Herndon, D.L.; McCune, D.C.; Meyerhofer, D.D.

    1986-07-01

    The horizontally scanning, multiangle charge-exchange analyzers on the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) were used to study the effects of MHD activity on the background ion distribution function and on the beam ion slowing-down process during high-power neutral injection. Sawtooth oscillations were observed in the fast ion flux on PLT and PDX, and measurements with neutral beams providing local neutral density enhancement indicate that ions are transported radially when these events occur. With near-perpendicular injection in PDX, at the lower toroidal fields necessary to maximize beta, rapid, repetitive bursts of greatly enhanced charge-exchange flux were observed. These are associated with the ''fishbone'' MHD instability, and a substantial depletion of the perpendicular slowing-down spectrum below the injection energy was seen. A simple phenomenological model for this loss mechanism was developed, and its use in simulation codes has been successful in providing good agreement with the data. The behavior and characteristics of this model are well matched by the direct theoretical calculations

  12. 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

  13. ROLE OF NEUTRALS IN CORE FUELING AND PEDESTAL STRUCTURE IN H-MODE DIII-D DISCHARGES

    International Nuclear Information System (INIS)

    WOLF, NS; PETRIE, TW; PORTER, GD; ROGNLIEN, TD; GROEBNER, RJ; MAKOWSKI, MA

    2002-01-01

    OAK A271 ROLE OF NEUTRALS IN CORE FUELING AND PEDESTAL STRUCTURE IN H-MODE DIII-D DISCHARGES. The 2-D fluid code UEDGE was used to analyze DIII-D experiments to determine the role of neutrals in core fueling, core impurities, and also the H-mode pedestal structure. The authors compared the effects of divertor closure on the fueling rate and impurity density of high-triangularity, H-mode plasmas. UEDGE simulations indicate that the decrease in both deuterium core fueling (∼ 15%-20%) and core carbon density (∼ 15%-30%) with the closed divertor compared to the open divertor configuration is due to greater divertor screening of neutrals. They also compared UEDGE results with a simple analytic model of the H-mode pedestal structure. The model predicts both the width and gradient of the transport barrier in n e as a function of the pedestal density. The more sophisticated UEDGE simulations of H-mode discharges corroborate the simple analytic model, which is consistent with the hypothesis that fueling processes play a role in H-mode transport barrier formation

  14. Exploration of the Super H-mode regime on DIII-D and potential advantages for burning plasma devices

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, W. M., E-mail: solomon@fusion.gat.com; Bortolon, A.; Grierson, B. A.; Nazikian, R.; Poli, F. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Snyder, P. B.; Burrell, K. H.; Garofalo, A. M.; Groebner, R. J.; Leonard, A. W.; Meneghini, O.; Osborne, T. H.; Petty, C. C. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Loarte, A. [ITER Organization, Route de Vinon-sur-Verdon - CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2016-05-15

    A new high pedestal regime (“Super H-mode”) has been predicted and accessed on DIII-D. Super H-mode was first achieved on DIII-D using a quiescent H-mode edge, enabling a smooth trajectory through pedestal parameter space. By exploiting Super H-mode, it has been possible to access high pedestal pressures at high normalized densities. While elimination of Edge localized modes (ELMs) is beneficial for Super H-mode, it may not be a requirement, as recent experiments have maintained high pedestals with ELMs triggered by lithium granule injection. Simulations using TGLF for core transport and the EPED model for the pedestal find that ITER can benefit from the improved performance associated with Super H-mode, with increased values of fusion power and gain possible. Similar studies demonstrate that the Super H-mode pedestal can be advantageous for a steady-state power plant, by providing a path to increasing the bootstrap current while simultaneously reducing the demands on the core physics performance.

  15. Validation of single-fluid and two-fluid magnetohydrodynamic models of the helicity injected torus spheromak experiment with the NIMROD code

    International Nuclear Information System (INIS)

    Akcay, Cihan; Victor, Brian S.; Jarboe, Thomas R.; Kim, Charlson C.

    2013-01-01

    We present a comparison study of 3-D pressureless resistive MHD (rMHD) and 3-D presureless two-fluid MHD models of the Helicity Injected Torus with Steady Inductive helicity injection (HIT-SI). HIT-SI is a current drive experiment that uses two geometrically asymmetric helicity injectors to generate and sustain toroidal plasmas. The comparable size of the collisionless ion skin depth d i to the resistive skin depth predicates the importance of the Hall term for HIT-SI. The simulations are run with NIMROD, an initial-value, 3-D extended MHD code. The modeled plasma density and temperature are assumed uniform and constant. The helicity injectors are modeled as oscillating normal magnetic and parallel electric field boundary conditions. The simulations use parameters that closely match those of the experiment. The simulation output is compared to the formation time, plasma current, and internal and surface magnetic fields. Results of the study indicate 2fl-MHD shows quantitative agreement with the experiment while rMHD only captures the qualitative features. The validity of each model is assessed based on how accurately it reproduces the global quantities as well as the temporal and spatial dependence of the measured magnetic fields. 2fl-MHD produces the current amplification (I tor /I inj ) and formation time τ f demonstrated by HIT-SI with similar internal magnetic fields. rMHD underestimates (I tor /I inj ) and exhibits much a longer τ f . Biorthogonal decomposition (BD), a powerful mathematical tool for reducing large data sets, is employed to quantify how well the simulations reproduce the measured surface magnetic fields without resorting to a probe-by-probe comparison. BD shows that 2fl-MHD captures the dominant surface magnetic structures and the temporal behavior of these features better than rMHD

  16. Torus type thermonuclear device

    International Nuclear Information System (INIS)

    Gomei, Yoshio.

    1982-01-01

    Purpose: To facilitate heat removal at limiters and enable helium discharge without using a diverter by the separate disposition of a main limiter receiving the heat from plasmas and an auxiliary limiter for helium discharge. Constitution: A main limiter for establishing and maintaining torus plasmas and an auxiliary limiter for helium discharge are disposed separately. The auxiliary limiter is disposed between the magnetic plane at the position where the plasmas in the confining region begin to contact the main limiter and the first blanket wall. Thus, a sufficient contact area with the plasmas can be taken for the main limiter disposed to the inside of the torus to thereby avoid excess heat concentration. Further, helium ions transported through a passage along the magnetic plane between the main limiter and the first blanket wall to the exhaust chamber are neutralized and thereafter discharged by the auxiliary limiter. (Moriyama, K.)

  17. Statistical study of TCV disruptivity and H-mode accessibility

    International Nuclear Information System (INIS)

    Martin, Y.; Deschenaux, C.; Lister, J.B.; Pochelon, A.

    1997-01-01

    Optimising tokamak operation consists of finding a path, in a multidimensional parameter space, which leads to the desired plasma characteristics and avoids hazards regions. Typically the desirable regions are the domain where an L-mode to H-mode transition can occur, and then, in the H-mode, where ELMs and the required high density< y can be maintained. The regions to avoid are those with a high rate of disruptivity. On TCV, learning the safe and successful paths is achieved empirically. This will no longer be possible in a machine like ITER, since only a small percentage of disrupted discharges will be tolerable. An a priori knowledge of the hazardous regions in ITER is therefore mandatory. This paper presents the results of a statistical analysis of the occurrence of disruptions in TCV. (author) 4 figs

  18. High temperature L- and H-mode confinement in JET

    International Nuclear Information System (INIS)

    Balet, B.; Boyd, D.A.; Campbell, D.J.

    1990-01-01

    The energy confinement properties of low density, high ion temperature L- and H-mode plasmas are investigated. For L-mode plasmas it is shown that, although the global confinement is independent of density, the energy confinement in the central region is significantly better at low densities than at higher densities. The improved confinement appears to be associated with the steepness of the density gradient. For the H-mode phase, although the confinement at the edge is dramatically improved, which is once again associated with the steep density gradient in the edge region, the central confinement properties are essentially the same as for the standard L-mode. The results are compared in a qualitative manner with the predictions of the ion temperature gradient instability theory and appear to be in disagreement with some aspects of this theory. (author). 13 refs, 15 figs

  19. Overview of H-mode studies in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Baker, D.R,; Allen, S.L.

    1994-01-01

    A major portion of the DIII-D program includes studies of the L-H transition, of the VH-mode, of particle transport and control and of the power-handling capability of a diverter. Significant progress has been made in all of these areas and the purpose of this paper is to summarize the major results obtained during the last two years. An increased understanding of the origin of improved confinement in H-mode and in VH-mode discharges has been obtained, good impurity control has been achieved in several operating scenarios, studies of helium transport provide encouraging results from the point of view of reactor design, an actively pumped diverter chamber has controlled the density in H-mode discharges and a radiative diverter is a promising technique for controlling the heat flux from the main plasma

  20. Theory of anomalous transport in H-mode plasmas

    International Nuclear Information System (INIS)

    Itoh, S.; Itoh, K.; Fukuyama, A.; Yagi, M.

    1993-05-01

    Theory of the anomalous transport is developed, and the unified formula for the L- and H-mode plasmas is presented. The self-sustained ballooning-mode turbulence is solved in the presence of the inhomogeneous radial electric field, E r . Reductions in transport coefficients and the amplitude and decorrelation length of fluctuations due to E r ' are quantitatively analyzed. Combined with the E r -bifurcation model, the thickness of the transport barrier is simultaneously determined. (author)

  1. Behaviour of impurities during the H-mode in JET

    International Nuclear Information System (INIS)

    Gianella, R.; Behringer, K.; Denne, B.; Gottardi, N.; Hellermann, M. von; Morgan, P.D.; Pasini, D.; Stamp, M.F.

    1989-01-01

    In additionally-heated tokamak discharges, the H-mode phases are reported to display, together with a better energy confinement, a longer global containment time for particles. In particular, steep gradients of electron density and temperature are sustained in the outer region of the plasma column. This enhanced performance is observed especially in discharges in which the activity of edge localized modes (ELMs) is low or absent. High confinement and accumulation of metallic impurities, which quickly give raise to terminal disruptions have been described under similar conditions. In JET H-modes very long impurity confinement times are also observed. However the experimental condition is somewhat more favourable since quiescent H-modes are obtained lasting much longer than the energy confinement times and the radiation from metals is generally negligible. The dominant impurities are normally carbon and oxygen, the latter generally accounting for half or more of the power radiated from the bulk plasma. During the X-point operation the effective influx of carbon into the discharge, which is normally in close correlation with that of deuterium, is substantially reduced while the influx of oxygen, whose production mechanisms is believed to be of a chemical nature, does not show significant variations. (author) 5 refs., 4 figs

  2. An emerging understanding of H-mode discharges in tokamaks

    International Nuclear Information System (INIS)

    Groebner, R.J.

    1992-12-01

    A remarkable degree of consistency of experimental results from tokamaks throughout the world has developed with regard to the phenomenology of the transition from L-mode to H-mode confinement in tokamaks. The transition is initiated in a narrow layer at the plasma periphery where density fluctuations are suppressed and steep gradients of temperature and density form in a region with large first and second radial derivatives in the υ E → = (E x B)/B 2 flow velocity. These results are qualitatively consistent with theories which predict suppression of fluctuations by shear or curvature in υE. The required υE flow is generated very rapidly when the magnitude of the heating power or of an externally imposed radial current exceed threshold values and several theoretical models have been developed to explain the observed changes in the υE flow. After the transition occurs, the altered boundary conditions enable the development of improved confinement in the plasma interior on a confinement time scale. The resulting H-mode discharge has typically twice the confinement of L-mode discharges and regimes of further improved confinement have been obtained in some H-mode scenarios

  3. Analysis of the H-mode density limit in the ASDEX upgrade tokamak using bolometry

    Energy Technology Data Exchange (ETDEWEB)

    Bernert, Matthias

    2013-10-23

    The high confinement mode (H-mode) is the operational scenario foreseen for ITER, DEMO and future fusion power plants. At high densities, which are favourable in order to maximize the fusion power, a back transition from the H-mode to the low confinement mode (L-mode) is observed. This H-mode density limit (HDL) occurs at densities on the order of, but below, the Greenwald density. In this thesis, the HDL is revisited in the fully tungsten walled ASDEX Upgrade tokamak (AUG). In AUG discharges, four distinct operational phases were identified in the approach towards the HDL. First, there is a stable H-mode, where the plasma density increases at steady confinement, followed by a degrading H-mode, where the core electron density is fixed and the confinement, expressed as the energy confinement time, reduces. In the third phase, the breakdown of the H-mode and transition to the L-mode, the overall electron density is fixed and the confinement decreases further, leading, finally, to an L-mode, where the density increases again at a steady confinement at typical L-mode values until the disruptive Greenwald limit is reached. These four phases are reproducible, quasi-stable plasma regimes and provide a framework in which the HDL can be further analysed. Radiation losses and several other mechanisms, that were proposed as explanations for the HDL, are ruled out for the current set of AUG experiments with tungsten walls. In addition, a threshold of the radial electric field or of the power flux into the divertor appears to be responsible for the final transition back to L-mode, however, it does not determine the onset of the HDL. The observation of the four phases is explained by the combination of two mechanisms: a fueling limit due to an outward shift of the ionization profile and an additional energy loss channel, which decreases the confinement. The latter is most likely created by an increased radial convective transport at the edge of the plasma. It is shown that the

  4. Analysis of the H-mode density limit in the ASDEX upgrade tokamak using bolometry

    International Nuclear Information System (INIS)

    Bernert, Matthias

    2013-01-01

    The high confinement mode (H-mode) is the operational scenario foreseen for ITER, DEMO and future fusion power plants. At high densities, which are favourable in order to maximize the fusion power, a back transition from the H-mode to the low confinement mode (L-mode) is observed. This H-mode density limit (HDL) occurs at densities on the order of, but below, the Greenwald density. In this thesis, the HDL is revisited in the fully tungsten walled ASDEX Upgrade tokamak (AUG). In AUG discharges, four distinct operational phases were identified in the approach towards the HDL. First, there is a stable H-mode, where the plasma density increases at steady confinement, followed by a degrading H-mode, where the core electron density is fixed and the confinement, expressed as the energy confinement time, reduces. In the third phase, the breakdown of the H-mode and transition to the L-mode, the overall electron density is fixed and the confinement decreases further, leading, finally, to an L-mode, where the density increases again at a steady confinement at typical L-mode values until the disruptive Greenwald limit is reached. These four phases are reproducible, quasi-stable plasma regimes and provide a framework in which the HDL can be further analysed. Radiation losses and several other mechanisms, that were proposed as explanations for the HDL, are ruled out for the current set of AUG experiments with tungsten walls. In addition, a threshold of the radial electric field or of the power flux into the divertor appears to be responsible for the final transition back to L-mode, however, it does not determine the onset of the HDL. The observation of the four phases is explained by the combination of two mechanisms: a fueling limit due to an outward shift of the ionization profile and an additional energy loss channel, which decreases the confinement. The latter is most likely created by an increased radial convective transport at the edge of the plasma. It is shown that the

  5. DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-09-13

    A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

  6. Acceleration of a compact torus

    International Nuclear Information System (INIS)

    Hartmann, C.W.; Eddleman, J.L.; Hammer, J.H.; Kusse, B.

    1987-01-01

    The authors report the first results of a study of acceleration of spheromak-type compact toruses in the RACE experiment (plasma Ring ACceleration Experiment). The RACE apparatus consists of (1) a magnetized, coaxial plasma gun 50 cm long, 35 cm OD, 20 cm ID, (2) 600 cm long coaxial acceleration electrodes 50 cm OD, 20 cm ID, (3) a 250 kJ electrolytic capacitor bank to drive the gun solenoid for initial magnetization, (4) a 200 kJ gun bank, (5) a 260 kJ accelerator bank, and (6) magnetic probes and other diagnostics, and vacuum apparatus. To outer acceleration electrode is an extension, at larger OD, of the gun outer electrode, and the inner acceleration electrode is supported and fed by a coaxial insert in the gun center electrode as shown

  7. Torus Breakdown in Noninvertible Maps

    DEFF Research Database (Denmark)

    Maistrenko, V.; Maistrenko, Yu.; Mosekilde, Erik

    2003-01-01

    We propose a criterion for the destruction of a two-dimensional torus through the formation of an infinite set of cusp points on the closed invariant curves defining the resonance torus. This mechanism is specific to noninvertible maps. The cusp points arise when the tangent to the torus at the p......We propose a criterion for the destruction of a two-dimensional torus through the formation of an infinite set of cusp points on the closed invariant curves defining the resonance torus. This mechanism is specific to noninvertible maps. The cusp points arise when the tangent to the torus...... at the point of intersection with the critical curve L-0 coincides with the eigendirection corresponding to vanishing eigenvalue for the noninvertible map. Further parameter changes lead typically to the generation of loops (self-intersections of the invariant manifolds) followed by the transformation...

  8. Formation and sustainment of a very low aspect ratio tokamak using coaxial helicity injection (the Helicity Injected Torus [HIT] experiment)

    International Nuclear Information System (INIS)

    Jarboe, T.R.; Nelson, B.A.

    1992-01-01

    In the paper we will detail the progress of the HIT experiment construction, including the following components: preliminary data and interpretation; diagnostic systems; vacuum vessel and pumping system; helicity source and power supplies; toroidal field coil and power supply; data acquisition system; collaboration with general atomics, with a brief summary given on each

  9. Plasma current dependence of the edge pedestal height in JET ELM-free H-modes

    International Nuclear Information System (INIS)

    Nave, M.; Lomas, P.; Gowers, C.

    2000-01-01

    Models for the suppression of turbulence in the L to H transition, suggest that the width of the H-mode edge barrier is either proportional or is of the order of the ion poloidal Larmor radius. This would require that the width of the edge barrier should depend on the plasma current. This dependence has been clearly verified at JET in experiments designed to control the edge MHD stability of ELM-free hot-ion H-mode plasmas. The effects of isotopic mass and the applicability of several edge barrier models to the hot-ion H-mode plasmas were analysed in using a large database containing both Deuterium-only (DD) and Deuterium-Tritium (DT) plasmas. This database has now been enlarged to include discharges from a plasma shape scan, allowing to study the dependence of the pedestal height on the edge shear. In addition the range of plasma currents was extended up to 6 MA. It is shown that the edge data is best described by a model where the edge barrier width is determined by the fast ions weighted towards the components with largest poloidal Larmor radii. However, it is not possible to eliminate conclusively the thermal ion model. (author)

  10. Simulation of electron thermal transport in H-mode discharges

    International Nuclear Information System (INIS)

    Rafiq, T.; Pankin, A. Y.; Bateman, G.; Kritz, A. H.; Halpern, F. D.

    2009-01-01

    Electron thermal transport in DIII-D H-mode tokamak plasmas [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] is investigated by comparing predictive simulation results for the evolution of electron temperature profiles with experimental data. The comparison includes the entire profile from the magnetic axis to the bottom of the pedestal. In the simulations, carried out using the automated system for transport analysis (ASTRA) integrated modeling code, different combinations of electron thermal transport models are considered. The combinations include models for electron temperature gradient (ETG) anomalous transport and trapped electron mode (TEM) anomalous transport, as well as a model for paleoclassical transport [J. D. Callen, Nucl. Fusion 45, 1120 (2005)]. It is found that the electromagnetic limit of the Horton ETG model [W. Horton et al., Phys. Fluids 31, 2971 (1988)] provides an important contribution near the magnetic axis, which is a region where the ETG mode in the GLF23 model [R. E. Waltz et al., Phys. Plasmas 4, 2482 (1997)] is below threshold. In simulations of DIII-D discharges, the observed shape of the H-mode edge pedestal is produced when transport associated with the TEM component of the GLF23 model is suppressed and transport given by the paleoclassical model is included. In a study involving 15 DIII-D H-mode discharges, it is found that with a particular combination of electron thermal transport models, the average rms deviation of the predicted electron temperature profile from the experimental profile is reduced to 9% and the offset to -4%.

  11. Elmo Bumpy Torus Reactor

    International Nuclear Information System (INIS)

    McAlees, D.G.; Uckan, N.A.; Lidsky, L.M.

    1976-01-01

    In the Elmo Bumpy Torus Reactor (EBTR) study the feasibility of achieving a fusion power plant based on the EBT confinement concept was evaluated. If the present understanding of the physics can be extrapolated to reactor scale devices the reactor could operate at high beta, high power density, and at steady state. The high aspect ratio of the device eases the accessibility, structural design and remote maintenance problems which are common to low aspect ratio machines. A version of the EBTR reference design described here could be constructed with only minor extrapolations in available technology

  12. H-mode edge rotation in W7-AS

    International Nuclear Information System (INIS)

    Hirsch, M.; Baldzuhn, J.; Ehmler, H.; Grigull, P.; Maassberg, H.; McCormick, K.; Wagner, F.; Wobig, H.

    2005-01-01

    In W7-AS three regimes of improved confinement exist which base on negative radial electric fields at the plasma edge resulting there from ion-root conditions of the ambipolar radial fluxes. Experimental control besides the magnetic configuration is given via the edge density profile i.e. the recycling and fuelling conditions. However, the ordering element seems to be the radial electric field profile (respectively its shear) and its interplay with the gradients of ion temperature and density. At low to medium densities the so called optimum confinement regime occurs with maximum density gradients located well inside the plasma boundary and large negative values of E r extending deep in the bulk plasma. For a large inner fraction of the bulk the ion temperature can be sufficiently high that ion transport conditions already can be explained by neoclassics. This regime delivers maximum values of T i , τ e and n τ e T i . Density gradients located right inside the plasma boundary result in the classical H-mode phenomena reminiscent to other toroidal devices with the capability of an edge layer with nearly complete suppression of turbulence either quasi stationary (in a quiescent H-mode) or intermittently (in between ELMs). At even higher densities and highly collisional plasmas with the maximum of ∇n shifted to or even out of the plasma boundary the High Density H-mode (HDH) opens access to steady state conditions with no measurable impurity accumulation. These improved confinement regimes are accessed and left via significant transitions of the transport properties albeit these transitions occur on rather different timescales. A comprehensive picture of improved edge confinement regimes in W7-AS is drawn based on the assumption that a weak edge bounded transport barrier resulting from the ion root conditions (thus E r <0) is the ground state of the (turbulent) edge plasma and already behaves as a barrier for anomalous transport. On top of that the classical H-mode

  13. Local Physics Basis of Confinement Degradation in JET ELMy H-Mode Plasmas and Implications for Tokamak Reactors

    International Nuclear Information System (INIS)

    Budny, R.V.; Alper, B.; Borba, D.; Cordey, J.G.; Ernst, D.R.; Gowers, C.

    2001-01-01

    First results of gyrokinetic analysis of JET [Joint European Torus] ELMy [Edge Localized Modes] H-mode [high-confinement modes] plasmas are presented. ELMy H-mode plasmas form the basis of conservative performance predictions for tokamak reactors of the size of ITER [International Thermonuclear Experimental Reactor]. Relatively high performance for long duration has been achieved and the scaling appears to be favorable. It will be necessary to sustain low Z(subscript eff) and high density for high fusion yield. This paper studies the degradation in confinement and increase in the anomalous heat transport observed in two JET plasmas: one with an intense gas puff and the other with a spontaneous transition between Type I to III ELMs at the heating power threshold. Linear gyrokinetic analysis gives the growth rate, gamma(subscript lin) of the fastest growing modes. The flow-shearing rate omega(subscript ExB) and gamma(subscript lin) are large near the top of the pedestal. Their ratio decreases approximately when the confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently high |gamma(subscript ExB)|/gamma(subscript lin) near the top of the pedestal for high confinement

  14. Low-n magnetohydrodynamic edge instabilities in quiescent H-mode plasmas with a safety-factor plateau

    International Nuclear Information System (INIS)

    Zheng, L.J.; Kotschenreuther, M.T.; Valanju, P.

    2013-01-01

    Low-n magnetohydrodynamic (MHD) modes in the quiescent high confinement mode (H-mode) pedestal are investigated in this paper. Here, n is the toroidal mode number. The low collisionality regime is considered, so that a safety-factor plateau arises in the pedestal region because of the strong bootstrap current. The JET-like (Joint European Torus) equilibria of quiescent H-mode discharges are generated numerically using the VMEC code. The stability of this type of equilibria is analysed using the AEGIS code, with subsonic rotation effects taken into account. The current investigation extends the previous studies of n = 1 modes to n = 2 and 3 modes. The numerical results show that the MHD instabilities in this type of equilibria have characteristic features of the infernal mode. We find that this type of mode tends to prevail when the safety-factor value in the shear-free region is slightly larger than an integer. In this case the frequencies (ω n ) of modes with toroidal mode number n roughly follow the rule ω n ∼ −nΩ p , where Ω p is the local rotation frequency where the infernal harmonic prevails. Since the infernal mode tends to develop near the pedestal top, where pressure driving is strong but magnetic shear stabilization is weak, this local rotation frequency tends to be close to the pedestal top value. These typical mode features bear close resemblance to the edge harmonic oscillations (or outer modes) at the quiescent H-mode discharges observed experimentally. (paper)

  15. Ohmic H-mode and confinement in TCV

    International Nuclear Information System (INIS)

    Moret, J.-M.; Anton, M.; Barry, S.

    1995-01-01

    The unique flexibility of TCV for the creation of a wide variety of plasma shapes has been exploited to address some aspects of tokamak physics for which the shape may play an important role. The electron energy confinement time in limited ohmic L-mode plasmas whose elongation and triangularity have been varied (κ = 1.3 - 1.9, δ 0.1 - 0.7) has been observed to improve with elongation as κ 0.5 but to degrade with triangularity as (1 - 0.8 δ), for fixed safety factor. Ohmic H-modes have been obtained in several diverted and limited configurations, with some of the diverted discharges featuring large ELMs whose effects on the global confinement have been quantified. These effects depend on the configuration: in double null (DN) equilibria, a single ELM expels on average 2%, 6% and 2.5% of the particle, impurity and thermal energy content respectively, whilst in single null (SN) configurations, the corresponding numbers are 3.5%, 7% and 9%, indicative of larger ELM effects. The presence of absence of large ELMs in DN discharges has been actively controlled in a single discharge by alternately forcing one or other of the two X-points to lie on the separatrix, permitting stationary density and impurity content (Z eff ∼ 1.6) in long H-modes (1.5 s). (Author)

  16. Ohmic H-mode and confinement in TCV

    International Nuclear Information System (INIS)

    Moret, J.M.; Anton, M.; Barry, S.

    1995-01-01

    The unique flexibility of TCV for the creation of a wide variety of plasma shapes has been exploited to address some aspects of tokamak physics for which the shape may play an important role. The electron energy confinement time in limited ohmic L-mode plasmas whose elongation and triangularity have been varied, has been observed to improve with elongation as κ 0.5 but to degrade with triangularity as (1-0.8 δ), for fixed safety factor. Ohmic H-modes have been obtained in several diverted and limited configurations, with some of the diverted discharges featuring large ELMs whose effects on the global confinement have been quantified. These effects depend on the configuration: in double null (DN) equilibria, a single ELM expels on average 2%, 6% and 2.5% of the particle, impurity and thermal energy content respectively, whilst in single null (SN) configurations, the corresponding numbers are 3.5%, 7% and 9%, indicative of larger ELM effects. The presence or absence of large ELMs in DN discharges has been actively controlled in a single discharge by alternately forcing one or other of the two X-points to lie on the separatrix, permitting stationary density and impurity content (Z eff ≅1.6) in long H-modes (1.5 s). (author) 9 figs., 9 refs

  17. Bifurcation to Enhanced Performance H-mode on NSTX

    Science.gov (United States)

    Battaglia, D. J.; Chang, C. S.; Gerhardt, S. P.; Kaye, S. M.; Maingi, R.; Smith, D. R.

    2015-11-01

    The bifurcation from H-mode (H98 Performance (EP)H-mode (H98 = 1.2 - 2.0) on NSTX is found to occur when the ion thermal (χi) and momentum transport become decoupled from particle transport, such that the ion temperature (Ti) and rotation pedestals increase independent of the density pedestal. The onset of the EPH-mode transition is found to correlate with decreased pedestal collisionality (ν*ped) and an increased broadening of the density fluctuation (dn/n) spectrum in the pedestal as measured with beam emission spectroscopy. The spectrum broadening at decreased ν*ped is consistent with GEM simulations that indicate the toroidal mode number of the most unstable instability increases as ν*ped decreases. The lowest ν*ped, and thus largest spectrum broadening, is achieved with low pedestal density via lithium wall conditioning and when Zeff in the pedestal is significantly reduced via large edge rotation shear from external 3D fields or a large ELM. Kinetic neoclassical transport calculations (XGC0) confirm that Zeff is reduced when edge rotation braking leads to a more negative Er that shifts the impurity density profiles inward relative to the main ion density. These calculations also describe the role kinetic neoclassical and anomalous transport effects play in the decoupling of energy, momentum and particle transport at the bifurcation to EPH-mode. This work was sponsored by the U.S. Department of Energy.

  18. Transport of impurities during H-mode pulses in JET

    International Nuclear Information System (INIS)

    Giannella, R.; Gottardi, N.; Mompean, F.; Mori, H.; Pasini, D.; Stork, D.; Barnsley, R.; Hawkes, N.C.; Lawson, K.

    1990-01-01

    The transport of impurities during the H-mode is very different from that observed in the other regimes. This is clearly evident in the quiescent discharges where the confinement time of impurities τ I are measured in all the quiescent H-mode discharges in spite of the variety of impurity behavior observed corresponding to different plasma parameters and operating scenarios. The condition of the machine has an influence on the role played by the various impurities, but this does not seem to affect the flow patterns of these ions substantially. In particular oxygen, which was often detected as the dominant radiator, can be reduced to a negligible fraction by He conditioning of the carbon X-point tiles or limiters or by evaporating beryllium in the vacuum vessel. Nevertheless the behaviour of the residual impurities in otherwise similar discharges remains substantially unchanged. The transport patterns appear in fact to be affected by the plasma parameters and their profiles. In particular, two extreme transport regimes are presented in the following. These discharges have been modelled with the aid of a recently developed fully time-dependent impurity transport code using heuristic profiles for the impurity diffusion D and the convection velocity v. (author) 4 refs., 5 figs

  19. Overview of long pulse H-mode operation on EAST

    Science.gov (United States)

    Gong, X.; Garofalo, A. M.; Wan, B.; Li, J.; Qian, J.; Li, E.; Liu, F.; Zhao, Y.; Wang, M.; Xu, H.; EAST Team

    2017-10-01

    The EAST research program aims to demonstrate steady-state long-pulse high-performance H-mode operations with ITER-like poloidal configuration and RF-dominated heating schemes. In the recent experimental campaign, a long pulse fully non-inductive H-mode discharge lasting over 100 seconds using the upper ITER-like tungsten divertor has been achieved in EAST. This scenario used only RF heating and current drive, but also benefitted from an integrated control of the wall conditioning, plasma configuration, divertor heat flux, particle exhaust, impurity management and superconducting coils safety. Maintaining effective coupling of multiple RF heating and current drive sources on EAST is a critical ingredient. This long pulse discharge had good energy confinement, H98,y2 1.1-1.2, and all of the plasma parameters reach a true steady-state. Power balance indicates that the confinement improvement is due partly to a significantly reduced core electron transport inside minor radius rho<0.4. This work was supported by the National Magnetic Confinement Fusion Program of China Contract No. 2015GB10200 and the US Department of Energy Contract No. DE-SC0010685.

  20. The incredible shrinking torus

    International Nuclear Information System (INIS)

    Fischler, W.; Susskind, L.

    1997-01-01

    Using M(atrix) theory, the dualities of toroidally compactified M-theory can be formulated as properties of super Yang Mills theories in various dimensions. We consider the cases of compactification on 1-, 2-, 3-, 4- and 5-dimensional tori. The dualities required by string theory lead to conjectures of remarkable symmetries and relations between field theories as well as extremely unusual dynamical properties. By studying the theories in the limit of vanishingly small tori, a wealth of information is obtained about strongly coupled fixed points of super Yang-Mills theory in various dimensions. Perhaps the most striking behavior, as noted by Rozali in this context, is the emergence of an additional dimension of space in the case of a 4-torus. (orig.)

  1. Torus type thermonuclear device

    International Nuclear Information System (INIS)

    Kitazawa, Hakaru; Saito, Ryusei.

    1981-01-01

    Purpose: To obtain toroidal coil supports structures capable of coping with the changes in the elasticity distribution due to thermal expansion and performing elastic support function corresponding to the distribution of stresses exerted on the toroidal coils, by providing elastic function to the inner circumference side of the coil support structures. Constitution: Support structures for supporting toroidal coils from above and below are formed at the torus inner circumference side thereof with ribs in contact with a central block and having elasticity coefficient corresponding to the distribution of stresses exerted on the toroidal coils, and the stresses exerted on the toroidal coils are elastically supported on the ribs. Accordingly, if the stress distribution varies due to the thermal expansion or the like, adequate supporting function can be obtained well-corresponding to such changes, whereby effective plasma confinement can be attained. (Moriyama, K.)

  2. Ballooning stability analysis of JET H-mode discharges

    International Nuclear Information System (INIS)

    O'Brien, D.P.; Galvao, R.; Keilhacker, M.; Lazzaro, E.; Watkins, M.L.

    1989-01-01

    Previous studies of the stability of a large aspect ratio model equilibrium to ideal MHD ballooning modes have shown that across the bulk of the plasma there exist two marginally stable values of the pressure gradient parameter α. These define an unstable zone which separates the first (small α) stable region from the second (large α) stable region. Close to the separatrix, however, the first and second regions can coalesce when the surface averaged current density, Λ, exceeds a critical value. The plasma in this region is then stable to ballooning modes at all values of the pressure gradient. In this paper we extend these results to JET H-mode equilibria using a finite aspect ratio ballooning formalism, and assess the relevance of ideal ballooning stability in these discharges. In particular we analyse shot 15894 at time 56 sec. which is 1.3 s into the H-phase. (author) 4 refs., 4 figs

  3. 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

  4. H-mode profile parametrization for extrapolation and control

    International Nuclear Information System (INIS)

    Imre, K.; Riedel, K.S.; Schissel, D.P.; Schunke, B.

    1996-01-01

    A steady-state ELMy H-mode profile data set of 68 DIII-D discharges and 74 JET discharges is fitted with an error of 7-8%. The advantages of a parametrization of the plasma profiles in terms of a semi-parametric representation, T(ρ, I p , n-bar, B t , P L , R), are described. The shape of the temperature profile depends almost exclusively upon the size, R and q 95 , with a secondary dependence on the heating power. The density profile depends primarily upon q95 with a secondary dependence on n-bar. The line-average temperature T-bar e scales as n-bar -0.31 instead of T-bar∼''n-bar'' -1.0 . The predicted ITER temperature is T-bar = 17.1 keV. (Author)

  5. Torus knots and mirror symmetry

    CERN Document Server

    Brini, Andrea; Marino, Marcos

    2012-01-01

    We propose a spectral curve describing torus knots and links in the B-model. In particular, the application of the topological recursion to this curve generates all their colored HOMFLY invariants. The curve is obtained by exploiting the full Sl(2, Z) symmetry of the spectral curve of the resolved conifold, and should be regarded as the mirror of the topological D-brane associated to torus knots in the large N Gopakumar-Vafa duality. Moreover, we derive the curve as the large N limit of the matrix model computing torus knot invariants.

  6. Experimental study of the β-limit in ohmic H-mode in the TUMAN-3M tokamak

    International Nuclear Information System (INIS)

    Lebedev, S.V.; Andreiko, M.V.; Askinazi, L.G.; Golant, V.E.; Kornev, V.A.; Krikunov, S.V.; Levin, L.S.; Rozhdestvensky, V.V.; Tukachinsky, A.S.; Yaroshevich, S.P.

    1998-01-01

    Because of its high confinement properties, the H-mode provides good opportunities to achieve high beta values in a tokamak. In this paper the results of an experimental study of β T and β N limits in the H-mode, obtained in a circular cross section tokamak without auxiliary heating are presented. The experiments were performed in the TUMAN-3M tokamak. The device has the following parameters: R 0 =0.53m, a s =0.22m (limiter configuration), B T ≤1.2T, I p ≤175kA, n-bar e ≤6.2x10 19 m -3 . The stored energy was measured using diamagnetic loops and compared with W calculated from kinetic data obtained by Thomson scattering and microwave interferometry. Measurements of the stored energy and of the β were performed in the ohmic H-mode before and after boronization and in the scenario with fast current ramp-down in ohmic H-mode. A maximum value of β T of 2.0% and β N of 2.0 were achieved. The β N limit achieved reveals itself as a 'soft' (non-disruptive) limit. The stored energy slowly decays after the current ramp-down. No correlation was found between beta restriction and MHD phenomena. Internal transport barrier (ITB) formation was observed in ohmic H-mode. An enhancement factor of 2.0 over ITER93H(ELM-free) was found in the ohmic H-mode with ITB. (author)

  7. Next Step Spherical Torus Design Studies

    International Nuclear Information System (INIS)

    Neumeyer, C.; Heitzenroeder, P.; Kessel, C.; Ono, M.; Peng, M.; Schmidt, J.; Woolley, R.; Zatz, I.

    2002-01-01

    Studies are underway to identify and characterize a design point for a Next Step Spherical Torus (NSST) experiment. This would be a ''Proof of Performance'' device which would follow and build upon the successes of the National Spherical Torus Experiment (NSTX) a ''Proof of Principle'' device which has operated at PPPL since 1999. With the Decontamination and Decommissioning (DandD) of the Tokamak Fusion Test Reactor (TFTR) nearly completed, the TFTR test cell and facility will soon be available for a device such as NSST. By utilizing the TFTR test cell, NSST can be constructed for a relatively low cost on a short time scale. In addition, while furthering spherical torus (ST) research, this device could achieve modest fusion power gain for short-pulse lengths, a significant step toward future large burning plasma devices now under discussion in the fusion community. The selected design point is Q=2 at HH=1.4, P subscript ''fusion''=60 MW, 5 second pulse, with R subscript ''0''=1.5 m, A=1.6, I subscript ''p''=10vMA, B subscript ''t''=2.6 T, CS flux=16 weber. Most of the research would be conducted in D-D, with a limited D-T campaign during the last years of the program

  8. Transport simulation of EAST long-pulse H-mode discharge with integrated modeling

    Science.gov (United States)

    Wu, M. Q.; Li, G. Q.; Chen, J. L.; Du, H. F.; Gao, X.; Ren, Q. L.; Li, K.; Chan, Vincent; Pan, C. K.; Ding, S. Y.; Jian, X.; Zhu, X.; Lian, H.; Qian, J. P.; Gong, X. Z.; Zang, Q.; Duan, Y. M.; Liu, H. Q.; Lyu, B.

    2018-04-01

    In the 2017 EAST experimental campaign, a steady-state long-pulse H-mode discharge lasting longer than 100 s has been obtained using only radio frequency heating and current drive, and the confinement quality is slightly better than standard H-mode, H98y2 ~ 1.1, with stationary peaked electron temperature profiles. Integrated modeling of one long-pulse H-mode discharge in the 2016 EAST experimental campaign has been performed with equilibrium code EFIT, and transport codes TGYRO and ONETWO under integrated modeling framework OMFIT. The plasma current is fully-noninductively driven with a combination of ~2.2 MW LHW, ~0.3 MW ECH and ~1.1 MW ICRF. Time evolution of the predicted electron and ion temperature profiles through integrated modeling agree closely with that from measurements. The plasma current (I p ~ 0.45 MA) and electron density are kept constantly. A steady-state is achieved using integrated modeling, and the bootstrap current fraction is ~28%, the RF drive current fraction is ~72%. The predicted current density profile matches the experimental one well. Analysis shows that electron cyclotron heating (ECH) makes large contribution to the plasma confinement when heating in the core region while heating in large radius does smaller improvement, also a more peaked LHW driven current profile is got when heating in the core. Linear analysis shows that the high-k modes instability (electron temperature gradient driven modes) is suppressed in the core region where exists weak electron internal transport barriers. The trapped electron modes dominates in the low-k region, which is mainly responsible for driving the electron energy flux. It is found that the ECH heating effect is very local and not the main cause to sustained the good confinement, the peaked current density profile has the most important effect on plasma confinement improvement. Transport analysis of the long-pulse H-mode experiments on EAST will be helpful to build future experiments.

  9. Particle and power deposition on divertor targets in EAST H-mode plasmas

    International Nuclear Information System (INIS)

    Wang, L.; Xu, G.S.; Guo, H.Y.; Chen, R.; Ding, S.; Gan, K.F.; Gao, X.; Gong, X.Z.; Jiang, M.; Liu, P.; Liu, S.C.; Luo, G.N.; Ming, T.F.; Wan, B.N.; Wang, D.S.; Wang, F.M.; Wang, H.Q.; Wu, Z.W.; Yan, N.; Zhang, L.

