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Sample records for alcator c-mod tokamak

  1. Alcator C-Mod Tokamak

    Data.gov (United States)

    Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...

  2. Twenty Years of Research on the Alcator C-Mod Tokamak

    Science.gov (United States)

    Greenwald, Martin

    2013-10-01

    Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

  3. Stationary density profiles in the Alcator C-mod tokamak

    International Nuclear Information System (INIS)

    Kesner, J.; Ernst, D.; Hughes, J.; Mumgaard, R.; Shiraiwa, S.; Whyte, D.; Scott, S.

    2012-01-01

    In the absence of an internal particle source, plasma turbulence will impose an intrinsic relationship between an inwards pinch and an outwards diffusion resulting in a stationary density profile. The Alcator C-mod tokamak utilizes RF heating and current drive so that fueling only occurs in the vicinity of the separatrix. Discharges that transition from L-mode to I-mode are seen to maintain a self-similar stationary density profile as measured by Thomson scattering. For discharges with negative magnetic shear, an observed rise of the safety factor in the vicinity of the magnetic axis appears to be accompanied by a decrease of electron density, qualitatively consistent with the theoretical expectations.

  4. 20 years of research on the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Greenwald, M.; Baek, S.; Barnard, H.; Beck, W.; Bonoli, P.; Brunner, D.; Burke, W.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Gao, C.; Golfinopoulos, T.; Granetz, R.; Hartwig, Z.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.

    2014-01-01

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

  5. Neutral particle dynamics in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Niemczewski, Artur P. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1995-08-01

    This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism.

  6. Neutral particle dynamics in the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Niemczewski, A.P.

    1995-08-01

    This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism. 146 refs., 82 figs., 14 tabs

  7. Numerical modelling of ICRF physics experiments in the Alcator C-mod tokamak

    International Nuclear Information System (INIS)

    Bonoli, P.T.; Boivin, R.L.; Brambilla, M.

    2001-01-01

    A full-wave spectral code (TORIC) has been used to simulate mode converted ion Bernstein wave (IBW) propagation and absorption for the first time at high poloidal mode number (-80< m<+80). Converged wave solutions for the mode converted wave are obtained in this limit and the predicted electron damping of the IBW is found to be consistent with experimental measurements from the Alcator C-Mod tokamak. The TORIC code has also been coupled to a bounce-averaged Fokker Planck module FPPRF and the combined codes are now run within the transport analysis tool TRANSP. This model was used to analyze off-axis hydrogen minority heating experiments in C-Mod where an internal transport barrier was obtained. (author)

  8. Modelling of advanced tokamak physics scenarios in ALCATOR C-Mod

    International Nuclear Information System (INIS)

    Bonoli, P.T.; Porkolab, M.; Ramos, J.

    2001-01-01

    Advanced tokamak modes of operation in Alcator C-Mod have been investigated using a simulation model which combines an MHD equilibrium and current profile control calculation with an ideal MHD stability analysis. Stable access to high β t operating modes with reversed shear current density profiles has been demonstrated using 2.4-3.0 MW of off-axis lower hybrid current drive (LHCD). Here β t =2μ 0 (p)/B 2 0 is the volume averaged toroidal plasma beta. Current profile control at the β-limit and beyond has also been demonstrated. The effects of LH power level as well as changes in the profiles of density and temperature on shear reversal radius have been quantified and are discussed. (author)

  9. Energetic ion loss detector on the Alcator C-Mod tokamak.

    Science.gov (United States)

    Pace, D C; Granetz, R S; Vieira, R; Bader, A; Bosco, J; Darrow, D S; Fiore, C; Irby, J; Parker, R R; Parkin, W; Reinke, M L; Terry, J L; Wolfe, S M; Wukitch, S J; Zweben, S J

    2012-07-01

    A scintillator-based energetic ion loss detector has been successfully commissioned on the Alcator C-Mod tokamak. This probe is located just below the outer midplane, where it captures ions of energies up to 2 MeV resulting from ion cyclotron resonance heating. After passing through a collimating aperture, ions impact different regions of the scintillator according to their gyroradius (energy) and pitch angle. The probe geometry and installation location are determined based on modeling of expected lost ions. The resulting probe is compact and resembles a standard plasma facing tile. Four separate fiber optic cables view different regions of the scintillator to provide phase space resolution. Evolving loss levels are measured during ion cyclotron resonance heating, including variation dependent upon individual antennae.

  10. The measurement of the intrinsic impurities of molybdenum and carbon in the Alcator C-Mod tokamak plasma using low resolution spectroscopy

    Science.gov (United States)

    May, M. J.; Finkenthal, M.; Regan, S. P.; Moos, H. W.; Terry, J. L.; Goetz, J. A.; Graf, M. A.; Rice, J. E.; Marmar, E. S.; Fournier, K. B.; Goldstein, W. H.

    1997-06-01

    The intrinsic impurity content of molybdenum and carbon was measured in the Alcator C-Mod tokamak using low resolution, multilayer mirror (MLM) spectroscopy ( Delta lambda ~1-10 AA). Molybdenum was the dominant high-Z impurity and originated from the molybdenum armour tiles covering all of the plasma facing surfaces (including the inner column, the poloidal divertor plates and the ion cyclotron resonant frequency (ICRF) limiter) at Alcator C-Mod. Despite the all metal first wall, a carbon concentration of 1 to 2% existed in the plasma and was the major low-Z impurity in Alcator C-Mod. Thus, the behaviour of intrinsic molybdenum and carbon penetrating into the main plasma and the effect on the plasma must be measured and characterized during various modes of Alcator C-Mod operation. To this end, soft X-ray extreme ultraviolet (XUV) emission lines of charge states, ranging from hydrogen-like to helium-like lines of carbon (radius/minor radius, r/a~1) at the plasma edge to potassium to chlorine-like (0.4Data Nucl. Data Tables 33 (1985) 149), which were incorporated into the collisional radiative model. The intrinsic i

  11. Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

    2015-09-15

    The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

  12. Validation of neutral point on JT-60U, Alcator C-Mod and ASDEX-Upgrade tokamaks

    International Nuclear Information System (INIS)

    Nakamura, Yukiharu; Yoshino, Ryuji; Pautasso, Gabriella; Gruber, Otto; Jardin, Stephen

    2002-01-01

    Validation studies of a neutrally balanced vertical plasma position, so-called ''neutral point'', have been carried out by computational simulations and experiments under trilateral Japan-US-EU collaborations. It was clarified that the neutral point, where VDEs (Vertical Displacement Events) are hardly occurred, does exit in the Alcator C-Mod and ASDEX-Upgrade tokamaks as well as the JT-60U, consistent with the simulations. Meanwhile, precise details of the VDE behavior exhibit their own characters according to the individual of the tokamaks such as an up-down asymmetry of plasma shape. Sensitivity of the neutral point to the plasma shape and current profile was also addressed in detail. (author)

  13. Measurement of impurity ion densities and energies in the divertor and edge regions of Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Griem, H.R.; Moreno, J.; Welch, B.L.

    1992-01-01

    A study to investigate impurity production and transport in the divertor and edge regions of the Alcator C-Mod tokamak through spectroscopic techniques is described. A 0.75-meter Czerny-Turner spectrometer with a 1200-g/mm grating and a 35-meter quartz optic bundle transmission line were tested. A high-resolution 2-meter spectrometer will be ordered. Data acquisition considerations are being addressed

  14. Edge turbulence imaging in the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Stotler, D.P.; Terry, J.L.; La Bombard, B.; Greenwald, M.; Muterspaugh, M.; Pitcher, C.S.; Hallatschek, K.; Maqueda, R.J.; Rogers, B.; Lowrance, J.L.; Mastrocola, V.J.; Renda, G.F.

    2002-01-01

    The two-dimensional (2D) radial vs poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I. H. Hutchinson, R. Boivin, P. T. Bonoli et al., Nucl. Fusion 41, 1391 (2001)] was measured using fast cameras and compared with three-dimensional numerical simulations of edge plasma turbulence. The main diagnostic is gas puff imaging, in which the visible D α emission from a localized D 2 gas puff is viewed along a local magnetic field line. The observed D α fluctuations have a typical radial and poloidal scale of ≅1 cm, and often have strong local maxima ('blobs') in the scrape-off layer. The motion of this 2D structure motion has also been measured using an ultrafast framing camera with 12 frames taken at 250 000 frames/s. Numerical simulations produce turbulent structures with roughly similar spatial and temporal scales and transport levels as that observed in the experiment; however, some differences are also noted, perhaps requiring diagnostic improvement and/or additional physics in the numerical model

  15. Alcator C-MOD proposal addendum

    International Nuclear Information System (INIS)

    Bonoli, P.; Greenwald, M.; Gwinn, D.

    1986-04-01

    Since the design concept and overall purpose of the Alcator C-MOD device are similar to that proposed in October 1985, we have chosen in this document only to highlight areas where changes or additions have been made. Chapters in the Addendum correspond to those in the Proposal, except Chapter 9 which describes a number of toroidal improvement concepts which are being considered for inclusion in the Alcator C-MOD experimental program. A description of the redesign and a discussion of the objectives of the experimental program are given

  16. Conceptual design Alcator C-MOD magnetic systems

    International Nuclear Information System (INIS)

    Schultz, J.H.; Becker, H.; Fertl, K.; Gwinn, D.; Montgomery, D.B.; Pierce, N.T.; Pillsbury, R.D. Jr.; Thome, R.J.

    1986-01-01

    The conceptual designs of the magnetic systems for Alcator C-MOD, a proposed tokamak at M.I.T., are described, including the toroidal magnet, the poloidal field coils and the cryogenic system. The toroidal magnet is constructed from rectangular plates, connected by sliding joints. Toroidal magnet forces are contained by a steel superstructure. Poloidal coil system options are largely or wholly inside the TF magnet, in order to control plasmas with high current, strong shaping, and expanded boundaries. All magnets are cryocooled by the natural circulation of boiling liquid nitrogen. 3 refs., 5 figs

  17. Gas jet disruption mitigation studies on Alcator C-Mod

    International Nuclear Information System (INIS)

    Granetz, R.; Whyte, D.G.; Izzo, V.A.; Biewer, T.; Reinke, M.L.; Terry, J.; Bader, A.; Bakhtiari, M.; Jernigan, T.; Wurden, G.

    2006-01-01

    Damaging effects of disruptions are a major concern for Alcator C-Mod, ITER and future tokamak reactors. High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the operational requirements of fast response time and reliability, while still being benign to subsequent discharges. Disruption mitigation experiments using an optimized gas jet injection system are being carried out on Alcator C-Mod to study the physics of gas jet penetration into high pressure plasmas, as well as the ability of the gas jet impurities to convert plasma energy into radiation on timescales consistent with C-Mod's fast quench times, and to reduce halo currents given C-Mod's high-current density. The dependence of impurity penetration and effectiveness on noble gas species (He, Ne, Ar, Kr) is also being studied. It is found that the high-pressure neutral gas jet does not penetrate deeply into the C-Mod plasma, and yet prompt core thermal quenches are observed on all gas jet shots. 3D MHD modelling of the disruption physics with NIMROD shows that edge cooling of the plasma triggers fast growing tearing modes which rapidly produce a stochastic region in the core of the plasma and loss of thermal energy. This may explain the apparent effectiveness of the gas jet in C-Mod despite its limited penetration. The higher-Z gases (Ne, Ar, Kr) also proved effective at reducing halo currents and decreasing thermal deposition to the divertor surfaces. In addition, noble gas jet injection proved to be benign for plasma operation with C-Mod's metal (Mo) wall, actually improving the reliability of the startup in the following discharge

  18. Rf modeling and design of a folded waveguide launcher for the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Bigelow, T.S.; Fogelman, C.F.; Baity, F.W.; Carter, M.D.; Hoffman, D.J.; Ryan, P.M.; Yugo, J.J.; Golovato, S.N.; Bonoli, P.

    1993-01-01

    The folded waveguide (FWG) launcher is being investigated as an improved antenna configuration for plasma heating in the ion cyclotron range of frequencies (ICRF). A development FWG launcher was successfully tested at Oak Ridge National Laboratory (ORNL) with a low-density plasma load and found to have significantly greater power density capability than current strap-type antennas operating in similar plasmas. To further test the concept on a high density tokamak plasma, a collaboration has been set up between ORNL and Massachusetts Institute of Technology (MIT) to develop and test an 80-MHz, 2-MW FWG on the Alcator C-Mod tokamak at MIT. The radio frequency (rf) electromagnetic modeling techniques and laboratory measurements used in the design of this antenna are described in this paper. A companion paper describes the mechanical design of the FWG

  19. Alcator C-Mod predictive modeling

    International Nuclear Information System (INIS)

    Pankin, Alexei; Bateman, Glenn; Kritz, Arnold; Greenwald, Martin; Snipes, Joseph; Fredian, Thomas

    2001-01-01

    Predictive simulations for the Alcator C-mod tokamak [I. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] are carried out using the BALDUR integrated modeling code [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1988)]. The results are obtained for temperature and density profiles using the Multi-Mode transport model [G. Bateman et al., Phys. Plasmas 5, 1793 (1998)] as well as the mixed-Bohm/gyro-Bohm transport model [M. Erba et al., Plasma Phys. Controlled Fusion 39, 261 (1997)]. The simulated discharges are characterized by very high plasma density in both low and high modes of confinement. The predicted profiles for each of the transport models match the experimental data about equally well in spite of the fact that the two models have different dimensionless scalings. Average relative rms deviations are less than 8% for the electron density profiles and 16% for the electron and ion temperature profiles

  20. Tungsten nano-tendril growth in the Alcator C-Mod divertor

    International Nuclear Information System (INIS)

    Wright, G.M.; Brunner, D.; Labombard, B.; Lipschultz, B.; Terry, J.L.; Whyte, D.G.; Baldwin, M.J.; Doerner, R.P.

    2012-01-01

    Growth of tungsten nano-tendrils (‘fuzz’) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The thickness of the individual nano-tendrils (50–100 nm) and the depth of the layer (600 ± 150 nm) are consistent with observations from experiments on linear plasma devices. The observation of tungsten fuzz in a tokamak may have important implications for material erosion, dust formation, divertor lifetime and tokamak operations in next-step devices. (letter)

  1. Molybdenum erosion measurements in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Wampler, W.R. [Sandia National Labs., Albuquerque, NM (United States); LaBombard, B.; Lipshultz, B.; Pappas, D.; Pitcher, C.S. [Massachusetts Inst. of Tech., Cambridge, MA (United States); McCracken, G.M. [JET Joint Undertaking, Abingdon (United Kingdom)

    1998-05-01

    Erosion of molybdenum was measured on a set of 21 tiles after a run campaign of 1,090 shots in the Alcator C-Mod tokamak. The net erosion of molybdenum, was determined from changes in the depth of a thin chromium marker layer measured by Rutherford backscattering. Net Mo erosion was found to be approximately 150 nm near the outer divertor strike point, and much less everywhere else. Gross erosion rates by sputtering were estimated using ion energies and fluxes obtained from Langmuir probe measurements of edge-plasma conditions. Predicted net erosion using calculated gross erosion with prompt redeposition and measured net erosion agree within a factor of 3. Sputtering by boron and molybdenum impurities dominates erosion.