    2012-01-01

    The effects of edge-localized modes (ELMs) on divertor particle and heat fluxes were investigated for the first time in the Experimental Advanced Superconducting Tokamak (EAST). The experiments were carried out with both double null and lower single null divertor configurations, and comparisons were made between the H-mode plasmas with lower hybrid current drive (LHCD) and those with combined ion cyclotron resonance heating (ICRH). The particle and heat flux profiles between and during ELMs were obtained from Langmuir triple-probe arrays embedded in the divertor target plates. And isolated ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM-free period. It was demonstrated that ELM-induced radial transport predominantly originated from the low-field side region, in good agreement with the ballooning-like transport model and experimental results of other tokamaks. ELMs significantly enhanced the divertor particle and heat fluxes, without significantly broadening the SOL width and plasma-wetted area on the divertor target in both LHCD and LHCD + ICRH H-modes, thus posing a great challenge for the next-step high-power, long-pulse operation in EAST. Increasing the divertor-wetted area was also observed to reduce the peak heat flux and particle recycling at the divertor target, hence facilitating long-pulse H-mode operation. The particle and heat flux profiles during ELMs appeared to exhibit multiple peak structures, and were analysed in terms of the behaviour of ELM filaments and the flux tubes induced by modified magnetic topology during ELMs. (paper)

  10. Density profile analysis during an ELM event in ASDEX Upgrade H-modes

    International Nuclear Information System (INIS)

    Nunes, I.; Manso, M.; Serra, F.; Horton, L.D.; Conway, G.D.; Loarte, A.

    2005-01-01

    This paper reports results on measurements of the density profiles. Here we analyse the behaviour of the electron density for a set of experiments in type I ELMy H-mode discharges in ASDEX Upgrade where the plasma current, plasma density, triangularity and input power were varied. Detailed measurements of the radial extent of the perturbation on the density profiles caused by the edge localized mode (ELM) crash (ELM affected depth), the velocity of the radial propagation of the perturbation as well as the width and gradient of the density pedestal are determined. The effect of a type I ELM event on the density profiles affects the outermost 20-40% of the plasma minor radius. At the scrape-off layer (SOL) the density profile broadens while in the pedestal region the density decreases resulting in a smaller density gradient. This change in the density profile defines a pivot point around which the density profile changes. The average radial velocity at the SOL is in the range 125-150 ms -1 and approximately constant for all the density layers far from the pivot point. The width of the density pedestal is approximately constant for all the ELMy H-mode discharges analysed, with values between 2 and 3.5 cm. These results are then compared with an analytical model where the width of the density is predominantly set by ionization (neutral penetration model). The width of the density profiles for L-mode discharges is included, since L- and H-mode have different particle transport. No agreement between the experimental results and the model is found

  11. Gyrokinetic Calculations of Microinstabilities and Transport During RF H-Modes on Alcator C-Mod

    International Nuclear Information System (INIS)

    Redi, M.H.; Fiore, C.; Bonoli, P.; Bourdelle, C.; Budny, R.; Dorland, W.D.; Ernst, D.; Hammett, G.; Mikkelsen, D.; Rice, J.; Wukitch, S.

    2002-01-01

    Physics understanding for the experimental improvement of particle and energy confinement is being advanced through massively parallel calculations of microturbulence for simulated plasma conditions. The ultimate goal, an experimentally validated, global, non-local, fully nonlinear calculation of plasma microturbulence is still not within reach, but extraordinary progress has been achieved in understanding microturbulence, driving forces and the plasma response in recent years. In this paper we discuss gyrokinetic simulations of plasma turbulence being carried out to examine a reproducible, H-mode, RF heated experiment on the Alcator CMOD tokamak3, which exhibits an internal transport barrier (ITB). This off axis RF case represents the early phase of a very interesting dual frequency RF experiment, which shows density control with central RF heating later in the discharge. The ITB exhibits steep, spontaneous density peaking: a reduction in particle transport occurring without a central particle source. Since the central temperature is maintained while the central density is increasing, this also suggests a thermal transport barrier exists. TRANSP analysis shows that ceff drops inside the ITB. Sawtooth heat pulse analysis also shows a localized thermal transport barrier. For this ICRF EDA H-mode, the minority resonance is at r/a * 0.5 on the high field side. There is a normal shear profile, with q monotonic

  12. Fusion advanced studies Torus

    International Nuclear Information System (INIS)

    2007-01-01

    The successful development of ITER and DEMO scenarios requires preparatory activities on devices that are smaller than ITER, sufficiently flexible and capable of investigating the peculiar physics of burning plasma conditions. The aim of the Fusion Advanced Studies Torus (FAST) proposal [2.1] (formerly FT3 [2.2]) is to show that the preparation of ITER scenarios and the development of new expertise for the DEMO design and RD can be effectively implemented on a new facility. FAST will a) operate with deuterium plasmas, thereby avoiding problems associated with tritium, and allow investigation of nonlinear dynamics (which are important for understanding alpha particle behaviour in burning plasmas) by using fast ions accelerated by heating and current drive systems; b) work in a dimensionless parameter range close to that of ITER; c) test technical innovative solutions, such as full-tungsten plasma-facing components and an advanced liquid metal divertor target for the first wall/divertor, directly relevant for ITER and DEMO; d) exploit advanced regimes with a much longer pulse duration than the current diffusion time; e) provide a test bed for ITER and DEMO diagnostics; f) provide an ideal framework for model and numerical code benchmarks, their verification and validation in ITER/ DEMO-relevant plasma conditions

  13. The H-mode power threshold in JET

    Energy Technology Data Exchange (ETDEWEB)

    Start, D F.H.; Bhatnagar, V P; Campbell, D J; Cordey, J G; Esch, H P.L. de; Gormezano, C; Hawkes, N; Horton, L; Jones, T T.C.; Lomas, P J; Lowry, C; Righi, E; Rimini, F G; Saibene, G; Sartori, R; Sips, G; Stork, D; Thomas, P; Thomsen, K; Tubbing, B J.D.; Von Hellermann, M; Ward, D J [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    New H-mode threshold data over a range of toroidal field and density values have been obtained from the present campaign. The scaling with n{sub e} B{sub t} is almost identical with that of the 91/92 period for the same discharge conditions. The scaling with toroidal field alone gives somewhat higher thresholds than the older data. The 1991/2 database shows a scaling of P{sub th} (power threshold) with n{sub e} B{sub t} which is approximately linear and agrees well with that observed on other tokamaks. For NBI and carbon target tiles the threshold power is a factor of two higher with the ion {Nu}B drift away from the target compared with the value found with the drift towards the target. The combination of ICRH and beryllium tiles appears to be beneficial for reducing P{sub th}. The power threshold is largely insensitive to plasma current, X-point height and distance between the last closed flux surface and the limiter, at least for values greater than 2 cm. (authors). 3 refs., 6 figs.

  14. Scaling studies of the H-mode pedestal

    International Nuclear Information System (INIS)

    Groebner, R.J.; Osborne, T.H.

    1998-01-01

    The structure and scaling of the H-mode pedestal are examined for discharges in the DIII-D tokamak. For typical conditions, the pedestal values of the ion and electron temperatures T i and T e are comparable. Measurements of main ion and C 6+ profiles indicate that the ion pressure gradient in the barrier is 50%--100% of the electron pressure gradient for deuterium plasmas. The magnitude of the pressure gradient in the barrier often exceeds the predictions of infinite-n ballooning mode theory by a factor of two. Moreover, via the bootstrap current, the finite pressure gradient acts to entirely remove ballooning stability limits for typical discharges. For a large dataset, the width of the pressure barrier δ is best described by the dimensionless scaling δ/R ∝ (β pol ped ) 0.4 where (β pol ped ) is the pedestal value of poloidal beta and R is the major radius. Scalings based on the poloidal ion gyroradius or the edge density gradient do not adequately describe overall trends in the data set and the propagation of the pressure barrier observed between edge-localized modes. The width of the T i barrier is quite variable and is not a good measure of the width of the pressure barrier

  15. H-mode pedestal and threshold studies over an expanded operating space on Alcator C-Moda)

    Science.gov (United States)

    Hubbard, A. E.; Hughes, J. W.; Bespamyatnov, I. O.; Biewer, T.; Cziegler, I.; LaBombard, B.; Lin, Y.; McDermott, R.; Rice, J. E.; Rowan, W. L.; Snipes, J. A.; Terry, J. L.; Wolfe, S. M.; Wukitch, S.

    2007-05-01

    This paper reports on studies of the edge transport barrier and transition threshold of the high confinement (H) mode of operation on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)], over a wide range of toroidal field (2.6-7.86T) and plasma current (0.4-1.7MA). The H-mode power threshold and edge temperature at the transition increase with field. Barrier widths, pressure limits, and confinement are nearly independent of field at constant current, but the operational space at high B shifts toward higher temperature and lower density and collisionality. Experiments with reversed field and current show that scrape-off-layer flows in the high-field side depend primarily on configuration. In configurations with the B ×∇B drift away from the active X-point, these flows lead to more countercurrent core rotation, which apparently contributes to higher H-mode thresholds. In the unfavorable case, edge temperature thresholds are higher, and slow evolution of profiles indicates a reduction in thermal transport prior to the transition in particle confinement. Pedestal temperatures in this case are also higher than in the favorable configuration. Both high-field and reversed-field results suggest that parameters at the L-H transition are influencing the evolution and parameters of the H-mode pedestal.

  16. Pedestal Temperature Model for Type III ELMy H-mode Plasma

    International Nuclear Information System (INIS)

    Buangam, W.; Suwanna, S.; Onjun, T.; Poolyarat, N.; Picha, R.; Singhsomroje, W.

    2009-07-01

    Full text: It is widely known that the improved performance of H-mode plasma results mainly from a formation of the pedestal, which is a narrow region of strong pressure gradient near the edge of plasma. A predictive capability for the conditions at the top of the pedestal is important, especially for predictive simulations of future experiments. New models for predicting the temperature values at the top of the pedestal for type III ELMy H-mode plasma are developed by using two different approaches: a theory-based approaches and an empirical approach. For a theory-based approach, a model is developed based on the calculation of thermal energy in the pedestal region and on accepted scaling laws of energy confinement time. For an empirical model, a scaling law for pedestal temperature in terms of plasma controlled parameters, such as plasma current, magnetic field, heating power, is deduced from experimental data. Predictions from these models are compared with experimental data from the Pedestal International Database. Statistical quantities, such as Root-Mean Square Error (RMSE) and offset values, are computed to quantify the predictive capability of the models. It is found that the theory-based model predicts the pedestal temperature values moderately well yielding RMSE between 30% and 40%. The IPB98(y,3) scaling law yields with best agreement with RMSE of 30.4%. The empirical model predicts the pedestal temperature value with better agreement, yield RMSE of 25.9%

  17. The ELMO Bumpy Torus: present and future

    International Nuclear Information System (INIS)

    Uckan, N.A.

    1978-01-01

    The ELMO Bumpy Torus (EBT) experiment at ORNL was designed to circumvent the MHD stability problem encountered in standard bumpy tori by using the ''minimum-B'' properties of annular high-beta, hot-electron plasmas formed by microwave heating. The EBT combines the attractive features of both mirrors and tokamaks. The feasibility of this combined system has been demonstrated successfully in the EBT-I experiment and discussed at greater length elsewhere. The present status of the EBT research activities at ORNL is described briefly

  18. 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

  19. Compact magnetic confinement fusion: Spherical torus and compact torus

    Directory of Open Access Journals (Sweden)

    Zhe Gao

    2016-05-01

    Full Text Available The spherical torus (ST and compact torus (CT are two kinds of alternative magnetic confinement fusion concepts with compact geometry. The ST is actually a sub-category of tokamak with a low aspect ratio; while the CT is a toroidal magnetic configuration with a simply-connected geometry including spheromak and field reversed pinch. The ST and CT have potential advantages for ultimate fusion reactor; while at present they can also provide unique fusion science and technology contributions for mainstream fusion research. However, some critical scientific and technology issues should be extensively investigated.

  20. Pedestal structure and stability in H-mode and I-mode: a comparative study on Alcator C-Mod

    International Nuclear Information System (INIS)

    Hughes, J.W.; Walk, J.R.; Davis, E.M.; LaBombard, B.; Baek, S.G.; Churchill, R.M.; Greenwald, M.; Hubbard, A.E.; Lipschultz, B.; Marmar, E.S.; Reinke, M.L.; Rice, J.E.; Theiler, C.; Terry, J.; White, A.E.; Whyte, D.G.; Snyder, P.B.; Groebner, R.J.; Osborne, T.; Diallo, A.

    2013-01-01

    New experimental data from the Alcator C-Mod tokamak are used to benchmark predictive modelling of the edge pedestal in various high-confinement regimes, contributing to greater confidence in projection of pedestal height and width in ITER and reactors. ELMy H-modes operate near stability limits for ideal peeling–ballooning modes, as shown by calculations with the ELITE code. Experimental pedestal width in ELMy H-mode scales as the square root of β pol at the pedestal top, i.e. the dependence expected from theory if kinetic ballooning modes (KBMs) were responsible for limiting the pedestal width. A search for KBMs in experiment has revealed a short-wavelength electromagnetic fluctuation in the pedestal that is a candidate driver for inter-edge localized mode (ELM) pedestal regulation. A predictive pedestal model (EPED) has been tested on an extended set of ELMy H-modes from C-Mod, reproducing pedestal height and width reasonably well across the data set, and extending the tested range of EPED to the highest absolute pressures available on any existing tokamak and to within a factor of three of the pedestal pressure targeted for ITER. In addition, C-Mod offers access to two regimes, enhanced D-alpha (EDA) H-mode and I-mode, that have high pedestals, but in which large ELM activity is naturally suppressed and, instead, particle and impurity transport are regulated continuously. Pedestals of EDA H-mode and I-mode discharges are found to be ideal magnetohydrodynamic (MHD) stable with ELITE, consistent with the general absence of ELM activity. Invocation of alternative physics mechanisms may be required to make EPED-like predictions of pedestals in these kinds of intrinsically ELM-suppressed regimes, which would be very beneficial to operation in burning plasma devices. (paper)

  1. Reduction of plasma density in the Helicity Injected Torus with Steady Inductance experiment by using a helicon pre-ionization source

    International Nuclear Information System (INIS)

    Hossack, Aaron C.; Jarboe, Thomas R.; Victor, Brian S.; Firman, Taylor; Prager, James R.; Ziemba, Timothy; Wrobel, Jonathan S.

    2013-01-01

    A helicon based pre-ionization source has been developed and installed on the Helicity Injected Torus with Steady Inductance (HIT-SI) spheromak. The source initiates plasma breakdown by injecting impurity-free, unmagnetized plasma into the HIT-SI confinement volume. Typical helium spheromaks have electron density reduced from (2–3) × 10 19 m −3 to 1 × 10 19 m −3 . Deuterium spheromak formation is possible with density as low as 2 × 10 18 m −3 . The source also enables HIT-SI to be operated with only one helicity injector at injector frequencies above 14.5 kHz. A theory explaining the physical mechanism driving the reduction of breakdown density is presented

  2. Reduction of plasma density in the Helicity Injected Torus with Steady Inductance experiment by using a helicon pre-ionization source

    Energy Technology Data Exchange (ETDEWEB)

    Hossack, Aaron C.; Jarboe, Thomas R.; Victor, Brian S. [Department of Aeronautics and Astronautics, University of Washington, Seattle, Washington 98195 (United States); Firman, Taylor; Prager, James R.; Ziemba, Timothy [Eagle Harbor Technologies, Inc., 119 W. Denny Way, Suite 210, Seattle, Washington 98119 (United States); Wrobel, Jonathan S. [979B West Moorhead Circle, Boulder, Colorado 80305 (United States)

    2013-10-15

    A helicon based pre-ionization source has been developed and installed on the Helicity Injected Torus with Steady Inductance (HIT-SI) spheromak. The source initiates plasma breakdown by injecting impurity-free, unmagnetized plasma into the HIT-SI confinement volume. Typical helium spheromaks have electron density reduced from (2–3) × 10{sup 19} m{sup −3} to 1 × 10{sup 19} m{sup −3}. Deuterium spheromak formation is possible with density as low as 2 × 10{sup 18} m{sup −3}. The source also enables HIT-SI to be operated with only one helicity injector at injector frequencies above 14.5 kHz. A theory explaining the physical mechanism driving the reduction of breakdown density is presented.

  3. Observation of precursor magnetic oscillations to the H-mode transition of ASDEX

    International Nuclear Information System (INIS)

    Toi, K.; Gernhardt, J.; Klueber, O.; Kornherr, M.

    1988-05-01

    Precursor oscillations to the H-mode transition are identified in magnetic fluctuations of the ASDEX H-mode discharges initiated without a sawtooth. This precursor is m=4/n=1 mode, rotating with f ≅ 10 kHz in the opposite direction to co-injected neutral beams. Time behaviour of the amplitude suggests that the H-mode transition is caused, not by the edge electron temperature, but by the edge current density. (orig.)

  4. Bifurcation structure of successive torus doubling

    International Nuclear Information System (INIS)

    Sekikawa, Munehisa; Inaba, Naohiko; Yoshinaga, Tetsuya; Tsubouchi, Takashi

    2006-01-01

    The authors discuss the 'embryology' of successive torus doubling via the bifurcation theory, and assert that the coupled map of a logistic map and a circle map has a structure capable of generating infinite number of torus doublings

  5. Study of H-mode threshold conditions in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Carlstrom, T.N.; Burrell, K.H.

    1996-10-01

    Studies have been conducted in DIII-D to determine the dependence of the power threshold P lh for the transition to the H-mode regime and the threshold P hl for the transition from H-mode to L-mode as functions of external parameters. There is a value of the line-averaged density n e at which P lh has a minimum and P lh tends to increase for lower and higher values of n e . Experiments conducted to separate the effect of the neutral density n 0 from the plasma density n e give evidence of a strong coupling between n 0 and n e . The separate effect of neutrals on the transition has not been determined. Coordinated experiments with JET made in the ITER shape show that P lh increases approximately as S 0.5 where S is the plasma surface area. For these discharges, the power threshold in DIII-D was high by normal standards, thus suggesting that effects other than plasma size may have affected the experiment. Studies of H-L transitions have been initiated and hysteresis of order 40% has been observed. Studies have also been done of the dependence of the L-H transition on local edge parameters. Characterization of the edge within a few ms prior to the transition shows that the range of edge temperatures at which the transition has been observed is more restrictive than the range of densities at which it occurs. These results suggest that some temperature function is important for controlling the transition

  6. Modification of H-Mode Pedestal Instabilities in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    J.R. Ferron; M.S. Chu; G.L. Jackson; L.L. Lao; R.L. Miller; T.H. Osborne; P.B. Snyder; E.J. Strait; T.S. Taylor; A.D. Turnbull; A.M. Garofalo; M.A. Makowski; B.W. Rice; M.S. Chance; L.R. Baylor; M. Murakami; M.R. Wade

    1999-01-01

    Through comparison of experiment and ideal magnetohydrodynamic (MHD) theory, modes driven in the edge region of tokamak H-mode discharges [Type I edge-localized modes (ELMs)] are shown to result from low toroidal mode number (n) instabilities driven by pressure gradient and current density. The mode amplitude and frequency are functions of the discharge shape. Reductions in mode amplitude are observed in discharge shapes with either high squareness or low triangularity where the low-n stability threshold in the edge pressure gradient is predicted to be reduced and the most unstable mode is expected to have higher values of n. The importance of access to the ballooning mode second stability regime is demonstrated through the changes in the ELM character that occur when second regime access is not available. An edge stability model is presented that predicts that there is a threshold value of n for second regime access and that the most unstable mode has n near this threshold

  7. Characteristics of edge pedestals in LHW and NBI heated H-mode plasmas on EAST

    Science.gov (United States)

    Zang, Q.; Wang, T.; Liang, Y.; Sun, Y.; Chen, H.; Xiao, S.; Han, X.; Hu, A.; Hsieh, C.; Zhou, H.; Zhao, J.; Zhang, T.; Gong, X.; Hu, L.; Liu, F.; Hu, C.; Gao, X.; Wan, B.; the EAST Team

    2016-10-01

    By using the recently developed Thomson scattering diagnostic, the pedestal structure of the H-mode with neutral beam injection (NBI) or/and lower hybrid wave (LHW) heating on EAST (Experimental Advanced Superconducting Tokamak) is analyzed in detail. We find that a higher ratio of the power of the NBI to the total power of the NBI and the lower hybrid wave (LHW) will produce a large and regular different edge-localized mode (ELM), and a lower ratio will produce a small and irregular ELM. The experiments show that the mean pedestal width has good correlation with β \\text{p,\\text{ped}}0.5 , The pedestal width appears to be wider than that on other similar machines, which could be due to lithium coating. However, it is difficult to draw any conclusion of correlation between ρ * and the pedestal width for limited ρ * variation and scattered distribution. It is also found that T e/\

  8. ELM suppression in low edge collisionality H-mode discharges using n = 3 magnetic perturbations

    Energy Technology Data Exchange (ETDEWEB)

    Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Evans, T E [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Doyle, E J [University of California, Los Angeles, California (United States); Fenstermacher, M E [Lawrence Livermore National Laboratory, Livermore, California (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Leonard, A W [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Moyer, R A [University of California, San Diego, California (United States); Osborne, T H; Schaffer, M J; Snyder, P B [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Thomas, P R [CEA Cadarache EURATOM Association, Cadarache (France); West, W P [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Boedo, J A [University of California, San Diego, California (United States); Garofalo, A M [Columbia University, New York, New York (United States); Gohil, P; Jackson, G L; La Haye, R J [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Lasnier, C J [Lawrence Livermore National Laboratory, Livermore, California (United States); Reimerdes, H [Columbia University, New York, New York (United States); Rhodes, T L [University of California, Los Angeles, California (United States); Scoville, J T [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Solomon, W M [Princeton Plasma Physics Laboratory, Princeton, New Jersey (United States); Thomas, D M [General Atomics, PO Box 85608, San Diego, CA 92186-9784 (United States); Wang, G [University of California, Los Angeles, California (United States); Watkins, J G [Sandia National Laboratories, Albuquerque, New Mexico (United States); Zeng, L [University of California, Los Angeles, California (United States)

    2005-12-15

    Using resonant magnetic perturbations with toroidal mode number n = 3, we have produced H-mode discharges without edge localized modes (ELMs) which run with constant density and radiated power for periods up to about 2550 ms (17 energy confinement times). These ELM suppression results are achieved at pedestal collisionalities close to those desired for next step burning plasma experiments such as ITER and provide a means of eliminating the rapid erosion of divertor components in such machines which could be caused by giant ELMs. The ELM suppression is due to an enhancement in the edge particle transport which reduces pedestal current density and maximum edge pressure gradient below the threshold for peeling-ballooning modes. These n = 3 magnetic perturbations provide a means of active control of edge plasma transport.

  9. ELM suppression in low edge collisionality H-mode discharges using n = 3 magnetic perturbations

    International Nuclear Information System (INIS)

    Burrell, K H; Evans, T E; Doyle, E J; Fenstermacher, M E; Groebner, R J; Leonard, A W; Moyer, R A; Osborne, T H; Schaffer, M J; Snyder, P B; Thomas, P R; West, W P; Boedo, J A; Garofalo, A M; Gohil, P; Jackson, G L; La Haye, R J; Lasnier, C J; Reimerdes, H; Rhodes, T L; Scoville, J T; Solomon, W M; Thomas, D M; Wang, G; Watkins, J G; Zeng, L

    2005-01-01

    Using resonant magnetic perturbations with toroidal mode number n = 3, we have produced H-mode discharges without edge localized modes (ELMs) which run with constant density and radiated power for periods up to about 2550 ms (17 energy confinement times). These ELM suppression results are achieved at pedestal collisionalities close to those desired for next step burning plasma experiments such as ITER and provide a means of eliminating the rapid erosion of divertor components in such machines which could be caused by giant ELMs. The ELM suppression is due to an enhancement in the edge particle transport which reduces pedestal current density and maximum edge pressure gradient below the threshold for peeling-ballooning modes. These n = 3 magnetic perturbations provide a means of active control of edge plasma transport

  10. SOLPS5 modelling of the type III ELMing H-mode on TCV

    International Nuclear Information System (INIS)

    Gulejova, B.; Pitts, R.A.; Wischmeier, M.; Behn, R.; Coster, D.; Horacek, J.; Marki, J.

    2007-01-01

    Although ohmic H-modes have long been produced on TCV and the effects of ELMs at the divertor target studied in some detail, no attempt has yet been made to model the scrape-off layer (SOL) in these plasmas. This paper describes details of the first such efforts in which simulations of the inter-ELM phases using the coupled fluid-Monte Carlo SOLPS5 code (without drifts) are constrained by careful upstream Thomson scattering and Langmuir probe profiles. Simulated divertor profiles are compared with Langmuir probes and fast IR camera measurements at the targets. To account for the very differing transport rates in the edge pedestal and main SOL regions, radial variation of edge transport coefficients has been introduced in the simulations. Similarly, it is found that transport in the main chamber and divertor regions must be separately adjusted to provide an acceptable code-experiment match

  11. 'Affine' algebras on the torus

    International Nuclear Information System (INIS)

    Zakkari, M.

    1993-07-01

    The analysis of the Kac-Moody ''like'' algebra L-circumflex 2 (G) on the torus is performed. It will be seen that the root systems construction leading to a Cartan matrix is not possible. Different twist of L-circumflex 2 λ (G) are discussed. Connections with known results are done. (author). 10 refs

  12. Papers presented at the IAEA technical committee meeting on H-mode physics

    International Nuclear Information System (INIS)

    TCV team

    1995-11-01

    The two papers contained in this report deal with ohmic H-modes and effect on confinement of edge localized modes in the TCV tokamak. They were presented by the TCV team at the 1995 IAEA technical committee meeting on H-mode physics. figs., tabs., refs

  13. Comparison of hybrid and baseline ELMy H-mode confinement in JET with the carbon wall

    NARCIS (Netherlands)

    Beurskens, M. N. A.; Frassinetti, L.; Challis, C.; Osborne, T.; Snyder, P. B.; Alper, B.; Angioni, C.; Bourdelle, C.; Buratti, P.; Crisanti, F.; Giovannozzi, E.; Giroud, C.; Groebner, R.; Hobirk, J.; Jenkins, I.; Joffrin, E.; Leyland, M. J.; Lomas, P.; Mantica, P.; McDonald, D.; Nunes, I.; Rimini, F.; Saarelma, S.; Voitsekhovitch, I.; P. de Vries,; Zarzoso, D.

    2013-01-01

    The confinement in JET baseline type I ELMy H-mode plasmas is compared to that in so-called hybrid H-modes in a database study of 112 plasmas in JET with the carbon fibre composite (CFC) wall. The baseline plasmas typically have beta(Nu) similar to 1.5-2, H-98 similar to 1, whereas the hybrid

  14. Investigation of lower hybrid current drive during H-mode in EAST tokamak

    International Nuclear Information System (INIS)

    Li Miao-Hui; Ding Bo-Jiang; Kong Er-Hua; Zhang Lei; Zhang Xin-Jun; Qian Jin-Ping; Yan Ning; Han Xiao-Feng; Shan Jia-Fang; Liu Fu-Kun; Wang Mao; Xu Han-Dong; Wan Bao-Nian

    2011-01-01

    H-mode discharges with lower hybrid current drive (LHCD) alone are achieved in EAST divertor plasma over a wide parameter range. These H-mode discharges are characterized by a sudden drop in D α emission and a spontaneous rise in main plasma density. Good lower hybrid (LH) coupling during H-mode is obtained by putting the plasma close to the antenna and by injecting D 2 gas from a pipe near the grill mouse. The analysis of lower hybrid current drive properties shows that the LH deposition profile shifts off axis during H-mode, and current drive (CD) efficiency decreases due to the increase in density. Modeling results of H-mode discharges with a general ray tracing code GENRAY are reported. (physics of gases, plasmas, and electric discharges)

  15. Plasma diagnostics in compact torus of UNICAMP (Campinas state university)

    International Nuclear Information System (INIS)

    Ueda, M; Doi, Y.; Aramaki, E.A.; Porto, P.; Berni, L.; Machida, M.

    1989-08-01

    This paper which describes experiments carried out in the Compact Torus of UNICAMP (TC-1) is divided into 3 parts: 1) summary of TC-1 characteristics and its operation mode; 2) description of diagnostics in use and ones to be installed, 3) recent experimental results using optical and electromagnetical diagnostics. (author)

  16. Effect of ripple-induced transport on H-mode performance in tokamaks

    International Nuclear Information System (INIS)

    Parail, V.; Vries, P. de; Lonnroth, J.; Kiviniemi, T.; Johnson, T.; Loarte, A.; Saibene, G.; Hatae, T.; Kamada, Y.; Konovalov, S.; Oyama, N.; Shinohara, K.; Tobita, K.; Urano, H.

    2005-01-01

    A number of experiments have shown that ripple-induced transport influences performance of ELMy H-modes in the tokamak. A noticeable difference in confinement, ELM frequency and amplitude was found between JET (with ripple amplitude δ∼0.1%) and JT-60U (with δ∼1%) in otherwise identical discharges. It was previously shown in JET experiments with enhanced ripple that a gradual increase in the ripple amplitude first leads to a modest improvement in plasma confinement, which is followed by the degradation of edge pedestal and further transition to the L-mode regime if δ increases further. The DIII-D team recently reported a marginal increase in confinement in experiments with an edge transport enhanced by the externally driven resonant magnetic perturbation. Numerical predictive modelling of the dynamics of ELMy H-mode JET plasma relevant to a JET/JT-60U similarity experiment has been conducted taking into account ripple-induced ion transport, which was computed using the orbit following code ASCOT. This predictive modelling reveals that, depending on plasma parameters, ripple amplitude and localisation (the latter depending on the toroidal coil design), this additional transport can either improve global plasma confinement or reduce it. These controlled ripple losses might be used as an effective tool for ELM mitigation and may provide an explanation for the difference between JET and JT-60U observed in the similarity experiments. A detailed comparison between ripple- induced transport and the alternative method of ELM mitigation by an externally driven edge magnetic perturbation is discussed. The fact that ripple losses mainly increase ion transport, while a stochastic magnetic layer increases electron transport indicates that it might be beneficial to use a combination of both methods in future experiments. This work was funded partly by the United Kingdom Engineering and Physical Sciences Research Council and by the European Communities under the contract of

  17. The Columbia Non-neutral Torus

    International Nuclear Information System (INIS)

    Pedersen, Thomas Sunn

    2009-01-01

    Final report for the Columbia Non-neutral Torus. This details the results from the design, construction and initial operation of the Columbia Non-neutral Torus. During the duration of this grant, I designed, built, and operated the Columbia Nonneutral Torus, the world's lowest aspect ratio stellarator, and arguably, the world's simplest stellarator. This demonstrates the ease and robustness of the chosen stellarator design and allowed us to commence the investigation of the physics of non-neutral plasmas confined on magnetic surfaces. These plasmas are unique in many ways and had not previously been studied in a stellarator. Our first results showed that it is possible to confine and study a relatively cold pure electron plasma in a stellarator. We confirmed that the plasma is stable, and that the plasma is reasonably well confined in a stellarator configuration. These results were published in Physics of Plasmas (2006) and Physical Review Letters (2006). They enabled the existing program which is resolving the underlying transport processes in a classical stellarator with intense self-electric fields and enable the next phase of operation, electron-positron plasma physics. During the period of this grant, two students were trained in experimental plasma physics and both received their PhD degrees shortly after the grant terminated. One student is now employed in the financial services industry, the other is a postdoctoral associate at Los Alamos National Laboratory. The chief goals were to build and begin operation of the Columbia Non-neutral Torus. These goals were achieved in the third year of funding. The development of diagnostic methods and the confirmation of stable equilibria were also achieved during the grant period. In summary, the main scientific goals were all met. The main educational goals were also met, as the experiment became the training ground not only for the two aforementioned graduate students but also for a number of undergraduate students

  18. Predictive modelling of edge transport phenomena in ELMy H-mode tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Loennroth, J.-S.

    2009-01-01

    This thesis discusses a range of work dealing with edge plasma transport in magnetically confined fusion plasmas by means of predictive transport modelling, a technique in which qualitative predictions and explanations are sought by running transport codes equipped with models for plasma transport and other relevant phenomena. The focus is on high confinement mode (H-mode) tokamak plasmas, which feature improved performance thanks to the formation of an edge transport barrier. H-mode plasmas are generally characterized by the occurrence of edge localized modes (ELMs), periodic eruptions of particles and energy, which limit confinement and may turn out to be seriously damaging in future tokamaks. The thesis introduces schemes and models for qualitative study of the ELM phenomenon in predictive transport modelling. It aims to shed new light on the dynamics of ELMs using these models. It tries to explain various experimental observations related to the performance and ELM-behaviour of H-mode plasmas. Finally, it also tries to establish more generally the potential effects of ripple-induced thermal ion losses on H-mode plasma performance and ELMs. It is demonstrated that the proposed ELM modelling schemes can qualitatively reproduce the experimental dynamics of a number of ELM regimes. Using a theory-motivated ELM model based on a linear instability model, the dynamics of combined ballooning-peeling mode ELMs is studied. It is shown that the ELMs are most often triggered by a ballooning mode instability, which renders the plasma peeling mode unstable, causing the ELM to continue in a peeling mode phase. Understanding the dynamics of ELMs will be a key issue when it comes to controlling and mitigating the ELMs in future large tokamaks. By means of integrated modelling, it is shown that an experimentally observed increase in the ELM frequency and deterioration of plasma confinement triggered by external neutral gas puffing might be due to a transition from the second to

  19. Application of the H-Mode, a Design and Interaction Concept for Highly Automated Vehicles, to Aircraft

    Science.gov (United States)

    Goodrich, Kenneth H.; Flemisch, Frank O.; Schutte, Paul C.; Williams, Ralph A.

    2006-01-01

    Driven by increased safety, efficiency, and airspace capacity, automation is playing an increasing role in aircraft operations. As aircraft become increasingly able to autonomously respond to a range of situations with performance surpassing human operators, we are compelled to look for new methods that help us understand their use and guide their design using new forms of automation and interaction. We propose a novel design metaphor to aid the conceptualization, design, and operation of highly-automated aircraft. Design metaphors transfer meaning from common experiences to less familiar applications or functions. A notable example is the "Desktop metaphor" for manipulating files on a computer. This paper describes a metaphor for highly automated vehicles known as the H-metaphor and a specific embodiment of the metaphor known as the H-mode as applied to aircraft. The fundamentals of the H-metaphor are reviewed followed by an overview of an exploratory usability study investigating human-automation interaction issues for a simple H-mode implementation. The envisioned application of the H-mode concept to aircraft is then described as are two planned evaluations.

  20. Discriminant analysis to predict the occurrence of ELMs in H-mode discharges

    International Nuclear Information System (INIS)

    Kardaun, O.J.W.F.; Itoh, S.; Itoh, K.; Kardaun, J.W.P.F.

    1993-08-01

    After an exposition of its theoretical background, discriminant analysis is applied to the H-mode confinement database to find the region in plasma parameter space in which H-mode with small ELMs (Edge Localized Modes) is likely to occur. The boundary of this region is determined by the condition that the probability of appearance of such a type of H-mode, as a function of the plasma parameters, should be (1) larger than some threshold value and (2) larger than the corresponding probability for other types of H-mode (i.e., H-mode without ELMs or with giant ELMs). In practice, the discrimination has been performed for the ASDEX, JET and JFT-2M tokamaks (a) using four instantaneous plasma parameters (injected power P inj , magnetic field B t , plasma current I p and line averaged electron density (n-bar e ) and (b) taking also memory effects of the plasma and the distance between the plasma and the wall into account, while using variables that are normalised with respect to machine size. Generally speaking, it is found that there is a substantial overlap between the region of H-mode with small ELMs and the region of the two other types of H-mode. However, the ELM-free and the giant ELM H-modes relatively rarely appear in the region, that, according to the analysis, is allocated to small ELMs. A reliable production of H-mode with only small ELMs seems well possible by choosing this regime in parameter space. In the present study, it was not attempted to arrive at a unified discrimination across the machines. So, projection from one machine to another remains difficult, and a reliable determination of the region where small ELMs occur still requires a training sample from the device under consideration. (author) 53 refs

  1. Recent Progress on Spherical Torus Research

    Energy Technology Data Exchange (ETDEWEB)

    Ono, Masayuki [PPPL; Kaita, Robert [PPPL

    2014-01-01

    The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ~ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ~ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of attractive fusion energy power source. Since the start of the two megaampere class ST facilities in 2000, National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all of fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.

  2. FLIT: Flowing LIquid metal Torus

    Science.gov (United States)

    Kolemen, Egemen; Majeski, Richard; Maingi, Rajesh; Hvasta, Michael

    2017-10-01

    The design and construction of FLIT, Flowing LIquid Torus, at PPPL is presented. FLIT focuses on a liquid metal divertor system suitable for implementation and testing in present-day fusion systems, such as NSTX-U. It is designed as a proof-of-concept fast-flowing liquid metal divertor that can handle heat flux of 10 MW/m2 without an additional cooling system. The 72 cm wide by 107 cm tall torus system consisting of 12 rectangular coils that give 1 Tesla magnetic field in the center and it can operate for greater than 10 seconds at this field. Initially, 30 gallons Galinstan (Ga-In-Sn) will be recirculated using 6 jxB pumps and flow velocities of up to 10 m/s will be achieved on the fully annular divertor plate. FLIT is designed as a flexible machine that will allow experimental testing of various liquid metal injection techniques, study of flow instabilities, and their control in order to prove the feasibility of liquid metal divertor concept for fusion reactors. FLIT: Flowing LIquid metal Torus. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

  3. Operational range and transport barrier of the H-mode in the stellarator W7-AS

    International Nuclear Information System (INIS)

    Hirsch, M.; Amadeo, P.; Anton, M.; Baldzuhn, J.; Brakel, R.; Bleuel, J.; Fiedler, S.; Geist, T.; Grigull, P.; Hartfuss, H.J.; Jaenicke, R.; Kick, M.; Kisslinger, J.; Koponen, J.; Wagner, F.; Weller, A.; Wobig, H.; Zoletnik, S.; Holzhauer, E.

    1998-01-01

    In W7-AS the H-mode is characterized by an edge transport barrier localized in the first 3-4 cm inside the separatrix. In the ELMy H-mode preceding the quiescent state ELMs appear as a sudden breakdown of the edge transport barrier in coincidence with bursts of fluctuations. Between ELMs fluctuations are identical to those of the quiescent H-mode. The operational range of the quiescent H-mode is determined by narrow windows of the edge rotational transform and a threshold edge electron density. In contrast, ELM-like events are observed for a variety of plasma conditions by far exceeding the narrow operational windows for the quiescent state. (author)

  4. X-Divertor Geometries for Deeper Detachment Without Degrading the DIII-D H-Mode

    Science.gov (United States)

    Covele, Brent; Kotschenreuther, M. T.; Valanju, P. M.; Mahajan, S. M.; Leonard, A. W.; Hyatt, A. W.; McLean, A. G.; Thomas, D. M.; Guo, H. Y.; Watkins, J. G.; Makowski, M. A.; Hill, D. N.

    2015-11-01

    Recent DIII-D experiments comparing the standard divertor (SD) and X-Divertor (XD) geometries show heat and particle flux reduction at the divertor target plate. The XD features large poloidal flux expansion, increased connection length, and poloidal field line flaring, quantified by the Divertor Index. Both SD and XD were pushed deep into detachment with increased gas puffing, until core energy confinement and pedestal pressure were substantially reduced. As expected, outboard target heat fluxes are significantly reduced in the XD compared to the SD under similar upstream plasma conditions, even at low Greenwald fraction. The high-triangularity (floor) XD cases show larger reduction in temperature, heat, and particle flux relative to the SD in all cases, while low-triangularity (shelf) XD cases show more modest reductions over the SD. Consequently, heat flux reduction and divertor detachment may be achieved in the XD with less gas puffing and higher pedestal pressures. Further causative analysis, as well as detailed modeling with SOLPS, is underway. These initial experiments suggest the XD as a promising candidate to achieve divertor heat flux control compatible with robust H-mode operation. Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-FG02-04ER54754, and DE-FG02-04ER54742.

  5. Structure, stability and ELM dynamics of the H-mode pedestal in DIII-D

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Leonard, A.W.; Osborne, T.H.

    2005-01-01

    Experiments are described that have increased understanding of the transport and stability physics that set the H-mode edge pedestal width and height, determine the onset of Type-I edge localized modes (ELMs), and produce the nonlinear dynamics of the ELM perturbation in the pedestal and scrape-off layer (SOL). Predictive models now exist for the n e pedestal profile and the p e height at the onset of Type-I ELMs, and progress has been made toward predictive models of the T e pedestal width and nonlinear ELM evolution. Similarity experiments between DIII-D and JET suggested that neutral penetration physics dominates in the relationship between the width and height of the n e pedestal while plasma physics dominates in setting the T e pedestal width. Measured pedestal conditions including edge current at ELM onset agree with intermediate-n peeling-ballooning (P-B) stability predictions. Midplane ELM dynamics data show the predicted (P-B) structure at ELM onset, large rapid variations of the SOL parameters, and fast radial propagation in later phases, similar to features in nonlinear ELM simulations. (author)

  6. Progress in quantifying the edge physics of the H mode regime in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Baker, D.R.; Burrell, K.H.

    2001-01-01

    Edge conditions in DIII-D are being quantified in order to provide insight into the physics of the H mode regime. Several studies show that electron temperature is not the key parameter that controls the L-H transition. Gradients of edge temperature and pressure are much more promising candidates for elements of such parameters. They systematically increase during the L phases of discharges which make a transition to H mode, and these increases are typically larger than the increases in the underlying quantities. The quality of H mode confinement is strongly correlated with the height of the H mode pedestal for the pressure. The gradient of the pressure is limited by MHD modes, in particular by ideal kink ballooning modes with finite mode number n. For a wide variety of discharges, the width of the barrier for electron pressure is well described by a relationship that is proportional to (β p ped ) 1/2 . A new regime of confinement, called the quiescent H mode, which provides steady state operation with no ELMs, low radiated power and normal H mode confinement, has been discovered. A coherent edge MHD mode provides adequate particle transport to control the plasma density while permitting the pressure pedestal to remain almost identical to that observed in ELMing discharges. (author)

  7. The 13th International Workshop on H-mode Physics and Transport Barriers (Oxford, UK, 2011) The 13th International Workshop on H-mode Physics and Transport Barriers (Oxford, UK, 2011)

    Science.gov (United States)

    Saibene, G.

    2012-11-01

    as to stimulate and lead the open discussion. Poster sessions were also organized to present specialist papers and provide a venue for continued discussion. The topics selected for this edition of the workshop were: 1. Integrated plasma scenarios for ITER and a reactor: experimental and theoretical studies, including the self-stabilizing transport approach. 2. Edge transport barrier control and plasma performance: physics of 3D stochastic magnetic fields for ELM suppression. 3. H-mode transition physics and H-mode pedestal structure: pedestal dynamics near transitions and requirements for high-confinement access and sustainment. 4. Energetic particle driven instabilities and related physics: H-mode and the transport barrier. 5. Role of and evidence for non-diffusive particle and toroidal momentum transport and impact of fuelling: experiments, theory and modelling. 6. Long-range correlation of plasma turbulence and interaction between edge and core transport. The choice of topics, and the amount of progress in the understanding of the complexity of transport barriers physics reflect the drive in the fusion community towards the preparation for the ITER tokamak operation. More than 100 scientists (including students) attended the three-day workshop, coming from all over the world to present their newest results, discuss with colleagues and enjoy the atmosphere of the beautiful Lady Margaret Hall. The preparation work of the International Advisory Committee (G. Saibene (EU - Chair), R. Groebner (US), T. S Hahm (KO), A. Hubbard (US), K. Ida (Japan), S. Lebedev (RF), N. Oyama (Japan), E Wolfrum (EU)) has been rewarded by the enthusiastic participation of scientists, experimentalist, modellers and theoreticians, and by the high level of the scientific discussion throughout the workshop, during lunch breaks and even at the conference dinner. The Committee is also grateful to EFDA for the support in the organization of the workshop and to the Local Organizing Committee (E

  8. Kinetic neoclassical transport in the H-mode pedestal

    Energy Technology Data Exchange (ETDEWEB)

    Battaglia, D. J.; Chang, C. S.; Ku, S.; Grierson, B. A. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08540 (United States); Burrell, K. H.; Grassie, J. S. de [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

    2014-07-15

    Multi-species kinetic neoclassical transport through the QH-mode pedestal and scrape-off layer on DIII-D is calculated using XGC0, a 5D full-f particle-in-cell drift-kinetic solver with self-consistent neutral recycling and sheath potentials. Quantitative agreement between the flux-driven simulation and the experimental electron density, impurity density, and orthogonal measurements of impurity temperature and flow profiles is achieved by adding random-walk particle diffusion to the guiding-center drift motion. The radial electric field (E{sub r}) that maintains ambipolar transport across flux surfaces and to the wall is computed self-consistently on closed and open magnetic field lines and is in excellent agreement with experiment. The E{sub r} inside the separatrix is the unique solution that balances the outward flux of thermal tail deuterium ions against the outward neoclassical electron flux and inward pinch of impurity and colder deuterium ions. Particle transport in the pedestal is primarily due to anomalous transport, while the ion heat and momentum transport are primarily due to the neoclassical transport. The full-f treatment quantifies the non-Maxwellian energy distributions that describe a number of experimental observations in low-collisionallity pedestals on DIII-D, including intrinsic co-I{sub p} parallel flows in the pedestal, ion temperature anisotropy, and large impurity temperatures in the scrape-off layer.

  9. Role of Density Gradient Driven Trapped Electron Modes in the H-Mode Inner Core with Electron Heating

    Science.gov (United States)

    Ernst, D.

    2015-11-01

    We present new experiments and nonlinear gyrokinetic simulations showing that density gradient driven TEM (DGTEM) turbulence dominates the inner core of H-Mode plasmas during strong electron heating. Thus α-heating may degrade inner core confinement in H-Mode plasmas with moderate density peaking. These DIII-D low torque quiescent H-mode experiments were designed to study DGTEM turbulence. Gyrokinetic simulations using GYRO (and GENE) closely match not only particle, energy, and momentum fluxes, but also density fluctuation spectra, with and without ECH. Adding 3.4 MW ECH doubles Te /Ti from 0.5 to 1.0, which halves the linear TEM critical density gradient, locally flattening the density profile. Density fluctuations from Doppler backscattering (DBS) intensify near ρ = 0.3 during ECH, displaying a band of coherent fluctuations with adjacent toroidal mode numbers. GYRO closely reproduces the DBS spectrum and its change in shape and intensity with ECH, identifying these as coherent TEMs. Prior to ECH, parallel flow shear lowers the effective nonlinear DGTEM critical density gradient 50%, but is negligible during ECH, when transport displays extreme stiffness in the density gradient. GS2 predictions show the DGTEM can be suppressed, to avoid degradation with electron heating, by broadening the current density profile to attain q0 >qmin > 1 . A related experiment in the same regime varied the electron temperature gradient in the outer half-radius (ρ ~ 0 . 65) using ECH, revealing spatially coherent 2D mode structures in the Te fluctuations measured by ECE imaging. Fourier analysis with modulated ECH finds a threshold in Te profile stiffness. Supported by the US DOE under DE-FC02-08ER54966 and DE-FC02-04ER54698.

  10. The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, R. J., E-mail: rperkins@pppl.gov; Hosea, J. C.; Jaworski, M. A.; Diallo, A.; Bell, R. E.; Bertelli, N.; Gerhardt, S.; Kramer, G. J.; LeBlanc, B. P.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Taylor, G.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Ahn, J.-W.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Sabbagh, S. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

    2015-04-15

    The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. Here, we demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heat flux transmission coefficient in the presence of the RF field. Although precise comparison between the computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. This work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor.

  11. The influence of gas fuelling location on H-mode access in the MAST spherical tokamak

    International Nuclear Information System (INIS)

    Field, A R; Carolan, P G; Conway, N J; Counsell, G F; Cunningham, G; Helander, P; Meyer, H; Taylor, D; Tournianski, M R; Walsh, M J

    2004-01-01

    The observation that high-field side (HFS) gas puff refuelling facilitates access to the improved confinement (H-mode) regime on the COMPASS-D and MAST tokamaks prompted a theoretical investigation of the role of the neutral gas dynamics in controlling the edge plasma rotation and radial E-field, E r . Within the framework of neo-classical theory, higher edge plasma flow, and hence E r , are predicted when fuelling from the HFS-rather than from the more usual low-field side (LFS)-provided neutral viscosity dominates the transport of toroidal angular momentum. Here, these predictions are compared with experiments on MAST, where the influence of the gas-puff location on the edge E r profile is measured spectroscopically. An increase in E r is indeed observed with HFS refuelling in the region where the edge transport barrier forms, provided the neutral density at the LFS is sufficiently low so as not to damp the toroidal flow

  12. Study of the conditions for spontaneous H-mode transitions in DIII-D

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Groebner, R.J.

    1996-01-01

    A series of scaling studies attempting to correlate the H(high)-mode power threshold (P TH ) with global parameters have been conducted. Data from these discharges is also being used to look for dependence of P TH on local edge parameters and to test theories of the transition. Boronization and better operational techniques have resulted in lower power thresholds and weaker density scaling. Neon impurity injection experiments show that radiation also plays a role in determining P TH . A low density threshold for the L(low)-H(high) transition has been linked with the locked mode low density limit, and can be reduced with the use of an error field correcting coil. Highly developed edge diagnostics, with spatial resolution as low as 5 mm, are used to evaluate how the power threshold depends on local edge conditions. Preliminary analysis of local edge conditions for parameter scans of n e , B T , and I p in single-null discharges, and the X-point imbalance in double-null discharges-show that, just before the transition to H-mode, the edge temperatures near the separatrix are approximately constant at 100 i e *i , varied from 2 to 17, demonstrating that a transition condition as simple as v *i = constant is inconsistent with the data. The local edge parameters of n e , T e , and T i do not always follow the same global scaling as P TH . Therefore, theories of the L-H transition need not be constrained by these scalings

  13. Fast wave current drive in H mode plasmas on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Grassie, J.S. de; Baity, F.W.

    1999-01-01

    Current driven by fast Alfven waves is measured in H mode and VH mode plasmas on the DIII-D tokamak for the first time. Analysis of the poloidal flux evolution shows that the fast wave current drive profile is centrally peaked but sometimes broader than theoretically expected. Although the measured current drive efficiency is in agreement with theory for plasmas with infrequent ELMs, the current drive efficiency is an order of magnitude too low for plasmas with rapid ELMs. Power modulation experiments show that the reduction in current drive with increasing ELM frequency is due to a reduction in the fraction of centrally absorbed fast wave power. The absorption and current drive are weakest when the electron density outside the plasma separatrix is raised above the fast wave cut-off density by the ELMs, possibly allowing an edge loss mechanism to dissipate the fast wave power since the cut-off density is a barrier for fast waves leaving the plasma. (author)

  14. H-mode pedestal characteristics in ITER shape discharges on DIII-D

    International Nuclear Information System (INIS)

    Osborne, T.H.; Burrell, K.H.; Groebner, R.J.

    1998-09-01

    Characteristics of the H-mode pedestal are studied in Type 1 ELM discharges with ITER cross-sectional shape and aspect ratio. The scaling of the width of the edge step gradient region, δ, which is most consistent with the data is with the normalized edge pressure, (β POL PED ) 0.4 . Fits of δ to a function of temperature, such as ρ POL , are ruled out in divertor pumping experiments. The edge pressure gradient is found to scale as would be expected from infinite n ballooning mode theory; however, the value of the pressure gradient exceeds the calculated first stable limit by more than a factor of 2 in some discharges. This high edge pressure gradient is consistent with access to the second stable regime for ideal ballooning for surfaces near the edge. In lower q discharges, including discharges at the ITER value of q, edge second stability requires significant edge current density. Transport simulations give edge bootstrap current of sufficient magnitude to open second stable access in these discharges. Ideal kink analysis using current density profiles including edge bootstrap current indicate that before the ELM these discharges may be unstable to low n, edge localized modes

  15. Edge operational space for high density/high confinement ELMY H-modes in JET

    International Nuclear Information System (INIS)

    Sartori, R.; Saibene, G.; Loarte, A.

    2002-01-01

    This paper discusses how the proximity to the L-H threshold affects the confinement of ELMy H-modes at high density. The largest reduction in confinement at high density is observed at the transition from the Type I to the Type III ELMy regime. At medium plasma triangularity, δ≅0.3 (where δ is the average triangularity at the separatrix), JET experiments show that by increasing the margin above the L-H threshold power and maintaining the edge temperature above the critical temperature for the transition to Type III ELMs, it is possible to avoid the degradation of the pedestal pressure with density, normally observed at lower power. As a result, the range of achievable densities (both in the core and in the pedestal) is increased. At high power above the L-H threshold power the core density was equal to the Greenwald limit with H97≅0.9. There is evidence that a mixed regime of Type I and Type II ELMs has been obtained at this intermediate triangularity, possibly as a result of this increase in density. At higher triangularity, δ≅0.5, the power required to achieve similar results is lower. (author)

  16. The influence of gas pressure on E↔H mode transition in argon inductively coupled plasmas

    Science.gov (United States)

    Zhang, Xiao; Zhang, Zhong-kai; Cao, Jin-xiang; Liu, Yu; Yu, Peng-cheng

    2018-03-01

    Considering the gas pressure and radio frequency power change, the mode transition of E↔H were investigated in inductively coupled plasmas. It can be found that the transition power has almost the same trend decreasing with gas pressure, whether it is in H mode or E mode. However, the transition density increases slowly with gas pressure from E to H mode. The transition points of E to H mode can be understood by the propagation of electromagnetic wave in the plasma, while the H to E should be illustrated by the electric field strength. Moreover, the electron density, increasing with the pressure and power, can be attributed to the multiple ionization, which changes the energy loss per electron-ion pair created. In addition, the optical emission characteristics in E and H mode is also shown. The line ratio of I750.4 and I811.5, taken as a proxy of the density of metastable state atoms, was used to illustrate the hysteresis. The 750.4 nm line intensity, which has almost the same trend with the 811.5 nm line intensity in H mode, both of them increases with power but decreases with gas pressure. The line ratio of 811.5/750.4 has a different change rule in E mode and H mode, and at the transition point of H to E, it can be one significant factor that results in the hysteresis as the gas pressure change. And compared with the 811.5 nm intensity, it seems like a similar change rule with RF power in E mode. Moreover, some emitted lines with lower rate constants don't turn up in E mode, while can be seen in H mode because the excited state atom density increasing with the electron density.

  17. W transport and accumulation control in the termination phase of JET H-mode discharges and implications for ITER

    Science.gov (United States)

    Köchl, F.; Loarte, A.; de la Luna, E.; Parail, V.; Corrigan, G.; Harting, D.; Nunes, I.; Reux, C.; Rimini, F. G.; Polevoi, A.; Romanelli, M.; Contributors, JET

    2018-07-01

    Tokamak operation with W PFCs is associated with specific challenges for impurity control, which may be particularly demanding in the transition from stationary H-mode to L-mode. To address W control issues in this phase, dedicated experiments have been performed at JET including the variation of the decrease of the power and current, gas fuelling and central ion cyclotron heating (ICRH), and applying active ELM control by vertical kicks. The experimental results obtained demonstrate the key role of maintaining ELM control to control the W concentration in the exit phase of H-modes with slow (ITER-like) ramp-down of the neutral beam injection power in JET. For these experiments, integrated fully predictive core+edge+SOL transport modelling studies applying discrete models for the description of transients such as sawteeth and ELMs have been performed for the first time with the JINTRAC suite of codes for the entire transition from stationary H-mode until the time when the plasma would return to L-mode focusing on the W transport behaviour. Simulations have shown that the existing models can appropriately reproduce the plasma profile evolution in the core, edge and SOL as well as W accumulation trends in the termination phase of JET H-mode discharges as function of the applied ICRH and ELM control schemes, substantiating the ambivalent effect of ELMs on W sputtering on one side and on edge transport affecting core W accumulation on the other side. The sensitivity with respect to NB particle and momentum sources has also been analysed and their impact on neoclassical W transport has been found to be crucial to reproduce the observed W accumulation characteristics in JET discharges. In this paper the results of the JET experiments, the comparison with JINTRAC modelling and the adequacy of the models to reproduce the experimental results are described and conclusions are drawn regarding the applicability of these models for the extrapolation of the applied W

  18. Fuzzy torus via q-Parafermion

    International Nuclear Information System (INIS)

    Aizawa, N; Chakrabarti, R

    2007-01-01

    We note that the recently introduced fuzzy torus can be regarded as a q-deformed parafermion. Based on this picture, classification of the Hermitian representations of the fuzzy torus is carried out. The result involves Fock-type representations and new finite-dimensional representations for q being a root of unity as well as already known finite-dimensional ones

  19. Pedestal characteristics and MHD stability of H-mode plasmas in TCV

    International Nuclear Information System (INIS)

    Pitzschke, A.

    2011-01-01

    temperature profile during the ELM cycle, the low repetition rate of the lasers used for Thomson scattering is a limiting. Although the system on TCV comprises 3 laser units that may be triggered in sequence with time separations down to 1 ms, time evolution over longer periods can only be reconstructed from repetitive events. In this context, an adjustment of the laser trigger to improve the synchronization with the ELM event is an advantage. A method was developed and implemented to generate a synchronizing trigger sequence, by a real-time monitoring of the D-alpha emission, which provides a marker for the ELM event. Recently, a ‘snowflake’ (SF) divertor configuration, proposed as a possible solution to reduce the plasma-wall interaction by changing the divertor’s poloidal magnetic field topology, was generated, for the first time, in TCV. A numerical code (KINX), based on a magnetohydrodynamic model (ideal MHD), was used to investigate the stability limits of this configuration under H-mode conditions and compare them with a similar standard single-null equilibrium. In a series of experiments, improved energy confinement was found and explained by improved stability of the edge region in the SF configuration. The influence of the pedestal structure in ELMy H-mode plasmas on the energy confinement and on ELM energy losses was investigated. The different ELM regimes found in TCV were analyzed, in particular the transition between type-III to type-I ELMs. The operational boundary of each ELM regime was characterized and verified by ideal MHD stability simulations for the ETB region. Recent studies on the scaling of the pedestal width with normalized poloidal pressure were confirmed. Using the capabilities of TCV, the influence of plasma shaping on pedestal parameters and MHD stability limits was investigated. In the past, models were developed to describe the onset of type-I ELMs, which are associated with modes in the ETB region arising from a coupling of pressure- and

  20. Thermo-mechanical and damage analyses of EAST carbon divertor under type-I ELMy H-mode operation

    Energy Technology Data Exchange (ETDEWEB)

    Li, W.X. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Ye, M.Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wu, S.T. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Qian, X.Y.; Zhu, C.C. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China)

    2016-04-15

    Highlights: • Type-I ELMy H-mode is one of the most severe operating environment in tokamak. • An actual time-history heat load has been used in thermo-mechanical analysis. • The analysis results are time-dependent during the whole discharge process. • The analysis could be very useful in evaluating the operational capability of the divertor. - Abstract: The lower carbon divertor has been used since 2008 in EAST, and many significant physical results, like the 410 s long pulse discharge and the 32 s H-mode operation, have been achieved. As the carbon divertor will still be used in the next few years while the injected auxiliary heating power would be increased gradually, it’s necessary to evaluate the operational capability of the carbon divertor under the heat loads during future operation. In this paper, an actual time-history heat load during type-I ELMy H-mode from EAST experiment, as one of the most severe operating environment in tokamak, has been used in the calculation and analysis. The finite element (FE) thermal and mechanical calculations have been carried out to analysis the stress and deformation of the carbon divertor during the heat loads. According to the results, the main impact on the overall temperature comes from the relative stable phase before and after the type-I ELMs and local peak load, and the transient thermal load such as type-I ELMy only has a significant effect on the surface temperature of the graphite tiles. The carbon divertor would work with high stress near the screw bolts in the current operational conditions, because of high preload and conservative frictional coefficient between the bolts and heatsink. For the future operation, new plasma facing materials (PFM) and divertor technology should be developed.

  1. First HIBP Measurement of Plasma Potential During the H-Mode Transition on the TUMAN-3M Tokamak

    International Nuclear Information System (INIS)

    Askinazi, L.G.; Golant, V.E.; Kornev, V.A.; Lebedev, S.V.; Shevkin, E.A.; Tukachinsky, A.S.; Zhubr, N.A.; Chmyga, A.A.; Dreval, N.B.; Khrebtov, S.M.; Komarov, A.S.; Krupnik, L.I.; Oost, G. van; Tendler, M.

    2003-01-01

    The difficulty of Heavy Ion Beam Probe (HIBP) application on the TUMAN-3M (R=0.53m, a=0.22m, BT=0.8T, Ip=140kA, Te=0.5keV, n<4 1019m-3) -- significant toroidal shift of beam trajectory -- is caused by high ratio of poloidal field to toroidal one. Strong UV radiation from the plasma loads the energy analyzer's detector and complicates the problem even more. This paper presents the results of first measurement of plasma potential evolution in the discharges performed in ohmic H-mode using 80 keV K+ beam and a Proca-Green secondary ion energy analyzer. Spatial region covered by the diagnostic in the experiments discussed was 0< r<0.6a. Spatial scan was performed utilizing the toroidal field decrease due to capacity power supply battery discharge. The change in plasma potential of the order of 100V has been measured during the H-mode formation. The potential in core plasma (r<0.6a) starts to change simultaneously with L-H transition, and than changes during ∼6-8ms after the transition. Thus, the potential changes rather slowly in a comparison with L-H transition timescale (∼2ms for TUMAN-3M ohmic H-mode). Possible explanation to the slow change in central plasma potential may be a formation of potential well structure at the plasma edge, in which radial electric field changes direction. This kind of structure is beneficial for the edge turbulent transport suppression because of high |∂Er/∂r|, but not necessary requires a strong change in central plasma potential to occur immediately. The results from microwave reflectometry support this hypothesis

  2. Comparison of L- and H-mode plasma edge fluctuations in MAST

    International Nuclear Information System (INIS)

    Dudson, B D; Dendy, R O; Kirk, A; Meyer, H; Counsell, G F

    2005-01-01

    Edge turbulence measurements from a reciprocating Langmuir probe in MAST are presented. A comparison of the range/standard deviation (R/S), growth of range, first moment and differencing and rescaling methods for calculating the Hurst exponent is made. The differencing and rescaling method is found to be the most useful for identifying scaling over long time-periods. A comparison is made between L-mode, dithering H-mode and H-mode plasma edge turbulence and evidence for self-similarity is found. Tests are performed and it is demonstrated that the results are due to properties of the data, and are not artefacts of the methods. A comparison of Hurst exponent methods with the autocorrelation function and power spectrum is used to demonstrate the presence of long-time correlation in L-mode data, and the absence of long-time correlation in the case of dithering H-mode

  3. Expression for the thermal H-mode energy confinement time under ELM-free conditions

    International Nuclear Information System (INIS)

    Ryter, F.; Gruber, O.; Kardaun, O.J.W.F.; Menzler, H.P.; Wagner, F.; Schissel, D.P.; DeBoo, J.C.; Kaye, S.M.

    1992-07-01

    The design of future tokamaks, which are supposed to reach ignition with the H-mode, requires a reliable scaling expression for the H-mode energy confinement time. In the present work, an H-mode scaling expression for the thermal plasma energy confinement time has been developed by combining data from four existing divertor tokamaks, ASDEX, DIII-D, JET and PBX-M. The plasma conditions, which were as similar as possible to ensure a coherent set of data, were ELM-free deuterium discharges heated by deuterium neutral beam injection. By combining four tokamaks, the parametric dependence of the thermal energy confinement on the main plasma parameters, including the three main geometrical variables, was determined. (orig./WL)

  4. Critical edge parameters for H-mode transition in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Carlstrom, T.N.

    1997-11-01

    Measurements in DIII-D of edge ion and electron temperatures (T i and T e ) just prior to the transition to H-mode are presented. A fitting model based on a hyperbolic tangent function is used in the analysis. The edge temperatures are observed to increase during the L-phase with the application of auxiliary heating. The temperature rise is small if the H-mode power threshold is close to the Ohmic power level in the absence of auxiliary heating and is large if the H-mode threshold is well above the Ohmic power level. The edge temperatures just prior to the transition are approximately proportional to the toroidal magnetic field Bt for the field either in the reversed or forward direction. However, for the reversed magnetic field, the temperatures are at least a factor of two higher than for the forward direction

  5. MHD-activity in ohmic, diverted and limited H-mode plasmas in TCV

    International Nuclear Information System (INIS)

    Pochelon, A.; Anton, M.; Buehlmann, F.; Dutch, M.J.; Duval, B.P.; Hirt, A.; Hofmann, F.; Joye, B.; Lister, J.B.; Llobet, X.; Martin, Y.; Moret, J.M.; Nieswand, C.; Pietrzyk, A.Z.; Tonetti, G.; Weisen, H.

    1994-01-01

    During its first year of operation the TCV tokamak has produced a variety of plasma configurations with currents in the range 150 to 800 kA and elongations in the range of 1.0 to 2.05. Ohmic H-modes have been obtained in diverted discharges and discharges limited on the graphite tiles inner wall. After boronisation in May 1994 H-modes with line average densities up to 1.7x10 20 m -3 , corresponding to a Murakami parameter of 10, were obtained. (author) 5 figs., 2 refs

  6. Observation of inverse hysteresis in the E to H mode transitions in inductively coupled plasmas

    International Nuclear Information System (INIS)

    Lee, Min-Hyong; Chung, Chin-Wook

    2010-01-01

    An inverse hysteresis is observed during the E mode to H mode transition in low pressure argon inductively coupled plasmas. The transition is accompanied by an evolution of electron energy distribution from the bi-Maxwellian to the Maxwellian distribution. The mechanism of this inversion is not clear. However, we think that the bi-Maxwellian electron energy distribution in E mode, where the proportion of high energy electron is much higher than the Maxwellian distribution, would be one of the reasons for the observed inverse hysteresis. As the gas pressure increases, the inverse hysteresis disappears and the E to H mode transition follows the scenario of usual hysteresis.

  7. The role of MHD instabilities in the improved H-mode scenario

    International Nuclear Information System (INIS)

    Flaws, Asher

    2009-01-01

    Recently a regime of tokamak operation has been discovered, dubbed the improved H-mode scenario, which simultaneously achieves increased energy confinement and stability with respect to standard H-mode discharges. It has been suggested that magnetohydrodynamic (MHD) instabilities play some role in establishing this regime. In this thesis MHD instabilities were identified, characterised, and catalogued into a database of improved H-mode discharges in order to statistically examine their behaviour. The onset conditions of MHD instabilities were compared to existing models based on previous H-mode studies. Slight differences were found, most notably a reduced β N onset threshold for the frequently interrupted regime for neoclassical tearing modes (NTM). This reduced threshold is due to the relatively low magnetic shear of the improved H-mode regime. This study also provided a first-time estimate for the seed island size of spontaneous onset NTMs, a phenomenon characteristic of the improved H-mode scenario. Energy confinement investigations found that, although the NTM impact on confinement follows the same model applicable to other operating regimes, the improved H-mode regime acts to mitigate the impact of NTMs by limiting the saturated island sizes for NTMs with toroidal mode number n ≥ 2. Surprisingly, although a significant loss in energy confinement is observed during the sawtooth envelope, it has been found that discharges containing fishbones and low frequency sawteeth achieve higher energy confinement than those without. This suggests that fishbone and sawtooth reconnection may indeed play a role in establishing the high confinement regime. It was found that the time evolution of the central magnetic shear consistently locks in the presence of sawtooth and fishbone reconnection. Presumably this is due to the periodic redistribution of the central plasma current, an effect which is believed to help establish and maintain the characteristic current profile

  8. H-mode confinement properties close to the power threshold in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Ryter, F; Fuchs, J; Schneider, W; Sips, A; Staebler, A; Stober, J

    2008-01-01

    Confinement properties close to the H-mode power threshold are studied in the ASDEX Upgrade tokamak. The results show that good confinement can be obtained close to the threshold with Type-I ELMs. The existence of Type-I ELMs does not necessarily require the heating power to be higher than the H-Mode power threshold, but it requires collisionality to be low enough. At higher collisionality Type-III ELMs replace the Type-I ELMs and confinement time is reduced by about 20%

  9. The role of MHD instabilities in the improved H-mode scenario

    Energy Technology Data Exchange (ETDEWEB)

    Flaws, Asher

    2009-02-16

    Recently a regime of tokamak operation has been discovered, dubbed the improved H-mode scenario, which simultaneously achieves increased energy confinement and stability with respect to standard H-mode discharges. It has been suggested that magnetohydrodynamic (MHD) instabilities play some role in establishing this regime. In this thesis MHD instabilities were identified, characterised, and catalogued into a database of improved H-mode discharges in order to statistically examine their behaviour. The onset conditions of MHD instabilities were compared to existing models based on previous H-mode studies. Slight differences were found, most notably a reduced {beta}{sub N} onset threshold for the frequently interrupted regime for neoclassical tearing modes (NTM). This reduced threshold is due to the relatively low magnetic shear of the improved H-mode regime. This study also provided a first-time estimate for the seed island size of spontaneous onset NTMs, a phenomenon characteristic of the improved H-mode scenario. Energy confinement investigations found that, although the NTM impact on confinement follows the same model applicable to other operating regimes, the improved H-mode regime acts to mitigate the impact of NTMs by limiting the saturated island sizes for NTMs with toroidal mode number n {>=} 2. Surprisingly, although a significant loss in energy confinement is observed during the sawtooth envelope, it has been found that discharges containing fishbones and low frequency sawteeth achieve higher energy confinement than those without. This suggests that fishbone and sawtooth reconnection may indeed play a role in establishing the high confinement regime. It was found that the time evolution of the central magnetic shear consistently locks in the presence of sawtooth and fishbone reconnection. Presumably this is due to the periodic redistribution of the central plasma current, an effect which is believed to help establish and maintain the characteristic current

  10. High performance H-mode plasmas at densities above the Greenwald limit

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Osborne, T.H.; Leonard, A.W.

    2001-01-01

    Densities up to 40 percent above the Greenwald limit are reproducibly achieved in high confinement (H ITER89p =2) ELMing H-mode discharges. Simultaneous gas fueling and divertor pumping were used to obtain these results. Confinement of these discharges, similar to moderate density H-mode, is characterized by a stiff temperature profile, and therefore sensitive to the density profile. A particle transport model is presented that explains the roles of divertor pumping and geometry for access to high densities. Energy loss per ELM at high density is a factor of five lower than predictions of an earlier scaling, based on data from lower density discharges. (author)

  11. PREFACE: 11th IAEA Technical Meeting on H-mode Physics and Transport Barriers

    Science.gov (United States)

    Takizuka, Tomonori

    2008-07-01

    This volume of Journal of Physics: Conference Series contains papers based on invited talks and contributed posters presented at the 11th IAEA Technical Meeting on H-mode Physics and Transport Barriers. This meeting was held at the Tsukuba International Congress Center in Tsukuba, Japan, on 26-28 September 2007, and was organized jointly by the Japan Atomic Energy Agency and the University of Tsukuba. The previous ten meetings in this series were held in San Diego (USA) 1987, Gut Ising (Germany) 1989, Abingdon (UK) 1991, Naka (Japan) 1993, Princeton (USA) 1995, Kloster Seeon (Germany) 1997, Oxford (UK) 1999, Toki (Japan) 2001, San Diego (USA) 2003, and St Petersburg (Russia) 2005. The purpose of the eleventh meeting was to present and discuss new results on H-mode (edge transport barrier, ETB) and internal transport barrier, ITB, experiments, theory and modeling in magnetic fusion research. It was expected that contributions give new and improved insights into the physics mechanisms behind high confinement modes of H-mode and ITBs. Ultimately, this research should lead to improved projections for ITER. As has been the tradition at the recent meetings of this series, the program was subdivided into six topics. The topics selected for the eleventh meeting were: H-mode transition and the pedestal-width Dynamics in ETB: ELM threshold, non-linear evolution and suppression, etc Transport relations of various quantities including turbulence in plasmas with ITB: rotation physics is especially highlighted Transport barriers in non-axisymmetric magnetic fields Theory and simulation on transport barriers Projections of transport barrier physics to ITER For each topic there was an invited talk presenting an overview of the topic, based on contributions to the meeting and on recently published external results. The six invited talks were: A Leonard (GA, USA): Progress in characterization of the H-mode pedestal and L-H transition N Oyama (JAEA, Japan): Progress and issues in

  12. Compact torus compression of microwaves

    International Nuclear Information System (INIS)

    Hewett, D.W.; Langdon, A.B.

    1985-01-01

    The possibility that a compact torus (CT) might be accelerated to large velocities has been suggested by Hartman and Hammer. If this is feasible one application of these moving CTs might be to compress microwaves. The proposed mechanism is that a coaxial vacuum region in front of a CT is prefilled with a number of normal electromagnetic modes on which the CT impinges. A crucial assumption of this proposal is that the CT excludes the microwaves and therefore compresses them. Should the microwaves penetrate the CT, compression efficiency is diminished and significant CT heating results. MFE applications in the same parameters regime have found electromagnetic radiation capable of penetrating, heating, and driving currents. We report here a cursory investigation of rf penetration using a 1-D version of a direct implicit PIC code

  13. Studies of accelerated compact toruses

    International Nuclear Information System (INIS)

    Hartman, C.W.; Eddleman, J.; Hammer, J.H.

    1983-01-01

    In an earlier publication we considered acceleration of plasma rings (Compact Torus). Several possible accelerator configurations were suggested and the possibility of focusing the accelerated rings was discussed. In this paper we consider one scheme, acceleration of a ring between coaxial electrodes by a B/sub theta/ field as in a coaxial rail-gun. If the electrodes are conical, a ring accelerated towards the apex of the cone undergoes self-similar compression (focusing) during acceleration. Because the allowable acceleration force, F/sub a/ = kappaU/sub m//R where (kappa - 2 , the accelerating distance for conical electrodes is considerably shortened over that required for coaxial electrodes. In either case, however, since the accelerating flux can expand as the ring moves, most of the accelerating field energy can be converted into kinetic energy of the ring leading to high efficiency

  14. Methanol in the L1551 Circumbinary Torus

    OpenAIRE

    White, Glenn J.; Fridlund, C. W. M.; Bergman, P.; Beardsmore, A.; Liseau, Rene; Phillips, R. R.

    2006-01-01

    We report observations of gaseous methanol in an edge-on torus surrounding the young stellar object L1551 IRS5. The peaks in the torus are separated by ~ 10,000 AU from L1551 IRS5, and contain ~ 0.03 earth masses of cold methanol. We infer that the methanol abundance increases in the outer part of the torus, probably as a result of methanol evaporation from dust grain surfaces heated by the shock luminosity associated with the shocks associated with the jets of an externally located x-ray sou...

  15. Quiescent H-mode operation using torque from non-axisymmetric, non-resonant magnetic fields

    International Nuclear Information System (INIS)

    Burrell, K.H.; Garofalo, A.M.; Osborne, T.H.; Snyder, P.B.; Solomon, W.M.; Park, J.-K.; Fenstermacher, M.E.; Orlov, D.M.

    2013-01-01

    Quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas including ITER. Using magnetic torque from n = 3 fields to replace counter-I p torque from neutral beam injection, we have achieved long duration, counter-rotating QH-mode operation with neutral beam injection (NBI) torque ranging continuously from counter-I p up to co-I p values of about 1 N m. This co-I p torque is about 3 times the scaled torque that ITER will have. This range also includes operation at zero net NBI torque, applicable to rf wave heated plasmas. These n = 3 fields have been created using coils either inside or, most recently, outside the toroidal coils. Experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values ν ped * ∼0.08, β T ped ∼ 1%$ and β N = 2. Discharges have confinement quality H 98y2 = 1.3, exceeding the value required for ITER. Initial work with low q 95 = 3.4 QH-mode plasmas transiently reached fusion gain values of G = β N H 89 /q 95 2 =0.4, which is the desired value for ITER; the limits on G have not yet been established. This paper also includes the most recent results on QH-mode plasmas run without n = 3 fields and with co-I p NBI; these shots exhibit co-I p plasma rotation and require NBI torque ⩾2 N m. The QH-mode work to date has made significant contact with theory. The importance of edge rotational shear is consistent with peeling–ballooning mode theory. We have seen qualitative and quantitative agreement with the predicted torque from neoclassical toroidal viscosity. (paper)

  16. Effect of variation in equilibrium shape on ELMing H-mode performance in DIII-D diverted plasmas

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Osborne, T.H.; Petrie, T.W.

    2001-01-01

    The changes in the performance of the core, pedestal, scrape-off-layer (SOL), and divertor plasmas as a result of changes in triangularity, δ, up/down magnetic balance, and secondary divertor volume were examined in shape variation experiments using ELMing H mode plasmas on DIII-D. In moderate density, unpumped plasmas, high δ∼0.7 increased the energy in the H mode pedestal and the global energy confinement of the core, primarily due to an increase in the margin by which the edge pressure gradient exceeded the value which would have been expected had it been limited by infinite-n ideal ballooning modes. In addition, a nearly balanced double-null (DN) shape was effective for sharing the peak heat flux in the divertor in these attached plasmas. For detached plasmas good heat flux sharing was obtained for a substantial range of unbalanced DN shapes. Finally, the presence of a second X-point in unbalanced DN shapes did not degrade the plasma performance if it was sufficiently far inside the vacuum vessel. These results indicate that a high δ unbalanced DN shape has some advantages over a single null shape for future high power tokamak operation. (author)

  17. Ion thermal conductivity and convective energy transport in JET hot-ion regimes and H-modes

    International Nuclear Information System (INIS)

    Tibone, F.; Balet, B.; Cordey, J.G.

    1989-01-01

    Local transport in a recent series of JET experiments has been studied using interpretive codes. Auxiliary heating, mainly via neutral beam injection, was applied on low-density target plasmas confined in the double-null X-point configuration. This has produced two-component plasmas with high ion temperature and neutron yield and, above a threshold density, H-modes characterised by peak density and power deposition profiles. H-mode confinement was also obtained for the first time with 25 MW auxiliary power, of which 10 MW was from ion cyclotron resonance heating. We have used profile measurements of electron temperature T e from electron cyclotron emission and LIDAR Thomson scattering, ion temperature T i from charge-exchange recombination spectroscopy (during NBI), electron density n e from LIDAR and Abel-inverted interferometer measurements. Only sparse information is, however, available to date concerning radial profiles of effective ionic charge and radiation losses. Deuterium depletion due to high impurity levels is an important effect in these discharges, and our interpretation of thermal ion energy content, neutron yield and ion particle fluxes needs to be confirmed using measured Z eff -profiles. (author) 4 refs., 4 figs

  18. 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.

  19. Poloidal rotation and the evolution of H-mode and VH-mode profiles

    International Nuclear Information System (INIS)

    Hinton, F.L.; Staebler, G.M.; Kim, Y.B.

    1993-12-01

    The physics which determines poloidal rotation, and its role in the development of profiles during H- and VH-modes, is discussed. A simple phenomenological transport model, which incorporates the rvec E x rvec B flow shear suppression of turbulence, is shown to predict profile evolution similar to that observed experimentally during H-mode and VH-mode

  20. Ubiquity of non-diffusive momentum transport in JET H-modes

    NARCIS (Netherlands)

    Weisen, H.; Camenen, Y.; Salmi, A.; Versloot, T. W.; de Vries, P. C.; Maslov, M.; Tala, T.; Beurskens, M.; Giroud, C.; JET-EFDA Contributors,

    2012-01-01

    A broad survey of the experimental database of neutral beam heated baseline H-modes and hybrid scenarios in the JET tokamak has established the ubiquity of non-diffusive momentum transport mechanisms in rotating plasmas. As a result of their presence, the normalized angular frequency gradient R

  1. New Edge Coherent Mode Providing Continuous Transport in Long Pulse H-mode Plasmas

    DEFF Research Database (Denmark)

    Wang, H.Q.; Xu, G.S.; Wan, B.N.

    2014-01-01

    An electrostatic coherent mode near the electron diamagnetic frequency (20–90 kHz) is observed in the steep-gradient pedestal region of long pulse H-mode plasmas in the Experimental Advanced Super-conducting Tokamak, using a newly developed dual gas-puff-imaging system and diamond-coated reciproc...

  2. Comparison of fusion alpha performance in JET advanced scenario and H-mode plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Asunta, O; Kurki-Suonio, T; Tala, T; Sipilae, S; Salomaa, R [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom)], E-mail: Otto.Asunta@tkk.fi

    2008-12-15

    Currently, plasmas with internal transport barriers (ITBs) appear the most likely candidates for steady-state scenarios for future fusion reactors. In such plasmas, the broad hot and dense region in the plasma core leads to high fusion gain, while the cool edge protects the integrity of the first wall. Economically desirable large bootstrap current fraction and low inductive current drive may, however, lead to degraded fast ion confinement. In this work the confinement and heating profile of fusion alphas were compared between H-mode and ITB plasmas in realistic JET geometry. The work was carried out using the Monte Carlo-based guiding-center-following code ASCOT. For the same plasma current, the ITB discharges were found to produce four to eight times more fusion power than a comparable ELMy H-mode discharge. Unfortunately, also the alpha particle losses were larger ({approx}16%) compared with the H-mode discharge (7%). In the H-mode discharges, alpha power was deposited to the plasma symmetrically around the magnetic axis, whereas in the current-hole discharge, the power was spread out to a larger volume in the plasma center. This was due to wider particle orbits, and the magnetic structure allowing for a broader hot region in the centre.