  2. Overview of recent results from a Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Marmar, E.; Batishchev, O.; Acedo, P.

    1999-01-01

    Recent results from the compact, high field, Alcator C-Mod tokamak program are summarized. H-mode threshold studies have demonstrated that the threshold appears to be closely related to local edge plasma parameters: for fixed field and plasma current, T e (ψ 95 ) takes on a density independent value at the transition. The Enhanced D-Alpha H-Mode (EDA) regime has been investigated. EDA is distinct from ELM free H-mode, in that there is no accumulation of impurities, and at the same time EDA does not exhibit large discrete ELMs. The energy confinement is degraded by only about 10%, compared to ELM free. Comparisons for EDA with ELMy H-Mode database scalings indicate τ EDA ∼1.2 τ ITER97H . Strong toroidal rotation is observed in ICRF-only auxiliary heated plasmas; the rotation increases with plasma pressure, and decreases with increasing plasma current. The inferred radial electric field reaches the order of 30 kV/m near the center of the plasma. Through feedback controlled nitrogen impurity puffing, steady state detached EDA H-Modes have been achieved with Z eff E is reduced by about 10% in the detached case, compared to the confinement before the N 2 puff begins. The heat load to the divertor is reduced by a factor of 4. Volume recombination rates are measured in the divertor, using 2-d tomography of Balmer series TV movies. Volume recombination can be a significant contributor to the overall reduction in ion current to the divertor plates which occurs in detachment. Particle balance measurements indicate that the divertor and main chamber plasmas are largely isolated from one another, at least with regard to particle recycling, with most of the main chamber (core plus scrape-off) fueling coming from neutrals in the main chamber volume. With the addition of Lower Hybrid Current Drive, C-Mod would be an ideal vehicle for investigation of advanced tokamak operation with fully relaxed current profiles. Detailed modeling indicates that discharges approaching the

  3. Overview of recent results from the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Marmar, E.; Batishchev, O.; Acedo, P.

    2001-01-01

    Recent results from the compact, high field, Alcator C-Mod tokamak program are summarized. H-mode threshold studies have demonstrated that the threshold appears to be closely related to local edge plasma parameters: for fixed field and plasma current, T e (ψ 95 ) takes on a density independent value at the transition. The Enhanced D-Alpha H-Mode (EDA) regime has been investigated. EDA is distinct from ELM free H mode, in that there is no accumulation of impurities, and at the same time EDA does not exhibit large discrete ELMs. The energy confinement is degraded by only about 10%, compared to ELM free. Comparisons for EDA with ELMy H-Mode database scalings indicate τEDA ∼ 1.2τ ITER97H . Strong toroidal rotation is observed in ICRF-only auxiliary heated plasmas; the rotation increases with plasma pressure, and decreases with increasing plasma current. The inferred radial electric field reaches the order of 30kV/m near the center of the plasma. Through feedback controlled nitrogen impurity puffing, steady state detached EDA H-Modes have been achieved with Z eff E is reduced by about 10% in the detached case, compared to the confinement before the N 2 puff begins. The heat load to the divertor is reduced by a factor of 4. Volume recombination rates are measured in the divertor, using 2-d tomography of Balmer series TV movies. Volume recombination can be a significant contributor to the overall reduction in ion current to the divertor plates which occurs in detachment. Particle balance measurements indicate that the divertor and main chamber plasmas are largely isolated from one another, at least with regard to particle recycling, with most of the main chamber (core plus scrape-off) fueling coming from neutrals in the main chamber volume. With the addition of Lower Hybrid Current Drive, C-Mod would be an ideal vehicle for investigation of advanced tokamak operation with fully relaxed current profiles. Detailed modeling indicates that discharges approaching the

  4. Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor

    International Nuclear Information System (INIS)

    Stotler, D.P.; Pitcher, C.S.; Boswell, C.J.; LaBombard, B.; Terry, J.L.; Elder, J.D.; Lisgo, S.

    2002-01-01

    A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration a re highlighted. Improvements to atomic and surface physics data and models will play a role in the latter

  5. 20 years of research on the Alcator C-Mod tokamaka)

    Science.gov (United States)

    Greenwald, M.; Bader, A.; Baek, S.; Bakhtiari, M.; Barnard, H.; Beck, W.; Bergerson, W.; Bespamyatnov, I.; Bonoli, P.; Brower, D.; Brunner, D.; Burke, W.; Candy, J.; Churchill, M.; Cziegler, I.; Diallo, A.; Dominguez, A.; Duval, B.; Edlund, E.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Garcia, O.; Gao, C.; Goetz, J.; Golfinopoulos, T.; Granetz, R.; Grulke, O.; Hartwig, Z.; Horne, S.; Howard, N.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.; Izzo, V.; Kessel, C.; LaBombard, B.; Lau, C.; Li, C.; Lin, Y.; Lipschultz, B.; Loarte, A.; Marmar, E.; Mazurenko, A.; McCracken, G.; McDermott, R.; Meneghini, O.; Mikkelsen, D.; Mossessian, D.; Mumgaard, R.; Myra, J.; Nelson-Melby, E.; Ochoukov, R.; Olynyk, G.; Parker, R.; Pitcher, S.; Podpaly, Y.; Porkolab, M.; Reinke, M.; Rice, J.; Rowan, W.; Schmidt, A.; Scott, S.; Shiraiwa, S.; Sierchio, J.; Smick, N.; Snipes, J. A.; Snyder, P.; Sorbom, B.; Stillerman, J.; Sung, C.; Takase, Y.; Tang, V.; Terry, J.; Terry, D.; Theiler, C.; Tronchin-James, A.; Tsujii, N.; Vieira, R.; Walk, J.; Wallace, G.; White, A.; Whyte, D.; Wilson, J.; Wolfe, S.; Wright, G.; Wright, J.; Wukitch, S.; Zweben, S.

    2014-11-01

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

  6. Alcator C-MOD final safety analysis

    International Nuclear Information System (INIS)

    Fiore, C.L.

    1989-06-01

    This document is designed to address the safety issues involved with the Alcator C-Mod project. This report will begin with a brief description of the experimental objectives which will be followed by information concerning the site. The Alcator C-Mod experiment is a pulsed fusion experiment in which a plasma formed from small amounts of hydrogen or deuterium gas is confined in a magnetic field for short periods (∼1 s). No radioactive fuels or fissile materials are used in the device, so that no criticality hazard exists and no credible nuclear accident can occur. During deuterium operation, the production of a small number of neutrons from a short pulse could result in a small amount of short- and intermediate-lived radioactive isotopes being produced inside the experimental cell. This report will demonstrate that this does not pose an additional hazard to the general population. The health and safety hazards resulting from Alcator C-Mod occur to the workers on the experiment, each of which is described in its own chapter with the steps taken to minimize the risk to employees. These hazards include fire, chemicals and cryogenics, air quality, electrical, electromagnetic radiation, ionizing radiation, and mechanical and natural phenomena. None of these hazards is unique to the facility, and methods of protection from them are well defined and are discussed in the chapter which describes each hazard. The quality assurance program, critical to ensuring the safety aspects of the program, will also be described

  7. Investigation of the critical edge ion heat flux for L-H transitions in Alcator C-Mod and its dependence on B T

    Science.gov (United States)

    Schmidtmayr, M.; Hughes, J. W.; Ryter, F.; Wolfrum, E.; Cao, N.; Creely, A. J.; Howard, N.; Hubbard, A. E.; Lin, Y.; Reinke, M. L.; Rice, J. E.; Tolman, E. A.; Wukitch, S.; Ma, Y.; ASDEX Upgrade Team; Alcator C-Mod Team

    2018-05-01

    This paper presents investigations on the role of the edge ion heat flux for transitions from L-mode to H-mode in Alcator C-Mod. Previous results from the ASDEX Upgrade tokamak indicated that a critical value of edge ion heat flux per particle is needed for the transition. Analysis of C-Mod data confirms this result. The edge ion heat flux is indeed found to increase linearly with density at given magnetic field and plasma current. Furthermore, the Alcator C-Mod data indicate that the edge ion heat flux at the L-H transition also increases with magnetic field. Combining the data from Alcator C-Mod and ASDEX Upgrade yields a general expression for the edge ion heat flux at the L-H transition. These results are discussed from the point of view of the possible physics mechanism of the L-H transition. They are also compared to the L-H power threshold scaling and an extrapolation for ITER is given.

  8. Edge Minority Heating Experiment in Alcator C-Mod

    International Nuclear Information System (INIS)

    Zweben, S.J.; Terry, J.L.; Bonoli, P.; Budny, R.; Chang, C.S.; Fiore, C.; Schilling, G.; Wukitch, S.; Hughes, J.; Lin, Y.; Perkins, R.; Porkolab, M.; Alcator C-Mod Team

    2005-01-01

    An attempt was made to control global plasma confinement in the Alcator C-Mod tokamak by applying ion cyclotron resonance heating (ICRH) power to the plasma edge in order to deliberately create a minority ion tail loss. In theory, an edge fast ion loss could modify the edge electric field and so stabilize the edge turbulence, which might then reduce the H-mode power threshold or improve the H-mode barrier. However, the experimental result was that edge minority heating resulted in no improvement in the edge plasma parameters or global stored energy, at least at power levels of P RF (le) 5.5 MW. A preliminary analysis of these results is presented and some ideas for improvement are discussed

  9. Particle size distribution of dust collected from Alcator C-MOD

    International Nuclear Information System (INIS)

    Gorman, S.V.; Carmack, W.J.; Hembree, P.B.

    1998-01-01

    There are important safety issues associated with tokamak dust, accumulated primarily from sputtering and disruptions. The dust may contain tritium, it may be activated, chemically toxic, and chemically reactive. The purpose of this paper is to present results from analyses of particulate collected from the Alcator C-MOD tokamak located at Massachusetts Institute of Technology (MIT) in Cambridge, Massachusetts. The sample obtained from C-MOD was not originally intended for examination outside of MIT. The sample was collected with the intent of performing only a composition analysis. However, MIT provided the INEEL with this sample for particle analysis. The sample was collected by vacuuming a section of the machine (covering approximately 1/3 of the machine surface) with a coarse fiber filter as the collection surface. The sample was then analyzed using an optical microscope, SEM microscope, Microtrac FRA particle size analyzer. The data fit a log-normal distribution. The count median diameter (CMD) of the samples ranged from 0.3 microm to 1.1 microm with geometric standard deviations (GSD) ranging from 2.8 to 5.2 and a mass median diameter (MMD) ranging from 7.22 to 176 microm

  10. Disruption Warning Database Development and Exploratory Machine Learning Studies on Alcator C-Mod

    Science.gov (United States)

    Montes, Kevin; Rea, Cristina; Granetz, Robert

    2017-10-01

    A database of about 1800 shots from the 2015 campaign on the Alcator C-Mod tokamak is assembled, including disruptive and non-disruptive discharges. The database consists of 40 relevant plasma parameters with data taken from 160k time slices. In order to investigate the possibility of developing a robust disruption prediction algorithm that is tokamak-independent, we focused machine learning studies on a subset of dimensionless parameters such as βp, n /nG , etc. The Random Forests machine learning algorithm provides insight on the available data set by ranking the relative importance of the input features. Its application on the C-Mod database, however, reveals that virtually no one parameter has more importance than any other, and that its classification algorithm has a low rate of successfully predicted samples, as well as poor false positive and false negative rates. Comparing the analysis of this algorithm on the C-Mod database with its application to a similar database on DIII-D, we conclude that disruption prediction may not be feasible on C-Mod. This conclusion is supported by empirical observations that most C-Mod disruptions are caused by radiative collapse due to molybdenum from the first wall, which happens on just a 1-2ms timescale. Supported by the US Dept. of Energy under DE-FC02-99ER54512 and DE-FC02-04ER54698.

  11. Experimental Study of Reversed Shear Alfven Eigenmodes During The Current Ramp In The Alcator C-Mod Tokamak

    International Nuclear Information System (INIS)

    Edlund, E.M.; Porkolab, M.; Kramer, G.J.; Lin, L.; Lin, Y.; Tsuji, N.; Wukitch, S.J.

    2010-01-01

    Experiments conducted in the Alcator C-Mod tokamak at MIT have explored the physics of reversed shear Alfven eigenmodes (RSAEs) during the current ramp. The frequency evolution of the RSAEs throughout the current ramp provides a constraint on the evolution of q min , a result which is important in transport modeling and for comparison with other diagnostics which directly measure the magnetic field line structure. Additionally, a scaling of the RSAE minimum frequency with the sound speed is used to derive a measure of the adiabatic index, a measure of the plasma compressibility. This scaling bounds the adiabatic index at 1.40 ± 0.15 used in MHD models and supports the kinetic calculation of separate electron and ion compressibilities with an ion adiabatic index close to 7/4.

  12. Tungsten impurity transport experiments in Alcator C-Mod to address high priority research and development for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Loarte, A.; Polevoi, A. R.; Hosokawa, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Reinke, M. L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Chilenski, M.; Howard, N.; Hubbard, A.; Hughes, J. W.; Rice, J. E.; Walk, J. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Köchl, F. [Technische Universität Wien, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Pütterich, T.; Dux, R. [Max-Planck-Institut für Plasmaphysik, Boltzmanstraße 2, D-85748 Garching (Germany); Zhogolev, V. E. [NRC “Kurchatov Institute,” Kurchatov Square 1, 123098 Moscow (Russian Federation)

    2015-05-15

    Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha H-mode regime with ITER-like mid-radius plasma density peaking and Ion Cyclotron Resonant heating, in which tungsten is introduced by the laser blow-off technique, have demonstrated that accumulation of tungsten in the central region of the plasma does not take place in these conditions. The measurements obtained are consistent with anomalous transport dominating tungsten transport except in the central region of the plasma where tungsten transport is neoclassical, as previously observed in other devices with dominant neutral beam injection heating, such as JET and ASDEX Upgrade. In contrast to such results, however, the measured scale lengths for plasma temperature and density in the central region of these Alcator C-Mod plasmas, with density profiles relatively flat in the core region due to the lack of core fuelling, are favourable to prevent inter and intra sawtooth tungsten accumulation in this region under dominance of neoclassical transport. Simulations of ITER H-mode plasmas, including both anomalous (modelled by the Gyro-Landau-Fluid code GLF23) and neoclassical transport for main ions and tungsten and with density profiles of similar peaking to those obtained in Alcator C-Mod show that accumulation of tungsten in the central plasma region is also unlikely to occur in stationary ITER H-mode plasmas due to the low fuelling source by the neutral beam injection (injection energy ∼ 1 MeV), which is in good agreement with findings in the Alcator C-Mod experiments.

  13. Advances in measurement and modeling of the high-confinement-mode pedestal on the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Hughes, J.W.; LaBombard, B.; Mossessian, D.A.; Hubbard, A.E.; Terry, J.; Biewer, T.