  3. The Effect of Plasma Shape on H-Mode Pedestal Characteristics on DIII-D

    International Nuclear Information System (INIS)

    T.H. Osborne; J.R. Ferron; R.J. Groebner; L.L. Lao; A.W. Leonard; R. Maingi; R.L. Miller; A.D. Turnbull; M.R. Wade; J.G. Watkins

    1999-01-01

    The characteristics of the H-mode are studied in discharges with varying triangularity and squareness. The pressure at the top of the H-mode pedestal increases strongly with triangularity primarily due to an increase in the margin by which the edge pressure gradient exceeds the ideal ballooning mode first stability limit. Two models are considered for how the edge may exceed the ballooning mode limit. In one model [1], access to the ballooning mode second stable regime allows the edge pressure gradient and associated bootstrap current to continue to increase until an edge localized, low toroidal mode number, ideal kink mode is destabilized. In the second model [2], the finite width of the H-mode transport barrier, and diamagnetic effects raise the pressure gradient limit above the ballooning mode limit. We observe a weak inverse dependence of the width of the H-mode transport barrier, Δ, on triangularity relative to the previously obtained [3] scaling Δ ∞ (β P PED ) 1/2 . The energy loss for Type I ELMs increases with triangularity in proportion to the pedestal energy increase. The temperature profile is found to respond stiffly to changes in T PED at low temperature, while at high temperature the response is additive. The response of the density profile is also found to play a role in the response of the total stored energy to changes in the W PED

  4. CORRELATION OF H-MODE BARRIER WIDTH AND NEUTRAL PENETRATION LENGTH

    International Nuclear Information System (INIS)

    GROEBNER, R.J.; MAHDAVI, M.A.; LEONARD, A.W.; OSBORNE, T.H.; WOLF, N.S.; PORTER, G.D.; STANGEBY, P.C.; BROOKS, N.H.; COLCHIN, R.J.; HEIDBRINK, W.W.; LUCE, T.C.; MCKEE, G.R.; OWEN, L.W.; WANG, G.; WHYTE, D.G.

    2002-01-01

    OAK A271 CORRELATION OF H-MODE BARRIER WIDTH AND NEUTRAL PENETRATION LENGTH. Pedestal studies in DIII-D find a good correlation between the width of the H-mode density barrier and the neutral penetration length. These results are obtained by comparing experimental density profiles to the predictions of an analytic model for the profile, obtained from the particle continuity equations for electrons and deuterium atoms. In its range of validity (edge temperature between 40-500 eV), the analytic model quantitatively predicts the observed decrease of the width as the pedestal density increases, the observed strong increase of the gradient of the density as the pedestal density increases and the observation that L-mode and H-mode profiles with the same pedestal density have very similar shapes. The width of the density barrier, measured from the edge of the electron temperature barrier, is the lower limit for the observed width of the temperature barrier. These results support the hypothesis that particle fueling provides the dominant control for the size of the H-mode transport barrier

  5. CORRELATION OF H-MODE BARRIER WIDTH AND NEUTRAL PENTRATION LENGTH

    International Nuclear Information System (INIS)

    GROEBNER, R.J.; MAHDAVI, M.A.; LEONARD, A.W.; OSBORNE, T.H.; WOLF, N.S.; PORTER, G.D.; STANGEBY, P.C.; BROOKS, N.H.; COLCHIN, R.J.; HEIDBRINK, W.W.; LUCE, T.C.; MCKEE, G.R.; OWEN, L.W.; WANG, G.; WHYTE, D.G.

    2002-01-01

    OAK A271 CORRELATION OF H-MODE BARRIER WIDTH AND NEUTRAL PENTRATION LENGTH. Pedestal studies in DIII-D find a good correlation between the width of the region of steep gradient in the H-mode density and the neutral penetration length. These results are obtained by comparing experimental density profiles to the predictions of an analytic model for the profile, obtained from the particle continuity equations for electrons and deuterium atoms. In its range of validity (edge temperature between 40-500 eV), the analytic model quantitatively predicts the observed decrease of the width as the pedestal density increases, the observed strong increase of the gradient of the density as the pedestal density increases and the observation that L-mode and H-mode profiles with the same pedestal density have very similar shapes. The width of the density barrier, measured from the edge of the electron temperature barrier, is the lower limit for the observed width of the temperature barrier. These results support the hypothesis that particle fueling provides a dominant control for the size of the H-mode transport barrier

  6. Progress in qualifying the edge physics of the H-mode regime in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Baker, D.R.; Boedo, J.A.

    2001-01-01

    Edge conditions in DIII-D are being quantified in order to provide insight into the physics of the H-mode regime. Electron temperature is not the key parameter that controls the L-H transition. Gradients of edge temperature and pressure are much more promising candidates for such parameters. The quality of H-mode confinement is strongly correlated with the height of the H-mode pedestal for the pressure. The gradient of the pressure appears to be controlled by MHD modes, in particular by kink-ballooning modes with finite mode number n. For a wide variety of discharges, the width of the barrier is well described with a relationship that is proportional to (β p ped ) 1/2 . An attractive regime of confinement has been discovered which provides steady-state operation with no ELMs, low impurity content and normal H-mode confinement. A coherent edge MHD-mode evidently provides adequate particle transport to control the plasma density and impurity content while permitting the pressure pedestal to remain almost identical to that observed in ELMing discharges. (author)

  7. Radiative type-III ELMy H-mode in all-tungsten ASDEX Upgrade

    NARCIS (Netherlands)

    Rapp, J.; Kallenbach, A.; Neu, R.; Eich, T.; Fischer, R.; Herrmann, A.; Potzel, S.; van Rooij, G. J.; Zielinski, J. J.; ASDEX Upgrade team,

    2012-01-01

    The type-III ELMy H-mode might be the solution for an integrated ITER operation scenario fulfilling the fusion power amplification factor (output fusion power to input heating power) of Q = 10 with simultaneous acceptable steady-state and transient power loads to the plasma-facing components. This

  8. L to H mode transitions and associated phenomena in divertor tokamaks

    International Nuclear Information System (INIS)

    Punjabi, A.

    1990-09-01

    This is the final report for the research project titled ''L to H Mode Transitions and Associated Phenomena in Divertor Tokamaks.'' The period covered by this project is the fiscal year 1990. This report covers the development of Advanced Two Chamber Model

  9. Tool path in torus tool CNC machining

    Directory of Open Access Journals (Sweden)

    XU Ying

    2016-10-01

    Full Text Available This paper is about tool path in torus tool CNC machining.The mathematical model of torus tool is established.The tool path planning algorithm is determined through calculation of the cutter location,boundary discretization,calculation of adjacent tool path and so on,according to the conversion formula,the cutter contact point will be converted to the cutter location point and then these points fit a toolpath.Lastly,the path planning algorithm is implemented by using Matlab programming.The cutter location points for torus tool are calculated by Matlab,and then fit these points to a toolpath.While using UG software,another tool path of free surface is simulated of the same data.It is drew compared the two tool paths that using torus tool is more efficient.

  10. ELM triggering conditions for the integrated modeling of H-mode plasmas

    International Nuclear Information System (INIS)

    Pankin, A.Y.; Schnack, D.D.; Bateman, G.; Kritz, A.H.; Brennan, D.P.; Snyder, P.B.; Voitsekhovitch, I.; Kruger, S.; Janeschitz, G.; Onjun, T.; Pacher, G.W.; Pacher, H.D.

    2005-01-01

    Recent advances in the integrated modeling of ELMy H-mode plasmas are presented. A new model for the H-mode pedestal and for the triggering of ELMs predicts the height, width, and shape of the H-mode pedestal and the frequency and width of ELMs. The model for the pedestal and ELMs is used in the ASTRA integrated transport code to follow the time evolution of tokamak discharges from L-mode through the transition from L-mode to H-mode, with the formation of the H-mode pedestal, and, subsequently, to the triggering of ELMs. Turbulence driven by the ion temperature gradient mode, resistive ballooning mode, trapped electron mode, and electron temperature gradient mode contributes to the anomalous thermal transport at the plasma edge in this model. Formation of the pedestal and the L-H transition is the direct result of E(vector) r x B(vector) flow shear suppression of anomalous transport. The periodic ELM crashes are triggered by MHD instabilities. Two mechanisms for triggering ELMs are considered: ELMs are triggered by ballooning modes if the pressure gradient exceeds the ballooning threshold or by peeling modes if the edge current density exceeds the peeling mode threshold. The BALOO, DCON, and ELITE ideal MHD stability codes are used to derive a new parametric expression for the peeling-ballooning threshold. The new dependence for the peeling-ballooning threshold is implemented in the ASTRA transport code. Results of integrated modeling of DIII-D like discharges are presented and compared with experimental observations. The results from the ideal MHD stability codes are compared with results from the resistive MHD stability code NIMROD. (author)

  11. Torus bifurcations in multilevel converter systems

    DEFF Research Database (Denmark)

    Zhusubaliyev, Zhanybai T.; Mosekilde, Erik; Yanochkina, Olga O.

    2011-01-01

    embedded one into the other and with their basins of attraction delineated by intervening repelling tori. The paper illustrates the coexistence of three stable tori with different resonance behaviors and shows how reconstruction of these tori takes place across the borders of different dynamical regimes....... The paper also demonstrates how pairs of attracting and repelling tori emerge through border-collision torus-birth and border-collision torus-fold bifurcations. © 2011 World Scientific Publishing Company....

  12. H-mode transition physics close to DN on MAST and its applications to other tokamaks

    International Nuclear Information System (INIS)

    Meyer, H.

    2004-01-01

    Full text: ELMy H-mode is the base-line operating scenario for the next step fusion device ITER. To improve active and passive pedestal control a deeper understanding of H- mode physics is desirable. MAST contributes towards this understanding with good edge diagnostics, and by accessing extreme parameter regimes. The first inter-machine comparisons with respect to the influence of the magnetic topology on the power threshold with ASDEX-Upgrade and NSTX reveal a reduction of the power threshold in true double null (C-DN) configuration opening new operation regimes in both devices. The 30% reduction in threshold power close to C-DN observed on ASDEX-Upgrade, though significant, is less than the factor of two or more observed in both large spherical tokamaks, MAST and NSTX. This points towards the importance of field line curvature for this effect. The power thresholds measured in C-DN on MAST and NSTX are very similar. Despite this strong effect on the power threshold, changes in most edge parameters in L-mode due to the different magnetic configurations are small. However, significant changes are seen in the toroidal impurity flow velocity, related to the radial electric field, and in the scrape-off-layer temperature decay length at the high field side. The statistical comparison of MAST data with various H-mode theories suggests that different instabilities need to be stabilised at different spatial positions in the region where the pedestal forms to access H-mode. Pedestal temperatures observed on MAST are two to five times lower than in MAST equivalent discharges at ASDEX-Upgrade. However, the pedestal densities are similar. The differences in L-mode are less significant. The usual DN operating regime with co current NBI in MAST has been extended to include single null (SN) configurations, to provide more direct comparisons with conventional tokamaks. The plasma edge in SN on MAST is more stable to ELMs and the typical type-III ELMs, often observed in C-DN, are

  13. Energy confinement and transport of H-mode plasmas in tokamak

    International Nuclear Information System (INIS)

    Urano, Hajime

    2005-02-01

    A characteristic feature of the high-confinement (H-mode) regime is the formation of a transport barrier near the plasma edge, where steepening of the density and temperature gradients is observed. The H-mode is expected to be a standard operation mode in a next-step fusion experimental reactor, called ITER-the International Thermonuclear Experimental Reactor. However, energy confinement in the H-mode has been observed to degrade with increasing density. This is a critical constraint for the operation domain in the ITER. Investigation of the main cause of confinement degradation is an urgent issue in the ITER Physics Research and Development Activity. A key element for solving this problem is investigation of the energy confinement and transport properties of H-mode plasmas. However, the influence of the plasma boundary characterized by the transport barrier in H-modes on the energy transport of the plasma core has not been examined sufficiently in tokamak research. The aim of this study is therefore to investigate the energy confinement properties of H-modes in a variety of density, plasma shape, seed impurity concentration, and conductive heat flux in the plasma core using the experimental results obtained in the JT-60U tokamak of Japan Atomic Energy Research Institute. Comparison of the H-mode confinement properties with those of other tokamaks using an international multi-machine database for extrapolation to the next step device was also one of the main subjects in this study. Density dependence of the energy confinement properties has been examined systematically by separating the thermal stored energy into the H-mode pedestal component determined by MHD stability called the Edge Localized Modes (ELMs) and the core component governed by gyro-Bohm-like transport. It has been found that the pedestal pressure imposed by the destabilization of ELM activities led to a reduction in the pedestal temperature with increasing density. The core temperature for each

  14. Formation of an internal transport barrier in the ohmic H-mode in the TUMAN-3M tokamak

    International Nuclear Information System (INIS)

    Andrejko, M.V.; Askinazi, L.G.; Golant, V.E.; Zhubr, N.A.; Kornev, V.A.; Krikunov, S.V.; Lebedev, S.V.; Levin, L.S.; Razdobarin, G.T.; Rozhdestvensky, V.V.; Smirnov, A.I.; Tukachinsky, A.S.; Yaroshevich, S.P.

    2000-01-01

    In experiments on studying the ohmic H-mode in the TUMAN-3M tokamak, it is found that, in high-current (I p ∼ 120-170 kA) discharges, a region with high electron-temperature and density gradients is formed in the plasma core. In this case, the energy confinement time τ E attains 9-18 ms, which is nearly twice as large as that predicted by the ELM-free ITER-93H scaling. This is evidence that the internal transport barrier in a plasma can exist without auxiliary heating. Calculations of the effective thermal diffusivity by the ASTRA transport code demonstrate a strong suppression of heat transport in the region where the temperature and density gradients are high

  15. 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.

  16. Thomson scattering on ELMO Bumpy Torus

    International Nuclear Information System (INIS)

    Cobble, J.A.

    1985-04-01

    Below 10 12 cm -3 density, a Thomson scattering experiment is an exacting task. Aside from the low signal level, the core plasma in this instance is bathed in high-energy x rays, surrounded by a glowing molecular surface plasma, and heated steady state by microwaves. This means that the noise level from radiation is high and the environment is extremely harsh-so harsh that much effort is required to overcome system damage. In spite of this, the ELMO Bumpy Torus (EBT) system has proven itself capable of providing reliable n/sub e/ and T/sub e/ measurements at densities as low as 2 x 10 11 cm -3 . Radial scans across 20 cm of the plasma diameter have been obtained on a routine basis, and the resulting information has been a great help in understanding confinement in the EBT plasma. The bulk electron properties are revealed as flat profiles of n/sub e/ and T/sub e/, with density ranging from 0.5 to 2.0 x 10 12 cm -3 and temperature decreasing from 100 to 20 eV as pressure in the discharge is increased at constant power. Evidence is presented for a suprathermal tail, which amounts to about 10% of the electron distribution at low pressures. The validity of this conclusion is supported by two independent sensitivity calibrations

  17. Scaling of H-mode pedestal characteristics in DIII-D and C-Mod

    International Nuclear Information System (INIS)

    Granetz, R.S.; Boivin, R.L.; Osborne, T.H.

    1999-01-01

    Since the H-mode edge pedestal effectively sets the boundary conditions for energy transport throughout the core, a better understanding of the pedestal region is necessary in order to fully predict H-mode performance. Pedestal characteristics in the DIII-D and Alcator C-Mod tokamaks are described, and scalings of the pedestal width with various plasma parameters are shown. The pedestal width in both tokamaks varies in an inverse sense with plasma current, and is independent of toroidal field. Other similarities, as well as differences, are discussed. It is also found that the pedestal widths of the various physical quantities involved (T e , T i , n e , n i ) may be different. (author)

  18. Coupling of an ICRF compact loop antenna to H-mode plasmas in DIII-D

    International Nuclear Information System (INIS)

    Mayberry, M.J.; Baity, F.W.; Hoffman, D.J.; Luxon, J.L.; Owens, T.L.; Prater, R.

    1987-01-01

    Low power coupling tests have been carried out with a prototype ICRF compact loop antenna on the DIII-D tokamak. During neutral-beam-heated L-mode discharges the antenna loading is typically R≅1-2Ω for an rf frequency of 32 MHz (B/sub T/ = 21 kG, ω = 2Ω/sub D/(0)). When a transition into the H-mode regime of improved confinement occurs, the loading drops to R≅0.5-1.0Ω. During ELMs, transient increases in loading up to several Ohms are observed. The apparent sensitivity of ICRF antenna coupling to changes in the edge plasma profiles associated with the H-mode regime could have important implications for the design of future high power systems

  19. New fluctuation phenomena in the H-mode regime of PDX tokamak plasmas

    International Nuclear Information System (INIS)

    Slusher, R.E.; Surko, C.M.; Valley, J.F.; Crowley, T.; Mazzucato, E.; McGuire, K.

    1984-05-01

    A new kind of quasi-coherent fluctuation is observed near the edge of plasmas in the PDX tokamak during H-mode operation. (The H-mode occurs in neutral beam heated divertor plasmas and is characterized by improved energy containment as well as large density and temperature gradients near the plasma edge.) These fluctuations are evidenced as VUV and density fluctuation bursts at well-defined frequencies (Δω/ω less than or equal to 0.1) in the frequency range between 50 and 180 kHz. They affect the edge temperature-density product, and therefore they may be important for understanding the relationship between the large edge density and temperature gradients and the improved energy confinement

  20. Parameter dependences of the separatrix density in nitrogen seeded ASDEX Upgrade H-mode discharges

    Science.gov (United States)

    Kallenbach, A.; Sun, H. J.; Eich, T.; Carralero, D.; Hobirk, J.; Scarabosio, A.; Siccinio, M.; ASDEX Upgrade Team; EUROfusion MST1 Team

    2018-04-01

    The upstream separatrix electron density is an important interface parameter for core performance and divertor power exhaust. It has been measured in ASDEX Upgrade H-mode discharges by means of Thomson scattering using a self-consistent estimate of the upstream electron temperature under the assumption of Spitzer-Härm electron conduction. Its dependence on various plasma parameters has been tested for different plasma conditions in H-mode. The leading parameter determining n e,sep was found to be the neutral divertor pressure, which can be considered as an engineering parameter since it is determined mainly by the gas puff rate and the pumping speed. The experimentally found parameter dependence of n e,sep, which is dominated by the divertor neutral pressure, could be approximately reconciled by 2-point modelling.

  1. Correlation of the tokamak H-mode density limit with ballooning stability at the separatrix

    Science.gov (United States)

    Eich, T.; Goldston, R. J.; Kallenbach, A.; Sieglin, B.; Sun, H. J.; ASDEX Upgrade Team; Contributors, JET

    2018-03-01

    We show for JET and ASDEX Upgrade, based on Thomson-scattering measurements, a clear correlation of the density limit of the tokamak H-mode high-confinement regime with the approach to the ideal ballooning instability threshold at the periphery of the plasma. It is shown that the MHD ballooning parameter at the separatrix position α_sep increases about linearly with the separatrix density normalized to Greenwald density, n_e, sep/n_GW for a wide range of discharge parameters in both devices. The observed operational space is found to reach at maximum n_e, sep/n_GW≈ 0.4 -0.5 at values for α_sep≈ 2 -2.5, in the range of theoretical predictions for ballooning instability. This work supports the hypothesis that the H-mode density limit may be set by ballooning stability at the separatrix.

  2. Effect of low density H-mode operation on edge and divertor plasma parameters

    International Nuclear Information System (INIS)

    Maingi, R.; Mioduszewski, P.K.; Cuthbertson, J.W.

    1994-07-01

    We present a study of the impact of H-mode operation at low density on divertor plasma parameters on the DIII-D tokamak. The line-average density in H-mode was scanned by variation of the particle exhaust rate, using the recently installed divertor cryo-condensation pump. The maximum decrease (50%) in line-average electron density was accompanied by a factor of 2 increase in the edge electron temperature, and 10% and 20% reductions in the measured core and divertor radiated power, respectively. The measured total power to the inboard divertor target increased by a factor of 3, with the major contribution coming from a factor of 5 increase in the peak heat flux very close to the inner strike point. The measured increase in power at the inboard divertor target was approximately equal to the measured decrease in core and divertor radiation

  3. Chapter 7: High-Density H-Mode Operation in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Stober, Joerg Karl; Lang, Peter Thomas; Mertens, Vitus

    2003-01-01

    Recent results are reported on the maximum achievable H-mode density and the behavior of pedestal density and central density peaking as this limit is approached. The maximum achievable H-mode density roughly scales as the Greenwald density, though a dependence on B t is clearly observed. In contrast to the stiff temperature profiles, the density profiles seem to allow more shape variation and especially with high-field-side pellet-injection, strongly peaked profiles with good confinement have been achieved. Also, spontaneous density peaking at high densities is observed in ASDEX Upgrade, which is related to the generally observed large time constants for the density profile equilibration. The equilibrated density profile shapes depend strongly on the heat-flux profile in the sense that central heating leads to significantly flatter profiles

  4. The physics of transport barrier formation in the PBX-M H-mode

    International Nuclear Information System (INIS)

    Tynan, G.R.; Schmitz, L.; Blush, L.

    1994-01-01

    Measurements of edge profiles, turbulence, and turbulent-driven transport were made inside the last-closed flux surface (LCFS) and in the scrape-off layer (SOL) of PBX-M L-mode and H-mode plasmas using a fast reciprocating Langmuir probe diagnostic. Direct measurements of the potential profile confirm the generation of a strong inward radial electric field (E r ∼ -100 V/cm) just inside the LCFS in H-mode. Density and potential fluctuations levels are reduced at the L-H transition, resulting in significantly lower turbulent transport. The reduction in turbulent transport occurs across the LCFS and SOL regions and is not localized to the region of strong radial electric field. (author)

  5. Influence of the wall material on the H-mode performance

    International Nuclear Information System (INIS)

    Itoh, K.; Itoh, S.

    1994-06-01

    Theory on the influence of the wall material on the level of the enhanced confinement in H-mode is discussed. When the high-Z material is employed as the wall, the reflection of the neutral particles causes the higher neutral particle density in the plasma. The increased neutral particles lead to the loss of the ion momentum, decrease the radial electric field and degrade the confinement improvement. (author)

  6. Coherent edge fluctuation measurements in H-mode discharges on JFT-2M

    International Nuclear Information System (INIS)

    Nagashima, Y; Shinohara, K; Hoshino, K; Ejiri, A; Tsuzuki, K; Ido, T; Uehara, K; Kawashima, H; Kamiya, K; Ogawa, H; Yamada, T; Shiraiwa, S; Ohara, S; Takase, Y; Asakura, N; Oyama, N; Fujita, T; Ide, S; Takenaga, H; Kusama, Y; Miura, Y

    2004-01-01

    Results of coherent edge fluctuation measurements using three diagnostics (a reciprocating Langmuir probe, a two channel O-mode reflectometer, and fast magnetic probes) in H-mode discharges on JFT-2M are presented. In discharges in which a high recycling steady (HRS) H-mode phase is obtained through a transient phase with slightly enhanced D α intensity, two types of coherent fluctuations are observed. The higher frequency mode (around 300 kHz) is the high frequency mode (HFM) observed in the HRS H-mode (Kamiya K et al 2003 9th IAEA Tech. Meeting H-mode Workshop Topic B-14). The lower frequency mode has a frequency of around 80 kHz. The HFM is detected by a Langmuir probe over a wide region in the SOL, as well as by the reflectometer and magnetic probes. However, the HFM is not detected by the higher frequency (38 GHz) channel of the reflectometer after the HRS transition, suggesting that the HFM is not located deeply inside the plasma. The 80 kHz mode is detected by both channels of the reflectometer and by a Langmuir probe, but not by magnetic probes, suggesting that it is an electrostatic mode. In contrast to the HFM, the 80 kHz mode is detected by the Langmuir probe only near the separatrix during the transient phase, which leads to either the HRS phase or the ELMy phase, and is similar to the fluctuations reported in Shinohara K et al (1998 J. Plasma Fusion Res. 74 607)

  7. Tungsten transport in JET H-mode plasmas in hybrid scenario, experimental observations and modelling

    Czech Academy of Sciences Publication Activity Database

    Angioni, C.; Mantica, P.; Pütterich, T.; Valisa, M.; Baruzzo, M.; Belli, A.E.; Belo, P.; Casson, F.J.; Challis, C.; Drewelow, P.; Giroud, C.; Hawkes, N.; Hender, T.C.; Hobirk, J.; Koskela, T.; Lauro Taroni, L.; Maggi, C.F.; Mlynář, Jan; Odstrčil, T.; Reinke, M.L.; Romanelli, M.

    2014-01-01

    Roč. 54, č. 8 (2014), 083028-083028 ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : heavy impurity transport * H-mode hybrid scenario * neoclassical and turbulent transport Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.062, year: 2014 http://iopscience.iop.org/0029-5515/54/8/083028/pdf/0029-5515_54_8_083028.pdf

  8. Observation of internal transport barrier in ELMy H-mode plasmas on the EAST tokamak

    Science.gov (United States)

    Yang, Y.; Gao, X.; Liu, H. Q.; Li, G. Q.; Zhang, T.; Zeng, L.; Liu, Y. K.; Wu, M. Q.; Kong, D. F.; Ming, T. F.; Han, X.; Wang, Y. M.; Zang, Q.; Lyu, B.; Li, Y. Y.; Duan, Y. M.; Zhong, F. B.; Li, K.; Xu, L. Q.; Gong, X. Z.; Sun, Y. W.; Qian, J. P.; Ding, B. J.; Liu, Z. X.; Liu, F. K.; Hu, C. D.; Xiang, N.; Liang, Y. F.; Zhang, X. D.; Wan, B. N.; Li, J. G.; Wan, Y. X.; EAST Team

    2017-08-01

    The internal transport barrier (ITB) has been obtained in ELMy H-mode plasmas by neutron beam injection and lower hybrid wave heating on the Experimental Advanced Superconducting Tokamak (EAST). The ITB structure has been observed in profiles of ion temperature, electron temperature, and electron density within ρ safety factor q(0) ˜ 1. Transport coefficients are calculated by particle balance and power balance analysis, showing an obvious reduction after the ITB formation.

  9. ELM triggering conditions for the integrated modeling of H-mode plasmas

    International Nuclear Information System (INIS)

    Pankin, A.Y.; Schnack, D.D.; Bateman, G.; Kritz, A.H.; Brennan, D.P.; Snyder, P.B.; Voitsekhovitch, I.; Kruger, S.; Janeschitz, G.; Onjun, T.; Pacher, G.W.; Pacher, H.D.

    2004-01-01

    Recent advances in the integrated modeling of ELMy (edge localized mode) H-mode plasmas are presented. A model for the H-mode pedestal and for the triggering of ELMs predicts the height, width, and shape of the H-mode pedestal and the frequency and width of ELMs. Formation of the pedestal and the L-H transition is the direct result of E r x B flow shear suppression of anomalous transport. The periodic ELM crashes are triggered by either the ballooning or peeling MHD instabilities. The BALOO, DCON, and ELITE ideal MHD stability codes are used to derive a new parametric expression for the peeling-ballooning threshold. The new dependence for the peeling-ballooning threshold is implemented in the ASTRA transport code. Results of integrated modeling of DIII-D like discharges are presented and compared with experimental observations. The results from the ideal MHD stability codes are compared with results from the resistive MHD stability code NIMROD. (authors)

  10. Edge ion dynamics in H-mode discharges in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Burrell, K.H.; Gohil, P.; Kim, J.; Seraydarian, R.P.

    1992-05-01

    The goal of this paper is to present detailed measurements of T i and E r at the plasma edge in L- and H-mode with high spatial resolution in order the study the edge ion dynamics. Of primary interest is the relationship between T i and E r and the behavior of the edge T i profile in H-mode. The principle findings are: there appears to be a threshold temperature for T i required for the transition to occur with T i at the LCFS in the range of 0.2--0.3 keV at the transition; a correlation between the edge E r profile and the edge T i profile has been observed; and values of T i of 2--3 keV within a few cm of the LCFS and of dT i /dr of up to 1 keV/cm are observed in the transport barrier in H-mode, with the scale length for T i being of the order of a poloidal gyroradius

  11. Correlation of H-mode barrier width and neutral penetration length

    International Nuclear Information System (INIS)

    Groebner, R.J.; Mahdavi, M.A.; Leonard, A.W.

    2003-01-01

    Pedestal studies in DIII-D find a good correlation between the width of the H-mode density barrier and the neutral penetration length. These results are obtained by comparing experimental density profiles to the predictions of an analytic model for the profile, obtained from the particle continuity equations for electrons and deuterium atoms. In its range of validity (edge temperature between 40-500 eV), the analytic model quantitatively predicts the observed decrease of the width as the pedestal density increases, the observed strong increase of the gradient of the density as the pedestal density increases and the observation that L-mode and H-mode profiles with the same pedestal density have very similar shapes. The width of the density barrier, measured from the edge of the electron temperature barrier, is the lower limit for the observed width of the temperature barrier. These results support the hypothesis that particle fueling provides the dominant control for the size of the H-mode transport barrier. (author)

  12. Comparison of H-mode barrier width with a model of neutral penetration length

    International Nuclear Information System (INIS)

    Groebner, R.J.; Mahdavi, M.A.; Leonard, A.W.; Osborne, T.H.; Brooks, N.S.; Wolf, N.S.; Porter, G.D.; Stangeby, P.C.; Colchin, R.J.; Owen, L.W.

    2004-01-01

    Pedestal studies in DIII-D find that the width of the region of steep gradient in the H-mode density is comparable with the neutral penetration length, as computed from a simple analytic model. This model has analytic solutions for the edge plasma and neutral density profiles, which are obtained from the coupled particle continuity equations for electrons and deuterium atoms. In its range of validity (edge temperature between 40 and 500 eV), the analytic model quantitatively predicts the observed decrease in the width as the pedestal density increases and the observed strong increase in the gradient of the density as the pedestal density increases. The model successfully predicts that L-mode and H-mode profiles with the same pedestal density have gradients that differ by less than a factor of 2. The width of the density barrier, measured from the edge of the electron temperature barrier, is the lower limit for the observed width of the temperature barrier. These results support the hypothesis that particle fuelling is an important part of the physics that determines the structure of the H-mode transport barrier. (author)

  13. Operational limits of high density H-modes in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Mertens, V.; Borrass, K.; Kaufmann, M.; Lang, P.T.; Lang, R.; Mueller, H.W.; Neuhauser, J.; Schneider, R.; Schweinzer, J.; Suttrop, W.

    2001-01-01

    Systematic investigations of H-mode density limit (H→L-mode back transition) plasmas with gas fuelling and alternatively with additional pellet injection from the magnetic high-field-side HFS are being performed in the new closed divertor configuration DV-II. The resulting database covering a wide range of the externally controllable plasma parameters I p , B t and P heat confirms that the H-mode threshold power exceeds the generally accepted prediction P L→H heat ∝B-bar t dramatically when one approaches Greenwald densities. Additionally, in contrast to the Greenwald scaling a moderate B t -dependence of the H-mode density limit is found. The limit is observed to coincide with divertor detachment and a strong increase of the edge thermal transport, which has, however, no detrimental effect on global τ E . The pellet injection scheme from the magnetic high-field-side HFS, developed recently on ASDEX Upgrade, leads to fast particle drifts which are, contrary to the standard injection from the low-field-side, directed into the plasma core. This improves markedly the pellet particle fuelling efficiency. The responsible physical mechanism, the diamagnetic particle drift of the pellet ablatant was successfully verified recently. Other increased particle losses on respectively different time scales after the ablation process, however, still persist. Generally, a clear gain in achievable density and plasma stored energy is achieved with stationary HFS pellet injection compared to gas-puffing. (author)

  14. Density fluctuation measurements via reflectometry on DIII-D during L- and H-mode operation

    International Nuclear Information System (INIS)

    Doyle, E.J.; Lehecka, T.; Luhmann, N.C. Jr.; Peebles, W.A.; Philipona, R.

    1990-01-01

    The unique ability of reflectometers to provide radial density fluctuation measurements with high spatial resolution (of the order of ≤ centimeters, is ideally suited to the study of the edge plasma modifications associated with H-mode operation. Consequently, attention has been focused on the study of these phenomena since an improved understanding of the physics of H-mode plasmas is essential if a predictive capability for machine performance is to be developed. In addition, DIII-D is ideally suited for such studies since it is a major device noted for its robust H-mode operation and excellent basic plasma profile diagnostic information. The reflectometer system normally used for fluctuation studies is an O-mode, homodyne, system utilizing 7 discrete channels spanning 15-75 GHz, with corresponding critical densities of 2.8x10 18 to 7x10 19 m -3 . The Gunn diode sources in this system are only narrowly tunable in frequency, so the critical densities are essentially fixed. An X-mode system, utilizing a frequency tunable BWO source, has also been used to obtain fluctuation data, and in particular, to 'fill in the gaps' between the discrete O-mode channels. (author) 12 refs., 5 figs

  15. A two term model of the confinement in Elmy H-modes using the global confinement and pedestal databases

    International Nuclear Information System (INIS)

    2003-01-01

    Two different physical models of the H-mode pedestal are tested against the joint pedestal-core database. These models are then combined with models for the core and shown to give a good fit to the ELMy H-mode database. Predictions are made for the next step tokamaks ITER and FIRE. (author)

  16. The quiescent H-mode regime for high performance edge localized mode-stable operation in future burning plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Garofalo, A. M., E-mail: garofalo@fusion.gat.com; Burrell, K. H.; Meneghini, O.; Osborne, T. H.; Paz-Soldan, C.; Smith, S. P.; Snyder, P. B.; Turnbull, A. D. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Eldon, D.; Grierson, B. A.; Solomon, W. M. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States); Hanson, J. M. [Columbia University, 2960 Broadway, New York, New York 10027-6900 (United States); Holland, C. [University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Huijsmans, G. T. A.; Liu, F.; Loarte, A. [ITER Organization, Route de Vinon sur Verdon, 13067 St Paul Lez Durance (France); Zeng, L. [University of California Los Angeles, P.O. Box 957099, Los Angeles, California 90095-7099 (United States)

    2015-05-15

    For the first time, DIII-D experiments have achieved stationary quiescent H-mode (QH-mode) operation for many energy confinement times at simultaneous ITER-relevant values of beta, confinement, and safety factor, in an ITER-like shape. QH-mode provides excellent energy confinement, even at very low plasma rotation, while operating without edge localized modes (ELMs) and with strong impurity transport via the benign edge harmonic oscillation (EHO). By tailoring the plasma shape to improve the edge stability, the QH-mode operating space has also been extended to densities exceeding 80% of the Greenwald limit, overcoming the long-standing low-density limit of QH-mode operation. In the theory, the density range over which the plasma encounters the kink-peeling boundary widens as the plasma cross-section shaping is increased, thus increasing the QH-mode density threshold. The DIII-D results are in excellent agreement with these predictions, and nonlinear magnetohydrodynamic analysis of reconstructed QH-mode equilibria shows unstable low n kink-peeling modes growing to a saturated level, consistent with the theoretical picture of the EHO. Furthermore, high density operation in the QH-mode regime has opened a path to a new, previously predicted region of parameter space, named “Super H-mode” because it is characterized by very high pedestals that can be more than a factor of two above the peeling-ballooning stability limit for similar ELMing H-mode discharges at the same density.

  17. Reactor assessments of advanced bumpy torus configurations

    International Nuclear Information System (INIS)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1984-02-01

    Recently, several innovative approaches were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator - snakey torus). Preliminary evaluations of reactor implications of each approach have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties deduced from provisional configurations that implement the approach but are not necessarily optimized. Further optimization is needed in all cases to evaluate the full potential of each approach. Results of these studies indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past

  18. Multicast Performance Analysis for High-Speed Torus Networks

    National Research Council Canada - National Science Library

    Oral, S; George, A

    2002-01-01

    ... for unicast-based and path-based multicast communication on high-speed torus networks. Software-based multicast performance results of selected algorithms on a 16-node Scalable Coherent Interface (SCI) torus are given...

  19. Magnetic perturbation experiments on MAST L- and H-mode plasmas using internal coils

    Czech Academy of Sciences Publication Activity Database

    Kirk, A.; Liu, Y.Q.; Nardon, E.; Tamain, P.; Cahyna, Pavel; Chapman, I.; Denner, P.; Meyer, H.; Mordijck, S.; Temple, D.

    2011-01-01

    Roč. 53, č. 6 (2011), 065011-065011 ISSN 0741-3335 R&D Projects: GA ČR GAP205/11/2341 Institutional research plan: CEZ:AV0Z20430508 Keywords : Resonant magnetic perturbations * L-H transition * spherical tokamaks * edge localized modes Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.425, year: 2011 http://dx.doi.org/10.1088/0741-3335/53/6/065011

  20. Various semiclassical limits of torus conformal blocks

    Energy Technology Data Exchange (ETDEWEB)

    Alkalaev, Konstantin [I.E. Tamm Department of Theoretical Physics, P.N. Lebedev Physical Institute,Leninsky ave. 53, Moscow, 119991 (Russian Federation); Department of General and Applied Physics, Moscow Institute of Physics and Technology,Institutskiy per. 7, Dolgoprudnyi, Moscow region, 141700 (Russian Federation); Geiko, Roman [Mathematics Department, National Research University Higher School of Economics,Usacheva str. 6, Moscow, 119048 (Russian Federation); Rappoport, Vladimir [I.E. Tamm Department of Theoretical Physics, P.N. Lebedev Physical Institute,Leninsky ave. 53, Moscow, 119991 (Russian Federation); Department of Quantum Physics, Institute for Information Transmission Problems,Bolshoy Karetny per. 19, Moscow, 127994 (Russian Federation)

    2017-04-12

    We study four types of one-point torus blocks arising in the large central charge regime. There are the global block, the light block, the heavy-light block, and the linearized classical block, according to different regimes of conformal dimensions. It is shown that the blocks are not independent being connected to each other by various links. We find that the global, light, and heavy-light blocks correspond to three different contractions of the Virasoro algebra. Also, we formulate the c-recursive representation of the one-point torus blocks which is relevant in the semiclassical approximation.

  1. Pro-torus actions on Poincaré duality spaces

    Indian Academy of Sciences (India)

    duality spaces, Borel's dimension formula and topological splitting principle to local weights, hold if 'torus' is replaced by 'pro-torus'. Keywords. Pro-torus; Poincaré duality space; local weight. 1. Introduction. In the theory of linear representations of compact connected Lie groups, the crucial first step is restriction to the ...

  2. Cirugía de torus mandibular

    Directory of Open Access Journals (Sweden)

    Manuel Ramon Osorio Castillo

    2014-06-01

    Full Text Available ResumenLos huesos maxilares no son ajenos a las patologías que se pueden presentar en el sistema esquelético. Algunas de esas condiciones y patologías son singulares por sus características clínicas, su distribución y prevalencia. Los torus palatinos, los torus mandibulares (TM y las exostosis de los maxilares son un claro ejemplo de ellos. Hasta la presente existen ideas especulativas acerca de su etiopatogenia, de los factores asociados, de su incidencia y prevalencia, de su necesidad de tratamiento, lo que puede crear confusión entre los clínicos tanto en diagnóstico como en el manejo.El torus como tumor óseo benigno puede localizarse en el maxilar a nivel del paladar, o en la mandíbula a nivel de las tablas internas; o puede aparecer en cualquier parte del esqueleto. El TM es una exostosis o crecimiento óseo en la superficie lingual de la mandíbula. Este crecimiento ocurre generalmente cerca de la línea milohioidea, opuesto a los premolares, pero se puede extender del canino al primer molar. La mucosa que los recubre tiende a ser fina y no tolera por lo general las fuerzas de las prótesis que se colocan encima de ellos. La incidencia del torus de la mandíbula es baja en el 6% a 12.5% entre caucásicos y en los habitantes de la llanura africana. De manera contraria, algunos autores reportan una prevalencia mucho más elevada en la Costa Atlántica Colombiana.Se presenta el caso de un paciente con torus mandibulares bilaterales, con muchos años de crecimiento, hasta que por situaciones tanto fonéticas como de ulceraciones repetitivas decidió someterse al acto quirúrgico de forma bilateral. Se presentan algunas consideraciones para el manejo de esta. (Duazary 2008; 111-114AbstractThe jawbone is not a strange to the pathologies that can occur in the skeletal system. Some of these terms and conditions are unique for their clinical features, distribution and prevalence. The torus palate, jawbone torus (TM in spanish and

  3. H-mode regimes and observators of central toroidal rotation in Alcator C-Mod

    International Nuclear Information System (INIS)

    Greenwald, M.; Rice, J.; Boivin, R.

    1999-01-01

    The Enhanced D α or EDA H-mode regime in Alcator C-Mod has been investigated and compared in detail to ELM-free plasmas. (In this paper, ELM-free will refer to discharges with no type I ELMs and with no sign of EDA, though technically, most EDA plasmas are ELM-free as well.) EDA discharges have only slightly lower energy confinement than comparable ELM-free ones, but show markedly reduced impurity confinement. Thus EDA discharges do not accumulate impurities and typically have a lower fraction of radiated power. EDA plasmas are seen to be more likely at low plasma current (q > 3.7 - 4), for moderate plasma shaping (0.35 - 0.55), and for high neutral pressures. No obvious trends were observed with input power or pressure (β). In both H-mode regimes, and in ICRF heated L-modes, central impurity toroidal rotation has been deduced, from the Doppler shifts of argon x-ray lines. Rotation velocities up to 1.3 x 10 5 m/s in the co-current direction have been observed in H-mode discharges that had no direct momentum input. There is a strong correlation between the increase in the central impurity rotation velocity and the increase in the plasma stored energy, induced by ICRF heating. In otherwise similar discharges with the same stored energy increase, plasmas with lower current rotate faster. The ion pressure gradient is an unimportant contributor to the central impurity rotation and the presence of a substantial core radial electric field is inferred during the ICRF pulse. An inward shift of ions induced by ICRF waves could give rise to a non-ambipolar electric field in the plasma core. Comparisons with a neo-classical ion orbit shift model show good agreement with the observations, both in magnitude, and in the scaling with plasma current. (author)

  4. Ball-Pen Probe Measurements in L-Mode and H-Mode on ASDEX Upgrade

    Czech Academy of Sciences Publication Activity Database

    Adámek, Jiří; Horáček, Jan; Müller, H. W.; Rohde, V.; Ionita, C.; Schrittwieser, R.; Mehlmann, F.; Kurzan, B.; Stöckel, Jan; Dejarnac, Renaud; Weinzettl, Vladimír; Seidl, Jakub; Peterka, M.

    2010-01-01

    Roč. 50, č. 9 (2010), s. 854-859 ISSN 0863-1042. [International Workshop on Electric Probes in Magnetized Plasmas/8th./. Innsbruck, 21.09.2009-24.09.2009] R&D Projects: GA AV ČR KJB100430901; GA ČR GA202/09/1467 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * ball- pen probe * electron temperature * L-mode * H-mode * ELMs Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.006, year: 2010 http://onlinelibrary.wiley.com/doi/10.1002/ctpp.201010145/pdf

  5. Crossbar H-mode drift-tube linac design with alternative phase focusing for muon linac

    Science.gov (United States)

    Otani, M.; Futatsukawa, K.; Hasegawa, K.; Kitamura, R.; Kondo, Y.; Kurennoy, S.

    2017-07-01

    We have developed a Crossbar H-mode (CH) drift-tube linac (DTL) design with an alternative phase focusing (APF) scheme for a muon linac, in order to measure the anomalous magnetic moment and electric dipole moment (EDM) of muons at the Japan Proton Accelerator Research Complex (J-PARC). The CH-DTL accelerates muons from β = v/c = 0.08 to 0.28 at an operational frequency of 324 MHz. The design and results are described in this paper.

  6. VH mode accessibility and global H-mode properties in previous and present JET configurations

    Energy Technology Data Exchange (ETDEWEB)

    Jones, T T.C.; Ali-Arshad, S; Bures, M; Christiansen, J P; Esch, H P.L. de; Fishpool, G; Jarvis, O N; Koenig, R; Lawson, K D; Lomas, P J; Marcus, F B; Sartori, R; Schunke, B; Smeulders, P; Stork, D; Taroni, A; Thomas, P R; Thomsen, K [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    In JET VH modes, there is a distinct confinement transition following the cessation of ELMs, observed in a wide variety of tokamak operating conditions, using both NBI and ICRF heating methods. Important factors which influence VH mode accessibility such as magnetic configuration and vessel conditions have been identified. The new JET pumped divertor configuration has much improved plasma shaping control and power and particle exhaust capability and should permit exploitation of plasmas with VH confinement properties over an even wider range of operating regimes, particularly at high plasma current; first H-modes have been obtained in the 1994 JET operating period and initial results are reported. (authors). 7 refs., 6 figs.

  7. Papers presented at the 6th H-mode workshop (Seeon, Germany)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    The 6th H-mode workshop was held at Kloster Seeon (Germany) during the period of September 22-24, 1997. Contribution to this workshop is reported. Reports include. 1. Role of Nonuniform Superthermal Ions for Internal Transport Barriers. 2. Electric Field Bifurcation and Transition in the Core Plasma of CHS. 3. Formation and Termination of High Ion Temperature Mode in Heliotron/torsatron Plasmas. 4. Transition to an Enhanced Internal Transport Barrier. 5. Physics of Collapses - Probabilistic Occurrence of ELMs and Crashes -. (J.P.N.)

  8. Confinement improvement in H-mode-like plasmas in helical systems

    International Nuclear Information System (INIS)

    Itoh, K.; Sanuki, H.; Itoh, S.; Fukuyama, A.; Yagi, M.

    1993-06-01

    The reduction of the anomalous transport due to the inhomogeneous radial electric field is theoretically studied for toroidal helical plasmas. The self-sustained interchange-mode turbulence is analysed for the system with magnetic shear and magnetic hill. For the system with magnetic well like conventional stellarators, the ballooning mode turbulence is studied. Influence of the radial electric field inhomogeneity on the transport coefficients and fluctuations are quantitatively shown. Unified theory of the transport coefficients in the L-mode and H-mode-like plasmas are presented. (author)

  9. Status of the COMPASS tokamak and characterization of the first H-mode

    Czech Academy of Sciences Publication Activity Database

    Pánek, Radomír; Adámek, Jiří; Aftanas, Milan; Bílková, Petra; Böhm, Petr; Brochard, F.; Cahyna, Pavel; Cavalier, Jordan; Dejarnac, Renaud; Dimitrova, Miglena; Grover, O.; Harrison, J.; Háček, Pavel; Havlíček, Josef; Havránek, Aleš; Horáček, Jan; Hron, Martin; Imríšek, Martin; Janky, Filip; Kirk, A.; Komm, Michael; Kovařík, Karel; Krbec, Jaroslav; Kripner, Lukáš; Markovič, Tomáš; Mitošinková, Klára; Mlynář, Jan; Naydenkova, Diana; Peterka, Matěj; Seidl, Jakub; Stöckel, Jan; Štefániková, Estera; Tomeš, Matěj; Urban, Jakub; Vondráček, Petr; Varavin, Mykyta; Varju, Jozef; Weinzettl, Vladimír; Zajac, Jaromír

    2016-01-01

    Roč. 58, č. 1 (2016), č. článku 014015. ISSN 0741-3335 R&D Projects: GA MŠk(CZ) LM2011021; GA ČR(CZ) GAP205/12/2327; GA ČR(CZ) GA15-10723S Institutional support: RVO:61389021 Keywords : COMPASS * ELM * tokamak * H-mode Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016

  10. Quiescent H-mode plasmas with strong edge rotation in the cocurrent direction.

    Science.gov (United States)

    Burrell, K H; Osborne, T H; Snyder, P B; West, W P; Fenstermacher, M E; Groebner, R J; Gohil, P; Leonard, A W; Solomon, W M

    2009-04-17

    For the first time in any tokamak, quiescent H-mode (QH-mode) plasmas have been created with strong edge rotation in the direction of the plasma current. This confirms the theoretical prediction that the QH mode should exist with either sign of the edge rotation provided the magnitude of the shear in the edge rotation is sufficiently large and demonstrates that counterinjection and counteredge rotation are not essential for the QH mode. Accordingly, the present work demonstrates a substantial broadening of the QH-mode operating space and represents a significant confirmation of the theory.

  11. Rotation characteristics of main ions and impurity ions in H-mode tokamak plasma

    International Nuclear Information System (INIS)

    Kim, J.; Burrell, K.H.; Gohil, P.; Groebner, R.J.; Kim, Y.; St. John, H.E.; Seraydarian, R.P.; Wade, M.R.

    1994-01-01

    Poloidal and toroidal rotation of the main ions (He 2+ ) and the impurity ions (C 6+ and B 5+ ) in H-mode helium plasmas have been measured via charge exchange recombination spectroscopy in the DIII-D tokamak. It was discovered that the main ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction, in qualitative agreement with the neoclassical theory. The deduced radial electric field in the edge is of the same negative-well shape regardless of which ion species is used, validating the fundamental nature of the electric field in L-H transition phenomenology

  12. Electron distribution functions in Io plasma torus

    International Nuclear Information System (INIS)

    Boev, A.G.

    2003-01-01

    Electron distribution functions measured by the Voyager 1 in different shares of the Io plasma torus are explained. It is proved that their suprathermal tails are formed by the electrical field induced by the 'Jupiter wind'. The Maxwellian parts of all these spectra characterize thermal equilibrium populations of electrons and the radiation of exited ions

  13. Surgical management of palatine Torus - case series

    Directory of Open Access Journals (Sweden)

    Thaís Sumie Nozu Imada

    Full Text Available INTRODUCTION: Torus palatinus is a specific name to identify exostoses developed in the hard palate along the median palatine suture. Despite of not being a pathological condition, its presence requires attention and knowledge regarding its management. Surgical removal of exostoses is indicated when the patient frequently traumatizes the area of palatine torus during mastication and speech or when it is necessary for the rehabilitation of the upper arcade with complete dentures. OBJECTIVE: The aim of this article is to present three cases of Torus palatinus and to discuss the management of them. CASE REPORT: In the first case, a 57-year-old Caucasian man sought oral rehabilitation of his edentulous maxilla but presented a hard nodules in the hard palate; in the second case, a 40-year-old Caucasian woman was referred for frequent trauma of palatal mucosa during mastication, aesthetic complaint, and discomfort caused by the trauma of her tongue in this area; and in the third case, a 45-year-old Caucasian woman presented with a lesion on the palate that caused difficulty swallowing. When the Torus palatinus was impairing the basic physiological functions of the patients, all cases were surgically treated, improving the patients' quality of life. FINAL CONSIDERATION: The dentist should be properly prepared to choose the best from among the existing surgical approaches for each individual lesion in order to improve the results and avoid possible complications.

  14. Induction effects of torus knots and unknots

    Science.gov (United States)

    Oberti, Chiara; Ricca, Renzo L.

    2017-09-01

    Geometric and topological aspects associated with induction effects of field lines in the shape of torus knots/unknots are examined and discussed in detail. Knots are assumed to lie on a mathematical torus of circular cross-section and are parametrized by standard equations. The induced field is computed by direct integration of the Biot-Savart law. Field line patterns of the induced field are obtained and several properties are examined for a large family of knots/unknots up to 51 crossings. The intensity of the induced field at the origin of the reference system (center of the torus) is found to depend linearly on the number of toroidal coils and reaches maximum values near the boundary of the mathematical torus. New analytical estimates and bounds on energy and helicity are established in terms of winding number and minimum crossing number. These results find useful applications in several contexts when the source field is either vorticity, electric current or magnetic field, from vortex dynamics to astrophysics and plasma physics, where highly braided magnetic fields and currents are present.

  15. Torus palatinus | Naidoo | SA Journal of Radiology

    African Journals Online (AJOL)

    Kupffer and Bessel-Hagen coined the term torus palatinus in 1879 for a benign osseous protuberance arising from the midline of the hard palate. Tori are present in approximately 20% of the population and are occult until adulthood. Recent advances in modern radiology have led to improved evaluation and diagnosis of ...

  16. Magnetostatics of the uniformly polarized torus

    DEFF Research Database (Denmark)

    Beleggia, Marco; De Graef, Marc; Millev, Yonko

    2009-01-01

    We provide an exhaustive description of the magnetostatics of the uniformly polarized torus and its derivative self-intersecting (spindle) shapes. In the process, two complementary approaches have been implemented, position-space analysis of the Laplace equation with inhomogeneous boundary condit...

  17. Spherical torus, compact fusion at low field

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1985-02-01

    A spherical torus is obtained by retaining only the indispensable components on the inboard side of a tokamak plasma, such as a cooled, normal conductor that carries current to produce a toroidal magnetic field. The resulting device features an exceptionally small aspect ratio (ranging from below 2 to about 1.3), a naturally elongated D-shaped plasma cross section, and ramp-up of the plasma current primarily by noninductive means. As a result of the favorable dependence of the tokamak plasma behavior to decreasing aspect ratio, a spherical torus is projected to have small size, high beta, and modest field. Assuming Mirnov confinement scaling, an ignition spherical torus at a field of 2 T features a major radius of 1.5 m, a minor radius of 1.0 m, a plasma current of 14 MA, comparable toroidal and poloidal field coil currents, an average beta of 24%, and a fusion power of 50 MW. At 2 T, a Q = 1 spherical torus will have a major radius of 0.8 m, a minor radius of 0.5 m, and a fusion power of a few megawatts

  18. Status of the COMPASS tokamak and characterization of the first H-mode

    Science.gov (United States)

    Pánek, R.; Adámek, J.; Aftanas, M.; Bílková, P.; Böhm, P.; Brochard, F.; Cahyna, P.; Cavalier, J.; Dejarnac, R.; Dimitrova, M.; Grover, O.; Harrison, J.; Háček, P.; Havlíček, J.; Havránek, A.; Horáček, J.; Hron, M.; Imríšek, M.; Janky, F.; Kirk, A.; Komm, M.; Kovařík, K.; Krbec, J.; Kripner, L.; Markovič, T.; Mitošinková, K.; Mlynář, J.; Naydenkova, D.; Peterka, M.; Seidl, J.; Stöckel, J.; Štefániková, E.; Tomeš, M.; Urban, J.; Vondráček, P.; Varavin, M.; Varju, J.; Weinzettl, V.; Zajac, J.; the COMPASS Team

    2016-01-01

    This paper summarizes the status of the COMPASS tokamak, its comprehensive diagnostic equipment and plasma scenarios as a baseline for the future studies. The former COMPASS-D tokamak was in operation at UKAEA Culham, UK in 1992-2002. Later, the device was transferred to the Institute of Plasma Physics of the Academy of Sciences of the Czech Republic (IPP AS CR), where it was installed during 2006-2011. Since 2012 the device has been in a full operation with Type-I and Type-III ELMy H-modes as a base scenario. This enables together with the ITER-like plasma shape and flexible NBI heating system (two injectors enabling co- or balanced injection) to perform ITER relevant studies in different parameter range to the other tokamaks (ASDEX-Upgrade, DIII-D, JET) and to contribute to the ITER scallings. In addition to the description of the device, current status and the main diagnostic equipment, the paper focuses on the characterization of the Ohmic as well as NBI-assisted H-modes. Moreover, Edge Localized Modes (ELMs) are categorized based on their frequency dependence on power density flowing across separatrix. The filamentary structure of ELMs is studied and the parallel heat flux in individual filaments is measured by probes on the outer mid-plane and in the divertor. The measurements are supported by observation of ELM and inter-ELM filaments by an ultra-fast camera.

  19. BURNING PLASMA PROJECTIONS USING DRIFT WAVE TRANSPORT MODELS AND SCALINGS FOR THE H-MODE PEDESTAL

    International Nuclear Information System (INIS)

    KINSEY, J.E.; ONJUN, T.; BATEMAN, G.; KRITZ, A.; PANKIN, A.; STAEBLER, G.M.; WALTZ, R.E.

    2002-01-01

    OAK-B135 The GLF23 and Multi-Mode (MM95) transport models are used along with a model for the H-mode pedestal to predict the fusion performance for the ITER, FIRE, and IGNITOR tokamak designs. The drift-wave predictive transport models reproduce the core profiles in a wide variety of tokamak discharges, yet they differ significantly in their response to temperature gradient (stiffness). Recent gyro-kinetic simulations of ITG/TEM and ETG modes motivate the renormalization of the GLF23 model. The normalizing coefficients for the ITG/TEM modes are reduced by a factor of 3.7 while the ETG mode coefficient is increased by a factor of 4.8 in comparison with the original model. A pedestal temperature model is developed for type I ELMy H-mode plasmas based on ballooning mode stability and a theory-motivated scaling for the pedestal width. In this pedestal model, the pedestal density is proportional to the line-averaged density and the pedestal temperature is inversely related to the pedestal density

  20. Characteristics of edge localized mode in JFT-2M H-mode

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi; Funahashi, Akimasa; Goldston, R.J.

    1989-03-01

    Characteristics of edge localized mode (ELM/ERP) during H-mode plasma of JFT-2M were investigated. It was found that ELM/ERP is mainly a density fluctuation phenomena in the edge, and electron temperature in the edge except just near the separatrix is not very much perturbed. Several experimental conditions to controll ELM/ERP are, plasma density, plasma ion species, heating power, and plasma current ramping. ELM/ERPs found in low density deuterium discharge are suppressed by raising the density. ELM/ERPs are pronounced in hydrogen plasma compared with deuterium plasma. ELM/ERPs seen in hydrogen plasma or in near marginal H-mode conditions are suppressed by increasing the heating power. ELM/ERPs are found to be suppressed by plasma current ramp down, whereas they are enhanced by current ramp up. MHD aspect of ELM/ERP was investigated. No clear MHD features of ELM/ERP were found. However, reversal of mode rotation seen imediately after ELM/ERP suggests the temporal return to L-mode during the ELM/ERP event. (author)

  1. Methane penetration in DIII-D ELMing H-mode plasmas

    International Nuclear Information System (INIS)

    West, W.P.; Lasnier, C.J.; Whyte, D.G.; Isler, R.C.; Evans, T.E.; Jackson, G.L.; Rudakov, D.; Wade, M.R.; Strachan, J.

    2003-01-01

    Carbon penetration into the core plasma during midplane and divertor methane puffing has been measured for DIII-D ELMing H-mode plasmas. The methane puffs are adjusted to a measurable signal, but global plasma parameters are only weakly affected (line average density, e > increases by E , drops by 6+ density profiles in the core measured as a function of time using charge exchange recombination spectroscopy. The methane penetration factor is defined as the difference in the core content with the puff on and puff off, divided by the carbon confinement time and the methane puffing rate. In ELMing H-mode discharges with ion ∇B drift direction into the X-point, increasing the line averaged density from 5 to 8x10 19 m -3 dropped the penetration factor from 6.6% to 4.6% for main chamber puffing. The penetration factor for divertor puffing was below the detection limit (<1%). Changing the ion ∇B drift to away from the X-point decreased the penetration factor by more than a factor of five for main chamber puffing

  2. Fuel ion rotation measurement and its implications on H-mode theories

    International Nuclear Information System (INIS)

    Kim, J.; Burrell, K.H.; Gohil, P.; Groebner, R.J.; Hinton, F.L.; Kim, Y.B.; Seraydarian, R.; Mandl, W.

    1993-10-01

    Poloidal and toroidal rotation of the fuel ions (He 2+ ) and the impurity ions (C 6+ and B 5+ ) in H-mode helium plasmas have been investigated in the DIII-D tokamak by means of charge exchange recombination spectroscopy, resulting in the discovery that the fuel ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction. The radial electric field obtained from radial force balance analysis of the measured pressure gradients and rotation velocities is shown to be the same regardless of which ion species is used and therefore is a more fundamental parameter than the rotation flows in studying H-mode phenomena. It is shown that the three contributions to the radial electric field (diamagnetic, poloidal rotation, and toroidal rotation terms) are comparable and consequently the poloidal flow does not solely represent the E x B flow. In the high-shear edge region, the density scale length is comparable to the ion poloidal gyroradius, and thus neoclassical theory is not valid there. In view of this new discovery that the fuel and impurity ions rotate in opposite sense, L-H transition theories based on the poloidal rotation may require improvement

  3. Correlation of H-mode density barrier width and neutral penetration length

    International Nuclear Information System (INIS)

    Groebner, R.J.

    2002-01-01

    Pedestal studies in DIII-D find a good correlation between the width of the H-mode particle barrier width(ne) and the neutral penetration length. These results are obtained by comparing experimental n e profiles to the predictions of an analytic model for the density profile, obtained from a solution of the particle continuity equations for electrons and deuterium atoms. Initial bench-marking shows that the model is consistent with the fluid neutrals model of the UEDGE code. In its range of validity (edge temperature between 0.02-0.3 keV), the model quantitatively predicts the observed values of width(ne), the observed decrease of width(ne) as the pedestal density n e,ped increases, the observed increase of the gradient of n e with the square of n e,ped , and the observation that L-mode and H-mode profiles with the same n e,ped have very similar widths. In the model, width(ne) depends on the fuelling source and on the plasma transport. Thus, these results provide evidence that the width of the particle barrier depends on both plasma physics and atomic physics. (author)

  4. Observational Constraints on a Pluto Torus of Circumsolar Neutral Gas

    Science.gov (United States)

    Hill, M. E.; Kollmann, P.; McNutt, R. L., Jr.; Smith, H. T.; Bagenal, F.; Brown, L. E.; Elliott, H. A.; Haggerty, D. K.; Horanyi, M.; Krimigis, S. M.; Kusterer, M. B.; Lisse, C. M.; McComas, D. J.; Piquette, M. R.; Sidrow, E. J.; Strobel, D. F.; Szalay, J.; Vandegriff, J. D.; Zirnstein, E.; Ennico Smith, K.; Olkin, C.; Weaver, H. A., Jr.; Young, L. A.; Stern, S. A.

    2015-12-01

    We present the concept of a neutral gas torus surrounding the Sun, aligned with Pluto's orbit, and place observational constraints based primarily on comparison of New Horizons (NH) measurements with a 3-D Monte Carlo model adapted from analogous satellite tori surrounding Saturn and Jupiter. Such a torus, or perhaps partial torus, should result from neutral N2 escaping from Pluto's exosphere. Unlike other more massive planets closer to the Sun, neutrals escape Pluto readily owing, e.g., to the high thermal speed relative to the escape velocity. Importantly, escaped neutrals have a long lifetime due to the great distance from the Sun, ~100 years for photoionization of N2 and ~180 years for photoionization of N, which results from disassociated N2. Despite the lengthy 248-year orbit, these long e-folding lifetimes may allow an enhanced neutral population to form an extended gas cloud that modifies the N2 spatial profile near Pluto. These neutrals are not directly observable by NH but once ionized N2+ or N+ are picked up by the solar wind, reaching ~50 keV, making these pickup ions (PUIs) detectable by NH's Pluto Energetic Particle Spectrometer Science Investigation (PEPSSI) instrument. PEPSSI observations analyzed to date may constrain the N2 density; the remaining ~95% of the encounter data, scheduled for downlink in August along with similarly anticipated data from the Solar Wind Around Pluto (SWAP) experiment, should help determine the Pluto outgassing rates. Measurements from SWAP include the solar wind speed, a quantity that greatly enhances PUI studies by enabling us to directly account for the PUI distribution's sensitive dependence on plasma speed. Note that anomalous cosmic ray Si observed at Voyager is overabundant by a factor of ~3000 relative to interstellar composition. This might be related to "outer source" PUIs, but the fact that N2 and Si are indistinguishable in many instruments could mean that N2 is actually driving this apparent Si discrepancy.

  5. A quantitative analysis of the effect of ELMs on H-mode thermal energy confinement in DIII-D

    International Nuclear Information System (INIS)

    Schissel, D.P.; Osborne, T.H.; Carlstrom, T.N.; Zohm, H.

    1992-06-01

    The desire to reach ignition in future tokamaks the energy confinement time critical parameter. The most promising enhanced (over L-mode) confinement regime is the H-mode, discovered on ASDEX with neutral beam heating, and then confirmed with various auxiliary heating sources on numerous machines. The knowledge of how H-mode τ E depends on different parameters is of chemical importance to the performance predictions for next generation devices. Inter-machine H-mode total and thermal energy confinement (τ th ) scalings, which are being utilized to predict ITER thermal energy confinement, have been created for discharges where the Edge Localized Mode (ELM) instability has not been present. Confinement scaling research hm concentrated on this ELM-free H-mode phase mostly owing to the difficulty of characterizing ELM behavior. To date, long pulse H-mode operation has only been achieved by utilizing ELMs to flush out unpurities and prevent radiative collapse of the discharge. Unfortunately, accompanying the ELMS is a decrease of the plasma stored energy due to the expulsion of particles near the edge of the discharge resulting in a reduction of the steep edge electron density gradient. In order to predict ITER's H-mode τ th in the presence of ELMS, an estimated 25% confinement degradation factor has been applied to the ELM-free predictions. Our work, summarized in this paper, indicates that this 25% reduction factor is too large and instead a value of approximately 15% would be more appropriate

  6. 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)

  7. Evolution of the radial electric field in a JET H-mode plasma

    International Nuclear Information System (INIS)

    Andrew, Y.; Hawkes, N.C.; Biewer, T.; Crombe, K.; Keeling, D.; De la Luna, E.; Giroud, C.; Korotkov, A.; Meigs, A.; Murari, A.; Nunes, I.; Sartori, R.; Tala, T.; Andrew, Y.; Hawkes, N.C.; Keeling, D.; Giroud, C.; Korotkov, A.; Meigs, A.; Biewer, T.; Crombe, K.; De la Luna, E.; Murari, A.; Nunes, I.; Sartori, R.; Tala, T.

    2008-01-01

    Results from recent measurements of carbon impurity ion toroidal and poloidal rotation velocities, ion temperature, ion density and the resulting radial electric field (E r ) profiles are presented from an evolving Joint European Torus (JET) tokamak plasma over a range of energy and particle confinement regimes. Significant levels of edge plasma poloidal rotation velocity have been measured for the first time on JET, with maximum values of ±9 km/s. Such values of poloidal rotation provide an important contribution to the total edge plasma E r profiles. Large values of shear in the measured E r profiles are observed to arise as a consequence of the presence of the edge transport barrier (ETB) and do not appear to be necessary for their formation or destruction. These results have an important impact on potential mechanisms for transport barrier triggering and sustainment in present-day and future high-performance fusion plasmas. (authors)

  8. Tokamak fluidlike equations, with applications to turbulence and transport in H mode discharges

    International Nuclear Information System (INIS)

    Kim, Y.B.; Biglari, H.; Carreras, B.A.; Diamond, P.H.; Groebner, R.J.; Kwon, O.J.; Spong, D.A.; Callen, J.D.; Chang, Z.; Hollenberg, J.B.; Sundaram, A.K.; Terry, P.W.; Wang, J.F.

    1990-01-01

    Significant progress has been made in developing tokamak fluidlike equations which are valid in all collisionality regimes in toroidal devices, and their applications to turbulence and transport in tokamaks. The areas highlighted in this paper include: the rigorous derivation of tokamak fluidlike equations via a generalized Chapman-Enskog procedure in various collisionality regimes and on various time scales; their application to collisionless and collisional drift wave models in a sheared slab geometry; applications to neoclassical drift wave turbulence; i.e. neoclassical ion-temperature-gradient-driven turbulence and neoclassical electron-drift-wave turbulence; applications to neoclassical bootstrap-current-driven turbulence; numerical simulation of nonlinear bootstrap-current-driven turbulence and tearing mode turbulence; transport in Hot-Ion H mode discharges. 20 refs., 3 figs

  9. Reciprocating Probe Measurements of L-H Transition in LHCD H-mode on EAST

    DEFF Research Database (Denmark)

    Peng, Liu; Guosheng, Xu; Huiqian, Wang

    2013-01-01

    that the power loss P loss was comparable during the L-H transition, by comparing the adjacent L-mode and H-mode discharge. The Dα emission, Te and ne decreased rapidly in the time scale of about 1 ms, and the radial electric field Er turned positive in this process near the last closed flux surface. Multiple L......-H-L transitions were observed during a single shot when the applied LHW power was marginal to the threshold. The floating potential (Vf) had negative spikes corresponding with the Dα signal, and Er oscillation evolved into several intermittent negative spikes just before the L-H transition. In some shots......, dithering was observed just before the L-H transition....

  10. Physics of the L-mode to H-mode transition in tokamaks

    International Nuclear Information System (INIS)

    Burrell, K.H.; Carlstrom, T.N.; Gohil, P.; Groebner, R.J.; Kim, J.; Osborne, T.H.; St. John, H.; Stambaugh, R.D.; Doyle, E.J.; Moyer, R.A.; Rettig, C.L.; Peebles, W.A.; Rhodes, T.L.; Finkenthal, D.; Hillis, D.L.; Wade, M.R.; Matsumoto, H.; Watkins, J.G.

    1992-07-01

    Combined theoretical and experimental work has resulted in the creation of a paradigm which has allowed semi-quantitative understanding of the edge confinement improvement that occurs in the H-mode. Shear in the E x B flow of the fluctuations in the plasma edge can lead to decorrelation of the fluctuations, decreased radial correlation lengths and reduced turbulent transport. Changes in the radial electric field, the density fluctuations and the edge transport consistent with shear stabilization of turbulence have been seen in several tokamaks. The purpose of this paper is to discuss the most recent data in the light of the basic paradigm of electric field shear stabilization and to critically compare the experimental results with various theories

  11. Dependence of H-mode power threshold on global and local edge parameters

    International Nuclear Information System (INIS)

    Groebner, R.J.; Carlstrom, T.N.; Burrell, K.H.

    1995-12-01

    Measurements of local electron density n e , electron temperature T e , and ion temperature T i have been made at the very edge of the plasma just prior to the transition into H-mode for four different single parameter scans in the DIII-D tokamak. The means and standard derivations of n e , T e , and T i under these conditions for a value of the normalized toroidal flux of 0.98 are respectively, 1.5 ± 0.7 x 10 19 m -3 , 0.051 ± 0.016 keV, and 0.14 ± 0.03 keV. The threshold condition for the transition is more sensitive to temperature than to density. The data indicate that the dependence is not as simple as a requirement for a fixed value of the ion collisionality

  12. Transport modeling of L- and H-mode discharges with LHCD on EAST

    Science.gov (United States)

    Li, M. H.; Ding, B. J.; Imbeaux, F.; Decker, J.; Zhang, X. J.; Kong, E. H.; Zhang, L.; Wei, W.; Shan, J. F.; Liu, F. K.; Wang, M.; Xu, H. D.; Yang, Y.; Peysson, Y.; Basiuk, V.; Artaud, J.-F.; Yuynh, P.; Wan, B. N.

    2013-04-01

    High-confinement (H-mode) discharges with lower hybrid current drive (LHCD) as the only heating source are obtained on EAST. In this paper, an empirical transport model of mixed Bohm/gyro-Bohm for electron and ion heat transport was first calibrated against a database of 3 L-mode shots on EAST. The electron and ion temperature profiles are well reproduced in the predictive modeling with the calibrated model coupled to the suite of codes CRONOS. CRONOS calculations with experimental profiles are also performed for electron power balance analysis. In addition, the time evolutions of LHCD are calculated by the C3PO/LUKE code involving current diffusion, and the results are compared with experimental observations.

  13. Nonlinear theory of trapped electron temperature gradient driven turbulence in flat density H-mode plasmas

    International Nuclear Information System (INIS)

    Hahm, T.S.

    1990-12-01

    Ion temperature gradient turbulence based transport models have difficulties reconciling the recent DIII-D H-mode results where the density profile is flat, but χ e > χ i in the core region. In this work, a nonlinear theory is developed for recently discovered ion temperature gradient trapped electron modes propagating in the electron diamagnetic direction. This instability is predicted to be linearly unstable for L Ti /R approx-lt κ θ ρ s approx-lt (L Ti /R) 1/4 . They are also found to be strongly dispersive even at these long wavelengths, thereby suggesting the importance of the wave-particle-wave interactions in the nonlinear saturation phase. The fluctuation spectrum and anomalous fluxes are calculated. In accordance with the trends observed in DIII-D, the predicted electron thermal diffusivity can be larger than the ion thermal diffusivity. 17 refs., 3 figs

  14. Helium Exhaust Studies in H-Mode Discharges in the DIII-D Tokamak Using an Argon-Frosted Divertor Cryopump

    International Nuclear Information System (INIS)

    Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Mahdavi, M.A.; Maingi, R.; West, W.P.; Brooks, N.H.; Burrell, K.H.; Groebner, R.J.; Jackson, G.L.; Klepper, C.C.; Laughon, G.; Menon, M.M.; Mioduszewski, P.K.

    1995-01-01

    The first experiments demonstrating exhaust of thermal helium in a diverted, H-mode deuterium plasma have been performed on the DIII-D tokamak. The helium, introduced via gas puffing, is observed to reach the plasma core, and then is readily removed from the plasma with a time constant of ∼10--20 energy-confinement times by an in-vessel cryopump conditioned with argon frosting. Detailed analysis of the helium profile evolution suggests that the exhaust rate is limited by the exhaust efficiency of the pump (∼5%) and not by the intrinsic helium-transport properties of the plasma

  15. Development of ITER 15 MA ELMy H-mode Inductive Scenario

    International Nuclear Information System (INIS)

    C. E. Kessel, D. Campbell, Y. Gribov, G. Saibene, G. Ambrosino, T. Casper, M. Cavinato, H. Fujieda, R. Hawryluk, L. D. Horton, A. Kavin, R. Kharyrutdinov, F. Koechl, J. Leuer, A. Loarte, P. J. Lomas, T. Luce, V. Lukash, M. Mattei, I.Nunes, V. Parail, A. Polevoi, A. Portone, R. Sartori, A.C.C. Sips, P. R. Thomas, A. Welander and J. Wesley

    2008-01-01

    The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low li (<0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation

  16. Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

    International Nuclear Information System (INIS)

    Lore, J. D.; Reinke, M. L.; Lipschultz, B.; Brunner, D.; LaBombard, B.; Terry, J.; Pitts, R. A.; Feng, Y.

    2015-01-01

    Experiments in Alcator C-Mod to assess the level of toroidal asymmetry in divertor conditions resulting from poloidally and toroidally localized extrinsic impurity gas seeding show a weak toroidal peaking (∼1.1) in divertor electron temperatures for high-power enhanced D-alpha H-mode plasmas. This is in contrast to similar experiments in Ohmically heated L-mode plasmas, which showed a clear toroidal modulation in the divertor electron temperature. Modeling of these experiments using the 3D edge transport code EMC3-EIRENE [Y. Feng et al., J. Nucl. Mater. 241, 930 (1997)] qualitatively reproduces these trends, and indicates that the different response in the simulations is due to the ionization location of the injected nitrogen. Low electron temperatures in the private flux region (PFR) in L-mode result in a PFR plasma that is nearly transparent to neutral nitrogen, while in H-mode the impurities are ionized in close proximity to the injection location, with this latter case yielding a largely axisymmetric radiation pattern in the scrape-off-layer. The consequences for the ITER gas injection system are discussed. Quantitative agreement with the experiment is lacking in some areas, suggesting potential areas for improving the physics model in EMC3-EIRENE

  17. Drift resonance and stability of the Io plasma torus

    Science.gov (United States)

    Zhan, Jie; Hill, T. W.

    2000-03-01

    The observed local time asymmetry of the Io plasma torus is generally attributed to the presence of a persistent dawn-to-dusk electric field in the Jovian magnetosphere. The local time asymmetry is modulated at the System 3 rotation period of Jupiter's magnetic field, suggesting that the dawn-to-dusk electric field may be similarly modulated. We argue that such a System 3 modulation would have a profound disruptive effect on the observed torus structure if the torus were to corotate at exactly the System 3 rate: the torus would be a resonantly forced harmonic oscillator, and would disintegrate in a few rotation periods, contrary to observations. This destabilizing effect is independent of, and in addition to, the more familiar effect of the centrifugal interchange instability, which is also capable of disrupting the torus in a few rotation periods in the absence of other effects. We conclude that the observed (few percent) corotation lag of the torus is essential to preserving the observed long-lived torus structure by detuning the resonant frequency (the torus drift frequency) relative to the forcing frequency (System 3). A possible outcome of this confinement mechanism is a residual radial oscillation of the torus at the beat period (~10 days) between System 3 and the torus drift period.

  18. A new equilibrium torus solution and GRMHD initial conditions

    Science.gov (United States)

    Penna, Robert F.; Kulkarni, Akshay; Narayan, Ramesh

    2013-11-01

    Context. General relativistic magnetohydrodynamic (GRMHD) simulations are providing influential models for black hole spin measurements, gamma ray bursts, and supermassive black hole feedback. Many of these simulations use the same initial condition: a rotating torus of fluid in hydrostatic equilibrium. A persistent concern is that simulation results sometimes depend on arbitrary features of the initial torus. For example, the Bernoulli parameter (which is related to outflows), appears to be controlled by the Bernoulli parameter of the initial torus. Aims: In this paper, we give a new equilibrium torus solution and describe two applications for the future. First, it can be used as a more physical initial condition for GRMHD simulations than earlier torus solutions. Second, it can be used in conjunction with earlier torus solutions to isolate the simulation results that depend on initial conditions. Methods: We assume axisymmetry, an ideal gas equation of state, constant entropy, and ignore self-gravity. We fix an angular momentum distribution and solve the relativistic Euler equations in the Kerr metric. Results: The Bernoulli parameter, rotation rate, and geometrical thickness of the torus can be adjusted independently. Our torus tends to be more bound and have a larger radial extent than earlier torus solutions. Conclusions: While this paper was in preparation, several GRMHD simulations appeared based on our equilibrium torus. We believe it will continue to provide a more realistic starting point for future simulations.

  19. 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.

  20. Exploring Torus Universes in Causal Dynamical Triangulations

    DEFF Research Database (Denmark)

    Budd, Timothy George; Loll, R.

    2013-01-01

    Motivated by the search for new observables in nonperturbative quantum gravity, we consider Causal Dynamical Triangulations (CDT) in 2+1 dimensions with the spatial topology of a torus. This system is of particular interest, because one can study not only the global scale factor, but also global...... shape variables in the presence of arbitrary quantum fluctuations of the geometry. Our initial investigation focusses on the dynamics of the scale factor and uncovers a qualitatively new behaviour, which leads us to investigate a novel type of boundary conditions for the path integral. Comparing large....... Apart from setting the stage for the analysis of shape dynamics on the torus, the new set-up highlights the role of nontrivial boundaries and topology....

  1. TORUS Annual Continuation and Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Arbanas, Goran [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Elster, Charlotte [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Echer, Jutta [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nunes, Filomena [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Thompson, Ian [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-02-24

    The TORUS collaboration derives its name from the research it focuses on, namely the Theory of Reactions for Unstable iSotopes. It is a Topical Collaboration in Nuclear Theory, and funded by the Nuclear Theory Division of the Office of Nuclear Physics in the Office of Science of the Department of Energy. The funding supports one postdoctoral researcher for the years 1 through 4. The collaboration brings together as Principal Investigators a large fraction of the nuclear reaction theorists currently active within the USA. The mission of the TORUS Topical Collaboration is to develop new methods that will advance nuclear reaction theory for unstable isotopes by using three-body techniques to improve directreaction calculations. This multi-institution collaborative effort is directly relevant to three areas of interest: the properties of nuclei far from stability; microscopic studies of nuclear input parameters for astrophysics, and microscopic nuclear reaction theory.

  2. Equilibrium-torus bifurcation in nonsmooth systems

    DEFF Research Database (Denmark)

    Zhusubahyev, Z.T.; Mosekilde, Erik

    2008-01-01

    Considering a set of two coupled nonautonomous differential equations with discontinuous right-hand sides describing the behavior of a DC/DC power converter, we discuss a border-collision bifurcation that can lead to the birth of a two-dimensional invariant torus from a stable node equilibrium...... point. We obtain the chart of dynamic modes and show that there is a region of parameter space in which the system has a single stable node equilibrium point. Under variation of the parameters, this equilibrium may disappear as it collides with a discontinuity boundary between two smooth regions...... in the phase space. The disappearance of the equilibrium point is accompanied by the soft appearance of an unstable focus period-1 orbit surrounded by a resonant or ergodic torus. Detailed numerical calculations are supported by a theoretical investigation of the normal form map that represents the piecewise...