    2006-01-01

    Edge transport barrier (ETB) studies on the Alcator C-Mod tokamak [Phys. Plasmas 1, 1511 (1994)] investigate pedestal scalings and the radial transport of plasma and neutrals. Pedestal profiles show trends with plasma operational parameters such as total current I P . A ballooning-like I P 2 dependence is seen in the pressure gradient, despite calculated stability to ideal ballooning modes. A similar scaling is seen in the near scrape-off layer for both low-confinement (L-mode) and H-mode discharges, possibly due to electromagnetic fluid drift turbulence setting transport near the separatrix. Neutral density diagnosis allows an examination of D 0 fueling in H-modes, yielding profiles of effective particle diffusivity in the ETB, which vary as I P is changed. Edge neutral transport is studied using a one-dimensional kinetic treatment. In both experiment and modeling, the C-Mod density pedestal exhibits a weakly increasing pedestal density and a nearly invariant density pedestal width as the D 0 source rate increases. Identical modeling performed on pedestal profiles typical of DIII-D [Nucl. Fusion 42, 614 (2002)] reveal differences in pedestal scalings qualitatively similar to experimental results

  14. Long Term Retention of Deuterium and Tritium in Alcator C-Mod

    International Nuclear Information System (INIS)

    FIORE, C.; LABOMBARD, B.; LIPSCHULTZ, B.; PITCHER, C.S.; SKINNER, C.H.; WAMPLER, WILLIAM R.

    1999-01-01

    We estimate the total in-vessel deuterium retention in Alcator C-Mod from a run campaign of about 1090 plasmas. The estimate is based on measurements of deuterium retained on 22 molybdenum tiles from the inner wall and divertor. The areal density of deuterium on the tiles was measured by nuclear reaction analysis. From these data, the in-vessel deuterium inventory is estimated to be about 0.1 gram, assuming the deuterium coverage is toroidally symmetric. Most of the retained deuterium is on the walls of the main plasma chamber, only about 2.5% of the deuterium is in the divertor. The D coverage is consistent with a layer saturated by implantation with ions and charge-exchange neutrals from the plasma. This contrasts with tokamaks with carbon plasma-facing components (PFC's) where long-term retention of tritium and deuterium is large and mainly in the divertor due to codeposition with carbon eroded by the plasma. The low deuterium retention in the C-Mod divertor is mainly due to the absence of carbon PFC's in C-Mod and the low erosion rate of Mo

  15. Boundary plasma heat flux width measurements for poloidal magnetic fields above 1 Tesla in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, Dan; Labombard, Brian; Kuang, Adam; Terry, Jim; Alcator C-Mod Team

    2017-10-01

    The boundary heat flux width, along with the total power flowing into the boundary, sets the power exhaust challenge for tokamaks. A multi-machine boundary heat flux width database found that the heat flux width in H-modes scaled inversely with poloidal magnetic field (Bp) and was independent of machine size. The maximum Bp in the database was 0.8 T, whereas the ITER 15 MA, Q =10 scenario will be 1.2 T. New measurements of the boundary heat flux width in Alcator C-Mod extend the international database to plasmas with Bp up to 1.3 T. C-Mod was the only experiment able to operate at ITER-level Bp. These new measurements are from over 300 plasma shots in L-, I-, and EDA H-modes spanning essentially the whole operating space in C-Mod. We find that the inverse-Bp dependence of the heat flux width in H-modes continues to ITER-level Bp, further reinforcing the empirical projection of 500 μm heat flux width for ITER. We find 50% scatter around the inverse-Bp scaling and are searching for the `hidden variables' causing this scatter. Supported by USDoE award DE-FC02-99ER54512.

  16. The measurement of the intrinsic impurities of molybdenum and carbon in the Alcator C-Mod tokamak plasma using low resolution spectroscopy

    International Nuclear Information System (INIS)

    May, M.J.; Finkenthal, M.; Regan, S.P.

    1997-01-01

    The intrinsic impurity content of molybdenum and carbon was measured in the Alcator C-Mod tokamak using low resolution, multilayer mirror (MLM) spectroscopy (Δλ ∼ 1-10 A). Molybdenum was the dominant high-Z impurity and originated from the molybdenum armour tiles covering all the plasma facing surfaces (including the inner column, the poloidal divertor plates and the ion cyclotron resonant frequency (ICRF) limiter) at Alcator C-Mod. Soft X ray extreme ultraviolet (XUV) emission, lines of charge states, ranging from hydrogen-like to helium-like lines of carbon (radius/minor radius, r/a ∼ 1) at the plasma edge to potassium- to chlorine-like (0.4 eff value, and the power losses through line radiation were estimated. For the diverted ohmically heated plasma examined, the intrinsic molybdenum and carbon concentrations in the core plasma were found to be ∼ 1.2 x 10 10 and ∼ 1.7 x 10 12 cm -3 , respectively. These measurements were obtained before the plasma facing components were boronized. The calculated radiated power from molybdenum was 170 kW; for carbon it was 45 kW. The contribution to the measured Z eff - 1 value of ∼ 0.8 was ∼ 0.11 for molybdenum and ∼ 0.5 for carbon. (author). 36 refs, 11 figs, 3 tabs

  17. Visible Spectrometer at the Compact Toroid Injection Experiment, the Sustained Spheromak Plasma Experiment and the Alcator C-Mod Tokamak for Doppler Width and Shift Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Graf, A; Howard, S; Horton, R; Hwang, D; May, M; Beiersdorfer, P; McLean, H; Terry, J

    2006-05-15

    A novel Doppler spectrometer is currently being used for ion or neutral velocity and temperature measurements on the Alcator C-Mod Tokamak. The spectrometer has an f/No. of {approx}3.1 and is appropriate for visible light (3500-6700 {angstrom}). The full width at half maximum from a line emitting calibration source has been measured to be as small as 0.4 {angstrom}. The ultimate time resolution is line brightness light limited and on the order of ms. A new photon efficient detector is being used for the setup at C-Mod. Time resolution is achieved by moving the camera during a plasma discharge in a perpendicular direction through the dispersion plane of the spectrometer causing a vertical streaking across the camera face. Initial results from C-Mod as well as previous measurements from the Compact Toroid Injection Experiment (CTIX) and the Sustained Spheromak Plasma Experiment (SSPX) are presented.

  18. Dissipative divertor operation in the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Lipschultz, B.; Goetz, J.; LaBombard, B.; McCracken, G.M.; Terry, J.L.; Graf, M.; Granetz, R.S.; Jablonski, D.; Kurz, C.; Niemczewski, A.; Snipes, J.

    1995-01-01

    The achievement of large volumetric power losses (dissipation) in the Alcator C-Mod divertor region is demonstrated in two operational modes: radiative divertor and detached divertor. During radiative divertor operation, the fraction of SOL power lost by radiation is P R /P SOL ∼0.8 with single null plasmas, n e 20 m -3 and I p e,div ≤6x10 20 m -3 . As the divertor radiation and density increase, the plasma eventually detaches abruptly from the divertor plates: I SAT drops at the target and the divertor radiation peak moves to the X-point region. Probe measurements at the divertor plate show that the transition occurs when T e ∼5 eV. The critical n e for detachment depends linearly on the input power. This abrupt divertor detachment is preceded by a comparatively long period ( similar 1-200 ms) where a partial detachment is observed to grow at the outer divertor plate. ((orig.))

  19. Molybdenum emission from impurity-induced m= 1 snake-modes on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Delgado-Aparicio, L. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Bitter, M.; Gates, D.; Hill, K.; Pablant, N. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Granetz, R.; Reinke, M.; Podpaly, Y.; Rice, J. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Beiersdorfer, P. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Sugiyama, L. [MIT - Laboratory for Nuclear Science, Cambridge, Massachusetts 02139 (United States)

    2012-10-15

    A suite of novel high-resolution spectroscopic imaging diagnostics has facilitated the identification and localization of molybdenum impurities as the main species during the formation and lifetime of m= 1 impurity-induced snake-modes on Alcator C-Mod. Such measurements made it possible to infer, for the first time, the perturbed radiated power density profiles from which the impurity density can be deduced.

  20. Comparison of tungsten nano-tendrils grown in Alcator C-Mod and linear plasma devices

    International Nuclear Information System (INIS)

    Wright, G.M.; Brunner, D.; Baldwin, M.J.; Bystrov, K.; Doerner, R.P.; Labombard, B.; Lipschultz, B.; De Temmerman, G.; Terry, J.L.; Whyte, D.G.; Woller, K.B.

    2013-01-01

    Growth of tungsten nano-tendrils (“fuzz”) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The depth of the W fuzz layer (600 ± 150 nm) is consistent with an empirical growth formula from the PISCES experiment. Re-creating the C-Mod exposures as closely as possible in Pilot-PSI experiment can produce nearly-identical nano-tendril morphology and layer thickness at surface temperatures that agree with uncertainties with the C-Mod W probe temperature data. Helium concentrations in W fuzz layers are measured at 1–4 at.%, which is lower than expected for the observed sub-surface voids to be filled with several GPa of helium pressure. This possibly indicates that the void formation is not pressure driven

  1. Lithium pellet injection experiments on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, Darren Thomas [Univ. of California, Berkeley, CA (United States)

    1996-06-01

    A pellet enhanced performance mode, showing significantly reduced core transport, is regularly obtained after the injection of deeply penetrating lithium pellets into Alcator C-Mod discharges. These transient modes, which typically persist about two energy confinement times, are characterized by a steep pressure gradient (ℓp ℓ a/5) in the inner third of the plasma, indicating the presence of an internal transport barrier. Inside this barrier, particle and energy diffusivities are greatly reduced, with ion thermal diffusivity dropping to near neoclassical values. Meanwhile, the global energy confinement time shows a 30% improvement over ITER89-P L-mode scaling. The addition of ICRF auxiliary heating shortly after the pellet injection leads to high fusion reactivity with neutron rates enhanced by an order of magnitude over L-mode discharges with similar input powers. A diagnostic system for measuring equilibrium current density profiles of tokamak plasmas, employing high speed lithium pellets, is also presented. Because ions are confined to move along field lines, imaging the Li+ emission from the toroidally extended pellet ablation cloud gives the direction of the magnetic field. To convert from temporal to radial measurements, the 3-D trajectory of the pellet is determined using a stereoscopic tracking system. These measurements, along with external magnetic measurements, are used to solve the Grad-Shafranov equation for the magnetic equilibrium of the plasma. This diagnostic is used to determine the current density profile of PEP modes by injection of a second pellet during the period of good confinement. This measurement indicates that a region of reversed magnetic shear exists at the plasma core. This current density profile is consistent with TRANSP calculations for the bootstrap current created by the pressure gradient. MHD stability analysis indicates that these plasmas are near the n = ∞ and the n = 1 marginal stability limits.

  2. Lithium pellet injection experiments on the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Garnier, D.T.

    1996-06-01

    A pellet enhanced performance mode, showing significantly reduced core transport, is regularly obtained after the injection of deeply penetrating lithium pellets into Alcator C-Mod discharges. These transient modes, which typically persist about two energy confinement times, are characterized by a steep pressure gradient (ell p ≤ a/5) in the inner third of the plasma, indicating the presence of an internal transport barrier. Inside this barrier, particle and energy diffusivities are greatly reduced, with ion thermal diffusivity dropping to near neoclassical values. Meanwhile, the global energy confinement time shows a 30% improvement over ITER89-P L-mode scaling. The addition of ICRF auxiliary heating shortly after the pellet injection leads to high fusion reactivity with neutron rates enhanced by an order of magnitude over L-mode discharges with similar input powers. A diagnostic system for measuring equilibrium current density profiles of tokamak plasmas, employing high speed lithium pellets, is also presented. Because ions are confined to move along field lines, imaging the Li + emission from the toroidally extended pellet ablation cloud gives the direction of the magnetic field. To convert from temporal to radial measurements, the 3-D trajectory of the pellet is determined using a stereoscopic tracking system. These measurements, along with external magnetic measurements, are used to solve the Grad-Shafranov equation for the magnetic equilibrium of the plasma. This diagnostic is used to determine the current density profile of PEP modes by injection of a second pellet during the period of good confinement. This measurement indicates that a region of reversed magnetic shear exists at the plasma core. This current density profile is consistent with TRANSP calculations for the bootstrap current created by the pressure gradient. MHD stability analysis indicates that these plasmas are near the n = ∞ and the n = 1 marginal stability limits

  3. Experiments and Simulations of ITER-like Plasmas in Alcator C-Mod

    International Nuclear Information System (INIS)

    Wilson, R.; Kessel, C.E.; Wolfe, S.; Hutchinson, I.H.; Bonoli, P.; Fiore, C.; Hubbard, A.E.; Hughes, J.; Lin, Y.; Ma, Y.; Mikkelsen, D.; Reinke, M.; Scott, S.; Sips, A.C.C.; Wukitch, S.

    2010-01-01

    Alcator C-Mod is performing ITER-like experiments to benchmark and verify projections to 15 MA ELMy H-mode Inductive ITER discharges. The main focus has been on the transient ramp phases. The plasma current in C-Mod is 1.3 MA and toroidal field is 5.4 T. Both Ohmic and ion cyclotron (ICRF) heated discharges are examined. Plasma current rampup experiments have demonstrated that (ICRF and LH) heating in the rise phase can save voltseconds (V-s), as was predicted for ITER by simulations, but showed that the ICRF had no effect on the current profile versus Ohmic discharges. Rampdown experiments show an overcurrent in the Ohmic coil (OH) at the H to L transition, which can be mitigated by remaining in H-mode into the rampdown. Experiments have shown that when the EDA H-mode is preserved well into the rampdown phase, the density and temperature pedestal heights decrease during the plasma current rampdown. Simulations of the full C-Mod discharges have been done with the Tokamak Simulation Code (TSC) and the Coppi-Tang energy transport model is used with modified settings to provide the best fit to the experimental electron temperature profile. Other transport models have been examined also.

  4. The effects of field reversal on the Alcator C-Mod divertor

    International Nuclear Information System (INIS)

    Hutchinson, I.H.; LaBombard, B.; Goetz, J.A.; Lipschultz, B.; McCracken, G.M.; Snipes, J.A.; Terry, J.L.

    1995-01-01

    Imbalances between the inboard and outboard legs of the single null divertor in tokamak Alcator C-Mod are observed to reverse when the direction of the toroidal field is reversed. These imbalances are measured by embedded probes in the target plates, tomographic reconstructions of bolometry and line radiation, and visible imaging. Density imbalances of about a factor of ten at the targets are observed at moderate density, decreasing as the density is raised until they are almost balanced. The data indicate that the electron pressure is not imbalanced, thus arguing against momentum imbalance as the cause of these drift-induced effects. Instead, power flux imbalance caused by E r ''and'' B convection, and enhanced by radiation, is suggested as the underlying cause. (Author)

  5. Neutral Transport Simulations of Gas Puff Imaging Experiments on Alcator C-Mod

    International Nuclear Information System (INIS)

    Stotler, D.P.; LaBombard, B.; Terry, J.L.; Zweben, S.J.