  3. Analysis of performance degradation in an electron heating dominant H-mode plasma after ECRH termination in EAST

    Science.gov (United States)

    Du, Hongfei; Ding, Siye; Chen, Jiale; Wang, Yifeng; Lian, Hui; Xu, Guosheng; Zhai, Xuemei; Liu, Haiqing; Zang, Qing; Lyu, Bo; Duan, Yanmin; Qian, Jinping; Gong, Xianzu

    2018-06-01

    In recent EAST experiments, significant performance degradation accompanied by a decrease of internal inductance is observed in an electron heating dominant H-mode plasma after the electron cyclotron resonance heating termination. The lower hybrid wave (LHW) deposition and effective electron heat diffusivity are calculated to explain this phenomenon. Analysis shows that the changes of LHW heating deposition rather than the increase of transport are responsible for the significant decrease in energy confinement (). The reason why the confinement degradation occurred on a long time scale could be attributed to both good local energy confinement in the core and also the dependence of LHW deposition on the magnetic shear. The electron temperature profile shows weaker stiffness in near axis region where electron heating is dominant, compared to that in large radius region. Unstable electron modes from low to high k in the core plasma have been calculated in the linear GYRO simulations, which qualitatively agree with the experimental observation. This understanding of the plasma performance degradation mechanism will help to find ways of improving the global confinement in the radio-frequency dominant scenario in EAST.

  4. Numerical model for radial transport in the ELMO Bumpy Torus

    International Nuclear Information System (INIS)

    Jaeger, E.F.; Hedrick, C.L.

    1977-11-01

    Neutral and charged particle densities and temperatures are calculated as functions of radius for the toroidal plasma in the ELMO Bumpy Torus (EBT) experiment. Energy dependent ionization and charge-exchange rates, ambipolar diffusion, and self-consistent radial electric field profiles are included. Variation in magnetic field due to finite plasma pressure, effects of energetic electron rings, and transport due to drift waves and magnetic field errors are neglected. Diffusion is assumed to be neoclassical with enhanced losses at low collisionalities. The model reproduces many of the observed features of EBT operation in the quiescent toroidal (T) mode. The self-consistently calculated electric field is everywhere positive (not as in experiments) unless enhanced electron collisionality is included. Solutions for advanced EBT's are obtained and confinement parameters predicted

  5. Torus-doubling process via strange nonchaotic attractors

    International Nuclear Information System (INIS)

    Mitsui, Takahito; Uenohara, Seiji; Morie, Takashi; Horio, Yoshihiko; Aihara, Kazuyuki

    2012-01-01

    Torus-doubling bifurcations typically occur only a finite number of times. It has been assumed that torus-doubling bifurcations in quasiperiodically forced systems are interrupted by the appearance of strange nonchaotic attractors (SNAs). In the present Letter, we study a quasiperiodically forced noninvertible map and report the occurrence of a torus-doubling process via SNAs. The mechanism of this process is numerically clarified. Furthermore, this process is experimentally demonstrated in a switched-capacitor integrated circuit. -- Highlights: ► We report the occurrence of a torus-doubling process via strange nonchaotic attractors (SNAs). ► The process consists of the gradual fractalization of a torus and the Heagy–Hammel transition. ► The torus-doubling process via SNAs is also experimentally demonstrated in an electronic circuit.

  6. Reactor-relevant quiescent H-mode operation using torque from non-axisymmetric, non-resonant magnetic fields

    International Nuclear Information System (INIS)

    Burrell, K. H.; Garofalo, A. M; Osborne, T. H.; Schaffer, M. J.; Snyder, P. B.; Solomon, W. M.; Park, J.-K.; Fenstermacher, M. E.

    2012-01-01

    Results from recent experiments demonstrate that quiescent H-mode (QH-mode) sustained by magnetic torque from non-axisymmetric magnetic fields is a promising operating mode for future burning plasmas. Using magnetic torque from n=3 fields to replace counter-I p torque from neutral beam injection (NBI), we have achieved long duration, counter-rotating QH-mode operation with NBI torque ranging from counter-I p to up to co-I p values of 1-1.3 Nm. This co-I p torque is 3 to 4 times the scaled torque that ITER will have. These experiments utilized an ITER-relevant lower single-null plasma shape and were done with ITER-relevant values of ν ped * and β N ped . These discharges exhibited confinement quality H 98y2 =1.3, in the range required for ITER. In preliminary experiments using n=3 fields only from a coil outside the toroidal coil, QH-mode plasmas with low q 95 =3.4 have reached fusion gain values of G=β N H 89 /q 95 2 =0.4, which is the desired value for ITER. Shots with the same coil configuration also operated with net zero NBI torque. The limits on G and co-I p torque have not yet been established for this coil configuration. QH-mode work to has made significant contact with theory. The importance of edge rotational shear is consistent with peeling-ballooning mode theory. Qualitative and quantitative agreements with the predicted neoclassical toroidal viscosity torque is seen.

  7. E-H mode transition in low-pressure inductively coupled nitrogen-argon and oxygen-argon plasmas

    International Nuclear Information System (INIS)

    Lee, Young Wook; Lee, Hye Lan; Chung, T. H.

    2011-01-01

    This work investigates the characteristics of the E-H mode transition in low-pressure inductively coupled N 2 -Ar and O 2 -Ar discharges using rf-compensated Langmuir probe measurements and optical emission spectroscopy (OES). As the ICP power increases, the emission intensities from plasma species, the electron density, the electron temperature, and the plasma potential exhibit sudden changes. The Ar content in the gas mixture and total gas pressure have been varied in an attempt to fully characterize the plasma parameters. With these control parameters varying, the changes of the transition threshold power and the electron energy distribution function (EEDF) are explored. In N 2 -Ar and O 2 -Ar discharges at low-pressures of several millitorr, the transition thresholds are observed to decrease with Ar content and pressure. It is observed that in N 2 -Ar plasmas during the transition, the shape of the EEDF changes from an unusual distribution with a flat hole near the electron energy of 3 eV in the E mode to a Maxwellian distribution in the H mode. However, in O 2 -Ar plasmas, the EEDFs in the E mode at low Ar contents show roughly bi-Maxwellian distributions, while the EEDFs in the H mode are observed to be nearly Maxwellian. In the E and H modes of O 2 -Ar discharges, the dissociation fraction of O 2 molecules is estimated using optical emission actinometry. During the E-H mode transition, the dissociation fraction of molecules is also enhanced.

  8. Dependence of helium transport on plasma current and ELM frequency in H-mode discharges in DIII-D

    International Nuclear Information System (INIS)

    Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Finkenthal, D.F.; West, W.P.; Burrell, K.H.; Seraydarian, R.P.

    1993-05-01

    The removal of helium (He) ash from the plasma core with high efficiency to prevent dilution of the D-T fuel mixture is of utmost importance for future fusion devices, such as the International Thermonuclear Experimental Reactor (ITER). A variety of measurements in L-mode conditions have shown that the intrinsic level of helium transport from the core to the edge may be sufficient to prevent sufficient dilution (i.e., τ He /τ E < 5). Preliminary measurements in biased-induced, limited H-mode discharges in TEXTOR suggest that the intrinsic helium transport properties may not be as favorable. If this trend is shown also in diverted H-mode plasmas, then scenarios based on ELMing H-modes would be less desirable. To further establish the database on helium transport in H-mode conditions, recent studies on the DIII-D tokamak have focused on determining helium transport properties in H-mode conditions and the dependence of these properties on plasma current and ELM frequency

  9. Studying uniform thickness II: Transversely nonsimple iterated torus knots

    DEFF Research Database (Denmark)

    LaFountain, Douglas

    2011-01-01

    We prove that an iterated torus knot type in the standard contact 3-sphere fails the uniform thickness property (UTP) if and only if it is formed from repeated positive cablings, which is precisely when an iterated torus knot supports the standard contact structure. This is the first complete UTP...... classification for a large class of knots. We also show that all iterated torus knots that fail the UTP support cabling knot types that are transversely non-simple....

  10. Predictive modelling of the impact of argon injection on H-mode plasmas in JET with the RITM code

    International Nuclear Information System (INIS)

    Unterberg, B; Kalupin, D; Tokar', M Z; Corrigan, G; Dumortier, P; Huber, A; Jachmich, S; Kempenaars, M; Kreter, A; Messiaen, A M; Monier-Garbet, P; Ongena, J; Puiatti, M E; Valisa, M; Hellermann, M von

    2004-01-01

    Self-consistent modelling of energy and particle transport of the plasma background and impurities has been performed with the code RITM for argon seeded high density H-mode plasmas in JET. The code can reproduce both the profiles in the plasma core and the structure of the edge pedestal. The impact of argon on core transport is found to be small; in particular, no significant change in confinement is observed in both experimental and modelling results. The same transport model, which has been used to reproduce density peaking in the radiative improved mode in TEXTOR, reveals a flat density profile in Ar seeded JET H-mode plasmas in agreement with the experimental observations. This behaviour is attributed to the rather flat profile of the safety factor in the bulk of H-mode discharges

  11. H-mode-like discharge under the presence of 1/1 rational surface at ergodic layer in LHD

    International Nuclear Information System (INIS)

    Morita, Shigeru; Morisaki, Tomohiro; Tanaka, Kenji

    2004-01-01

    H-mode-like discharge was found in LHD with a full B t field of 2.5T at an outwardly shifted configuration of R ax = 4.00 m where the m/n = 1/1 rational surface is located at the ergodic layer. The H-mode-like discharge was triggered by changing the P NBI from 9MW to 5 MW in a density range of 4-8 x 10 13 cm -3 , followed by a clear density rise, ELM-like H α bursts, and a reduction of magnetic fluctuation. These H-mode-like features vanished with a small radial movement of the 1/1 surface. (author)

  12. H-mode edge stability of Alcator C-mod plasmas

    International Nuclear Information System (INIS)

    Mossessian, D.A.; Hubbard, A.; Hughes, J.W.; Greenwald, M.; LaBombard, B.; Snipes, J.A.; Wolfe, S.; Snyder, P.; Wilson, H.; Xu, X.; Nevins, W.

    2003-01-01

    For steady state H-mode operation, a relaxation mechanism is required to limit build-up of the edge gradient and impurity content. C-Mod sees two such mechanisms - EDA and grassy ELMs, but not large type I ELMs. In EDA the edge relaxation is provided by an edge localized quasi coherent electromagnetic mode that exists at moderate pedestal temperature T 3.5 and does not limit the build up of the edge pressure gradient. The mode is not observed in the ideal MHD stability analysis, but is recorded in the nonlinear real geometry fluctuations modeling based on fluid equations and is thus tentatively identified as a resistive ballooning mode. At high edge pressure gradients and temperatures the mode is replaced by broadband fluctuations (f< 50 kHz) and small irregular ELMs are observed. Based on ideal MHD calculations that include the effects of edge bootstrap current, these ELMs are identified as medium n (10 < n < 50) coupled peeling/ballooning modes. The stability thresholds, its dependence on the plasma shape and the modes structure are studied experimentally and with the linear MHD stability code ELITE. (author)

  13. Review of DIII-D H-Mode Density Limit Studies

    International Nuclear Information System (INIS)

    Maingi, R.; Mahdavi, M.A.

    2005-01-01

    Density limit studies over the past 10 yr on DIII-D have successfully identified several processes that limit plasma density in various operating modes. The recent focus of these studies has been on maintenance of the high-density operational window with good H-mode level energy confinement. We find that detachment and onset of multifaceted axisymmetric radiation from the edge (MARFE), fueling efficiency, particle confinement, and magnetohydrodynamic activity can impose density limits in certain regimes. By studying these processes, we have devised techniques with either pellets or gas fueling and divertor pumping to achieve line average density above Greenwald scaling, relying on increasing the ratio of pedestal to separatrix density, as well as density profile peaking. The scaling of several of these processes to next-step devices (e.g., the International Thermonuclear Experimental Reactor) has indicated that sufficiently high pedestal density can be achieved with conventional fueling techniques while ensuring divertor partial detachment needed for heat flux reduction. One density limit process requiring further study is neoclassical tearing mode (NTM) onset, and techniques for avoidance/mitigation of NTMs need additional development in present-day devices operated at high density

  14. Development of superconducting crossbar-H-mode cavities for proton and ion accelerators

    Directory of Open Access Journals (Sweden)

    F. Dziuba

    2010-04-01

    Full Text Available The crossbar-H-mode (CH structure is the first superconducting multicell drift tube cavity for the low and medium energy range operated in the H_{21} mode. Because of the large energy gain per cavity, which leads to high real estate gradients, it is an excellent candidate for the efficient acceleration in high power proton and ion accelerators with fixed velocity profile. A prototype cavity has been developed and tested successfully with a gradient of 7  MV/m. A few new superconducting CH cavities with improved geometries for different high power applications are under development at present. One cavity (f=325  MHz, β=0.16, seven cells is currently under construction and studied with respect to a possible upgrade option for the GSI UNILAC. Another cavity (f=217  MHz, β=0.059, 15 cells is designed for a cw operated energy variable heavy ion linac application. Furthermore, the EUROTRANS project (European research program for the transmutation of high level nuclear waste in an accelerator driven system, 600 MeV protons, 352 MHz is one of many possible applications for this kind of superconducting rf cavity. In this context a layout of the 17 MeV EUROTRANS injector containing four superconducting CH cavities was proposed by the Institute for Applied Physics (IAP Frankfurt. The status of the cavity development related to the EUROTRANS injector is presented.

  15. A model for a scrape-off-layer low-high (L-H) mode transition

    International Nuclear Information System (INIS)

    Cohen, R.H.; Xu, X.

    1995-01-01

    Increasing the radial mode number has a stabilizing effect on the conducting-wall and curvature-driven interchange modes in a tokamak scrape-off layer (SOL), arising from the increased polarization response. Such an effect is naturally imposed as the SOL width is decreased, and for a narrow-enough SOL, the stabilizing effect is stronger than the increase in the instability drives. By combining a mixing-length estimate for the thermal diffusivity with energy conservation and heat conduction equations and the condition of continuity of the heat flux at the separatrix, it is found that the resultant turbulence-transport system admits two solutions, one stable and one unstable, at different SOL widths; the inclusion of additional physics can add a second stable root at lower width. These roots are plausibly identified with SOL behavior in low (L) and high (H) modes. Particularly when a model is introduced for finite-β, finite-k parallel effects on the modes, a power threshold for transition to the narrower root is obtained, suggesting a possible L-H transition mechanism. The non-monotonic dependence of the turbulent heat flux vs SOL width and the possibility of multiple solutions for the equilibrium SOL width are verified with nonlinear simulations. copyright 1995 American Institute of Physics

  16. Plasma interaction with tungsten samples in the COMPASS tokamak in ohmic ELMy H-modes

    International Nuclear Information System (INIS)

    Dimitrova, M; Weinzettl, V; Matejicek, J; Dejarnac, R; Stöckel, J; Havlicek, J; Panek, R; Popov, Tsv; Marinov, S; Costea, S

    2016-01-01

    This paper reports experimental results on plasma interaction with tungsten samples with or without pre-grown He fuzz. Under the experimental conditions, arcing was observed on the fuzzy tungsten samples, resulting in localized melting of the fuzz structure that did not extend into the bulk. The parallel power flux densities were obtained from the data measured by Langmuir probes embedded in the divertor tiles on the COMPASS tokamak. Measurements of the current-voltage probe characteristics were performed during ohmic ELMy H-modes reached in deuterium plasmas at a toroidal magnetic field B T = 1.15 T, plasma current I p = 300 kA and line-averaged electron density n e = 5×10 19 m -3 . The data obtained between the ELMs were processed by the recently published first-derivative probe technique for precise determination of the plasma potential and the electron energy distribution function (EEDF). The spatial profile of the EEDF shows that at the high-field side it is Maxwellian with a temperature of 5 -- 10 eV. The electron temperatures and the ion-saturation current density obtained were used to evaluate the radial distribution of the parallel power flux density as being in the order of 0.05 -- 7 MW/m 2 . (paper)

  17. Comparison of H-mode pedestals in different confinement regimes in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Groebner, R J [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Leonard, A W [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Luce, T C [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Fenstermacher, M E [Lawrence Livermore National Laboratory, Livermore, California (United States); Jackson, G L [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Osborne, T H [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Thomas, D M [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Wade, M R [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States)

    2006-05-15

    A survey of global performance parameters and their correlation with pedestal parameters is performed for standard H-mode, QH-mode and the enhanced confinement regimes of VH-mode, hybrid and advanced tokamak in the DIII-D tokamak. This study shows that there is a trend for global confinement quality or global beta to increase as the pedestal electron pressure or beta increases. However, there are also improvements in core confinement and beta, observed at fixed pedestal pressure or beta, which indicate that factors other than pedestal parameters also contribute to the best core performance. Several other pedestal structure parameters are found to be similar among these regimes. The scale lengths for electron pressure in the pedestal are in the range 0.8-1.6 cm at the outer midplane, most {eta}{sub e} values are in the range 1-3 in the middle of the T{sub e} pedestal and the T{sub e} and n{sub e} pedestals tend to penetrate the same distance into the plasma.

  18. H-mode threshold power scaling and the ∇B drift effect

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Burrell, K.H.; Groebner, R.J.; Staebler, G.M.

    1997-06-01

    One of the largest influences on the H-mode power threshold (P TH ) is the direction of the ion ∇B drift relative to the X-point location, where factors of 2--3 increase in P TH are observed for the ion ∇B drift away from the X-point. It is proposed that the threshold power scaling observed in single-null configurations with the ion ∇B drift toward the X-point location (P TH ∼ nB, where n is the plasma density, and B is the toroidal field) is due to the scaling of the magnitude of the ∇B drift effect. Hinton and later Hinton and Stebler have modeled this effect as neoclassical cross field fluxes of both heat and particles driven by poloidal temperature gradients on the open field lines in the scrape-off layer (SOL). The ∇B drift effect influences the power threshold by affecting the edge conditions needed for the L-H transition. It is not essential for the L-H transition itself since transitions are observed with either direction of B. Predictions of this model include saturation of the B scaling of P TH at high field, 1/B scaling of P TH with reverse B, and no B scaling of P TH in balanced double-null configurations. This last prediction is consistent with the observed scaling of p TH in double-null plasma sin DIII-D

  19. Global gyrokinetic simulations of the H-mode tokamak edge pedestal

    Energy Technology Data Exchange (ETDEWEB)

    Wan, Weigang; Parker, Scott E.; Chen, Yang [Department of Physics, University of Colorado, Boulder, Colorado 80309 (United States); Groebner, Richard J. [General Atomics, Post Office Box 85068, San Diego, California 92186 (United States); Yan, Zheng [University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Pankin, Alexei Y.; Kruger, Scott E. [Tech-X Corporation, 5621 Arapahoe Ave., Boulder, Colorado 80305 (United States)

    2013-05-15

    Global gyrokinetic simulations of DIII-D H-mode edge pedestal show two types of instabilities may exist approaching the onset of edge localized modes: an intermediate-n, high frequency mode which we identify as the “kinetic peeling ballooning mode (KPBM),” and a high-n, low frequency mode. Our previous study [W. Wan et al., Phys. Rev. Lett. 109, 185004 (2012)] has shown that when the safety factor profile is flattened around the steep pressure gradient region, the high-n mode is clearly kinetic ballooning mode and becomes the dominant instability. Otherwise, the KPBM dominates. Here, the properties of the two instabilities are studied by varying the density and temperature profiles. It is found that the KPBM is destabilized by density and ion temperature gradient, and the high-n mode is mostly destabilized by electron temperature gradient. Nonlinear simulations with the KPBM saturate at high levels. The equilibrium radial electric field (E{sub r}) reduces the transport. The effect of the parallel equilibrium current is found to be weak.

  20. Operational conditions and characteristics of ELM-events during H-mode plasmas in the stellarator W7-AS

    International Nuclear Information System (INIS)

    Hirsch, M.; Grigull, P.; Wobig, H.; Kisslinger, J.; McCormick, K.; Anton, M.; Baldzuhn, J.; Fiedler, S.; Fuchs, Ch.; Geiger, J.; Giannone, L.; Hartfuss, H.-J.; Holzhauer, E.; Hirsch, M.; Jaenicke, R.; Kick, M.; Maassberg, H.; Wagner, F.; Weller, A.

    2000-01-01

    H-mode operation in the low-shear stellarator W7-AS is achieved for specific plasma edge topologies characterized by three 'operational windows' of the edge rotational transform. An explanation for this strong influence of the magnetic configuration could be the increase of viscous damping if rational surfaces and thus island structures occur within the relevant plasma edge layer, thereby impeding the development of an edge transport barrier. Prior to the final transition to a quiescent state, the plasma edge passes a rich phenomenology of dynamic behaviour such as dithering and ELMs. Plasma edge parameters indicate that a quiescent H-mode occurs if a certain edge pressure is achieved. (author)

  1. Role of zonal flow predator-prey oscillations in triggering the transition to H-mode confinement.

    Science.gov (United States)

    Schmitz, L; Zeng, L; Rhodes, T L; Hillesheim, J C; Doyle, E J; Groebner, R J; Peebles, W A; Burrell, K H; Wang, G

    2012-04-13

    Direct evidence of zonal flow (ZF) predator-prey oscillations and the synergistic roles of ZF- and equilibrium E×B flow shear in triggering the low- to high-confinement (L- to H-mode) transition in the DIII-D tokamak is presented. Periodic turbulence suppression is first observed in a narrow layer at and just inside the separatrix when the shearing rate transiently exceeds the turbulence decorrelation rate. The final transition to H mode with sustained turbulence and transport reduction is controlled by equilibrium E×B shear due to the increasing ion pressure gradient.

  2. Transient waveform acquisition system for the ELMO Bumpy Torus

    International Nuclear Information System (INIS)

    Young, K.G.; Burris, R.D.; Hillis, D.H.; Overbey, D.R.

    1984-10-01

    The transient waveform system described in this report is designed to acquire analog waveforms from the ELMO Bumpy Torus (EBT) diagnostic experiments. Pressure, density, synchrotron radiation, etc., are acquired and digitized with a Kinetic Systems TR812 transient recorder and associated modules located in a CAMAC crate. The system can simultaneously acquire, display, and transmit sets of data consisting of identification parameters and up to 1024 data points for 1 to 64 input signals (frequency range = 0.01 pulse/s to 100 kHz) of data every one or more minutes; thus, it can run continuously without operator intervention. The data are taken on a VAX 11/780 and transmitted to a data base on a DECSystem-10. To aid the programmer in making future modifications to the system, detailed documentation using the Yourdon structural methods has been given

  3. Latitudinal oscillations of plasma within the Io torus

    Science.gov (United States)

    Cummings, W. D.; Dessler, A. J.; Hill, T. W.

    1980-01-01

    The equilibrium latitude and the period of oscillations about this equilibrium latitude are calculated for a plasma in a centrifugally dominated tilted dipole magnetic field representing Jupiter's inner magnetosphere. It is found that for a hot plasma the equilibrium latitude in the magnetic equator, for a cold plasma it is the centrifugal equator, and for a warm plasma it is somewhere in between. An illustrative model is adopted in which atoms are sputtered from the Jupiter-facing hemisphere of Io and escape Io's gravity to be subsequently ionized some distance from Io. Finally, it is shown that ionization generally does not occur at the equilibrium altitude, and that the resulting latitudinal oscillations provide an explanation for the irregularities in electron concentration within the torus, as reported by the radioastronomy experiment aboard Voyager I.

  4. Two dimensional critical models on a torus

    International Nuclear Information System (INIS)

    Saleur, H.; Di Francesco, P.

    1987-01-01

    After the general developments of conformal invariance in two dimensions, it was realized that the study of critical models in finite geometries, in addition to the practical information it could provide through finite size scaling, was also of great conceptual interest. The simplest example is the case of the torus, a genus 1 surface which is thus not conformally equivalent to the plane. This geometry appears quite frequently in lattice calculations for systems with periodic boundary conditions, and is also very natural from the point of view of string theory. We will discuss briefly in these notes the main results obtained so far in this simple case

  5. The Schwinger Model on the torus

    International Nuclear Information System (INIS)

    Azakov, S.

    1996-08-01

    The classical and quantum aspects of the Schwinger model on the torus are considered. First we find explicitly all zero modes of the Dirac operator in the topological sectors with nontrivial Chern index and its spectrum. In the second part we determine the regularized effective action and discuss the propagators related to it. Finally we calculate the gauge invariant averages of the fermion bilinears and correlation functions of currents and densities. We show that in the infinite volume limit the well-known result for the chiral condensate can be obtained and the clustering property can be established. (author). 23 refs

  6. Simulations of particle and heat fluxes in an ELMy H-mode discharge on EAST using BOUT++ code

    Science.gov (United States)

    Wu, Y. B.; Xia, T. Y.; Zhong, F. C.; Zheng, Z.; Liu, J. B.; team3, EAST

    2018-05-01

    In order to study the distribution and evolution of the transient particle and heat fluxes during edge-localized mode (ELM) bursts on the Experimental Advanced Superconducting Tokamak (EAST), the BOUT++ six-field two-fluid model is used to simulate the pedestal collapse. The profiles from the EAST H-mode discharge #56129 are used as the initial conditions. Linear analysis shows that the resistive ballooning mode and drift-Alfven wave are two dominant instabilities for the equilibrium, and play important roles in driving ELMs. The evolution of the density profile and the growing process of the heat flux at divertor targets during the burst of ELMs are reproduced. The time evolution of the poloidal structures of T e is well simulated, and the dominant mode in each stage of the ELM crash process is found. The studies show that during the nonlinear phase, the dominant mode is 5, and it changes to 0 when the nonlinear phase goes to saturation after the ELM crash. The time evolution of the radial electron heat flux, ion heat flux, and particle density flux at the outer midplane (OMP) are obtained, and the corresponding transport coefficients D r, χ ir, and χ er reach maximum around 0.3 ∼ 0.5 m2 s‑1 at ΨN = 0.9. The heat fluxes at outer target plates are several times larger than that at inner target plates, which is consistent with the experimental observations. The simulated profiles of ion saturation current density (j s) at the lower outboard (LO) divertor target are compared to those of experiments by Langmuir probes. The profiles near the strike point are similar, and the peak values of j s from simulation are very close to the measurements.

  7. Investigation of peeling-ballooning stability prior to transient outbursts accompanying transitions out of H-mode in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Eldon, D., E-mail: deldon@princeton.edu [University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093-0964 (United States); Princeton University, Princeton, New Jersey 08543 (United States); Boivin, R. L.; Groebner, R. J.; Osborne, T. H.; Snyder, P. B.; Turnbull, A. D.; Burrell, K. H. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Tynan, G. R.; Boedo, J. A. [University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093-0964 (United States); Kolemen, E. [Princeton University, Princeton, New Jersey 08543 (United States); Schmitz, L. [University of California Los Angeles, Los Angeles, California 90095-7099 (United States); Wilson, H. R. [University of York, Heslington, York YO10 5DD (United Kingdom)

    2015-05-15

    The H-mode transport barrier allows confinement of roughly twice as much energy as in an L-mode plasma. Termination of H-mode necessarily requires release of this energy, and the timescale of that release is of critical importance for the lifetimes of plasma facing components in next step tokamaks such as ITER. H-L transition sequences in modern tokamaks often begin with a transient outburst which appears to be superficially similar to and has sometimes been referred to as a type-I edge localized mode (ELM). Type-I ELMs have been shown to be consistent with ideal peeling ballooning instability and are characterized by significant (up to ∼50%) reduction of pedestal height on short (∼1 ms) timescales. Knowing whether or not this type of instability is present during H-L back transitions will be important of planning for plasma ramp-down in ITER. This paper presents tests of pre-transition experimental data against ideal peeling-ballooning stability calculations with the ELITE code and supports those results with secondary experiments that together show that the transient associated with the H-L transition is not triggered by the same physics as are type-I ELMs.

  8. On some Closed Magnetic Curves on a 3-torus

    Energy Technology Data Exchange (ETDEWEB)

    Munteanu, Marian Ioan, E-mail: marian.ioan.munteanu@gmail.com [Alexandru Ioan Cuza University of Iaşi, Faculty of Mathematics (Romania); Nistor, Ana Irina, E-mail: ana.irina.nistor@gmail.com [Gh. Asachi Technical University of Iaşi, Department of Mathematics and Informatics (Romania)

    2017-06-15

    We consider two magnetic fields on the 3-torus obtained from two different contact forms on the Euclidean 3-space and we study when their corresponding normal magnetic curves are closed. We obtain periodicity conditions analogues to those for the closed geodesics on the torus.

  9. Ballooning mode stability for self-consistent pressure and current profiles at the H-mode edge

    International Nuclear Information System (INIS)

    Miller, R.L.; Lin-Liu, Y.R.; Osborne, T.H.; Taylor, T.S.

    1997-11-01

    The edge pressure gradient (H-mode pedestal) for computed equilibria in which the current density profile is consistent with the bootstrap current may not be limited by the first regime ballooning limit. The transition to second stability is easier for: higher elongation, intermediate triangularity, larger ratio, pedestal at larger radius, narrower pedestal width, higher q 95 , and lower collisionality

  10. H-mode transition physics close to double null on MAST and its applications to other tokamaks

    International Nuclear Information System (INIS)

    Meyer, H.; Carolan, P.G.; Cunningham, G.; Kirk, A.; Lloyd, B.; Saarelma, S.; Wilson, H.R.; Conway, G.D.; Horton, L.D.; Ryter, F.; Schirmer, J.; Suttrop, W.; Maingi, R.

    2005-01-01

    By accessing extreme parameter regimes combined with well diagnosed edge MAST data contribute towards the understanding of H-mode physics. The first inter-machine comparisons with respect to the influence of the magnetic topology on the power threshold with ASDEX Upgrade and NSTX reveal a reduction of the power threshold in true double null (C-DN) configuration opening new operation regimes in both devices. In L-mode, the negative radial electric field close to the separatrix was found to be more negative in C-DN than in single null (SN), whilst most of the other edge parameters are similar. Pedestal temperatures in MAST are lower than in ASDEX Upgrade in MAST-equivalent discharges, whereas the pedestal densities can be similar, although in long inter ELM periods the MAST density pedestal is higher than on ASDEX Upgrade. In order to test four leading H-mode theories MAST data are compared statistically to their H-mode access criteria. The usual DN operating regime with co current NBI in MAST has been extended to include single null (SN) configurations, to provide more direct comparisons with conventional tokamaks. The plasma edge in SN on MAST is more stable to ELMs and the typical type-III ELMs, often observed in C-DN, are absent, despite input powers close to the H-mode threshold power. In this respect, the stability of measured plasma edge profiles in SN and DN against ideal peeling-ballooning modes will be discussed. (author)

  11. Metal impurity transport control in JET H-mode plasmas with central ion cyclotron radiofrequency power injection

    DEFF Research Database (Denmark)

    Valisa, M.; Carraro, L.; Predebon, I.

    2011-01-01

    The scan of ion cyclotron resonant heating (ICRH) power has been used to systematically study the pump out effect of central electron heating on impurities such as Ni and Mo in H-mode low collisionality discharges in JET. The transport parameters of Ni and Mo have been measured by introducing...

  12. H-mode pedestal characteristics, ELMs, and energy confinement in ITER shape discharges on DIII-D

    International Nuclear Information System (INIS)

    Osborne, T.H.; Groebner, R.J.; Lao, L.L.; Leonard, A.W.; Miller, R.L.; Thomas, D.M.; Waltz, R.E.; Maingi, R.; Porter, G.D.

    1997-12-01

    The H-mode confinement enhancement factor, H, is found to be strongly correlated with the height of the edge pressure pedestal in ITER shape discharges. In discharges with Type I ELMs the pedestal pressure is set by the maximum pressure gradient before the ELM and the width of the H-mode transport barrier. The pressure gradient before Type I ELMs is found to scale as would be expected for a stability limit set by ideal ballooning modes, but with values significantly in excess of that predicted by stability code calculations. The width of the H-mode transport barrier is found to scale equally well with pedestal P(POL)(2/3) or B(POL)(1/2). The improved H value in high B(POL) discharges may be due to a larger edge pressure gradient and wider H-mode transport barrier consistent with their higher edge ballooning mode limit. Deuterium puffing is found to reduce H consistent with the smaller pedestal pressure which results from the reduced barrier width and critical pressure gradient. Type I ELM energy loss is found to be proportional to the change in the pedestal energy

  13. Differences in the H-mode pedestal width of temperature and density

    International Nuclear Information System (INIS)

    Schneider, P A; Wolfrum, E; Günter, S; Kurzan, B; Lackner, K; Zohm, H; Groebner, R J; Osborne, T H; Ferron, J R; Snyder, P B; Beurskens, M N A; Dunne, M G

    2012-01-01

    A pedestal database was built using data from type-I ELMy H-modes of ASDEX Upgrade, DIII-D and JET. ELM synchronized pedestal data were analysed with the two-line method. The two-line method is a bilinear fit which shows better reproducibility of pedestal parameters than a modified hyperbolic tangent fit. This was tested with simulated and experimental data. The influence of the equilibrium reconstruction on pedestal parameters was investigated with sophisticated reconstructions from CLISTE and EFIT including edge kinetic profiles. No systematic deviation between the codes could be observed. The flux coordinate system is influenced by machine size, poloidal field and plasma shape. This will change the representation of the width in different coordinates, in particular, the two normalized coordinates Ψ N and r/a show a very different dependence on the plasma shape. The scalings derived for the pedestal width, Δ, of all machines suggest a different scaling for the electron temperature and the electron density. Both cases show similar dependence with machine size, poloidal magnetic field and pedestal electron temperature and density. The influence of ion temperature and toroidal magnetic field is different on each of Δ T e and Δ n e . In dimensionless form the density pedestal width in Ψ N scales with ρ 0.6 i* , the temperature pedestal width with β p,ped 0.5 . Both widths also show a strong correlation with the plasma shape. The shape dependence originates from the coordinate transformation and is not visible in real space. The presented scalings predict that in ITER the temperature pedestal will be appreciably wider than the density pedestal. (paper)

  14. First-wall heat-flux measurements during ELMing H-mode plasma

    International Nuclear Information System (INIS)

    Lasnier, C.J.; Allen, S.L.; Hill, D.N.; Leonard, A.W.; Petrie, T.W.

    1994-01-01

    In this report we present measurements of the diverter heat flux in DIII-D for ELMing H-mode and radiative diverter conditions. In previous work we have examined heat flux profiles in lower single-null diverted plasmas and measured the scaling of the peak heat flux with plasma current and beam power. One problem with those results was our lack of good power accounting. This situation has been improved to better than 80--90% accountability with the installation of new bolometer arrays, and the operation of the entire complement of 5 Infrared (IR) TV cameras using the DAPS (Digitizing Automated Processing System) video processing system for rapid inter-shot data analysis. We also have expanded the scope of our measurements to include a wider variety of plasma shapes (e.g., double-null diverters (DND), long and short single-null diverters (SND), and inside-limited plasmas), as well as more diverse discharge conditions. Double-null discharges are of particular interest because that shape has proven to yield the highest confinement (VH-mode) and beta of all DIII-D plasmas, so any future diverter modifications for DIII-D will have to support DND operation. In addition, the proposed TPX tokamak is being designed for double-null operation, and information on the magnitude and distribution of diverter heat flux is needed to support the engineering effort on that project. So far, we have measured the DND power sharing at the target plates and made preliminary tests of heat flux reduction by gas injection

  15. Pellet injection into H-mode ITER plasma with the presence of internal transport barriers

    Science.gov (United States)

    Leekhaphan, P.; Onjun, T.

    2011-04-01

    The impacts of pellet injection into ITER type-1 ELMy H-mode plasma with the presence of internal transport barriers (ITBs) are investigated using self-consistent core-edge simulations of 1.5D BALDUR integrated predictive modeling code. In these simulations, the plasma core transport is predicted using a combination of a semi-empirical Mixed B/gB anomalous transport model, which can self-consistently predict the formation of ITBs, and the NCLASS neoclassical model. For simplicity, it is assumed that toroidal velocity for ω E× B calculation is proportional to local ion temperature. In addition, the boundary conditions are predicted using the pedestal temperature model based on magnetic and flow shear stabilization width scaling; while the density of each plasma species, including both hydrogenic and impurity species, at the boundary are assumed to be a large fraction of its line averaged density. For the pellet's behaviors in the hot plasma, the Neutral Gas Shielding (NGS) model by Milora-Foster is used. It was found that the injection of pellet could result in further improvement of fusion performance from that of the formation of ITB. However, the impact of pellet injection is quite complicated. It is also found that the pellets cannot penetrate into a deep core of the plasma. The injection of the pellet results in a formation of density peak in the region close to the plasma edge. The injection of pellet can result in an improved nuclear fusion performance depending on the properties of pellet (i.e., increase up to 5% with a speed of 1 km/s and radius of 2 mm). A sensitivity analysis is carried out to determine the impact of pellet parameters, which are: the pellet radius, the pellet velocity, and the frequency of injection. The increase in the pellet radius and frequency were found to greatly improve the performance and effectiveness of fuelling. However, changing the velocity is observed to exert small impact.

  16. Temporal evolution of H-mode pedestal in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Osborne, T.H.; Leonard, A.W.; Fenstermacher, M.E.

    2009-01-01

    The temporal evolution of pedestal parameters is examined in the initial edge localized mode (ELM)-free phase and inter-ELM phases of H-mode discharges in the DIII-D tokamak. These discharges are heated by deuterium neutral beam injection and achieve type-I ELMing conditions. Pedestal parameters exhibit qualitatively similar behaviour in both the ELM-free and inter-ELM phases. There is a trend for the widths and heights of pedestals for electron density, temperature and pressure to increase during these phases; the increase in width is most pronounced in the density and least pronounced in electron temperature. Near the separatrix, the ion temperature achieves higher values but a flatter profile as compared with the electron temperature. Higher heating powers lead to a faster evolution of the pedestal and to a shorter period until the onset of an ELM. For sufficiently long ELM-free or inter-ELM periods, some parameters, particularly gradients, approach a steady state. However, a simultaneous steady state in all parameters is not observed. The simultaneous increase in density width and pedestal density is opposite to the predictions of a simple model, which predicts that the density width is set by neutral penetration. Thus, additional physics must be added to the simple model to provide a more general description of pedestal behaviour. However, the barrier growth is qualitatively consistent with time-dependent theoretical models that predict a self-consistent temporal growth of the pedestal due to E x B shearing effects. In addition, an approximate linear correlation is observed between the density width and the square root of the pedestal ion temperature and also between the density width and the square root of the pedestal beta poloidal. These pedestal studies suggest that a complete model of the pedestal width in type-I ELMing discharges must be time dependent, include transport physics during inter-ELM periods and include the limits to pedestal evolution

  17. Arithmetic functions in torus and tree networks

    Science.gov (United States)

    Bhanot, Gyan; Blumrich, Matthias A.; Chen, Dong; Gara, Alan G.; Giampapa, Mark E.; Heidelberger, Philip; Steinmacher-Burow, Burkhard D.; Vranas, Pavlos M.

    2007-12-25

    Methods and systems for performing arithmetic functions. In accordance with a first aspect of the invention, methods and apparatus are provided, working in conjunction of software algorithms and hardware implementation of class network routing, to achieve a very significant reduction in the time required for global arithmetic operation on the torus. Therefore, it leads to greater scalability of applications running on large parallel machines. The invention involves three steps in improving the efficiency and accuracy of global operations: (1) Ensuring, when necessary, that all the nodes do the global operation on the data in the same order and so obtain a unique answer, independent of roundoff error; (2) Using the topology of the torus to minimize the number of hops and the bidirectional capabilities of the network to reduce the number of time steps in the data transfer operation to an absolute minimum; and (3) Using class function routing to reduce latency in the data transfer. With the method of this invention, every single element is injected into the network only once and it will be stored and forwarded without any further software overhead. In accordance with a second aspect of the invention, methods and systems are provided to efficiently implement global arithmetic operations on a network that supports the global combining operations. The latency of doing such global operations are greatly reduced by using these methods.

  18. Torus actions and their applications in topology and combinatorics

    CERN Document Server

    Buchstaber, Victor M

    2002-01-01

    The book presents the study of torus actions on topological spaces is presented as a bridge connecting combinatorial and convex geometry with commutative and homological algebra, algebraic geometry, and topology. This established link helps in understanding the geometry and topology of a space with torus action by studying the combinatorics of the space of orbits. Conversely, subtle properties of a combinatorial object can be realized by interpreting it as the orbit structure for a proper manifold or as a complex acted on by a torus. The latter can be a symplectic manifold with Hamiltonian torus action, a toric variety or manifold, a subspace arrangement complement, etc., while the combinatorial objects include simplicial and cubical complexes, polytopes, and arrangements. This approach also provides a natural topological interpretation in terms of torus actions of many constructions from commutative and homological algebra used in combinatorics. The exposition centers around the theory of moment-angle comple...

  19. Stability of Microturbulent Drift Modes during Internal Transport Barrier Formation in the Alcator C-Mod Radio Frequency Heated H-mode

    International Nuclear Information System (INIS)

    Redi, M.H.; Fiore, C.L.; Dorland, W.; Mikkelsen, D.R.; Rewoldt, G.; Bonoli, P.T.; Ernst, D.R.; Rice, J.E.; Wukitch, S.J.