    2002-01-01

    Visible imaging of gas puffs has been used on the Alcator C-Mod tokamak to characterize edge plasma turbulence, yielding data that can be compared with plasma turbulence codes. Simulations of these experiments with the DEGAS 2 Monte Carlo neutral transport code have been carried out to explore the relationship between the plasma fluctuations and the observed light emission. By imposing two-dimensional modulations on the measured time-average plasma density and temperature profiles, we demonstrate that the spatial structure of the emission cloud reflects that of the underlying turbulence. However, the photon emission rate depends on the plasma density and temperature in a complicated way, and no simple scheme for inferring the plasma parameters directly from the light emission patterns is apparent. The simulations indicate that excited atoms generated by molecular dissociation are a significant source of photons, further complicating interpretation of the gas puff imaging results.Visibl e imaging of gas puffs has been used on the Alcator C-Mod tokamak to characterize edge plasma turbulence, yielding data that can be compared with plasma turbulence codes. Simulations of these experiments with the DEGAS 2 Monte Carlo neutral transport code have been carried out to explore the relationship between the plasma fluctuations and the observed light emission. By imposing two-dimensional modulations on the measured time-average plasma density and temperature profiles, we demonstrate that the spatial structure of the emission cloud reflects that of the underlying turbulence. However, the photon emission rate depends on the plasma density and temperature in a complicated way, and no simple scheme for inferring the plasma parameters directly from the light emission patterns is apparent. The simulations indicate that excited atoms generated by molecular dissociation are a significant source of photons, further complicating interpretation of the gas puff imaging results

  6. Two dimensional radiated power diagnostics on Alcator C-Mod

    International Nuclear Information System (INIS)

    Reinke, M. L.; Hutchinson, I. H.

    2008-01-01

    The radiated power diagnostics for the Alcator C-Mod tokamak have been upgraded to measure two dimensional structure of the photon emissivity profile in order to investigate poloidal asymmetries in the core radiation. Commonly utilized unbiased absolute extreme ultraviolet (AXUV) diode arrays view the plasma along five different horizontal planes. The layout of the diagnostic set is shown and the results from calibrations and recent experiments are discussed. Data showing a significant, 30%-40%, inboard/outboard emissivity asymmetry during ELM-free H-mode are presented. The ability to use AXUV diode arrays to measure absolute radiated power is explored by comparing diode and resistive bolometer-based emissivity profiles for highly radiative L-mode plasmas seeded with argon. Emissivity profiles match in the core but disagree radially outward resulting in an underprediction of P rad of nearly 50% by the diodes compared to P rad determined using resistive bolometers.

  7. High speed movies of turbulence in Alcator C-Mod

    International Nuclear Information System (INIS)

    Terry, J.L.; Zweben, S.J.; Bose, B.; Grulke, O.; Marmar, E.S.; Lowrance, J.; Mastrocola, V.; Renda, G.

    2004-01-01

    A high speed (250 kHz), 300 frame charge coupled device camera has been used to image turbulence in the Alcator C-Mod Tokamak. The camera system is described and some of its important characteristics are measured, including time response and uniformity over the field-of-view. The diagnostic has been used in two applications. One uses gas-puff imaging to illuminate the turbulence in the edge/scrape-off-layer region, where D 2 gas puffs localize the emission in a plane perpendicular to the magnetic field when viewed by the camera system. The dynamics of the underlying turbulence around and outside the separatrix are detected in this manner. In a second diagnostic application, the light from an injected, ablating, high speed Li pellet is observed radially from the outer midplane, and fast poloidal motion of toroidal striations are seen in the Li + light well inside the separatrix

  8. Lower hybrid wave edge power loss quantification on the Alcator C-Mod tokamak

    Science.gov (United States)

    Faust, I. C.; Brunner, D.; LaBombard, B.; Parker, R. R.; Terry, J. L.; Whyte, D. G.; Baek, S. G.; Edlund, E.; Hubbard, A. E.; Hughes, J. W.; Kuang, A. Q.; Reinke, M. L.; Shiraiwa, S.; Wallace, G. M.; Walk, J. R.

    2016-05-01

    For the first time, the power deposition of lower hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal, and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt ( t Radiofrequency (LHRF) power. LHRF power was found to absorb more strongly in the edge at higher densities. It is found that a majority of this edge-deposited power is promptly conducted to the divertor. This correlates with the loss of current drive efficiency at high density previously observed on Alcator C-Mod, and displaying characteristics that contrast with the local RF edge absorption seen on other tokamaks. Measurements of ionization in the active divertor show dramatic changes due to LHRF power, implying that divertor region can be a key for the LHRF edge power deposition physics. These observations support the existence of a loss mechanism near the edge for LHRF at high density ( n e > 1.0 × 10 20 (m-3)). Results will be shown addressing the distribution of power within the SOL, including the toroidal symmetry and radial distribution. These characteristics are important for deducing the cause of the reduced LHCD efficiency at high density and motivate the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch.

  9. Edge Zonal Flows and Blob Propagation in Alcator C-Mod

    International Nuclear Information System (INIS)

    Zweben, S.; Terry, J.L.; Agostini, M.; Davis, B.; Grulke, O.; Hager, R.; Hughes, J.; LaBombard, B.; D'Ippolito, D.A.; Myra, J.R.; Russell, D.A.

    2011-01-01

    Here we describe recent measurements of the 2-D motion of turbulence in the edge and scrape-off layer (SOL) of the Alcator C-Mod tokamak. This data was taken using the outer midplane gas puff imaging (GPI) camera, which views a 6 cm radial by 6 cm poloidal region near the separatrix just below the outer midplane [1]. The data were taken in Ohmic or RF heated L-mode plasmas at 400,000 frames/sec for ∼50 msec/shot using a Phantom 710 camera in a 64 x 64 pixel format. The resulting 2-D vs. time movies [2] can resolve the structure and motion of the turbulence on a spatial scale covering 0.3-6 cm. The images were analyzed using either a 2-D cross-correlation code (Sec. 2) or a 2-D blob tracking code (Sec. 3).

  10. A Lower Hybrid Current Drive System for Alcator C-Mod

    International Nuclear Information System (INIS)

    Bernabei, S.; Hosea, J.C.; Loesser, D.; Rushinski, J.; Wilson, J.R.; Bonoli, P.; Grimes, M.; Parker, R.; Porkolab, M.; Terry, D.; Woskov, P.

    2001-01-01

    A Lower Hybrid Current Drive system is being constructed jointly by Plasma Science and Fusion Center (PSFC) and Princeton Plasma Physics Laboratory (PPPL) for installation on the Alcator C-Mod tokamak, with the primary goal of driving plasma current in the outer region of the plasma. The Lower Hybrid (LH) system consists of 3 MW power at 4.6 GHz with a maximum pulse length of 5 seconds. Twelve klystrons will feed an array of 4-vertical and 24-horizontal waveguides mounted in one equatorial port. The coupler will incorporate some compact characteristics of the multijunction power splitting while retaining full control of the toroidal phase. In addition a dynamic phase control system will allow feedback stabilization of MHD modes. The desire to avoid possible waveguide breakdown and the need for compactness have resulted in some innovative technical solution which will be presented

  11. Internal transport barriers on Alcator C-Mod

    International Nuclear Information System (INIS)

    Fiore, C.L.; Rice, J.E.; Bonoli, P.T.; Boivin, R.L.; Goetz, J.A.; Hubbard, A.E.; Hutchinson, I.H.; Granetz, R.S.; Greenwald, M.J.; Marmar, E.S.; Mossessian, D.; Porkolab, M.; Taylor, G.; Snipes, J.; Wolfe, S.M.; Wukitch, S.J.

    2001-01-01

    The formation of internal transport barriers (ITBs) has been observed in the core region of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under a variety of conditions. The improvement in core confinement following pellet injection (pellet enhanced performance or PEP mode) has been well documented on Alcator C-Mod in the past. Recently three new ITB phenomena have been observed which require no externally applied particle or momentum input. Short lived ITBs form spontaneously following the high confinement to low confinement mode transition and are characterized by a large increase in the global neutron production (enhanced neutron or EN modes). Experiments with ion cyclotron range of frequencies power injection to the plasma off-axis on the high field side results in the central density rising abruptly and becoming peaked. The ITB formed at this time lasts for ten energy confinement times. The central toroidal rotation velocity decreases and changes sign as the density rises. Similar spontaneous ITBs have been observed in ohmically heated H-mode plasmas. All of these ITB events have strongly peaked density profiles with a minimum in the density scale length occurring near r/a=0.5 and have improved confinement parameters in the core region of the plasma

  12. Production of internal transport barriers via self-generated mean flows in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Fiore, C. L.; Ernst, D. R.; Podpaly, Y. A.; Howard, N. T.; Lee, Jungpyo; Reinke, M. L.; Rice, J. E.; Hughes, J. W.; Ma, Y. [MIT-PSFC, 77 Mass. Ave., Cambridge, Massachusetts 02139 (United States); Mikkelsen, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States); Rowan, W. L.; Bespamyatnov, I. [FRC, U of Texas at Austin, Austin, Texas 78712 (United States)

    2012-05-15

    New results suggest that changes observed in the intrinsic toroidal rotation influence the internal transport barrier (ITB) formation in the Alcator C-Mod tokamak [E. S. Marmar and Alcator C-Mod group, Fusion Sci. Technol. 51, 261 (2007)]. These arise when the resonance for ion cyclotron range of frequencies (ICRF) minority heating is positioned off-axis at or outside of the plasma half-radius. These ITBs form in a reactor relevant regime, without particle or momentum injection, with Ti Almost-Equal-To Te, and with monotonic q profiles (q{sub min} < 1). C-Mod H-mode plasmas exhibit strong intrinsic co-current rotation that increases with increasing stored energy without external drive. When the resonance position is moved off-axis, the rotation decreases in the center of the plasma resulting in a radial toroidal rotation profile with a central well which deepens and moves farther off-axis when the ICRF resonance location reaches the plasma half-radius. This profile results in strong E Multiplication-Sign B shear (>1.5 Multiplication-Sign 10{sup 5} rad/s) in the region where the ITB foot is observed. Gyrokinetic analyses indicate that this spontaneous shearing rate is comparable to the linear ion temperature gradient (ITG) growth rate at the ITB location and is sufficient to reduce the turbulent particle and energy transport. New and detailed measurement of the ion temperature demonstrates that the radial profile flattens as the ICRF resonance position moves off axis, decreasing the drive for the ITG the instability as well. These results are the first evidence that intrinsic rotation can affect confinement in ITB plasmas.

  13. Flux-driven turbulence GDB simulations of the IWL Alcator C-Mod L-mode edge compared with experiment

    Science.gov (United States)

    Francisquez, Manaure; Zhu, Ben; Rogers, Barrett

    2017-10-01

    Prior to predicting confinement regime transitions in tokamaks one may need an accurate description of L-mode profiles and turbulence properties. These features determine the heat-flux width upon which wall integrity depends, a topic of major interest for research aid to ITER. To this end our work uses the GDB model to simulate the Alcator C-Mod edge and contributes support for its use in studying critical edge phenomena in current and future tokamaks. We carried out 3D electromagnetic flux-driven two-fluid turbulence simulations of inner wall limited (IWL) C-Mod shots spanning closed and open flux surfaces. These simulations are compared with gas puff imaging (GPI) and mirror Langmuir probe (MLP) data, examining global features and statistical properties of turbulent dynamics. GDB reproduces important qualitative aspects of the C-Mod edge regarding global density and temperature profiles, within reasonable margins, and though the turbulence statistics of the simulated turbulence follow similar quantitative trends questions remain about the code's difficulty in exactly predicting quantities like the autocorrelation time A proposed breakpoint in the near SOL pressure and the posited separation between drift and ballooning dynamics it represents are examined This work was supported by DOE-SC-0010508. This research used resources of the National Energy Research Scientific Computing Center (NERSC).

  14. The LHCD Launcher for Alcator C-Mod - Design, Construction, Calibration and Testing

    International Nuclear Information System (INIS)

    Hosea, J.; Beals, D.; Beck, W.; Bernabei, S.; Burke, W.; Childs, R.; Ellis, R.; Fredd, E.; Greenough, N.; Grimes, M.; Gwinn, D.; Irby, J.; Jurczynski, S.; Koert, P.; Kung, C.C.; Loesser, G.D.; Marmar, E.; Parker, R.; Rushinski, J.; Schilling, G.; Terry, D.; Vieira, R.; Wilson, J.R.; Zaks, J.

    2005-01-01

    MIT and PPPL have joined together to fabricate a high-power lower hybrid current drive (LHCD) system for supporting steady-state AT regime research on Alcator C-Mod. The goal of the first step of this project is to provide 1.5 MW of 4.6 GHz rf [radio frequency] power to the plasma with a compact launcher which has excellent spectral selectivity and fits into a single C-Mod port. Some of the important design, construction, calibration and testing considerations for the launcher leading up to its installation on C-Mod are presented here

  15. The Ar{sup 17+} Ly{sub {alpha}2}/Ly{sub {alpha}1} ratio in Alcator C-Mod tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Rice, J E; Reinke, M L; Ince-Cushman, A C; Podpaly, Y A [Plasma Science and Fusion Center, MIT, Cambridge, MA (United States); Ashbourn, J M A [Mathematical Institute, University of Oxford, Oxford (United Kingdom); Gu, M F [SSL, University of California Berkeley, CA (United States); Bitter, M; Hill, K [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Rachlew, E, E-mail: rice@psfc.mit.edu [KTH, Stockholm (Sweden)

    2011-08-28

    High-quality spectra of hydrogen-like Ar{sup 17+} have been obtained from Alcator C-Mod tokamak plasmas using a spatially imaging high-resolution x-ray spectrometer system in an extensive study of the underlying high-n satellite lines. The ratio of Ly{sub {alpha}2} (1S{sub 1/2}-2P{sub 1/2}) to Ly{sub {alpha}1} (1S{sub 1/2}-2P{sub 3/2}) was found to be {approx}0.52 regardless of plasma parameters, which is somewhat greater than the ratio of the statistical weights of the upper n = 2 levels, 0.5. This difference is mainly due to the effects of collisional excitation of fine-structure sub-levels. For the observations presented here, electron densities were in an extended range from 3x10{sup 19} to 4x10{sup 20} m{sup -3} with electron and ion temperatures between 1 and 4 keV. Experimental results are compared to calculations from COLRAD, a collisional-radiative modelling code, and good agreement is shown.

  16. Three-dimensional Simulation of Gas Conductance Measurement Experiments on Alcator C-Mod

    International Nuclear Information System (INIS)

    Stotler, D.P.; LaBombard, B.

    2004-01-01

    Three-dimensional Monte Carlo neutral transport simulations of gas flow through the Alcator C-Mod subdivertor yield conductances comparable to those found in dedicated experiments. All are significantly smaller than the conductance found with the previously used axisymmetric geometry. A benchmarking exercise of the code against known conductance values for gas flow through a simple pipe provides a physical basis for interpreting the comparison of the three-dimensional and experimental C-Mod conductances

  17. Analysis of 4-strap ICRF Antenna Performance in Alcator C-Mod

    International Nuclear Information System (INIS)

    Schilling, G.; Wukitch, S.J.; Boivin, R.L.; Goetz, J.A.; Hosea, J.C.; Irby, J.H.; Lin, Y.; Parisot, A.; Porkolab, M.; Wilson, J.R.