    2003-01-01

    Recent H-mode experiments on Alcator C-Mod [I.H. Hutchinson, et al., Phys. Plasmas 1 (1994) 1511] which exhibit an internal transport barrier (ITB), have been examined with flux tube geometry gyrokinetic simulations, using the massively parallel code GS2 [M. Kotschenreuther, G. Rewoldt, and W.M. Tang, Comput. Phys. Commun. 88 (1995) 128]. The simulations support the picture of ion/electron temperature gradient (ITG/ETG) microturbulence driving high xi/ xe and that suppressed ITG causes reduced particle transport and improved ci on C-Mod. Nonlinear calculations for C-Mod confirm initial linear simulations, which predicted ITG stability in the barrier region just before ITB formation, without invoking E x B shear suppression of turbulence. Nonlinear fluxes are compared to experiment, which both show low heat transport in the ITB and higher transport within and outside of the barrier region

  20. Pellet injection into H-mode ITER plasma with the presence of internal transport barriers

    Energy Technology Data Exchange (ETDEWEB)

    Leekhaphan, P. [Thammasat University, School of Bio-Chemical Engineering and Technology, Sirindhorn International Institute of Technology (Thailand); Onjun, T. [Thammasat University, School of Manufacturing Systems and Mechanical Engineering, Sirindhorn International Institute of Technology (Thailand)

    2011-04-15

    The impacts of pellet injection into ITER type-1 ELMy H-mode plasma with the presence of internal transport barriers (ITBs) are investigated using self-consistent core-edge simulations of 1.5D BALDUR integrated predictive modeling code. In these simulations, the plasma core transport is predicted using a combination of a semi-empirical Mixed B/gB anomalous transport model, which can self-consistently predict the formation of ITBs, and the NCLASS neoclassical model. For simplicity, it is assumed that toroidal velocity for {omega}{sub E Multiplication-Sign B} calculation is proportional to local ion temperature. In addition, the boundary conditions are predicted using the pedestal temperature model based on magnetic and flow shear stabilization width scaling; while the density of each plasma species, including both hydrogenic and impurity species, at the boundary are assumed to be a large fraction of its line averaged density. For the pellet's behaviors in the hot plasma, the Neutral Gas Shielding (NGS) model by Milora-Foster is used. It was found that the injection of pellet could result in further improvement of fusion performance from that of the formation of ITB. However, the impact of pellet injection is quite complicated. It is also found that the pellets cannot penetrate into a deep core of the plasma. The injection of the pellet results in a formation of density peak in the region close to the plasma edge. The injection of pellet can result in an improved nuclear fusion performance depending on the properties of pellet (i.e., increase up to 5% with a speed of 1 km/s and radius of 2 mm). A sensitivity analysis is carried out to determine the impact of pellet parameters, which are: the pellet radius, the pellet velocity, and the frequency of injection. The increase in the pellet radius and frequency were found to greatly improve the performance and effectiveness of fuelling. However, changing the velocity is observed to exert small impact.

  1. An FPGA-based torus communication network

    Energy Technology Data Exchange (ETDEWEB)

    Pivanti, Marcello; Schifano, Sebastiano Fabio [INFN, Ferrara (Italy); Ferrara Univ. (Italy); Simma, Hubert [DESY, Zeuthen (Germany). John von Neumann-Institut fuer Computing NIC

    2011-02-15

    We describe the design and FPGA implementation of a 3D torus network (TNW) to provide nearest-neighbor communications between commodity multi-core processors. The aim of this project is to build up tightly interconnected and scalable parallel systems for scientific computing. The design includes the VHDL code to implement on latest FPGA devices a network processor, which can be accessed by the CPU through a PCIe interface and which controls the external PHYs of the physical links. Moreover, a Linux driver and a library implementing custom communication APIs are provided. The TNW has been successfully integrated in two recent parallel machine projects, QPACE and AuroraScience. We describe some details of the porting of the TNW for the AuroraScience system and report performance results. (orig.)

  2. An FPGA-based torus communication network

    International Nuclear Information System (INIS)

    Pivanti, Marcello; Schifano, Sebastiano Fabio; Simma, Hubert

    2011-02-01

    We describe the design and FPGA implementation of a 3D torus network (TNW) to provide nearest-neighbor communications between commodity multi-core processors. The aim of this project is to build up tightly interconnected and scalable parallel systems for scientific computing. The design includes the VHDL code to implement on latest FPGA devices a network processor, which can be accessed by the CPU through a PCIe interface and which controls the external PHYs of the physical links. Moreover, a Linux driver and a library implementing custom communication APIs are provided. The TNW has been successfully integrated in two recent parallel machine projects, QPACE and AuroraScience. We describe some details of the porting of the TNW for the AuroraScience system and report performance results. (orig.)

  3. Space Propulsion via Spherical Torus Fusion Reactor

    International Nuclear Information System (INIS)

    Williams, Craig H.; Juhasz, Albert J.; Borowski, Stanley K.; Dudzinski, Leonard A.

    2003-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 204 days, with an initial mass in low Earth orbit of 1630 mt. Engineering conceptual design, analysis, and assessment were performed on all major systems including nuclear fusion reactor, magnetic nozzle, power conversion, fast wave plasma heating, fuel pellet injector, startup/re-start fission reactor and battery, and other systems. Detailed fusion reactor design included analysis of plasma characteristics, power balance and utilization, first wall, toroidal field coils, heat transfer, and neutron/X-ray radiation

  4. Effect of robust torus on the dynamical transport

    International Nuclear Information System (INIS)

    Martins, C G L; Carvalho, R Egydio de; Caldas, I L; Roberto, M

    2010-01-01

    In the present work, we quantify the fraction of trajectories that reach a specific region of the phase space when we vary a control parameter using two symplectic maps: one non-twist and another one twist. The two maps were studied with and without a robust torus. We compare the obtained patterns and we identify the effect of the robust torus on the dynamical transport. We show that the effect of meandering-like barriers loses importance in blocking the radial transport when the robust torus is present.

  5. A new boundary control scheme for simultaneous achievement of H-mode and radiative cooling (SHC boundary)

    International Nuclear Information System (INIS)

    Ohyabu, N.

    1995-05-01

    We have proposed a new boundary control scheme (SHC boundary), which could allow simultaneous achievement of the H-mode type confinement improvement and radiative cooling with wide heat flux distribution. In our proposed configuration, a low m island layer sharply separates a plasma confining region from an open 'ergodic' boundary. The degree of openness in the ergodic boundary must be high enough to make the plasma pressure constant along the field line, which in turn separates low density plasma just outside the plasma confining region (the key external condition for achieving a good H-mode discharge) from very high density, cold radiative plasma near the wall (required for effective edge radiative cooling). Examples of such proposed SHC boundaries for Heliotron typed devices and tokamaks are presented. (author)

  6. Characterisation of the ELM synchronized H-mode edge pedestal in ASDEX upgrade and DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Philip A.; Wolfrum, Elisabeth; Guenter, Sibylle; Kurzan, Bernd; Zohm, Hartmut [Max Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Groebner, Rich; Osborne, Tom H.; Ferron, John; Snyder, Philip B. [General Atomics, San Diego, CA (United States); Dunne, Mike G. [Department of Physics, University College Cork, Association Euratom-DCU, Cork (Ireland); Collaboration: ASDEX Upgrade Team; DIII-D Team

    2011-07-01

    The results of a large database of edge pedestal data from type-I ELMy H-mode discharges from ASDEX Upgrade and DIII-D are presented. The data from high resolution edge diagnostics of both devices is analysed with the same analysis code in order to avoid systematic differences. Furthermore, sophisticated equilibrium reconstructions are used to asses uncertainties which arise during mapping from 2D real space coordinates to 1D flux coordinates. ELM synchronization allows the study of the pedestal structure at the ELM stability boundary. The pedestal is characterized by its top value, the gradient and the width. A large parameter range is covered by the two devices. Over this parameter range the profile shape of edge electron density differs from that of the temperature, irrespective of the device. However, the resulting electron pressure profile shape remains similar for all analysed H-Mode discharges.

  7. Scaling of ELM and H-mode pedestal characteristics in ITER shape discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Osborne, T.H.; Groebner, R.J.; Lao, L.L.; Leonard, A.W.; Miller, R.L.; Thomas, D.M.; Waltz, R.E.; Maingi, R.; Porter, G.D.

    1997-07-01

    The authors have shown a correlation between the H-mode pressure pedestal height and the energy confinement enhancement in ITER shape discharges on DIII-D which is consistent with the behavior of H in different ELM classes. The width of the steep gradient region was found to equally well fit the scalings δ/R ∝ (ρ POL /R) 2/3 and δ/R ∝ (β POL PED /R) 1/2 . The normalized pressure gradient α MHD was found to be relatively constant just before a type I ELM. An estimate of T PED for ITER gave 1 to 5 keV. They also estimate ΔE ELM ≅ 26 MJ for ITER. They identified a distinct class of type III ELM at low density which may play a role in setting H at powers near the H-mode threshold power

  8. Measurement of peripheral electron temperature by electron cyclotron emission during the H-mode transition in JFT-2M tokamak

    International Nuclear Information System (INIS)

    Hoshino, Katsumichi; Yamamoto, Takumi; Kawashima, Hisato

    1987-01-01

    Time evolution and profile of peripheral electron temperature during the H-mode like transition in a tokamak plasma is measured using the second and third harmonic of electron cyclotron emission (ECE). The so called ''H-mode'' state which has good particle/energy confinement is characterized by sudden decrease in the spectral line intensity of deuterium molecule. Such a sudden decrease in the line intensity of D α with good energy confinement is found not only in divertor discharges, but also in limiter dischargs in JFT-2M tokamak. It is found by the measurement of ECE that the peripheral electron temperature suddenly increases in both of such phases. The relation between H-transition and the peripheral electron temperature or its profile is investigated. (author)

  9. 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.

  10. Suppression of tungsten accumulation during ELMy H-mode by lower hybrid wave heating in the EAST tokamak

    Directory of Open Access Journals (Sweden)

    L. Zhang

    2017-08-01

    Full Text Available EAST tokamak has been equipped with upper tungsten divertor since 2014. The tungsten accumulation has been often observed in NBI-heated H-mode discharges suggesting deleterious tungsten confinement in the plasma core. It causes not only H-L back transition but also plasma disruption in several discharges. Suppression of the tungsten accumulation is therefore the most important issue in EAST to achieve a long pulse H-mode discharge. In order to study the tungsten behavior in the long pulse discharge, tungsten spectra have been measured at 20–140Å. The tungsten density, nw, is evaluated from the intensity of tungsten unresolved transition array (W-UTA in a wavelength range of 45–70Å which is composed of several ionization stages of tungsten, e.g. W27+-W45+ at Te0∼2.5keV. It is found that the tungsten accumulation can be suppressed when the 4.6GHz LHW with PLHW∼0.8MW is superimposed on the NBI phase (PNBI= 1.9MW. During the superimposed phase the ELM frequency, fELM, increases from ∼30Hz to ∼60Hz and the tungsten density is halved compared to the NBI-heated discharge. The H-mode discharge can be thus steadily sustained for longer period. It is found that the nw is a large function of the ratio of LHW power to the total injection power, PLHW/(PLHW+PNBI, and the nw can be reduced, at least, in an order of magnitude smaller than that in NBI-heated discharges at PLHW/(PLHW+PNBI≥0.8. The result strongly suggests a possible way toward the steady H-mode discharge.

  11. Effect of Wave Accessibility on Lower Hybrid Wave Current Drive in Experimental Advanced Superconductor Tokamak with H-Mode Operation

    International Nuclear Information System (INIS)

    Li Xin-Xia; Xiang Nong; Gan Chun-Yun

    2015-01-01

    The effect of the wave accessibility condition on the lower hybrid current drive in the experimental advanced superconductor Tokamak (EAST) plasma with H-mode operation is studied. Based on a simplified model, a mode conversion layer of the lower hybrid wave between the fast wave branch and the slow wave branch is proved to exist in the plasma periphery for typical EAST H-mode parameters. Under the framework of the lower hybrid wave simulation code (LSC), the wave ray trajectory and the associated current drive are calculated numerically. The results show that the wave accessibility condition plays an important role on the lower hybrid current drive in EAST plasma. For wave rays with parallel refractive index n ‖ = 2.1 or n ‖ = 2.5 launched from the outside midplane, the wave rays may penetrate the core plasma due to the toroidal geometry effect, while numerous reflections of the wave ray trajectories in the plasma periphery occur. However, low current drive efficiency is obtained. Meanwhile, the wave accessibility condition is improved if a higher confined magnetic field is applied. The simulation results show that for plasma parameters under present EAST H-mode operation, a significant lower hybrid wave current drive could be obtained for the wave spectrum with peak value n ‖ = 2.1 if a toroidal magnetic field B T = 2.5 T is applied. (paper)

  12. Vertical injection of compact torus into the STOR-M tokamak

    International Nuclear Information System (INIS)

    Liu, D.; Singh, A.K.; Hirose, A.; Xiao, C.

    2005-01-01

    Vertical compact torus injection into the STOR-M tokamak has been conducted with the University of Saskatchewan Compact Torus Injector (USCTI). The injector stayed at the horizontal position and the CT was bent by 90 deg. using a curved conducting drift tube. The curved drift tube did not have significant effects on the CT velocity. Furthermore, the curved drift tube did not change the magnetic field topology. Preliminary vertical CT injection experiments have been carried out on the STOR-M tokamak. CT injection induced prompt increase in the electron density and in the soft x-ray radiation level. Further modifications of the 90 deg. are underway to improve the CT parameters and to further study the effects of CT injection on the tokamak plasma parameters. (author)

  13. Plasma dynamics with second and third-harmonic ECRH and access to quasi-stationary ELM-free H-mode on TCV

    International Nuclear Information System (INIS)

    Porte, L.; Coda, S.; Alberti, S.; Arnoux, G.; Blanchard, P.; Bortolon, A.; Fasoli, A.; Goodman, T.P.; Klimanov, Y.; Martin, Y.; Maslov, M.; Scarabosio, A.; Weisen, H.

    2007-01-01

    Intense electron cyclotron resonance heating (ECRH) and electron cyclotron current drive (ECCD) are employed on the Tokamak a Configuration Variable (TCV) both in second- and third-harmonic X-mode (X2 and X3). The plasma behaviour under such conditions is driven largely by the electron dynamics, motivating extensive studies of the heating and relaxation phenomena governing both the thermal and suprathermal electron populations. In particular, the dynamics of suprathermal electrons are intimately tied to the physics of X2 ECCD. ECRH is also a useful tool for manipulating the electron distribution function in both physical and velocity space. Fundamental studies of the energetic electron dynamics have been performed using periodic, low-duty-cycle bursts of ECRH, with negligible average power injection, and with electron cyclotron emission (ECE). The characteristic times of the dynamical evolution are clearly revealed. Suprathermal electrons have also been shown to affect the absorption of X3 radiation. Thermal electrons play a crucial role in high density plasmas where indirect ion heating can be achieved through ion-electron collisions. In recent experiments ∼ 1.35 MW of vertically launched X3 ECRH was coupled to a diverted ELMy H-mode plasma. In cases where ≥ 1.1 MW of ECRH power was coupled, the discharge was able to transition into a quasi-stationary ELM-free H-mode regime. These H-modes operated at β N ∼ 2, n-bar e /n G approx. 0.25 and had high energy confinement, H IPB98(y,2) up to ∼ 1.6. Despite being purely electron heated and having no net particle source these discharges maintained a density peaking factor (n e,o /(n e ) ∼ 1.6). They also exhibited spontaneous toroidal momentum production in the co-current direction. The momentum production is due to a transport process as there is no external momentum input. This process supports little or no radial gradient of the toroidal velocity

  14. Study on H-mode access at low density with lower hybrid current drive and lithium-wall coatings on the EAST superconducting tokamak

    DEFF Research Database (Denmark)

    Xu, G.S.; Wan, B.N.; Li, J.G.

    2011-01-01

    The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of HIPB98(y,2) ~ 1 has been obtained with about 1 MW lower hybrid wave power on the EAST superconducting tokamak. The first H-mode plasma appeared after wall conditioning by lithium (Li) evaporation before ...

  15. Topological T-duality for torus bundles with monodromy

    Science.gov (United States)

    Baraglia, David

    2015-05-01

    We give a simplified definition of topological T-duality that applies to arbitrary torus bundles. The new definition does not involve Chern classes or spectral sequences, only gerbes and morphisms between them. All the familiar topological conditions for T-duals are shown to follow. We determine necessary and sufficient conditions for existence of a T-dual in the case of affine torus bundles. This is general enough to include all principal torus bundles as well as torus bundles with arbitrary monodromy representations. We show that isomorphisms in twisted cohomology, twisted K-theory and of Courant algebroids persist in this general setting. We also give an example where twisted K-theory groups can be computed by iterating T-duality.

  16. Torus as phase space: Weyl quantization, dequantization, and Wigner formalism

    Energy Technology Data Exchange (ETDEWEB)

    Ligabò, Marilena, E-mail: marilena.ligabo@uniba.it [Dipartimento di Matematica, Università di Bari, I-70125 Bari (Italy)

    2016-08-15

    The Weyl quantization of classical observables on the torus (as phase space) without regularity assumptions is explicitly computed. The equivalence class of symbols yielding the same Weyl operator is characterized. The Heisenberg equation for the dynamics of general quantum observables is written through the Moyal brackets on the torus and the support of the Wigner transform is characterized. Finally, a dequantization procedure is introduced that applies, for instance, to the Pauli matrices. As a result we obtain the corresponding classical symbols.

  17. On the energy crisis in the Io plasma torus

    Science.gov (United States)

    Smith, Robert A.; Bagenal, Fran; Cheng, Andrew F.; Strobel, Darrell

    1988-01-01

    Recent calculations of the energy balance of the Io plasma torus show that the observed UV and EUV radiation cannot be maintained solely via energy input by the ion pickup mechanism. Current theoretical models of the torus must be modified to include non-local energy input. It is argued that the required energy may be supplied by inward diffusion of energetic heavy ions with energies less than about 20 keV.

  18. The geometric Schwinger model on the torus. Pt. 1

    International Nuclear Information System (INIS)

    Joos, H.

    1990-01-01

    The author analyzes the Euclidean version of the geometric Schwinger model on the torus. After the calculation of the zero mode wave functions associated with the different topological sectors, which can be expressed by θ functions defined on the two-dimensional torus, he determines the regularized effective action and discusses the propagator related to it. Finally he studies applications to the standard questions like the particle spectrum, the screening of the static potential, and the appearance of the anomaly. (HSI)

  19. 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

  20. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost

  1. Hirzebruch genera of manifolds with torus action

    International Nuclear Information System (INIS)

    Panov, T E

    2001-01-01

    A quasitoric manifold is a smooth 2n-manifold M 2n with an action of the compact torus T n such that the action is locally isomorphic to the standard action of T n on C n and the orbit space is diffeomorphic, as a manifold with corners, to a simple polytope P n . The name refers to the fact that topological and combinatorial properties of quasitoric manifolds are similar to those of non-singular algebraic toric varieties (or toric manifolds). Unlike toric varieties, quasitoric manifolds may fail to be complex. However, they always admit a stably (or weakly almost) complex structure, and their cobordism classes generate the complex cobordism ring. Buchstaber and Ray have recently shown that the stably complex structure on a quasitoric manifold is determined in purely combinatorial terms, namely, by an orientation of the polytope and a function from the set of codimension-one faces of the polytope to primitive vectors of the integer lattice. We calculate the χ y -genus of a quasitoric manifold with a fixed stably complex structure in terms of the corresponding combinatorial data. In particular, this gives explicit formulae for the classical Todd genus and the signature. We also compare our results with well-known facts in the theory of toric varieties

  2. ADHM construction of instantons on the torus

    International Nuclear Information System (INIS)

    Ford, C.; Pawlowski, J.M.; Tok, T.; Wipf, A.

    2001-01-01

    We apply the ADHM instanton construction to SU(2) gauge theory on T n xR 4-n for n=1,2,3,4. To do this we regard instantons on T n xR 4-n as periodic (modulo gauge transformations) instantons on R 4 . Since the R 4 topological charge of such instantons is infinite the ADHM algebra takes place on an infinite dimensional linear space. The ADHM matrix M is related to a Weyl operator (with a self-dual background) on the dual torus T-tilde n . We construct the Weyl operator corresponding to the one-instantons on T n xR 4-n . In order to derive the self-dual potential on T n xR 4-n it is necessary to solve a specific Weyl equation. This is a variant of the Nahm transformation. In the case n=2 (i.e., T 2 xR 2 ) we essentially have an Aharonov-Bohm problem on T-tilde 2 . In the one-instanton sector we find that the scale parameter, λ, is bounded above, λ 2 V-tilde 2

  3. Study of density fluctuation in L-mode and H-mode plasmas on JFT-2M by microwave reflectometer

    International Nuclear Information System (INIS)

    Shinohara, Kouji

    1997-08-01

    We propose the model which can explain the runaway phase. The model takes account of the scattered wave which is caused by the density fluctuation near the cut-off layer. We should take a new approach instead of the conventional phase measurement in order to derive the information of the density fluctuation from the data with the runaway phase. The complex spectrum and the rotary spectrum analyses are useful tools to analyze such data. The density fluctuation in L-mode and H-mode plasmas is discussed by using this new approach. We have observed that the reduction of the density fluctuation is localized in the edge region where the sheared electric field is produced. The fluctuations in the range of frequency lower than 100 kHz are mainly reduced. Two interesting features have been observed. One is the detection of the coherent mode around 100 kHz in H-mode. This mode appears about 10 ms after L to H transition. The timing corresponds to the formation of a steep density and temperature gradient in the edge region. The other is the enhancement of the fluctuations with the frequency higher than 300 kHz in H-mode in contrast to the reduction of the fluctuations with the frequency lower than 100 kHz. The Doppler shift is observed in the complex auto-power spectrum of the reflected wave when the plasma is actively moved. We have confirmed that the movement of the plasma is appropriately measured by using the low pass filter. The reflectometer can be used to measure the density profile by using a low pass filter even when the runaway phase phenomenon occurs. (author). 150 refs

  4. New features of L-H transition in limiter H-modes of JIPP T-IIU

    International Nuclear Information System (INIS)

    Toi, K.; Morita, S.; Kawahata, K.

    1992-09-01

    In limiter H-modes of JIPP T-IIU, a new type of L-H transition preceded by an ELM is observed. The preceding ELM (pre-ELM) appears just prior to the L-H transition. This type of transition is usually observed in H-modes of JIPP T-IIU. The L-H transition without the pre-ELM is triggered only in the case when a sufficiently large rapid current ramp down is emploied. In H-modes with constant q(a)∼3.5-4.5, coherent magnetic oscillations with m=3/n=1 destabilized during L-phase are further enhanced at the pre-ELM, and suppressed suddenly at the transition. This mode is situated in the region of the transport barrier. Propagation frequency of the m=3/n=1 mode, which may be affected by plasma mass rotation, rises appreciably (by ∼ 10 %) during H-phase with frequent ELMs, but remains unchanged for at least 200 μs after the transition. Behaviours of the m=3/n=1 and m=2/n=1 modes are well explained by quasi-linear resistive tearing mode analysis for modelled toroidal current density profiles slightly detached from the limiter. These experimental results suggest that the transition is controlled by the change of a magnetic field structure relating to the modification of a toroidal current density profile near the edge. The possibility for the development of edge radial electric field as a consequence of the transition is discussed. (author)

  5. Long sustainment of quasi-steady-state high βp H mode discharges in JT-60U

    International Nuclear Information System (INIS)

    Isayama, A.; Kamada, Y.; Ozeki, T.; Ide, S.; Fujita, T.; Oikawa, T.; Suzuki, T.; Neyatani, Y.; Isei, N.; Hamamatsu, K.; Ikeda, Y.; Takahashi, K.; Kajiwara, K.

    2001-01-01

    Quasi-steady-state high β p H mode discharges performed by suppressing neoclassical tearing modes (NTMs) are described. Two operational scenarios have been developed for long sustainment of the high β p H mode discharge: NTM suppression by profile optimization, and NTM stabilization by local electron cyclotron current drive (ECCD)/electron cyclotron heating (ECH) at the magnetic island. Through optimization of pressure and safety factor profiles, a high β p H mode plasma with H 89PL = 2.8, HH y,2 = 1.4, β p ∼ 2.0 and β N ∼ 2.5 has been sustained for 1.3 s at small values of collisionality ν e* and ion Larmor radius ρ i* without destabilizing the NTMs. Characteristics of the NTMs destabilized in the region with central safety factor above unity are investigated. The relation between the beta value at the mode onset β N on and that at the mode disappearance β N off can be described as β N off /β N on =0.05-0.4, which shows the existence of hysteresis. The value of β N /ρ i* at the onset of an m/n = 3/2 NTM has a collisionality dependence, which is empirically given by β N /ρ i* ∝ ν e* 0.36 . However, the profile effects such as the relative shapes of pressure and safety factor profiles are equally important. The onset condition seems to be affected by the strength of the pressure gradient at the mode rational surface. Stabilization of the NTM by local ECCD/ECH at the magnetic island has been attempted. A 3/2 NTM has been completely stabilized by EC wave injection of 1.6 MW. (author)

  6. Evaluation of Particle Pinch and Diffusion Coefficients in the Edge Pedestal of DIII-D H-mode Discharges

    Science.gov (United States)

    Stacey, W. M.; Groebner, R. J.

    2009-11-01

    Momentum balance requires that the radial particle flux satisfy a pinch-diffusion relationship. The pinch can be evaluated in terms of measurable quantities (rotation velocities, Er, etc.) by the use of momentum and particle balance [1,2], the radial particle flux can be determined by momentum balance, and then the diffusion coefficient can be evaluated from the pinch diffusion relation using the measured density gradient. Applications to several DIII-D H-mode plasmas are presented. 6pt [1] W.M. Stacey, Contr. Plasma Phys. 48, 94 (2008). [2] W.M. Stacey and R.J. Groebner, Phys. Plasmas 15, 012503 (2008).

  7. Kinetic transport model for the ELMO Bumpy Torus

    International Nuclear Information System (INIS)

    Jaeger, E.F.; Hedrick, C.L.; Tolliver, J.S.

    1978-05-01

    A bounce-averaged drift kinetic equation is solved for the toroidal plasma in the ELMO Bumpy Torus (EBT). The distribution function is assumed isotropic in pitch angle and calculated as a function of radius and speed using finite differences on a two-dimensional grid. A Fokker-Planck representation of the collision operator includes Coulomb, microwave, ionizing, and charge-exchange collisions. Ion and electron fluxes, computed as integrals of the distribution function, are of comparable magnitude for ambipolar potentials which are approximately self-consistent. Initial results assume an unperturbed distribution function which is Maxwellian; however, this is not a necessary assumption in the model. Careful accounting of loss regions where electric and magnetic poloidal drifts cancel (super banana particle orbits) leads to ion loss rates which are in some cases two orders of magnitude greater than electron rates. In these cases, radially inward pointing self-consistent electric fields occur with potentials on the order of a few times the ion temperature. These negative field results are in approximate agreement with experiment and appear to be stable to the electric field runaway encountered in positive field cases

  8. Neutral beam power measurements inside the ASDEX torus

    International Nuclear Information System (INIS)

    Zengliang, Y.; Staebler, A.; Vollmer, O.

    1982-11-01

    Neutral beam power measurements inside the ASDEX torus are done with a retractable calorimeter which is only radiation cooled. The calorimeter plate made from Molybdenum is subdivided into nine segments whose increase in energy content due to a shot yields the absorbed beam power. Different models for the backward extrapolation of the measured temperature curves are examined for a series of low energy shots with the result that pure radiation cooling is a valid assumption. Furthermore, a temperature correction to the measured power is derived from these experiments. The evaluation of the shots onto this calorimeter is done by a computer program. The application of this program to a few full power shots shows that a neutral power up to 3.2 MW has been injected into the ASDEX vessel by the two injectors with an overall efficiency of up to 40%. Reionization losses due to the ASDEX stray field are less than 10%; they do not show any dependence upon the pulse length for shots up to 200 ms. (orig.)

  9. Nonlinear coupling of tearing fluctuations in the Madison Symmetric Torus

    International Nuclear Information System (INIS)

    Sarff, J.S.; Almagri, A.F.; Cekic, M.; Den Hartog, D.J.; Fiksel, G.; Hokin, S.A.; Ji, H.; Prager, S.C.; Shen, W.; Stoneking, M.R.; Assadi, S.; Sidikman, K.L.

    1992-11-01

    Three-wave, nonlinear, tearing mode coupling has been measured in the Madison Symmetric Torus (MST) reversed-field pinch (RFP) [Fusion Technol. 19, 131 (1991)] using bispectral analysis of edge magnetic fluctuations resolved in ''k-space. The strength of nonlinear three-wave interactions satisfying the sum rules m 1 + m 2 = m 3 and n 1 + n 2 = n 3 is measured by the bicoherency. In the RFP, m=l, n∼2R/a (6 for MST) internally resonant modes are linearly unstable and grow to large amplitude. Large values of bicoherency occur for two m=l modes coupled to an m=2 mode and the coupling of intermediate toroidal modes, e.g., n=6 and 7 coupled to n=13. These experimental bispectral features agree with predicted bispectral features derived from MHD computation. However, in the experiment, enhanced coupling occurs in the ''crash'' phase of a sawtooth oscillation concomitant with a broadened mode spectrum suggesting the onset of a nonlinear cascade

  10. Internal transport barriers in the National Spherical Torus Experimenta)

    Science.gov (United States)

    Yuh, H. Y.; Levinton, F. M.; Bell, R. E.; Hosea, J. C.; Kaye, S. M.; LeBlanc, B. P.; Mazzucato, E.; Peterson, J. L.; Smith, D. R.; Candy, J.; Waltz, R. E.; Domier, C. W.; Luhmann, N. C.; Lee, W.; Park, H. K.

    2009-05-01

    In the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 41, 1435 (2001)], internal transport barriers (ITBs) are observed in reversed (negative) shear discharges where diffusivities for electron and ion thermal channels and momentum are reduced. While neutral beam heating can produce ITBs in both electron and ion channels, high harmonic fast wave heating can also produce electron ITBs (e-ITBs) under reversed magnetic shear conditions without momentum input. Interestingly, the location of the e-ITB does not necessarily match that of the ion ITB (i-ITB). The e-ITB location correlates best with the magnetic shear minima location determined by motional Stark effect constrained equilibria, whereas the i-ITB location better correlates with the location of maximum E ×B shearing rate. Measured electron temperature gradients in the e-ITB can exceed critical gradients for the onset of electron thermal gradient microinstabilities calculated by linear gyrokinetic codes. A high-k microwave scattering diagnostic shows locally reduced density fluctuations at wave numbers characteristic of electron turbulence for discharges with strongly negative magnetic shear versus weakly negative or positive magnetic shear. Reductions in fluctuation amplitude are found to be correlated with the local value of magnetic shear. These results are consistent with nonlinear gyrokinetic simulations predicting a reduction in electron turbulence under negative magnetic shear conditions despite exceeding critical gradients.

  11. 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.

  12. Dimensionally similar studies of confinement and H-mode transition in ASDEX Upgrade and JET

    International Nuclear Information System (INIS)

    Ryter, F.; Stober, J.; Suttrop, W.

    2001-01-01

    Joint experiments on confinement and L-H transition were performed in ASDEX Upgrade and JET. The confinement experiments suggest that the invariance principle is not always fulfilled at high density. For the L-H transition studies, the dimensionless variables taken at the plasma edge can be in general only made identical per pair, due to the condition imposed by the L-H transition. This new approach to investigate the L-H physics suggests a weak dependence of the L-H transition mechanism on collisionality. (author)

  13. Dimensionally similar studies of confinement and H-mode transition in ASDEX Upgrade and JET

    International Nuclear Information System (INIS)

    Ryter, F.; Stober, J.; Suttrop, W.

    1999-01-01

    Joint experiments on confinement and L-H transition were performed in ASDEX Upgrade and JET. The confinement experiments suggest that the invariance principle is not always fulfilled at high density. For the L-H transition studies, the dimensionless variables taken at the plasma edge can be in general only made identical per pair, due to the condition imposed by the L-H transition. This new approach to investigate the L-H physics suggests a weak dependence of the L-H transition mechanism on collisionality. (author)

  14. TFTR centralized torus interface valve control system

    International Nuclear Information System (INIS)

    Pearson, G.G.; Olsen, D.H.

    1983-01-01

    A system developed especially for the TFTR to monitor and control the interface between the vacuum vessel and associated diagnostics will be described in this paper. Diagnostics which must be connected to the machine vacuum are required to do so through a Torus Interface Valve (TIV). Two types of TIV's are used on TFTR. The first type is a non-latching valve which must be held in the opened position by a sustained OPEN command, returning automatically to the closed position when the OPEN command is removed. This type of TIV is used on all systems which never insert a probe into the vacuum vessel through the TIV. The second type of TIV is a latching valve which requires a momentary OPEN command to open and a momentary CLOSE command to close. Each TIV is linked to its own dedicated logic controller. Each logic controller is hardwired to the appropriate TIV OPEN/CLOSED limit switches, probe IN/OUT limit switches, TFTR vacuum vessel pressure setpoint switches, and diagnostic pressure setpoint switches. The logic controller can be configured for local (push-button) or remote (computer) control. Each controller has a uniquely coded keyswitch to determine the configuration. Whether under local or remote control, all OPEN and CLOSE commands must be approved by the TIV controller (TIVC). In the case of systems with probes, the controller must receive a positive indication that the probe is completely backed out before a CLOSE command will be transmitted from the TIVC to the TIV. Before a valve will be opened by a controller, the differential pressure across the valve must be within certain limits

  15. Heuristic Drift-based Model of the Power Scrape-off width in H-mode Tokamaks

    International Nuclear Information System (INIS)

    Goldston, Robert J.

    2011-01-01

    An heuristic model for the plasma scrape-off width in H-mode plasmas is introduced. Grad B and curv B drifts into the SOL are balanced against sonic parallel flows out of the SOL, to the divertor plates. The overall particle flow pattern posited is a modification for open field lines of Pfirsch-Shlueter flows to include sinks to the divertors. These assumptions result in an estimated SOL width of ∼ 2αρ p /R. They also result in a first-principles calculation of the particle confinement time of H-mode plasmas, qualitatively consistent with experimental observations. It is next assumed that anomalous perpendicular electron thermal diffusivity is the dominant source of heat flux across the separatrix, investing the SOL width, defined above, with heat from the main plasma. The separatrix temperature is calculated based on a two-point model balancing power input to the SOL with Spitzer-Haerm parallel thermal conduction losses to the divertor. This results in a heuristic closed-form prediction for the power scrape-off width that is in reasonable quantitative agreement both in absolute magnitude and in scaling with recent experimental data from deuterium plasmas. Further work should include full numerical calculations, including all magnetic and electric drifts, as well as more thorough comparison with experimental data.

  16. An Heuristic Drift-Based Model of the Power Scrape-Off Width in H-Mode Tokamaks

    International Nuclear Information System (INIS)

    Goldston, Robert J.

    2011-01-01

    An heuristic model for the plasma scrape-off width in H-mode plasmas is introduced. Grad B and curv B drifts into the SOL are balanced against sonic parallel flows out of the SOL, to the divertor plates. The overall mass flow pattern posited is a modification for open field lines of Pfirsch-Shlueter flows to include sinks to the divertors. These assumptions result in an estimated SOL width of 2αρ p /R. They also result in a first-principles calculation of the particle confinement time of H-mode plasmas, qualitatively consistent with experimental observations. It is next assumed that anomalous perpendicular electron thermal diffusivity is the dominant source of heat flux across the separatrix, investing the SOL width, defined above, with heat from the main plasma. The separatrix temperature is calculated based on a two-point model balancing power input to the SOL with Spitzer-Haerm parallel thermal conduction losses to the divertor. This results in an heuristic closed-form prediction for the power scrape-off width that is in remarkable quantitative agreement both in absolute magnitude and in scaling with recent experimental data. Further work should include full numerical calculations, including all magnetic and electric drifts, as well as more thorough comparison with experimental data.

  17. Investigation of the hydrogen fluxes in the plasma edge of W7-AS during H-mode discharges

    International Nuclear Information System (INIS)

    Langer, U.; Taglauer, E.; Fischer, R.

    2001-01-01

    In the stellarator W7-AS the H-mode is characterized by an edge transport barrier which is localized within a few centimeters inside the separatrix. The corresponding L-H transition shows well-known features such as the steepening of the temperature and density profiles in the region of the separatrix. With a so-called sniffer probe the temporal development of the hydrogen and deuterium fluxes has been studied in the plasma edge during different H-mode discharges with deuterium gas puffing. Prior to the transition a significant reduction of the deuterium and also the hydrogen fluxes can be observed. This fact confirms the assumption that the steepening of the density profiles starts at the outermost edge of the plasma. Moreover, sniffer probe measurements in the plasma edge could therefore identify a precursor for the L-H transition. The analysis of the hydrogen neutral gases shows a distinct change of the hydrogen isotope ratio during the transition. This observation is in agreement with the change in the particle fluxes onto the targets and can also be seen in the reduced H α signals from the limiters. It is further demonstrated that significant improvement in the time resolution of the measured data can be obtained by deconvolution of the data with the apparatus function using Bayesian probability theory and the Maximum Entropy method with adaptive kernels

  18. Effects of triangularity on confinement, density limit and profile stiffness of H-modes on ASDEX upgrade

    International Nuclear Information System (INIS)

    Stober, J.; Gruber, O.; Kallenbach, A.; Mertens, V.; Ryter, F.; Staebler, A.; Suttrop, W.; Treutterer, W.

    2000-01-01

    At ASDEX Upgrade the influence of triangularity on the H-mode performance has been studied intensively. It has been found that confinement increases with δ for a fixed line-averaged density. Though confinement decreases with increasing density for all analysed values of δ, H-factors (ITERH-98P) at the Greenwald density could be raised to 1 for the highest δ values achieved so far. The H-mode density limit could be increased by approx. 20%. There is a scatter of about 30% on the confinement data, which is anti-correlated to the average density in the scrape-off layer or the neutral fluxes outside the plasma. For nearly all discharges analysed so far, the temperature profiles are self-similar. This indication of profile stiffness could be verified by changing the heat-flux profile by changing the beam-voltage of the neutral-beam injection (NBI) at high density. At low density, first results indicate a deviation from this stiff behaviour. (author)

  19. Vortex flow in a torus - a method for arc stabilization

    International Nuclear Information System (INIS)

    Polman, R.W.

    1976-08-01

    Experiments on ring vortices inside a torus and experiments on semi-toroidal arcs stabilized by such vortices are described. The studies were performed in two separate devices. One of the toroidal vortex chambers - 'Cogion', with R = 0.45 m and r = 0.10 m - permits the establishment of a gas flow only. In the other device - 'Tovorex', with R = 0.19 m and r = 0.04 m - it is also possible to draw a semi-toroidal arc. The measurements surprisingly show that it is possible to describe the radial distribution of the poloidal flow in terms of a plane turbulent wall jet discharging in an external stream. The velocity profile and the growth of the width of the jet are in accordance with experimental data on this subject. A different behaviour is found for the decay of the maximum velocity. The core of the flow proves to be almost stagnant; the axis of rotation is displaced outwardly with respect to the centre of the cross-section over a distance of 6 mm. In 'Tovorex' information about the rotating flow is obtained without the presence of an arc in the vortex core. The velocity profiles prove to be independent of the pressure (50-400 Torr). For experiments with arcs nitrogen is used. It has been found that the semi-toroidal DC-arc, surrounded by a continuous metal wall can be stabilized by the toroidal vortices in the experimental range30 -1 , 0.6 -1 , depends on the current and on the pressure and is independent of Usub(j) and a. Temperatures of the discharge are estimated at approximately 6000 K. The velocity profiles in both vortex chambers are obtained with a hot-wire anemometer operated at constant resistance

  20. Partially collisional model of the Titan hydrogen torus

    International Nuclear Information System (INIS)

    Hilton, D.A.