    2003-01-01

    A 4-strap ICRF antenna was designed and fabricated for plasma heating and current drive in the Alcator C-Mod tokamak. Initial upgrades were carried out in 2000 and 2001, which eliminated surface arcing between the metallic protection tiles and reduced plasma-wall interactions at the antenna front surface. A boron nitride septum was added at the antenna midplane to intersect electric fields resulting from radio-frequency sheath rectification, which eliminated antenna corner heating at high power levels. The current feeds to the radiating straps were reoriented from an E||B to E parallel B geometry, avoiding the empirically observed ∼15 kV/cm field limit and raising antenna voltage holding capability. Further modifications were carried out in 2002 and 2003. These included changes to the antenna current strap, the boron nitride tile mounting geometry, and shielding the BN-metal interface from the plasma. The antenna heating efficiency, power, and voltage characteristics under these various configurations will be presented

  18. The multi-spectral line-polarization MSE system on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Mumgaard, R. T., E-mail: mumgaard@psfc.mit.edu; Khoury, M. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Scott, S. D. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States)

    2016-11-15

    A multi-spectral line-polarization motional Stark effect (MSE-MSLP) diagnostic has been developed for the Alcator C-Mod tokamak wherein the Stokes vector is measured in multiple wavelength bands simultaneously on the same sightline to enable better polarized background subtraction. A ten-sightline, four wavelength MSE-MSLP detector system was designed, constructed, and qualified. This system consists of a high-throughput polychromator for each sightline designed to provide large étendue and precise spectral filtering in a cost-effective manner. Each polychromator utilizes four narrow bandpass interference filters and four custom large diameter avalanche photodiode detectors. Two filters collect light to the red and blue of the MSE emission spectrum while the remaining two filters collect the beam pi and sigma emission generated at the same viewing volume. The filter wavelengths are temperature tuned using custom ovens in an automated manner. All system functions are remote controllable and the system can be easily retrofitted to existing single-wavelength line-polarization MSE systems.

  19. The multi-spectral line-polarization MSE system on Alcator C-Mod

    International Nuclear Information System (INIS)

    Mumgaard, R. T.; Khoury, M.; Scott, S. D.

    2016-01-01

    A multi-spectral line-polarization motional Stark effect (MSE-MSLP) diagnostic has been developed for the Alcator C-Mod tokamak wherein the Stokes vector is measured in multiple wavelength bands simultaneously on the same sightline to enable better polarized background subtraction. A ten-sightline, four wavelength MSE-MSLP detector system was designed, constructed, and qualified. This system consists of a high-throughput polychromator for each sightline designed to provide large étendue and precise spectral filtering in a cost-effective manner. Each polychromator utilizes four narrow bandpass interference filters and four custom large diameter avalanche photodiode detectors. Two filters collect light to the red and blue of the MSE emission spectrum while the remaining two filters collect the beam pi and sigma emission generated at the same viewing volume. The filter wavelengths are temperature tuned using custom ovens in an automated manner. All system functions are remote controllable and the system can be easily retrofitted to existing single-wavelength line-polarization MSE systems.

  20. Assessment of ICRF Antenna Performance in Alcator C-Mod

    International Nuclear Information System (INIS)

    Schilling, G.; Wukitch, S.J.; Lin, Y.; Basse, N.; Bonoli, P.T.; Edlund, E.; Lin, L.; Parisot, A.; Porkolab, M.

    2004-01-01

    The Alcator C-Mod has presented a challenge to install high-power ICRF antennas in a tight space. Modifications have been made to the antenna plasma-facing surfaces and the internal current-carrying structure in order to overcome performance limitations. At the present time, the antennas have exceeded 5 MW into plasma with heating phasing, up to 2.7 MW with current-drive phasing, with good efficiency and no deleterious effects

  1. The alcator C-MOD control system

    International Nuclear Information System (INIS)

    Bosco, J.; Fairfax, S.

    1992-01-01

    The Alcator C-MOD experiment includes over 30 engineering and diagnostic subsystems. The control system hardware and software is a mixture of custom and commercial products which includes sensors, signal conditioners, hard-wired controls, programmable logic controllers, displays, a hybrid analog/digital computer, networked personal computers, and networked VAX workstations. This paper describes the computer-based portions of the control system. The control system coordinates all C-MOD systems including power, vacuum, heating and cooling, access control, plasma shape and position control, and diagnostics. Programmable logic controllers (PLC's) are located near each subsystem. The control room is isolated by fiber optics. Functions that are essential to personnel or equipment safety (e.g. access control) are implemented in hardwired logic and monitored but not controlled by the PLC's. The initial configuration will include over 25 Allen-Bradley PLC-5 units. The PLCs in each subsystem are connected to personal computers (PC's) in the control room. The PC's provide graphical displays and operator interface. The Pc's are networked and share process data with each other and with a master control console and a large mimic panel

  2. Effect of reflection on Hα emissions in Alcator C-MOD

    International Nuclear Information System (INIS)

    Karney, C.F.; Stotler, D.P.; Skinner, C.H.; Terry, J.L.; Pappas, D.A.

    1999-01-01

    In order to explain anomalous intensity ratios which have been observed in Alcator C-MOD, the H α emissions in that experiment have been modeled with the DEGAS 2 code including the effects of wall reflection. By assuming that the first wall has different reflection coefficients for the two polarizations, we have qualitatively reproduced the observed anomaly. copyright 1999 American Institute of Physics

  3. Electron critical gradient scale length measurements of ICRF heated L-mode plasmas at Alcator C-Mod tokamak

    Science.gov (United States)

    Houshmandyar, S.; Hatch, D. R.; Horton, C. W.; Liao, K. T.; Phillips, P. E.; Rowan, W. L.; Zhao, B.; Cao, N. M.; Ernst, D. R.; Greenwald, M.; Howard, N. T.; Hubbard, A. E.; Hughes, J. W.; Rice, J. E.

    2018-04-01

    A profile for the critical gradient scale length (Lc) has been measured in L-mode discharges at the Alcator C-Mod tokamak, where electrons were heated by an ion cyclotron range of frequency through minority heating with the intention of simultaneously varying the heat flux and changing the local gradient. The electron temperature gradient scale length (LTe-1 = |∇Te|/Te) profile was measured via the BT-jog technique [Houshmandyar et al., Rev. Sci. Instrum. 87, 11E101 (2016)] and it was compared with electron heat flux from power balance (TRANSP) analysis. The Te profiles were found to be very stiff and already above the critical values, however, the stiffness was found to be reduced near the q = 3/2 surface. The measured Lc profile is in agreement with electron temperature gradient (ETG) models which predict the dependence of Lc-1 on local Zeff, Te/Ti, and the ratio of the magnetic shear to the safety factor. The results from linear Gene gyrokinetic simulations suggest ETG to be the dominant mode of turbulence in the electron scale (k⊥ρs > 1), and ion temperature gradient/trapped electron mode modes in the ion scale (k⊥ρs < 1). The measured Lc profile is in agreement with the profile of ETG critical gradients deduced from Gene simulations.

  4. Blob sizes and velocities in the Alcator C-Mod scrape-off layer

    DEFF Research Database (Denmark)

    Kube, R.; Garcia, O.E.; LaBombard, B.

    A new blob-tracking algorithm for the GPI diagnostic installed in the outboard-midplane of Alcator C-Mod is developed. I t tracks large-amplitude fluctuations propagating through the scrape-off layer and calculates blob sizes and velocities. We compare the results of this method to a blob velocity...

  5. Highlights of the Alcator C-Mod Research Campaign

    Science.gov (United States)

    Greenwald, Martin; Alcator Team

    2011-10-01

    Alcator C-Mod has completed an experimental campaign focusing on broad scientific issues with particular emphasis on ITER needs and requests. Experiments with no NBI torque have investigated spontaneous flow reversal, creation of transport barriers aided by the shear of intrinsic rotation and a variety of RF flow drive schemes. Studies of I-mode have found conditions where a wide operating regime opens up, allowing easy access to long-lived, high-performance discharges with L-mode like particle confinement. We are validating the EPED and BOUT++ models for pedestal height/width and ELM onset using extended parameter scans in ELMy H-mode. The challenge of high-Z impurity generation with ICRF is being addressed first by deployment of a novel antenna whose current straps and antenna box are perpendicular to the total magnetic field -second by studies of the modification of edge impurity transport, where fine-scale Er structures in the SOL in the presence of ICRF heating have been found. LH current drive has produced non-inductive reversed shear regimes at n ~ 5x1019 which exhibit electron temperature ITBs. The first observations have been made of in-tokamak production of divertor tungsten nano-structures (fuzz), which had previously been seen only in linear laboratory experiments. Supported by DoE DE-FC02-99ER54512.

  6. Upgrades to the 4-strap ICRF Antenna in Alcator C-Mod

    International Nuclear Information System (INIS)

    G. Schilling; J.C. Hosea; J.R. Wilson; W. Beck; R.L. Boivin; P.T. Bonoli; D. Gwinn; W.D. Lee; E. Nelson-Melby; M. Porkolab; R. Vieira; S.J. Wukitch; J.A. Goetz

    2001-01-01

    A 4-strap ICRF antenna suitable for plasma heating and current drive has been designed and fabricated for the Alcator C-Mod tokamak. Initial operation in plasma was limited by high metallic impurity injection resulting from front surface arcing between protection tiles and from current straps to Faraday shields. Antenna modifications were made in February 2000, resulting in impurity reduction, but low-heating efficiency was observed when the antenna was operated in its 4-strap rather than a 2-strap configuration. Further modifications were made in July 2000, with the installation of BN plasma-facing tiles and radio- frequency bypassing of the antenna backplane edges and ends to reduce potential leakage coupling to plasma surface modes. Good heating efficiency was now observed in both heating configurations, but coupled power was limited to 2.5 MW in H-mode, 3 MW in L-mode, by plasma-wall interactions. Additional modifications were started in February 2001 and will be completed by this meeting. All the above upgrades and their effect on antenna performance will be presented

  7. Nonaxisymmetric field effects on Alcator C-Mod

    International Nuclear Information System (INIS)

    Wolfe, S.M.; Hutchinson, I.H.; Granetz, R.S.; Rice, J.; Hubbard, A.; Lynn, A.; Phillips, P.; Hender, T.C.; Howell, D.F.; La Haye, R.J.; Scoville, J.T.

    2005-01-01

    A set of external coils (A-coils) capable of producing nonaxisymmetric, predominantly n=1, fields with different toroidal phase and a range of poloidal mode m spectra has been used to determine the threshold amplitude for mode locking over a range of plasma parameters in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda, P. Bonoli, S. Fairfax, C. Fiore, J. Goetz, S. Golovato, R. Granetz, M. Greenwald et al., Phys. Plasmas 1, 1511 (1994)]. The threshold perturbations and parametric scalings, expressed in terms of (B 21 /B T ), are similar to those observed on larger, lower field devices. The threshold is roughly linear in density, with typical magnitudes of order 10 -4 . This result implies that locked modes should not be significantly more problematic for the International Thermonuclear Experimental Reactor [I. P. B. Editors, Nucl. Fusion 39, 2286 (1999)] than for existing devices. Coordinated nondimensional identity experiments on the Joint European Torus [Fusion Technol. 11, 13 (1987)], DIII-D [Fusion Technol. 8, 441 (1985)], and C-Mod, with matching applied mode spectra, have been carried out to determine more definitively the field and size scalings. Locked modes on C-Mod are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement, frequently leading to disruptions. Intrinsic error fields inferred from the threshold studies are found to be consistent in amplitude and phase with a comprehensive model of the sources of field errors based on 'as-built' coil and bus-work details and coil imperfections inferred from measurements using in situ magnetic diagnostics on dedicated test pulses. Use of the A-coils to largely cancel the 2/1 component of the intrinsic nonaxisymmetric field has led to expansion of the accessible operating space in C-Mod, including operation up to 2 MA plasma current at 8 T

  8. Full-wave and Fokker Planck analysis of ICRF heating experiments in the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Bonoli, P.T.; Golovato, S.; Porkolab, M.; Takase, Y.

    1996-01-01

    The Alcator C-Mod device is a high field, high density, shaped tokamak with parameters a = 0.22 m, R 0 = 0.67 m, B 0 ≤ 9.0 T, κ ≤ 1.8, δ ≤ 0.8, and 1.0 x 10 20 m -3 n e (0) ≤ 1.0 x 10 21 m -3 . Four megawatt of ICRF power is available at 80 MHz. The wide operating range in magnetic field makes several heating schemes possible: (i) Second harmonic heating of hydrogen (f 0 = 2f CH ) at 2.6 T in (D-H); (ii) Fundamental heating of (H) (f 0 = f CH ) at 5.3T in a D-(H) plasma; and (iii) Fundamental heating of ( 3 He) (f 0 = f C 3 He ) at 7.9 T in a D-( 3 He) plasma. The most successful heating regime to date has been (H)-minority heating at 5.3 T. Pellet enhanced performance (PEP) modes have also been achieved in C-Mod in D-(H) at 5.3 T and in D-( 3 He) at 7.9 T, with a combination of intense ICRF heating and Li-pellet injection. A variety of numerical models are used to analyze these heating schemes. A 1-D full-wave code (FELICE) is used to study open-quotes single passclose quotes damping of the ICRF wavefront and damping of mode-converted ion Bernstein waves. A toroidal full-wave code (FISIC) is used to study interference and focussing effects of the ICRF waves as well as damping of the ICRF power upon multiple passes of the ICRF wavefront. A combined bounce averaged Fokker Planck and toroidal full-wave code (FPPRF) is used to study the ion tail formation, orbit losses, and the power partition of the ICRF tail to the background electrons and ions. Full-wave and Fokker Planck analyses confirm the strong single pass absorption of the ICRF power in D-(H) at 5.3 T. Analysis of PEP-mode plasmas in D-( 3 He) indicates improved wave focussing and 3 He-cyclotron absorption of the ICRF waves relative to L-mode. A dramatic increase in the transfer of 3 He tail power to the background deuterium is also found for PEP-mode plasmas

  9. High performance discharges and capabilities in Alcator C-Mod

    International Nuclear Information System (INIS)

    Porkolab, M.

    1996-01-01

    Alcator C-Mod is a compact, diverted, shaped, high magnetic field (B = 9 T) tokamak operating at the Massachusetts Institute of Technology Plasma Fusion Center. The machine interior is all metallic, and the walls and divertor region are covered with molybdenum tiles. The vacuum vessel is a continuous, thick wall stainless steel construction, prototypical of future fusion devices (e.g., ITER). Typical discharge cleaning utilizes ECDC, or electron-cyclotron discharge cleaning, in the steady state at low magnetic field (0.0875 T). While its dimensions are compact (R = 0.67 m, a = 0.22 m, K = 1.8), C-Mod is designed to operate up to 2.5 MA at 9.0 T magnetic field. To present date the machine has operated at currents up to 1.5 MA at B = 5.3 T, and magnetic fields up to 8.0 T at I p = 1.2 MA. Due to the high current density, line average densities of 4.0 x 10 20 m -3 are obtained with gas fueling, and peak densities in excess of 1.0 x 10 21 m -3 have been obtained with pellet fueling. Typical pulse lengths are up to 2.0 seconds, with a flat-top of typically 1.0 sec. Presently the device is equipped with 4.0 MW of ICRF heating power operating at 80 MHz, but this capability is being upgraded to 8.0 MW with the addition of 4.0 MW of tunable ICRF power operating at 40.80 MHz. A 20 pellet/pulse deuterium injector is operational, and a 4 pellet Li injector is also operational. To reduce the influx of metallic impurities during high power operation, recently boronization of the machine interior was begun prior to plasma discharges, this allowed plasma operation with full auxiliary power capability without excessive radiative power losses from the plasma core. 7 refs

  10. Correlation ECE diagnostic in Alcator C-Mod

    International Nuclear Information System (INIS)

    Sung, C.; Irby, J.; Leccacorvi, R.; Vieira, R.; Oi, C.; Rice, J.; Reinke, M.; Gao, C.; Ennever, P.; Porkolab, M.; Churchill, R.; Theiler, C.; Walk, J.; Hughes, J.; Hubbard, A.; Greenwald, M.

    2015-01-01

    Correlation ECE (CECE) is a diagnostic technique that allows measurement of small amplitude electron temperature, Te, fluctuations through standard cross-correlation analysis methods. In Alcator C-Mod, a new CECE diagnostic has been installed[Sung RSI 2012], and interesting phenomena have been observed in various plasma conditions. We find that local Te fluctuations near the edge (ρ ~ 0:8) decrease across the linearto- saturated ohmic confinement transition, with fluctuations decreasing with increasing plasma density [Sung NF 2013], which occurs simultaneously with rotation reversals [Rice NF 2011]. Te fluctuations are also reduced across core rotation reversals with an increase of plasma density in RF heated L-mode plasmas, which implies that the same physics related to the reduction of Te fluctuations may be applied to both ohmic and RF heated L-mode plasmas. In I-mode plasmas, we observe the reduction of core Te fluctuations, which indicates changes of turbulence occur not only in the pedestal region but also in the core across the L/I transition [White NF 2014]. The present CECE diagnostic system in C-Mod and these experimental results are described in this paper

  11. Overview of recent Alcator C-Mod research

    International Nuclear Information System (INIS)

    Marmar, E.S.; Bai, B.; Boivin, R.L.

    2003-01-01

    Research on the Alcator C-Mod tokamak is focused on high particle- and power-density plasma regimes to understand particle and energy transport in the core, the dynamics of the H-mode pedestal, and scrape-off layer and divertor physics. The auxiliary heating is provided exclusively by RF waves, and both the physics and technology of RF heating and current drive are studied. The momentum which is manifested in strong toroidal rotation, in the absence of direct momentum input, has been shown to be transported in from the edge of the plasma following the L to H transition, with time scale comparable to that for energy transport. In discharges which develop internal transport barriers (ITBs), the rotation slows first inside the barrier region, and then subsequently outside of the barrier foot. Heat pulse propagation studies using sawteeth indicate a very narrow region of strongly reduced energy transport, located near r/a = 0.5. Addition of on-axis ICRF heating arrests the buildup of density and impurities, leading to quasi-steady conditions. The quasi-coherent mode associated with EDA H-mode appears to be due to a resistive ballooning instability. As the pedestal pressure gradient and temperature are increased in EDA H-mode, small ELMs appear; detailed modeling indicates that these are due to intermediate n peeling-ballooning modes. Phase Contrast Imaging (PCI) has been used to directly detect density fluctuations driven by ICRF waves in the core of the plasma, and mode conversion to an intermediate wavelength Ion Cyclotron Wave has been observed for the first time. The bursty turbulent density fluctuations, observed to drive rapid cross-field particle transport in the edge plasma, appear to play a key role the dynamics of the density limit. Preparations for quasi-steady-state Advanced Tokamak studies with lower hybrid current drive are well underway, and time dependent modeling indicates that regimes with high bootstrap fraction can be produced. (author)

  12. VUV study of impurity generation during ICRF heating experiments on the Alcator C tokamak

    International Nuclear Information System (INIS)

    Manning, H.L.

    1986-06-01

    A 2.2 meter grazing incidence VUV monochromator has been converted into a time-resolving spectrograph by the addition of a new detector system, based on a microchannel plate image intensifier linked to a 1024-element linear photodiode array. The system covers the wavelength range 15 to 1200 A (typically 40 A at a time) with resolution of up to .3 A FWHM. Time resolution is selectable down to 0.5 msec. The system sensitivity was absolutely calibrated below 150 A by a soft x-ray calibration facility. The spectrograph was installed on the Alcator C tokamak at MIT to monitor plasma impurity emission. There, cross-calibration with a calibrated EUV monochromator was performed above 400 A. Calibration results, system performance characteristics, and data from Alcator C are presented. Observations of impurity behavior are presented from a series of ICRF heating experiments (180 MHz, 50 to 400 kW) performed on the Alcator C tokamak, using graphite limiters and stainless steel antenna Faraday shields

  13. Investigation of RF-enhanced plasma potentials on Alcator C-Mod

    International Nuclear Information System (INIS)

    Ochoukov, R.; Whyte, D.G.; Brunner, D.; Cziegler, I.; LaBombard, B.; Lipschultz, B.; Myra, J.; Terry, J.; Wukitch, S.

    2013-01-01

    Radio frequency (RF) sheath rectification is a leading mechanism suspected of causing anomalously high erosion of plasma facing materials in RF-heated plasmas on Alcator C-Mod. An extensive experimental survey of the plasma potential (Φ P ) in RF-heated discharges on C-Mod reveals that significant Φ P enhancement (>100 V) is found on outboard limiter surfaces, both mapped and not mapped to active RF antennas. Surfaces that magnetically map to active RF antennas show Φ P enhancement that is, in part, consistent with the recently proposed slow wave rectification mechanism. Surfaces that do not map to active RF antennas also experience significant Φ P enhancement, which strongly correlates with the local fast wave intensity. In this case, fast wave rectification is a leading candidate mechanism responsible for the observed enhancement

  14. Overview of the Alcator C-MOD Research Program

    International Nuclear Information System (INIS)

    Scott; S.; Bader, A.; Bakhtiari, M.; Basse, N.; Beck, W.; Biewer, T.; Bernabei, S.; Bonoli, P.

    2007-01-01

    Recent research on the high-field, high-density diverted Alcator C-MOD tokamak has focused on the plasma physics and plasma engineering required for ITER and for attractive fusion reactors. Experimental campaigns over the past two years have focused on understanding the physical mechanisms that affect the plasma performance realized with all-molybdenum walls versus walls with low-Z coatings. RF sheath rectification along flux tubes that intersect the RF antenna is found to be a major cause of localized boron erosion and impurity generation. Initial lower-hybrid current drive (LHCD) experiments (PLH p ∼ 1.0 MA with good efficiency, I drive = 0.4P LH /n eo R (MA,MW, 10 20 m -3 ,m). Disruption mitigation via massive gas-jet impurity puffing has proven successful at high plasma pressure, indicating this technique has promise for implementation on ITER. Pressure gradients in the near SOL of Ohmic L-mode plasmas are observed to scale consistently as I p 2 , and show a significant dependence on X-point topology. Modeling of H-mode edge fueling indicates high self-screening to neutrals in the pedestal and scrape-off layer (SOL), and reproduces experimental density pedestal response to changes in neutral source. Detailed measurements of the temperature and density profiles in the near sol and fast framing movies of the turbulent structures provide improved understanding of the mechanisms that control transport in the edge region.

  15. Migration of alcator C-Mod computer infrastructure to Linux

    International Nuclear Information System (INIS)

    Fredian, T.W.; Greenwald, M.; Stillerman, J.A.

    2004-01-01

    The Alcator C-Mod fusion experiment at MIT in Cambridge, Massachusetts has been operating for twelve years. The data handling for the experiment during most of this period was based on MDSplus running on a cluster of VAX and Alpha computers using the OpenVMS operating system. While the OpenVMS operating system provided a stable reliable platform, the support of the operating system and the software layered on the system has deteriorated in recent years. With the advent of extremely powerful low cost personal computers and the increasing popularity and robustness of the Linux operating system a decision was made to migrate the data handling systems for C-Mod to a collection of PC's running Linux. This paper will describe the new system configuration, the effort involved in the migration from OpenVMS, the results of the first run campaign under the new configuration and the impact the switch may have on the rest of the MDSplus community

  16. Investigation of RF-enhanced plasma potentials on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Ochoukov, R., E-mail: ochoukov@psfc.mit.edu [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); Whyte, D.G.; Brunner, D. [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); Cziegler, I. [Center for Energy Research, UCSD, 9500 Gilman Drive, La Jolla, CA 92093 (United States); LaBombard, B.; Lipschultz, B. [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); Myra, J. [Lodestar Research Corporation, 2400 Central Avenue P-5, Boulder, CO 80301 (United States); Terry, J.; Wukitch, S. [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States)

    2013-07-15

    Radio frequency (RF) sheath rectification is a leading mechanism suspected of causing anomalously high erosion of plasma facing materials in RF-heated plasmas on Alcator C-Mod. An extensive experimental survey of the plasma potential (Φ{sub P}) in RF-heated discharges on C-Mod reveals that significant Φ{sub P} enhancement (>100 V) is found on outboard limiter surfaces, both mapped and not mapped to active RF antennas. Surfaces that magnetically map to active RF antennas show Φ{sub P} enhancement that is, in part, consistent with the recently proposed slow wave rectification mechanism. Surfaces that do not map to active RF antennas also experience significant Φ{sub P} enhancement, which strongly correlates with the local fast wave intensity. In this case, fast wave rectification is a leading candidate mechanism responsible for the observed enhancement.

  17. An assessment of ion temperature measurements in the boundary of the Alcator C-Mod tokamak and implications for ion fluid heat flux limiters

    International Nuclear Information System (INIS)

    Brunner, D; LaBombard, B; Churchill, R M; Hughes, J; Lipschultz, B; Ochoukov, R; Theiler, C; Walk, J; Rognlien, T D; Umansky, M V; Whyte, D

    2013-01-01

    The ion temperature is not frequently measured in the boundary of magnetic fusion devices. Comparisons among different ion temperature techniques and simulations are even rarer. Here we present a comparison of ion temperature measurements in the boundary of the Alcator C-Mod tokamak from three different diagnostics: charge exchange recombination spectroscopy (CXRS), an ion sensitive probe (ISP), and a retarding field analyzer (RFA). Comparison between CXRS and the ISP along with close examination of the ISP measurements reveals that the ISP is space charge limited. It is thus unable to measure ion temperature in the high density (>10 19 m −3 ) boundary plasma of C-Mod with its present geometry. Comparison of ion temperatures measured by CXRS and the RFA shows fair agreement. Ion and electron parallel heat flow is analyzed with a simple 1D fluid code. The code takes divertor measurements as input and results are compared to the measured ratios of upstream ion to electron temperature, as inferred respectively by CXRS and a Langmuir probe. The analysis reveals the limits of the fluid model at high Knudsen number. The upstream temperature ratio is under predicted by a factor of 2. Heat flux limiters (kinetic corrections) to the fluid model are necessary to match experimental data. The values required are found to be close to those reported in kinetic simulations. The 1D code is benchmarked against the 2D plasma fluid code UEDGE with good agreement. (paper)

  18. An assessment of ion temperature measurements in the boundary of the Alcator C-Mod tokamak and implications for ion fluid heat flux limiters

    Science.gov (United States)

    Brunner, D.; LaBombard, B.; Churchill, R. M.; Hughes, J.; Lipschultz, B.; Ochoukov, R.; Rognlien, T. D.; Theiler, C.; Walk, J.; Umansky, M. V.; Whyte, D.

    2013-09-01

    The ion temperature is not frequently measured in the boundary of magnetic fusion devices. Comparisons among different ion temperature techniques and simulations are even rarer. Here we present a comparison of ion temperature measurements in the boundary of the Alcator C-Mod tokamak from three different diagnostics: charge exchange recombination spectroscopy (CXRS), an ion sensitive probe (ISP), and a retarding field analyzer (RFA). Comparison between CXRS and the ISP along with close examination of the ISP measurements reveals that the ISP is space charge limited. It is thus unable to measure ion temperature in the high density (>1019 m-3) boundary plasma of C-Mod with its present geometry. Comparison of ion temperatures measured by CXRS and the RFA shows fair agreement. Ion and electron parallel heat flow is analyzed with a simple 1D fluid code. The code takes divertor measurements as input and results are compared to the measured ratios of upstream ion to electron temperature, as inferred respectively by CXRS and a Langmuir probe. The analysis reveals the limits of the fluid model at high Knudsen number. The upstream temperature ratio is under predicted by a factor of 2. Heat flux limiters (kinetic corrections) to the fluid model are necessary to match experimental data. The values required are found to be close to those reported in kinetic simulations. The 1D code is benchmarked against the 2D plasma fluid code UEDGE with good agreement.

  19. Edge Turbulence Imaging on NSTX and Alcator C-Mod

    International Nuclear Information System (INIS)

    S.J. Zweben; R.A. Maqueda; J.L. Terry; B. Bai; C.J. Boswell; C.E. Bush; D. D'Ippolito; E.D. Fredrickson; M. Greenwald; K. Hallatschek; S. Kaye; B. LaBombard; R. Maingi; J. Myra; W.M. Nevins; B.N. Rogers; D.P. Stotler; J. Wilgen; and X.Q. Xu

    2002-01-01

    Edge turbulence images have been made using an ultra-high speed CCD camera on both NSTX and Alcator C-Mod. In both cases, the D-alpha or HeI (587.6 nm) line emission from localized deuterium or helium gas puffs was viewed along a local magnetic field line near the outer midplane. Fluctuations in this line emission reflect fluctuations in electron density and/or electron temperature through the atomic excitation rates, which can be modeled using the DEGAS-2 code. The 2-D structure of the measured turbulence can be compared with theoretical simulations based on 3-D fluid models

  20. Overview of Alcator C-Mod Research

    Science.gov (United States)

    White, A. E.

    2017-10-01

    Alcator C-Mod, a compact (R =0.68m, a =0.21m), high magnetic field, Bt Research spans the topics of core transport and turbulence, RF heating and current drive, pedestal physics, scrape-off layer, divertor and plasma wall interactions. In the last experimental campaign, Super H-mode was explored and featured the highest pedestal pressures ever recorded, pped 90 kPa (90% of ITER target), consistent with EPED predictions. Optimization of naturally ELM-suppressed EDA H-modes accessed the highest volume averaged pressures ever achieved (〈p〉>2 atm), with pped 60 kPa. The SOL heat flux width has been measured at Bpol = 1.25T, confirming the Eich scaling over a broader poloidal field range than before. Multi-channel transport studies focus on the relationship between momentum transport and heat transport with perturbative experiments and new multi-scale gyrokinetic simulation validation techniques were developed. U.S. Department of Energy Grant No. DE-FC02-99ER54512.

  1. Direct detection of lower hybrid wave using a reflectometer on Alcator C-Moda)

    Science.gov (United States)

    Shiraiwa, S.; Baek, S.; Dominguez, A.; Marmar, E.; Parker, R.; Kramer, G. J.