    1987-01-01

    A numerical model was developed for atomic hydrogen densities in the Titan hydrogen torus. The effects of occasional collisions were included in order to accurately simulate physical conditions inferred from the Voyager 1 and 2 Ultraviolet Spectrometer (UVS) results of Broadfoot et al. (1981) and Sandel et al. (1982). The model employed Lagrangian perturbation of orbital elements of hydrogen atoms launched from Titan and Monte Carlo simulation of collisions and loss mechanisms. The torus is found to be azimuthally symmetric with the density sharply peaked at Titan's orbit, and decreasing rapidly in the outward and perpendicular directions and more gradually inward from 17 to 5 R/sub s/. The energetic hydrogen atoms from Saturn's upper atmosphere, first predicted by Shemansky and Smith (1982), were also investigated. Collisions of these Saturnian atoms with the torus population do not contribute to the torus density, and will lead to a net loss of torus atoms if their launch speeds from Saturn extend above 40 km/sec. The Saturnian atoms produce a corona which was modeled using the theory of Chamberlain (1963)

  1. Alpha particle studies during JET DT experiments

    International Nuclear Information System (INIS)

    1999-01-01

    The 1997 DT experiment (DTE1) at the Joint European Torus included studies of the behaviour of alpha particles in high temperature plasmas. Clear alpha particle heating was observed in a series of otherwise similar 10MW hot-ion H-modes by scanning the DT mixture from 0%T to 93%T. Maxima in central temperature and energy content were obtained which corresponded with the maximum in fusion yield. Alfven Eigenmodes (AEs) have been detected in JET, driven by NBI or ICRH fast ions. However, in agreement with theory, no AE activity was observed in DT plasmas which could be attributed to alpha particle drive, except in the afterglow of some Optimised Shear pulses. Ion Cyclotron Emission (ICE) was detected at harmonics of the alpha particle cyclotron frequency at the outer edge of the plasma. The ICE is interpreted as being close to magnetoacoustic cyclotron instability, driven by inverted alpha distributions at the plasma edge. The high-energy neutral particle spectra showed features, which are ascribed to a mixture of alphas, neutralised by helium-like impurities, and deuterons, born from elastic collisions with alpha particles and neutralised by hydrogen-like impurities. The results of all these studies are consistent with classical alpha particle trapping and slowing-down. Future DT experiments will aim to increase alpha particle pressure, so interactions with plasma instabilities can be studied. The measurement of knock-on neutral triton spectra offers a clean way to determine confined alpha densities in these future experiments. (author)

  2. Alpha particle studies during JET DT experiments

    International Nuclear Information System (INIS)

    2001-01-01

    The 1997 DT experiment (DTE1) at the Joint European Torus included studies of the behaviour of alpha particles in high temperature plasmas. Clear alpha particle heating was observed in a series of otherwise similar 10MW hot-ion H-modes by scanning the DT mixture from 0%T to 93%T. Maxima in central temperature and energy content were obtained which corresponded with the maximum in fusion yield. Alfven Eigenmodes (AEs) have been detected in JET, driven by NBI or ICRH fast ions. However, in agreement with theory, no AE activity was observed in DT plasmas which could be attributed to alpha particle drive, except in the afterglow of some Optimised Shear pulses. Ion Cyclotron Emission (ICE) was detected at harmonics of the alpha particle cyclotron frequency at the outer edge of the plasma. The ICE is interpreted as being close to magnetoacoustic cyclotron instability, driven by inverted alpha distributions at the plasma edge. The high-energy neutral particle spectra showed features, which are ascribed to a mixture of alphas, neutralised by helium-like impurities, and deuterons, born from elastic collisions with alpha particles and neutralised by hydrogen-like impurities. The results of all these studies are consistent with classical alpha particle trapping and slowing-down. Future DT experiments will aim to increase alpha particle pressure, so interactions with plasma instabilities can be studied. The measurement of knock-on neutral triton spectra offers a clean way to determine confined alpha densities in these future experiments. (author)

  3. JET experiments with tritium and deuterium–tritium mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Horton, Lorne, E-mail: Lorne.Horton@jet.uk [JET Exploitation Unit, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); European Commission, B-1049 Brussels (Belgium); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Batistoni, P. [Unità Tecnica Fusione - ENEA C. R. Frascati - via E. Fermi 45, Frascati (Roma), 00044, Frascati (Italy); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Boyer, H.; Challis, C.; Ćirić, D. [CCFE, Culham Science Centre, Abingdon OX14 3DB, Oxon (United Kingdom); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Donné, A.J.H. [EUROfusion Programme Management Unit, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); FOM Institute DIFFER, PO Box 1207, NL-3430 BE Nieuwegein (Netherlands); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Eriksson, L.-G. [European Commission, B-1049 Brussels (Belgium); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Garcia, J. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Garzotti, L.; Gee, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB, Oxon (United Kingdom); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Hobirk, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Joffrin, E. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); and others

    2016-11-01

    Highlights: • JET is preparing for a series of experiments with tritium and deuterium–tritium mixtures. • Physics objectives include integrated demonstration of ITER operating scenarios, isotope and alpha physics. • Technology objectives include neutronics code validation, material studies and safety investigations. • Strong emphasis on gaining experience in operation of a nuclear tokamak and training scientists and engineers for ITER. - Abstract: Extensive preparations are now underway for an experiment in the Joint European Torus (JET) using tritium and deuterium–tritium mixtures. The goals of this experiment are described as well as the progress that has been made in developing plasma operational scenarios and physics reference pulses for use in deuterium–tritium and full tritium plasmas. At present, the high performance plasmas to be tested with tritium are based on either a conventional ELMy H-mode at high plasma current and magnetic field (operation at up to 4 MA and 4 T is being prepared) or the so-called improved H-mode or hybrid regime of operation in which high normalised plasma pressure at somewhat reduced plasma current results in enhanced energy confinement. Both of these regimes are being re-developed in conjunction with JET's ITER-like Wall (ILW) of beryllium and tungsten. The influence of the ILW on plasma operation and performance has been substantial. Considerable progress has been made on optimising performance with the all-metal wall. Indeed, operation at the (normalised) ITER reference confinement and pressure has been re-established in JET albeit not yet at high current. In parallel with the physics development, extensive technical preparations are being made to operate JET with tritium. The state and scope of these preparations is reviewed, including the work being done on the safety case for DT operation and on upgrading machine infrastructure and diagnostics. A specific example of the latter is the planned calibration at

  4. Edge radial electric field structure in quiescent H-mode plasmas in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); West, W P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Doyle, E J [University of California, Los Angeles, CA 90095-1597 (United States); Austin, M E [University of Texas at Austin, Austin, TX 78712 (United States); DeGrassie, J S [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Gohil, P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Greenfield, C M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Jayakumar, R [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); Kaplan, D H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lao, L L [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Leonard, A W [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M A [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); McKee, G R [University of Wisconsin, Madison, WI 53706-1687 (United States); Solomon, W M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Thomas, D M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Rhodes, T L [University of California, Los Angeles, CA 90095-1597 (United States); Wade, M R [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wang, G [University of California, Los Angeles, CA 90095-1597 (United States); Watkins, J G [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Zeng, L [University of California, Los Angeles, CA 90095-1597 (United States)

    2004-05-01

    H-mode operation is the choice for next step tokamak devices based on either conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the {beta} limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D over the past four years have demonstrated a new operating regime, the quiescent H-mode (QH-mode) regime, that solves these problems. QH-mode plasmas have now been run for over 4 s (>30 energy confinement times). Utilizing the steady-state nature of the QH-mode edge allows us to obtain unprecedented spatial resolution of the edge ion profiles and the edge radial electric field, E{sub r}, by sweeping the edge plasma slowly past the view points of the charge exchange spectroscopy system. We have investigated the effects of direct edge ion orbit loss on the creation and sustainment of the QH-mode. Direct loss of ions injected into the velocity-space loss cone at the plasma edge is not necessary for creation or sustainment of the QH-mode. The direct ion orbit loss has little effect on the edge E{sub r} well. The E{sub r} at the bottom of the well in these cases is about -100 kV m{sup -1} compared with -20 to -30 kV m{sup -1} in the standard H-mode. The well is about 1 cm wide, which is close to the diameter of the deuteron gyro-orbit. We also have investigated the effect of changing edge triangularity by changing the plasma shape from upwardly biased single null to magnetically balanced double null. We have now achieved the QH-mode in these double-null plasmas. The increased triangularity allows us to increase pedestal density in QH-mode plasmas by a factor of about 2.5 and overall pedestal pressure by a factor of 2. Pedestal {beta} and {nu}{sup *} values matching the values desired for ITER have been achieved. In

  5. Smooth invariant densities for random switching on the torus

    Science.gov (United States)

    Bakhtin, Yuri; Hurth, Tobias; Lawley, Sean D.; Mattingly, Jonathan C.

    2018-04-01

    We consider a random dynamical system obtained by switching between the flows generated by two smooth vector fields on the 2d-torus, with the random switchings happening according to a Poisson process. Assuming that the driving vector fields are transversal to each other at all points of the torus and that each of them allows for a smooth invariant density and no periodic orbits, we prove that the switched system also has a smooth invariant density, for every switching rate. Our approach is based on an integration by parts formula inspired by techniques from Malliavin calculus.

  6. The CLAS12 Torus Detector Magnet at Jefferson Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Luongo, Cesar [Jefferson Lab; Ballard, Joshua [Jefferson Lab; Biallas, George [Jefferson Lab; Elouadrhiri, Latifa [Jefferson Lab; Fair, Ruben [Jefferson Lab; Ghoshal, Probir [Jefferson Lab; Kashy, Dave [Jefferson Lab; Legg, Robert [Jefferson Lab; Pastor, Orlando [Jefferson Lab; Rajput-Ghoshal, Renuka [Jefferson Lab; Rode, Claus [Jefferson Lab; Wiseman, Mark [Jefferson Lab; Young, Glenn [Jefferson Lab; Elementi, Luciano [Fermilab; Krave, Steven [Fermilab; Makarov, Alexander [Fermilab; Nobrega, Fred [Fermilab; Velev, George [Fermilab

    2015-12-17

    The CLAS12 Torus is a toroidal superconducting magnet, which is part of the detector for the 12-GeV accelerator upgrade at Jefferson Laboratory (JLab). The coils were wound/fabricated by Fermilab, with JLab responsible for all other parts of the project scope, including design, integration, cryostating the individual coils, installation, cryogenics, I&C, etc. This paper provides an overview of the CLAS12 Torus magnet features and serves as a status report of its installation in the experimental hall. Completion and commissioning of the magnet is expected in 2016.

  7. Maass Cusp Forms on Singly Punctured Two-Torus

    International Nuclear Information System (INIS)

    Siddig, Abubaker Ahmed Mohamed; Shah, Nurisya Mohd; Zainuddin, Hishamuddin

    2009-01-01

    Quantum mechanical systems on punctured surfaces modeled by hyperbolic spaces can play an interesting role in exploring quantum chaos and in studying behaviour of future quantum nano-devices. The case of singly-punctured two-torus, for example, has been well-studied in the literature particularly for its scattering states. However, the bound states on the punctured torus given by Maass cusp forms are lesser known. In this note, we report on the algorithm of numerically computing these functions and we present ten lower-lying eigenvalues for each odd and even Maass cusp forms.

  8. Transport analysis of the edge zone of H-mode plasmas by computer simulation

    International Nuclear Information System (INIS)

    Becker, G.; Murmann, H.

    1988-01-01

    Local transport and ideal ballooning stability in the L-phase and ELM-free H-phase in ASDEX are analysed by computer modelling. It is found that the diffusivities χ e and D at the edge are reduced by a factor of six a few milliseconds after the H-transition. Local transport in the inner plasma improves at an early stage by a typical factor of two. A change in the collisionality regime of electrons and ions does not take place. During the L-phase and the quiescent H-phase ideal ballooning modes are found to be stable. Computer experiments further show that a significant reduction in the particle flux at the separatrix takes place which is closely connected with the H-transition process. This explains the observed buildup of a density shoulder on a millisecond time-scale and the drop of the particle flow into the divertor. A strong decrease of the electron heat conduction flux at the separatrix is, however, ruled out in ELM-free periods. On the assumption of electrostatic turbulence induced transport, these results are consistent with measured density fluctuation levels near the separatrix. (author). 20 refs, 9 figs

  9. Global Particle Balance Measurements in DIII-D H-mode Discharges

    International Nuclear Information System (INIS)

    Unterberg, Ezekial A.; Allen, S.L.; Brooks, N.; Evans, T.E.; Leonard, A.W.; McLean, A.; Watkins, J.G.; Whyte, D.G.

    2011-01-01

    Experiments are performed on the DIII-D tokamak to determine the retention rate in an all graphite first-wall tokamak. A time-dependent particle balance analysis shows a majority of the fuel retention occurs during the initial Ohmic and L-mode phase of discharges, with peak fuel retention rates typically similar to 2 x 10(21) D/s. The retention rate can be zero within the experimental uncertainties (<3 x 10(20) D/s) during the later stationary phase of the discharge. In general, the retention inventory can decrease in the stationary phase by similar to 20-30% from the initial start-up phase of the discharge. Particle inventories determined as a function of time in the discharge, using a 'dynamic' particle balance analysis, agree with more accurate particle inventories directly measured after the discharge, termed 'static' particle balance. Similarly, low stationary retention rates are found in discharges with heating from neutral-beams, which injects particles, and from electron cyclotron waves, which does not inject particles. Detailed analysis of the static and dynamic balance methods provide an estimate of the DIII-D global co-deposition rate of <= 0.6-1.2 x 10(20) D/s. Dynamic particle balance is also performed on discharges with resonant magnetic perturbation ELM suppression and shows no additional retention during the ELM-suppressed phase of the discharge.

  10. H-mode access during plasma current ramp-up in TCV

    International Nuclear Information System (INIS)

    Martin, Y.; Behn, R.; Furno, I.; Labit, B.; Reimerdes, H.

    2014-01-01

    A recent TCV experiment has investigated the dependence of the L–H transition threshold power on the plasma current ramp-rate and the X-point height above the divertor target, which both have previously been seen to affect the transition behaviour. Systematic scans in ohmically heated plasmas do not show any dependence on the plasma current ramp-up rate. In contrast, the threshold power is found to increase by a factor of two while the X-point is moved from about 10 cm up to 35 cm above the vessel floor. However, further increase, up to 60 cm, does not lead to any further increase of the required power. The Fundamenski et al model is tested against the measurements. Estimates of the Wagner number (Wa) at L–H transitions are generally close to unity, in accordance with the model. In contrast, estimates of Wa before the L–H transition, i.e. in L-mode, do not show the expected evolution towards unity. (paper)

  11. ITER operational space for full plasma current H-mode operation

    Energy Technology Data Exchange (ETDEWEB)

    Mattei, M. [Assoc. Euratom-ENEA-CREATE, Seconda University di Napoli, Aversa (Italy)], E-mail: massimiliano.mattei@unirc.it; Cavinato, M.; Saibene, G.; Portone, A. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain); Albanese, R.; Ambrosino, G. [Assoc. Euratom-ENEA-CREATE, University Napoli Federico II, Napoli (Italy); Horton, L.D. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Kessel, C. [Princeton Plasma Physics Laboratory, Princeton University (United States); Koechl, F. [Assoc. EURATOM-OAW/ATI, Vienna (Austria); Lomas, P.J. [Euratom/UKAEA Fusion Assoc., Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Nunes, I. [Assoc. EURATOM/IST, Centro de Fusao Nuclear, Lisbon (Portugal); Parail, V. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Sartori, R. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain); Sips, A.C.C. [Max Planck-Institut fur Plasmaphysik, EURATOM-Association, Garching (Germany); Thomas, P.R. [Fusion for Energy Joint Undertaking, 08019 Barcelona (Spain)

    2009-06-15

    Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I{sub p}{sup 4}. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current I{sub p} of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7-0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.

  12. Initial Results from the Lost Alpha Diagnostics on Joint European Torus

    Energy Technology Data Exchange (ETDEWEB)

    Darrow, Doug; Cecil, Ed; Ellis, Bob; Fullard, Keith; Hill, Ken; Horton, Alan; Kiptily, Vasily; Pedrick, Les; Reich, Matthias

    2007-07-25

    Two devices have been installed in the Joint European Torus (JET) vacuum vessel near the plasma boundary to investigate the loss of energetic ions and fusion products in general and alpha particles in particular during the upcoming JET experiments. These devices are (i) a set of multichannel thin foil Faraday collectors, and (ii) a well collimated scintillator which is optically connected to a charge-coupled device. Initial results, including the radial energy and poloidal dependence of lost ions from hydrogen and deuterium plasmas during the 2005–06 JET restart campaign, will be presented.

  13. Initial Results from the Lost Alpha Diagnostics on Joint European Torus

    International Nuclear Information System (INIS)

    Darrow, Doug; Baeumel, Stefan; Cecil, Ed; Ellis, Bob; Fullard, Keith; Hill, Ken; Horton, Alan; Kiptily, Vasily; Pedrick, Les; Reich, Matthias; Werner, Andreas

    2007-01-01

    Two devices have been installed in the Joint European Torus (JET) vacuum vessel near the plasma boundary to investigate the loss of energetic ions and fusion products in general and alpha particles in particular during the upcoming JET experiments. These devices are (i) a set of multichannel thin foil Faraday collectors, and (ii) a well collimated scintillator which is optically connected to a charge-coupled device. Initial results, including the radial energy and poloidal dependence of lost ions from hydrogen and deuterium plasmas during the 2005-06 JET restart campaign, will be presented.

  14. Kinetic equilibrium reconstruction for the NBI- and ICRH-heated H-mode plasma on EAST tokamak

    Science.gov (United States)

    Zhen, ZHENG; Nong, XIANG; Jiale, CHEN; Siye, DING; Hongfei, DU; Guoqiang, LI; Yifeng, WANG; Haiqing, LIU; Yingying, LI; Bo, LYU; Qing, ZANG

    2018-04-01

    The equilibrium reconstruction is important to study the tokamak plasma physical processes. To analyze the contribution of fast ions to the equilibrium, the kinetic equilibria at two time-slices in a typical H-mode discharge with different auxiliary heatings are reconstructed by using magnetic diagnostics, kinetic diagnostics and TRANSP code. It is found that the fast-ion pressure might be up to one-third of the plasma pressure and the contribution is mainly in the core plasma due to the neutral beam injection power is primarily deposited in the core region. The fast-ion current contributes mainly in the core region while contributes little to the pedestal current. A steep pressure gradient in the pedestal is observed which gives rise to a strong edge current. It is proved that the fast ion effects cannot be ignored and should be considered in the future study of EAST.

  15. Characteristics of heat flux and particle flux to the divertor in H-mode of JT-60U

    International Nuclear Information System (INIS)

    Itami, K.; Hosogane, N.; Asakura, N.; Kubo, H.; Tsuji, S.; Shimada, M.

    1995-01-01

    Heat flux and particle flux behavior in H-mode is studied in a comparative manner. It was confirmed that the multiple peak structure of heat flux during ELM activity has a role in reducing the average value of a peak heat flux at the divertor. In order to characterize heat and particle flux during ELM activity, the ELM part and the steady state part of heat flux and particle flux were determined and statistically analyzed. A large in-out asymmetry of peak ELM heat flux density was found. The asymmetry is almost unaffected by the ion grad-B drift direction. In-out asymmetry of both ELM and steady-state parts of the particle flux were found to be similar. ((orig.))

  16. High resolution main-ion charge exchange spectroscopy in the DIII-D H-mode pedestal.

    Science.gov (United States)

    Grierson, B A; Burrell, K H; Chrystal, C; Groebner, R J; Haskey, S R; Kaplan, D H

    2016-11-01

    A new high spatial resolution main-ion (deuterium) charge-exchange spectroscopy system covering the tokamak boundary region has been installed on the DIII-D tokamak. Sixteen new edge main-ion charge-exchange recombination sightlines have been combined with nineteen impurity sightlines in a tangentially viewing geometry on the DIII-D midplane with an interleaving design that achieves 8 mm inter-channel radial resolution for detailed profiles of main-ion temperature, velocity, charge-exchange emission, and neutral beam emission. At the plasma boundary, we find a strong enhancement of the main-ion toroidal velocity that exceeds the impurity velocity by a factor of two. The unique combination of experimentally measured main-ion and impurity profiles provides a powerful quasi-neutrality constraint for reconstruction of tokamak H-mode pedestals.

  17. High-frequency coherent edge fluctuations in a high-pedestal-pressure quiescent H-mode plasma.

    Science.gov (United States)

    Yan, Z; McKee, G R; Groebner, R J; Snyder, P B; Osborne, T H; Burrell, K H

    2011-07-29

    A set of high frequency coherent (HFC) modes (f=80-250 kHz) is observed with beam emission spectroscopy measurements of density fluctuations in the pedestal of a strongly shaped quiescent H-mode plasma on DIII-D, with characteristics predicted for kinetic ballooning modes (KBM): propagation in the ion-diamagnetic drift direction; a frequency near 0.2-0.3 times the ion-diamagnetic frequency; inferred toroidal mode numbers of n∼10-25; poloidal wave numbers of k(θ)∼0.17-0.4 cm(-1); and high measured decorrelation rates (τ(c)(-1)∼ω(s)∼0.5×10(6) s(-1)). Their appearance correlates with saturation of the pedestal pressure. © 2011 American Physical Society

  18. H-mode WEST tungsten divertor operation: deuterium and nitrogen seeded simulations with SOLEDGE2D-EIRENE

    Directory of Open Access Journals (Sweden)

    G. Ciraolo

    2017-08-01

    Full Text Available Simulations of WEST H-mode divertor scenarios have been performed with SOLEDGE2D-EIRENE edge plasma transport code, both for pure deuterium and nitrogen seeded discharge. In the pure deuterium case, a target heat flux of 8 MW/m2 is reached, but misalignment between heat and the particle outflux yields 50 eV plasma temperature at the target plates. With nitrogen seeding, the heat and particle outflux are observed to be aligned so that lower plasma temperatures at the target plates are achieved together with the required high heat fluxes. This change in heat and particle outflux alignment is analysed with respect to the role of divertor geometry and the impact of vertical vs horizontal target plates on neutrals spreading.

  19. Quasiperiodicity and Torus Breakdown in a Power Electronic DC/DC Converter

    DEFF Research Database (Denmark)

    Zhusubaliyev, Zhanybai; Soukhoterin, Evgeniy; Mosekilde, Erik

    2007-01-01

    This paper discusses the mechanisms of torus formation and torus destruction in a dc/dc converter with relay control and hysteresis. We establish a chart of the dynamical modes in the input voltage versus load resistance parameter plane. This chart displays several different torus bifurcations...

  20. Theory of Rapid Formation of Pedestal and Pedestal width due to Anomalous Particle Pinch in the Edge of H-mode Discharges

    Energy Technology Data Exchange (ETDEWEB)

    Kaw, P.K., E-mail: kaw@ipr.res.in [Institute for Plasma Research, Bhat (India); Singh, R. [Institute for Plasma Research, Bhat (India); ITER Organization, Saint Paul-lez-Durance [France; Nordman, H. [Chamlers Institute of Technology, Goteborg (Sweden); Garbet, X.; Bourdelle, C. [CEA, Saint Paul-lez-Durance (France); Campbell, D.; Loarte, A.; Bora, D. [ITER Organization, Saint Paul-lez-Durance (France)

    2012-09-15

    Full text: A theory based on a turbulent particle pinch is proposed to explain the rapid formation of sharp density gradients in tokamak edge plasmas, in particular the pedestal region. The inward radial particle flux in the pedestal results from the interaction between small scale electron temperature gradient driven (ETG) turbulence and self-consistently formed 'electron geodesic acoustic modes' (el-GAMs). To address this phenomenon, the el-GAM modulational instability driven by the ETG turbulence background is studied. The ETG level of fluctuations and particle pinch are estimated through the back reaction of eGAMs on ETG turbulence. It is found that the particle pinch is quite sensitive to magnetic shear, safety factor, ratio of electron to ion temperatures and atomic mass number. In the absence of particle source in the pedestal, the density gradient length scale, of the order of the pedestal width, is estimated. It is shown that it is proportional to the major radius, up to some dependence on the poloidal beta. Moreover it does not depend on the normalized gyro-radius. This scaling agrees with DIII-D and JET similarity experiments. This dependence is favorable when extrapolated to the pedestal width in ITER in spite of its low normalized gyro radius. It is also shown that the density scale length becomes sharper by increasing the magnetic shear. A new H-mode pedestal pressure scaling is derived assuming that the pressure gradient is limited by the ballooning instability. (author)

  1. A statistical methodology to derive the scaling law for the H-mode power threshold using a large multi-machine database

    International Nuclear Information System (INIS)

    Murari, A.; Lupelli, I.; Gaudio, P.; Gelfusa, M.; Vega, J.

    2012-01-01

    In this paper, a refined set of statistical techniques is developed and then applied to the problem of deriving the scaling law for the threshold power to access the H-mode of confinement in tokamaks. This statistical methodology is applied to the 2010 version of the ITPA International Global Threshold Data Base v6b(IGDBTHv6b). To increase the engineering and operative relevance of the results, only macroscopic physical quantities, measured in the vast majority of experiments, have been considered as candidate variables in the models. Different principled methods, such as agglomerative hierarchical variables clustering, without assumption about the functional form of the scaling, and nonlinear regression, are implemented to select the best subset of candidate independent variables and to improve the regression model accuracy. Two independent model selection criteria, based on the classical (Akaike information criterion) and Bayesian formalism (Bayesian information criterion), are then used to identify the most efficient scaling law from candidate models. The results derived from the full multi-machine database confirm the results of previous analysis but emphasize the importance of shaping quantities, elongation and triangularity. On the other hand, the scaling laws for the different machines and at different currents are different from each other at the level of confidence well above 95%, suggesting caution in the use of the global scaling laws for both interpretation and extrapolation purposes. (paper)

  2. Legendrian and transverse cables of positive torus knots

    DEFF Research Database (Denmark)

    B. Etnyre, John; la Fountain, Douglas James; Tosun, Bulent

    In this paper we classify Legendrian and transverse knots in the knot types obtained from positive torus knots by cabling. This classification allows us to demonstrate several new phenomena. Specifically, we show there are knot types that have non-destabilizable Legendrian representatives whose T...

  3. Recursive representation of the torus 1-point conformal block

    Science.gov (United States)

    Hadasz, Leszek; Jaskólski, Zbigniew; Suchanek, Paulina

    2010-01-01

    The recursive relation for the 1-point conformal block on a torus is derived and used to prove the identities between conformal blocks recently conjectured by Poghossian in [1]. As an illustration of the efficiency of the recurrence method the modular invariance of the 1-point Liouville correlation function is numerically analyzed.

  4. Global solvability for involutive systems on the torus

    Directory of Open Access Journals (Sweden)

    Cleber de Medeira

    2013-11-01

    Full Text Available In this article, we consider a class of involutive systems of n smooth vector fields on the torus of dimension n+1. We prove that the global solvability of this class is related to an algebraic condition involving Liouville forms and the connectedness of all sublevel and superlevel sets of the primitive of a certain 1-form associated with the system.

  5. Atomic force microscopy of torus-bearing pit membranes

    Science.gov (United States)

    Roland R. Dute; Thomas Elder

    2011-01-01

    Atomic force microscopy was used to compare the structures of dried, torus-bearing pit membranes from four woody species, three angiosperms and one gymnosperm. Tori of Osmanthus armatus are bipartite consisting of a pustular zone overlying parallel sets of microfibrils that form a peripheral corona. Microfibrils of the corona form radial spokes as they traverse the...

  6. Linear pinch driven by a moving compact torus

    International Nuclear Information System (INIS)

    Hartman, C.W.; Hammer, J.H.; Eddleman, J.L.

    1984-01-01

    In principle, a Z-pinch of sufficiently large aspect ratio can provide arbitrarily high magnetic field intensity for the confinement of plasma. In practice, however, achievable field intensities and timescales are limited by parasitic inductances, pulse driver power, current, voltage, and voltage standoff of nearby insulating surfaces or surrounding gas. Further, instabilities may dominate to prevent high fields (kink mode) or enhance them (sausage mode) but in a nonuniform and uncontrollable way. In this paper we discuss an approach to producing a high-field-intensity pinch using a moving compact torus. The moving torus can serve as a very high power driver and may be used to compress a pre-established pinch field, switch on an accelerating pinch field, or may itself be reconfigured to form an intense pinch. In any case, the high energy, high energy density, and high velocity possible with an accelerated compact torus can provide extremely high power to overcome, by a number of orders of magnitude, the limitations to pinch formation described earlier. In this paper we will consider in detail pinches formed by reconfiguration of the compact torus

  7. Modular differential equations for torus one-point functions

    International Nuclear Information System (INIS)

    Gaberdiel, Matthias R; Lang, Samuel

    2009-01-01

    It is shown that in a rational conformal field theory every torus one-point function of a given highest weight state satisfies a modular differential equation. We derive and solve these differential equations explicitly for some Virasoro minimal models. In general, however, the resulting amplitudes do not seem to be expressible in terms of standard transcendental functions

  8. Spherical torus (ST) concept and its reactor implications

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Lazarus, E.A.; Miller, R.L.; Carreras, B.A.; Hogan, J.T.; Krakowski, R.A.; Seed, T.J.; Zubrin, R.M.; Schnurr, N.M.

    1986-01-01

    A brief description of the spherical torus design is given. The design concept includes resistive demountable toroidal field coils, poloidal divertor for impurity control, oscillating-field current maintenance, RF initiation and ramp-up of the plasma current, and flowing liquid-metal breeding blanket. 4 refs., 6 figs

  9. Five-dimensional gauge theory and compactification on a torus

    NARCIS (Netherlands)

    Haghighat, B.; Vandoren, S.J.G.

    2011-01-01

    We study five-dimensional minimally supersymmetric gauge theory compactified on a torus down to three dimensions, and its embedding into string/M-theory using geometric engineering. The moduli space on the Coulomb branch is hyperkaehler equipped with a metric with modular transformation properties.

  10. ELMO Bumpy Torus fusion-reactor design study

    International Nuclear Information System (INIS)

    Bathke, C.G.; Krakowski, R.A.

    1981-01-01

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is described that emphasizes those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are generic to magnetic fusion being adopted from past, more extensive tokamak reactor designs

  11. EDITORIAL: Special issue containing papers presented at the 12th International Workshop on H-mode Physics and Transport Barriers Special issue containing papers presented at the 12th International Workshop on H-mode Physics and Transport Barriers

    Science.gov (United States)

    Hahm, T. S.

    2010-06-01

    The 12th International Workshop on H-mode Physics and Transport Barriers was held at the Princeton Plasma Physics Laboratory, Princeton, New Jersey, USA between September 30 and October 2, 2009. This meeting was the continuation of a series of previous meetings which was initiated in 1987 and has been held bi-annually since then. Following the recent tradition at the last few meetings, the program was sub- divided into six sessions. At each session, an overview talk was presented, followed by two or three shorter oral presentations which supplemented the coverage of important issues. These talks were followed by discussion periods and poster sessions of contributed papers. The sessions were: Physics of Transition to/from Enhanced Confinement Regimes, Pedestal and Edge Localized Mode Dynamics, Plasma Rotation and Momentum Transport, Role of 3D Physics in Transport Barriers, Transport Barriers: Theory and Simulations and High Priority ITER Issues on Transport Barriers. The diversity of the 90 registered participants was remarkable, with 22 different nationalities. US participants were in the majority (36), followed by Japan (14), South Korea (7), and China (6). This special issue of Nuclear Fusion consists of a cluster of 18 accepted papers from submitted manuscripts based on overview talks and poster presentations. The paper selection procedure followed the guidelines of Nuclear Fusion which are essentially the same as for regular articles with an additional requirement on timeliness of submission, review and revision. One overview paper and five contributed papers report on the H-mode pedestal related results which reflect the importance of this issue concerning the successful operation of ITER. Four papers address the rotation and momentum transport which play a crucial role in transport barrier physics. The transport barrier transition condition is the main focus of other four papers. Finally, four additional papers are devoted to the behaviour and control of

  12. Nonlinear evolution of magnetic islands in a two fluid torus

    International Nuclear Information System (INIS)

    Sugiyama, L.E.; Park, W.

    1996-01-01

    A numerical model MH3D-T for the two fluid description of macroscopic evolution in a full three dimensional torus has been developed. Based on the perturbative drift ordering, generalized to arbitrary perturbation size, the model follows the full temperature evolution, including the thermal equilibration along the magnetic field. It contains the diamagnetic drifts, ion gyroviscous stress tensor, and the Hall term in Ohm's law. Electron inertia is neglected. The numerical model solves the same equations in a torus and in several simplified configurations. It has been benchmarked against the diamagnetic ω* i stabilization of the resistive m = 1, n = 1 reconnecting mode in a cylinder. The nonlinear evolution of resistive magnetic islands with m,n ≠ 1,1 in a cylinder is found to agree with previous analytic and reduced-torus results, which show that the diamagnetic rotation vanishes early in the island evolution and the saturated island size is determined by the same external driving factor Δ' as in MHD. The two fluid evolution in a full torus, however, differs from that in a cylinder and from the resistive MHD evolution. The poloidal rotation velocity undergoes a degree of poloidal momentum damping in the torus, even without neoclassical effects. The two fluid magnetic island grows faster, nonlinearly, than the resistive MHD island, and also couples different toroidal harmonics more effectively. Plasma compressibility and processes operating along the magnetic field play a much more important role than in MHD or in simple geometry. The two fluid model contains all the important neoclassical fluid effects except for the b circ ∇ circ Π parallelj viscous force terms. The addition of these terms is in progress

  13. Midplane neutral density profiles in the National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Stotler, D. P., E-mail: dstotler@pppl.gov; Bell, R. E.; Diallo, A.; LeBlanc, B. P.; Podestà, M.; Roquemore, A. L.; Ross, P. W. [Princeton Plasma Physics Laboratory, Princeton University, P. O. Box 451, Princeton, New Jersey 08543-0451 (United States); Scotti, F. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2015-08-15

    Atomic and molecular density data in the outer midplane of NSTX [Ono et al., Nucl. Fusion 40, 557 (2000)] are inferred from tangential camera data via a forward modeling procedure using the DEGAS 2 Monte Carlo neutral transport code. The observed Balmer-β light emission data from 17 shots during the 2010 NSTX campaign display no obvious trends with discharge parameters such as the divertor Balmer-α emission level or edge deuterium ion density. Simulations of 12 time slices in 7 of these discharges produce molecular densities near the vacuum vessel wall of 2–8 × 10{sup 17 }m{sup −3} and atomic densities ranging from 1 to 7 × 10{sup 16 }m{sup −3}; neither has a clear correlation with other parameters. Validation of the technique, begun in an earlier publication, is continued with an assessment of the sensitivity of the simulated camera image and neutral densities to uncertainties in the data input to the model. The simulated camera image is sensitive to the plasma profiles and virtually nothing else. The neutral densities at the vessel wall depend most strongly on the spatial distribution of the source; simulations with a localized neutral source yield densities within a factor of two of the baseline, uniform source, case. The uncertainties in the neutral densities associated with other model inputs and assumptions are ≤50%.

  14. 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

  15. 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.

  16. Integrated simulations of H-mode operation in ITER including core fuelling, divertor detachment and ELM control

    Science.gov (United States)

    Polevoi, A. R.; Loarte, A.; Dux, R.; Eich, T.; Fable, E.; Coster, D.; Maruyama, S.; Medvedev, S. Yu.; Köchl, F.; Zhogolev, V. E.

    2018-05-01

    ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1  +  SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65 T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5–7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W re-deposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on

  17. Measurement of The Magnetic Field in a Spherical Torus Plasma via Electron Bernstein Wave Emission Harmonic Overlap

    International Nuclear Information System (INIS)

    Jones, B.; Taylor, G.; Efthimion, P.C.; Munsat, T.

    2004-01-01

    Measurement of the magnetic field in a spherical torus by observation of harmonic overlap frequencies in the electron Bernstein wave (EBW) spectrum has been previously suggested [V.F. Shevchenko, Plasma Phys. Reports 26 (2000) 1000]. EBW mode conversion to X-mode radiation has been studied in the Current Drive Experiment-Upgrade spherical torus, [T. Jones, Ph.D. thesis, Princeton University, 1995] with emission measured at blackbody levels [B. Jones et al., Phys. Rev. Lett. 90 (2003) article no. 165001]. Sharp transitions in the thermally emitted EBW spectrum have been observed for the first two harmonic overlaps. These transition frequencies are determined by the magnetic field and electron density at the mode conversion layer in accordance with hot-plasma wave theory. Prospects of extending this measurement to higher harmonics, necessary in order to determine the magnetic field profile, and high beta equilibria are discussed for this proposed magnetic field diagnostic

  18. Impurity transport model for the normal confinement and high density H-mode discharges in Wendelstein 7-AS

    International Nuclear Information System (INIS)

    Ida, K; Burhenn, R; McCormick, K; Pasch, E; Yamada, H; Yoshinuma, M; Inagaki, S; Murakami, S; Osakabe, M; Liang, Y; Brakel, R; Ehmler, H; Giannone, L; Grigull, P; Knauer, J P; Maassberg, H; Weller, A

    2003-01-01

    An impurity transport model based on diffusivity and the radial convective velocity is proposed as a first approach to explain the differences in the time evolution of Al XII (0.776 nm), Al XI (55 nm) and Al X (33.3 nm) lines following Al-injection by laser blow-off between normal confinement discharges and high density H-mode (HDH) discharges. Both discharge types are in the collisional regime for impurities (central electron temperature is 0.4 keV and central density exceeds 10 20 m -3 ). In this model, the radial convective velocity is assumed to be determined by the radial electric field, as derived from the pressure gradient. The diffusivity coefficient is chosen to be constant in the plasma core but is significantly larger in the edge region, where it counteracts the high local values of the inward convective velocity. Under these conditions, the faster decay of aluminium in HDH discharges can be explained by the smaller negative electric field in the bulk plasma, and correspondingly smaller inward convective velocity, due to flattening of the density profiles

  19. Experimental evidence for the suitability of ELMing H-mode operation in ITER with regard to core transport of helium

    International Nuclear Information System (INIS)

    Wade, M.R.; Hillis, D.L.; Burrell, K.H.

    1996-09-01

    Studies have been conducted in DIII-D to assess the viability of the ITER design with regard to helium ash removal, including both global helium exhaust studies and detailed helium transport studies. With respect to helium ash accumulation, the results are encouraging for successful operation of ITER in ELMing H-mode plasmas with conventional high-recycling divertor operation. Helium can be removed from the plasma core with a characteristic time constant of ∼ 8 energy confinement times, even with a central source of helium. Furthermore, the exhaust rate is limited by the pumping efficiency of the system and not by transport of helium within the plasma core. Helium transport studies have shown that D He /X eff ∼ 1 in all confinement regimes studied to date and there is little dependence of D He /X eff on normalized gyroradius in dimensionless scaling studies, suggesting that D He /X eff will be ∼ 1 in ITER. These observations suggest that helium transport within the plasma core should be sufficient to prevent unacceptable fuel dilution in ITER. However, helium exhaust is also strongly dependent on many factors (e.g., divertor plasma conditions, plasma and baffling geometry, flux amplification, pumping speed, etc.) that are difficult to extrapolate. Studies have revealed the helium diffusivity decreases as the plasma density increases, which is unfavorable to ITER's extremely high density operation

  20. Initial results of H-mode edge pedestal turbulence evolution with quadrature reflectometer measurements on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Wang, G. [University of California, Los Angeles, CA 90095 (United States)]. E-mail: wangg@fusion.gat.com; Peebles, W.A. [University of California, Los Angeles, CA 90095 (United States); Doyle, E.J. [University of California, Los Angeles, CA 90095 (United States); Rhodes, T.L. [University of California, Los Angeles, CA 90095 (United States); Zeng, L. [University of California, Los Angeles, CA 90095 (United States); Nguyen, X. [University of California, Los Angeles, CA 90095 (United States); Osborne, T.H. [General Atomics, San Diego, CA 92186-5608 (United States); Snyder, P.B. [General Atomics, San Diego, CA 92186-5608 (United States); Kramer, G.J. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nazikian, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Groebner, R.J. [General Atomics, San Diego, CA 92186-5608 (United States); Burrell, K.H. [General Atomics, San Diego, CA 92186-5608 (United States); Leonard, A.W. [General Atomics, San Diego, CA 92186-5608 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Strait, E.J. [General Atomics, San Diego, CA 92186-5608 (United States)

    2007-06-15

    High-resolution quadrature reflectometer measurements of density fluctuation levels have been obtained on DIII-D for H-mode edge pedestal studies. Initial results are presented from the L-H transition to the first ELM for two cases: (i) a low pedestal beta discharge, in which density turbulence in the pedestal has little change during the ELM-free phase, and (ii) a high pedestal beta discharge in which both density and magnetic turbulence are observed to increase before the first ELM. These high beta data are consistent with the existence of electromagnetic turbulence suggested by some transport models. During Type-I ELM cycles, when little magnetic turbulence can be observed, pedestal turbulence increases just after an ELM crash and then decreases before next ELM strikes, in contrast to a drop after ELM crash and then it re-grows when strong magnetic turbulence shows similar behavior. Clear ELM precursors are observed on {<=}20% of Type-I ELMs observed to date.