    2010-10-01

    The possibility of directly detecting a density perturbation produced by lower hybrid (LH) waves using a reflectometer is presented. We investigate the microwave scattering of reflectometer probe beams by a model density fluctuation produced by short wavelength LH waves in an Alcator C-Mod experimental condition. In the O-mode case, the maximum response of phase measurement is found to occur when the density perturbation is approximately centimeters in front of the antenna, where Bragg scattering condition is satisfied. In the X-mode case, the phase measurement is predicted to be more sensitive to the density fluctuation close to the cut-off layer. A feasibility test was carried out using a 50 GHz O-mode reflectometer on the Alcator C-Mod tokamak, and positive results including the detection of 4.6 GHz pump wave and parametric decay instabilities were obtained.

  2. Ion and electron parameters in the alcator C tokamak scrape-off region

    International Nuclear Information System (INIS)

    Wan, A.S.H.

    1986-05-01

    Janus is a bi-directional, multi-functional edge probe used to diagnose the ion and electron parameters in the Alcator C tokamak scrape-off region. Two mirror image sets of diagnostics are aligned to face the electron and ion sides along magnetic field lines. Each set of diagnostics consists of a retarding-field energy analyzer (RFEA), a Langmuir probe, and a calorimeter. The RFEA can alternatively sample both the ion and electron parallel energy distribution functions during a tokamak discharge. From the Langmuir probe, one can infer electron temperature, density, and the plasma floating potential. Simple Langmuir probe theory is found to yield the best agreement between the measured Langmuir probe characteristics and the RFEA-inferred T/sub e/. The calorimeter independently detects the total parallel heat flux incident to an electrically floating plate. The measured sheath transmission coefficient, however, is typically lower than the theoretically predicted value by a factor of approx.3. Together these diagnostics enable detailed, localized edge plasma characterization on Alcator C

  3. Fluctuating zonal flows in the I-mode regime in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Cziegler, I.; Diamond, P. H.; Fedorczak, N.; Manz, P.; Tynan, G. R.; Xu, M. [Center for Momentum Transport and Flow Organization, University of California, San Diego, La Jolla, California 92093 (United States); Churchill, R. M.; Hubbard, A. E.; Lipschultz, B.; Sierchio, J. M.; Terry, J. L.; Theiler, C. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02138 (United States)

    2013-05-15

    Velocity fields and density fluctuations of edge turbulence are studied in I-mode [F. Ryter et al., Plasma Phys. Controlled Fusion 40, 725 (1998)] plasmas of the Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] tokamak, which are characterized by a strong thermal transport barrier in the edge while providing little or no barrier to the transport of both bulk and impurity particles. Although previous work showed no clear geodesic-acoustic modes (GAM) on C-Mod, using a newly implemented, gas-puff-imaging based time-delay-estimate velocity inference algorithm, GAM are now shown to be ubiquitous in all I-mode discharges examined to date, with the time histories of the GAM and the I-mode specific [D. Whyte et al., Nucl. Fusion 50, 105005 (2010)] Weakly Coherent Mode (WCM, f = 100–300 kHz, Δf/f≈0.5, and k{sub θ}≈1.3 cm{sup −1}) closely following each other through the entire duration of the regime. Thus, the I-mode presents an example of a plasma state in which zero frequency zonal flows and GAM continuously coexist. Using two-field (density-velocity and radial-poloidal velocity) bispectral methods, the GAM are shown to be coupled to the WCM and to be responsible for its broad frequency structure. The effective nonlinear growth rate of the GAM is estimated, and its comparison to the collisional damping rate seems to suggest a new view on I-mode threshold physics.

  4. Experimental and computational evaluation of neutrals in the Alcator C-Mod edge pedestal

    Science.gov (United States)

    Hughes, J. W.; Mossessian, D.; Labombard, B.; Terry, J.

    2004-11-01

    Pedestal-forming edge transport barriers (ETBs) in tokamak plasmas and the physics governing them are linked to the enhancement of confinement obtained in H-mode plasmas. Studies on Alcator C-Mod employ experimental measurements and simple 1-D transport models in order to better understand ETB physics. We examine the influences of ionization and charge exchange on the pedestals in electron density and temperature. Routine measurements from edge Thomson scattering (ETS) give pedestal scalings with global plasma parameters, while individual ETS profiles are combined with scanning Langmuir probe data and optical D_α emissivity measurements to give atomic density profiles and the associated radial distribution of the ionization source rate. From H-mode profiles of these quantities a well in effective plasma diffusivity is calculated, and is shown to systematically vary as the confinement regime is varied from ELM-free to EDA. Experimental work is supplemented with modeling and computation of edge neutral transport via KN1D, a kinetic solver for atomic and molecular distribution functions in slab geometry. The level of agreement between experiment and model is encouraging.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  6. Rotation and transport in Alcator C-Mod ITB plasmas

    Science.gov (United States)

    Fiore, C. L.; Rice, J. E.; Podpaly, Y.; Bespamyatnov, I. O.; Rowan, W. L.; Hughes, J. W.; Reinke, M.

    2010-06-01

    Internal transport barriers (ITBs) are seen under a number of conditions in Alcator C-Mod plasmas. Most typically, radio frequency power in the ion cyclotron range of frequencies (ICRFs) is injected with the second harmonic of the resonant frequency for minority hydrogen ions positioned off-axis at r/a > 0.5 to initiate the ITBs. They can also arise spontaneously in ohmic H-mode plasmas. These ITBs typically persist tens of energy confinement times until the plasma terminates in radiative collapse or a disruption occurs. All C-Mod core barriers exhibit strongly peaked density and pressure profiles, static or peaking temperature profiles, peaking impurity density profiles and thermal transport coefficients that approach neoclassical values in the core. The strongly co-current intrinsic central plasma rotation that is observed following the H-mode transition has a profile that is peaked in the centre of the plasma and decreases towards the edge if the ICRF power deposition is in the plasma centre. When the ICRF resonance is placed off-axis, the rotation develops a well in the core region. The central rotation continues to decrease as long as the central density peaks when an ITB develops. This rotation profile is flat in the centre (0 ITB density profile is observed (0.5 ITB foot that is sufficiently large to stabilize ion temperature gradient instabilities that dominate transport in C-Mod high density plasmas.

  7. 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)

  8. Lower Hybrid Current Drive Experiments in Alcator C-Mod

    International Nuclear Information System (INIS)

    Wilson, J. R.; Bonoli, P.; Hubbard, A.; Parker, R.; Schmidt, A.; Wallace, G.; Wright, J.; Bernabei, S.

    2007-01-01

    A Lower Hybrid Current Drive (LHCD) system has been installed on the Alcator C-MOD tokamak at MIT. Twelve klystrons at 4.6 GHz feed a 4x22 waveguide array. This system was designed for maximum flexibility in the launched parallel wave-number spectrum. This flexibility allows tailoring of the lower hybrid deposition under a variety of plasma conditions. Power levels up to 900 kW have been injected into the tokomak. The parallel wave number has been varied over a wide range, n parallel ∼1.6-4. Driven currents have been inferred from magnetic measurements by extrapolating to zero loop voltage and by direct comparison to Fisch-Karney theory, yielding an efficiency of n 20 IR/P∼0.3. Modeling using the CQL3D code supports these efficiencies. Sawtooth oscillations vanish, accompanied with peaking of the electron temperature (T e0 rises from 2.8 to 3.8 keV). Central q is inferred to rise above unity from the collapse of the sawtooth inversion radius, indicating off-axis cd as expected. Measurements of non-thermal x-ray and electron cyclotron emission confirm the presence of a significant fast electron population that varies with phase and plasma density. The x-ray emission is observed to be radialy broader than that predicted by simple ray tracing codes. Possible explanations for this broader emission include fast electron diffusion or broader deposition than simple ray tracing predictions (perhaps due to diffractive effects)

  9. Gas jet disruption mitigation studies on Alcator C-Mod and DIII-D

    International Nuclear Information System (INIS)

    Granetz, R.S.; Hollmann, E.M.; Whyte, D.G.; Izzo, V.A.; Antar, G.Y.; Bader, A.; Bakhtiari, M.; Biewer, T.; Boedo, J.A.; Evans, T.E.; Hutchinson, I.H.; Jernigan, T.C.; Gray, D.S.; Groth, M.; Humphreys, D.A.; Lasnier, C.J.; Moyer, R.A.; Parks, P.B.; Reinke, M.L.; Rudakov, D.L.; Strait, E.J.; Terry, J.L.; Wesley, J.; West, W.P.; Wurden, G.; Yu, J.

    2007-01-01

    High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects. In both machines, the sequence of events following gas injection is observed to be quite similar: the jet neutrals stop near the plasma edge, the edge temperature collapses and large MHD modes are quickly destabilized, mixing the hot plasma core with the edge impurity ions and radiating away the plasma thermal energy. High radiated power fractions are achieved, thus reducing the conducted heat loads to the chamber walls and divertor. A significant (2 x or more) reduction in halo current is also observed. Runaway electron generation is small or absent. These similar results in two quite different tokamaks are encouraging for the applicability of this disruption mitigation technique to ITER

  10. Microturbulent drift mode suppression as a trigger mechanism for internal transport barriers on Alcator C-Mod

    International Nuclear Information System (INIS)

    Zhurovich, K.; Fiore, C.L.; Ernst, D.R.; Bonoli, P.T.; Greenwald, M.J.; Hubbard, A.E.; Hughes, J.W.; Marmar, E.S.; Mikkelsen, D.R.; Phillips, P.; Rice, J.E.

    2007-01-01

    Internal transport barriers (ITBs) can be routinely produced in enhanced D α (EDA) H-mode discharges on the Alcator C-Mod tokamak by putting the minority ion cyclotron resonance layer at vertical bar r/a vertical bar ≥ 0.5 during the current flat top phase of the discharge. These ITBs are characterized by density peaking at constant temperature and are therefore both particle and energy transport barriers. The ITB formation appears to result from widening the region near the magnetic axis in which toroidal drift modes are stable, allowing the Ware pinch to peak the density profile. Experimental evidence shows that shifting the ICRF resonance off-axis results in a local flattening of ion and electron temperature profiles. TRANSP calculations of ion temperature profiles support this experimentally observed trend. Stability analysis of ion temperature gradient (ITG) and electron temperature gradient modes at times before ITB formation is done using the linear gyrokinetic code GS2. These gyrokinetic calculations find that the most unstable modes in the C-Mod EDA H-mode core, prior to ITB onset, are the toroidal ITG driven type. These modes are suppressed in the ITB region through a temperature gradient reduction when the ICRF resonance is shifted off-axis

  11. Comparison of Scrape-off Layer Turbulence in Alcator C-Mod with Three Dimensional Gyrofluid Computations

    International Nuclear Information System (INIS)

    Zweben, S.J.; Scott, B.D.; Terry, J.L.; LaBombard, B.; Hughes, J.W.; Stotler, D.P.

    2009-01-01

    This paper describes quantitative comparisons between turbulence measured in the scrape-off layer (SOL) of Alcator C-Mod (S. Scott, A. Bader, M. Bakhtiari et al., Nucl. Fusion 47, S598 (2007)) and three dimensional computations using electromagnetic gyrofluid equations in a two-dimensional tokamak geometry. These comparisons were made for the outer midplane SOL for a set of inner-wall limited, near-circular Ohmic plasmas. The B field and plasma density were varied to assess gyroradius and collisionality scaling. The poloidal and radial correlation lengths in the experiment and computation agreed to within a factor of 2 and did not vary significantly with either B or density. The radial and poloidal propagation speeds and the frequency spectra and poloidal k-spectra also agreed fairly well. However, the autocorrelation times and relative Da fluctuation levels were higher in the experiment by more than a factor of 2. Possible causes for these disagreements are discussed.

  12. Design of a New Optical System for Alcator C-Mod Motional Stark Effect Diagnostic

    International Nuclear Information System (INIS)

    Ko, Jinseok; Scott, Steve; Bitter, Manfred; Lerner, Scott

    2009-01-01

    The motional Stark effect (MSE) diagnostic on Alcator C-Mod uses an in-vessel optical system (five lenses and three mirrors) to relay polarized light to an external polarimeter because port access limitations on Alcator C-Mod preclude a direct view of the diagnostic beam. The system experiences unacceptable, spurious drifts of order several degrees in measured pitch angle over the course of a run day. Recent experiments illuminated the MSE diagnostic with polarized light of fixed orientation as heat was applied to various optical elements. A large change in measured angle was observed as two particular lenses were heated, indicating that thermal-stress-induced birefringence is a likely cause of the spurious variability. Several new optical designs have been evaluated to eliminate the affected in-vessel lenses and to replace the focusing they provide with curved mirrors; however, ray tracing calculations imply that this method is not feasible. A new approach is under consideration that utilizes in situ calibrations with in-vessel reference polarized light sources. 2008 American Institute of Physics.

  13. Observations of core modes during RF-generated internal transport barriers in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lynn, A G [Fusion Research Center, University of Texas at Austin, Austin, TX (United States); Phillips, P E [Fusion Research Center, University of Texas at Austin, Austin, TX (United States); Hubbard, A E [Plasma Science and Fusion Center, MIT, Cambridge, MA (United States); Wukitch, S J [Plasma Science and Fusion Center, MIT, Cambridge, MA (United States)

    2004-05-01

    In the Alcator C-Mod tokamak, a high-resolution heterodyne ECE radiometer has been used to measure the electron temperature in plasma discharges with internal transport barriers (ITBs). ITBs are formed by the application of off-axis (r/a {approx} 0.5) ICRF power. Strong density peaking indicates the formation of the ITB. When the ITB is formed, the ECE radiometer detects a small amplitude mode localized near the magnetic axis. Surprisingly, as this mode amplitude grows a dip in the temperature profile is clearly observed at the same location. If sawteeth are present, the mode amplitude appears to be suppressed by the sawtooth crash and no dip in the temperate profiles is observed. TORAY, a ray-tracing code, has been used to investigate the possible refractive effects of the steep density gradients in the ITB and its effects on the ECE observations. The results show that refractive effects can explain the observed local changes in temperature. Ray-tracing also indicates that the observed modes are density fluctuations. Observations of broadband density fluctuations during 4.5 T ITBs are also described.

  14. Real-time sensing and gas jet mitigation of VDEs on Alcator C-Mod

    Science.gov (United States)

    Granetz, R. S.; Wolfe, S. M.; Izzo, V. A.; Reinke, M. L.; Terry, J. L.; Hughes, J. W.; Zhurovich, K.; Whyte, D. G.; Bakhtiari, M.; Wurden, G.

    2006-10-01

    Experiments have been carried out in Alcator C-Mod to test the effectiveness of gas jet disruption mitigation of VDEs with real-time detection and triggering by the C-Mod digital plasma control system (DPCS). The DPCS continuously computes the error in the plasma vertical position from the magnetics diagnostics. When this error exceeds an adjustable preset value, the DPCS triggers the gas jet valve (with a negligible latency time). The high-pressure gas (argon) only takes a few milliseconds to enter the vacuum chamber and begin affecting the plasma, but this is comparable to the VDE timescale on C-Mod. Nevertheless, gas jet injection reduced the halo current, increased the radiated power fraction, and reduced the heating of the divertor compared to unmitigated disruptions, but not quite as well as in earlier mitigation experiments with vertically stable plasmas. Presumably a faster overall response time would be beneficial, and several ways to achieve this will also be discussed.

  15. Grazing incidence EUV study of the Alcator tokamaks

    International Nuclear Information System (INIS)

    Castracane, J.

    1982-01-01

    The use of impurity radiation to examine plasma conditions is a well known technique. To gain access, however, to the hot, central portion of the plasma created in the present confinement machines it is necessary to be able to observe radiation from medium and heavy elements such as molybdenum and iron. These impurities radiate primarily in the extreme ultra violet region of the spectrum and can play a role in the power balance of the tokamak. Radiation from highly ionized molybdenum was examined on the Alcator A and C tokamaks using a photometrically calibrated one meter grazing incidence monochromator. On Alcator A, a pseudo-continuum of Mo emissions in the 60 to 100 A ranges were seen to comprise 17% of the radiative losses from the plasma. This value closely matched measurements by a broad band bolometer array. Following these preliminary measurements, the monochromator was transferred to Alcator C for a more thorough examination of EUV emissions. Deviations from predicted scaling laws for energy confinement time vs density were observed on this machine

  16. Disruption Neutral Point Experiment on Alcator C-Mod

    Science.gov (United States)

    Granetz, R. S.; Nakamura, Y.

    2000-10-01

    Disruptions of single-null elongated plasmas generally result in loss of vertical position control, leading to a current quench occurring at the top or bottom of the machine, with all the attendant problems of halo and eddy currents flowing in divertor structures. On JT-60U, it has been found that if the plasma is operated with its magnetic axis at a particular height, called the neutral point, the initial vertical drift after a thermal quench is significantly slower than usual, and sometimes can even be arrested, thereby avoiding a current quench in the divertor region entirely. In an ongoing collaboration between MIT and JAERI, the neutral point concept is being tested in Alcator C-Mod, which has a significantly higher plasma elongation than JT-60U (1.65 vs 1.3). Calculations using TSC predict a neutral point at z~=+1 cm above the midplane (a=22 cm). The existence of a neutral point has now been experimentally confirmed, albeit at a height of z=+2.7 cm. The plasma has remained vertically stable for up to 9 ms after the disruption thermal quench, which in principle, is long enough for the PF control system to respond, if programmed appropriately. In addition, the physics of the neutral point stability on C-Mod appears to be somewhat different than that on JT-60U.

  17. Electron Temperature Fluctuation Measurements and Transport Model Validation at Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    White, Anne [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2017-06-22

    for studying core turbulence are needed in order to assess the accuracy of gyrokinetic models for turbulent-driven particle, heat and momentum transport. New core turbulence diagnostics at the world-class tokamaks Alcator C-Mod at MIT and ASDEX Upgrade at the Max Planck Institute for Plasma Physics have been designed, developed, and operated over the course of this project. These new instruments are capable of measuring electron temperature fluctuations and the phase angle between density and temperature fluctuations locally and quantitatively. These new data sets from Alcator C-Mod and ASDEX Upgrade are being used to fill key gaps in our understanding of turbulent transport in tokamaks. In particular, this project has results in new results on the topics of the Transport Shortfall, the role of ETG turbulence in tokamak plasmas, profile stiffness, the LOC/SOC transition, and intrinsic rotation reversals. These data are used in a rigorous process of “Transport model validation”, and this group is a world-leader on using turbulence models to design new hardware and new experiments at tokamaks. A correlation electron cyclotron emission (CECE) diagnostic is an instrument used to measure micro-scale fluctuations (mm-scale, compared to the machine size of meters) of electron temperature in magnetically confined fusion plasmas, such as those in tokamaks and stellarators. These micro-scale fluctuations are associated with drift-wave type turbulence, which leads to enhanced cooling and mixing of particles in fusion plasmas and limits achieving the required temperatures and densities for self-sustained fusion reactions. A CECE system can also be coupled with a reflectometer system that measured micro-scale density fluctuations, and from these simultaneous measurements, one can extract the phase between the density (n) and temperature (T) fluctuations, creating an nT phase diagnostic. Measurements of the fluctuations and the phase angle between them are extremely useful for

  18. Ohmic ITBs in Alcator C-Mod

    Science.gov (United States)

    Fiore, C. L.; Rowan, W. L.; Dominguez, A.; Hubbard, A. E.; Ince-Cushman, A.; Greenwald, M. J.; Lin, L.; Marmar, E. S.; Reinke, M.; Rice, J. E.; Zhurovich, K.

    2007-11-01

    Internal transport barrier plasmas can arise spontaneously in ohmic Alcator C-Mod plasmas where an EDA H-mode has been developed by magnetic field ramping. These ohmic ITBs share the hallmarks of ITBs created with off-axis ICRF injection in that they have highly peaked density and pressure profiles and the peaking can be suppressed by on-axis ICRF. There is a reduction of particle and thermal flux in the barrier region which then allows the neoclassical pinch to peak the central density. Recent work on ITB onset conditions [1] which was motivated by turbulence studies [2] points to the broadening of the Ti profile with off-axis ICRF acting to reduce the ion temperature gradient. This suppresses ITG instability driven particle fluxes, which is thought to be the primary mechanism for ITB formation. The object of this study is to examine the characteristics of ohmic ITBs to find whether the stability of plasmas and the plasma parameters support the onset model. [1]K. Zhurovich, et al., To be published in Nuclear Fusion [2] D. R. Ernst, et al., Phys. Plasmas 11, 2637 (2004)

  19. Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

    2016-02-01

    Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

  20. Neutral particle diagnostics for ALCATOR C-Mod

    International Nuclear Information System (INIS)

    Kurz, C.; Fiore, C.L.

    1990-01-01

    The ALCATOR C-Mod experiment will be equipped with two PPPL charge exchange neutral particle analyzers (CENAs), one of which views the plasma tangentially (R tan /R 0 =1.05), whereas the second has a horizontally scannable sight line (0≤R tan /R 0 ≤0.51). The perpendicularly viewing CENA will be capable of analyzing neutrals up to 600 keV amu for up to three separate species simultaneously. Thus high-energy tails can be observed together with the bulk ion temperature. The operation of both analyzers will allow simultaneous measurements from both the perpendicular and tangential chords. The CENAs will be used to study the effect of ICRF heating on the ion energy distribution with emphasis on the high-energy tail. A Fokker--Planck code (FPPRF) [Hammett, Ph.D. thesis, Princeton (1986)] is used to assess the appropriate operating regime of the analyzer (n≤4x10 20 m -3 for T i =2 keV, for Maxwellian ion energy distribution). The experimental design and computer simulations will be detailed

  1. Comparison of measured and modeled gas-puff emissions on Alcator C-Mod

    Science.gov (United States)

    Baek, Seung-Gyou; Terry, J. L.; Stotler, D. P.; Labombard, B. L.; Brunner, D. F.

    2017-10-01

    Understanding neutral transport in tokamak boundary plasmas is important because of its possible effects on the pedestal and scrape-off layer (SOL). On Alcator C-Mod, measured neutral line emissions from externally-puffed deuterium and helium gases are compared with the synthetic results of a neutral transport code, DEGAS 2. The injected gas flow rate and the camera response are absolutely calibrated. Time-averaged SOL density and temperature profiles are input to a steady-state simulation. An updated helium atomic model is employed in DEGAS2. Good agreement is found for the D α peak brightness and profile shape. However, the measured helium I line brightness is found to be lower than that in the simulation results by a roughly a factor of three over a wide range of density particularly in the far SOL region. Two possible causes for this discrepancy are reviewed. First, local cooling due to gas puff may suppress the line emission. Second, time-dependent turbulence effect may impact the helium neutral transport. Unlike deuterium atoms that gain energy from charge exchange and dissociation processes, helium neutrals remain cold and have a relatively short mean free path, known to make them prone to turbulence based on the Kubo number criterion. Supported by USDoE awards: DE-FC02-99ER54512, DE-SC0014251, and DE-AC02-09CH11466.

  2. Production of internal transport barriers via self-generated mean flows in Alcator C-Moda)

    Science.gov (United States)

    Fiore, C. L.; Ernst, D. R.; Podpaly, Y. A.; Mikkelsen, D.; Howard, N. T.; Lee, Jungpyo; Reinke, M. L.; Rice, J. E.; Hughes, J. W.; Ma, Y.; Rowan, W. L.; Bespamyatnov, I.

    2012-05-01

    New results suggest that changes observed in the intrinsic toroidal rotation influence the internal transport barrier (ITB) formation in the Alcator C-Mod tokamak [E. S. Marmar and Alcator C-Mod group, Fusion Sci. Technol. 51, 261 (2007)]. These arise when the resonance for ion cyclotron range of frequencies (ICRF) minority heating is positioned off-axis at or outside of the plasma half-radius. These ITBs form in a reactor relevant regime, without particle or momentum injection, with Ti ≈ Te, and with monotonic q profiles (qmin 1.5 × 105 rad/s) in the region where the ITB foot is observed. Gyrokinetic analyses indicate that this spontaneous shearing rate is comparable to the linear ion temperature gradient (ITG) growth rate at the ITB location and is sufficient to reduce the turbulent particle and energy transport. New and detailed measurement of the ion temperature demonstrates that the radial profile flattens as the ICRF resonance position moves off axis, decreasing the drive for the ITG the instability as well. These results are the first evidence that intrinsic rotation can affect confinement in ITB plasmas.

  3. Soft x-ray tomography on the Alcator C tokamak

    International Nuclear Information System (INIS)

    Camacho, J.F.

    1985-06-01

    A soft x-ray tomography experiment has been performed on the Alcator C tokamak. An 80-chord array of detectors consisting of miniature PIN photodiodes was used to obtain tomographic reconstructions of the soft x-ray emissivity function's poloidal cross-section. The detectors are located around the periphery of the plasma at one toroidal location (top and bottom ports) and are capable of yielding useful information over a wide range of plasma operating parameters and conditions. The reconstruction algorithm employed makes no assumption whatsoever about plasma rotation, position, or symmetry. Its performance was tested, and it was found to work well and to be fairly insensitive to estimated levels of random and systematic errors in the data

  4. A CO2 laser polarimeter for measurement of plasma current profile in Alcator C-Mod

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Richards, R.K.; Irby, J.; Luke, T.

    1994-01-01

    A multichannel infrared polarimeter system for measurement of the plasma current profile in Alcator C-Mod has been designed, constructed, and tested. The system utilizes a cw CO 2 , laser at a wavelength of 10.6 μm. An electro-optic polarization-modulation technique has been used to achieve the high sensitivity required for the measurement. The recent results of the measurements as well as the feasibility of its application on ITER are presented

  5. Transport of light, trace impurities in Alcator C-Mod

    Science.gov (United States)

    Rowan, W. L.; Bespamyatnov, I. O.; Liao, K. T.; Horton, W.; Fu, X. R.; Hughes, J. W.

    2012-10-01

    Light impurity profiles for boron were measured in ITB, H-mode, L-mode, and I-mode discharges in Alcator C-Mod. Within this wide range of modes, the profiles varied from peaked to hollow to flat. Specifically, hollow profiles are often observed in H-mode, while ITBs produce strong peaking, and L-mode produces moderate peaking. I-mode discharges are characterized by flat impurity profiles. For the study reported here, the profiles were measured with charge exchange recombination spectroscopy. The dependences of Rv/D were sought on dimensionless quantities including ion density scale length, effective charge, collisionality, and temperature scale length. We find that neoclassical transport consistently underestimates the measured transport. The excess measured transport is assumed to be turbulent. The strongest dependence of Rv/D is with temperature scale length. In addition, the measured transport was compared with the prediction of an analytical theory of drift wave turbulence that identifies transport implications for drift waves driven by ion and impurity density gradients.

  6. Alcator C vertical viewing electron cyclotron emission diagnostic

    International Nuclear Information System (INIS)

    Kato, K.; Hutchinson, I.H.

    1986-03-01

    Electron cyclotron emission measured vertically through the center of a tokamak plasma yields detailed information about the electron velocity distribution. A diagnostic developed for this purpose on Alcator C tokamak uses specialized focusing optics to obtain a well collimated viewing chord, a compact viewing dump made of pyrex or Macor to reduce the effects of wall reflection and depolarization, and a rapid-scan polarizing Michelson interferometer - InSb detector system for the spectrum measurement; all constrained by the limited access and the compact size of Alcator C. Results of diffraction analysis are used to evaluate the theoretical performance of the optical system

  7. Upgraded PMI diagnostic capabilities using Accelerator-based In-situ Materials Surveillance (AIMS) on Alcator C-Mod

    Science.gov (United States)

    Kesler, Leigh; Barnard, Harold; Hartwig, Zachary; Sorbom, Brandon; Lanza, Richard; Terry, David; Vieira, Rui; Whyte, Dennis

    2014-10-01

    The AIMS diagnostic was developed to rapidly and non-invasively characterize in-situ plasma material interactions (PMI) in a tokamak. Recent improvements are described which significantly expand this measurement capability on Alcator C-Mod. The detection time at each wall location is reduced from about 10 min to 30 s, via improved hardware and detection geometry. Detectors are in an augmented re-entrant tube to maximize the solid angle between detectors and diagnostic locations. Spatial range is expanded by using beam dynamics simulation to design upgraded B-field power supplies to provide maximal poloidal access, including a ~20° toroidal range in the divertor. Measurement accuracy is improved with angular and energy resolved cross section measurements obtained using a separate 0.9 MeV deuteron ion accelerator. Future improvements include the installation of recessed scintillator tiles as beam targets for calibration of the diagnostic. Additionally, implanted depth marker tiles will enable AIMS to observe the in-situ erosion and deposition of high-Z plasma-facing materials. This work is supported by U.S. DOE Grant No. DE-FG02-94ER54235 and Cooperative Agreement No. DE-FC02-99ER54512.

  8. Web based electronic logbook and experiment run database viewer for Alcator C-Mod

    International Nuclear Information System (INIS)

    Fredian, T.W.; Stillerman, J.A.

    2006-01-01

    Since 1991, the scientists and engineers at the Alcator C-Mod experiment at MIT have been recording text entries about the experiments being performed in an electronic logbook. In addition, separate documents such as run plans, run summaries and experimental proposals have been created and stored in a variety of formats in computer files. This information has now been organized and made available via any modern web browser. The new web based interface permits the user to browse through all the logbook entries, run information and even view some key data traces of the experiment. Since this information is being catalogued by Internet search engines, these tools can also be used to quickly locate information. The web based logbook and run information interface provides some additional capabilities. Once logged into the web site, users can add, delete or modify logbook entries directly from their browser. The logbook window on their browser also provides dynamic updating when any new logbook entries are made. There is also live C-Mod operation status information with optional audio announcements available. The user can receive the same state change announcements such as 'entering init' or 'entering pulse' as they would if they were sitting in the C-Mod control room. This paper will describe the functionality of the web based logbook and how it was implemented

  9. A thermal transport coefficient for ohmic and ICRF plasmas in alcator C-mode

    International Nuclear Information System (INIS)

    Daughton, W.; Coppi, B.; Greenwald, M.

    1996-01-01

    The energy confinement in plasmas produced by Alcator C-Mod machine is markedly different from that observed by previous high field compact machines such as Alcator A and C, FT, and more recently FTU. For ohmic plasmas at low and moderate densities, the confinement times routinely exceed those expected from the so-called open-quotes neo-Alcatorclose quotes scaling by a factor as high as three. For both ohmic and ICRF heated plasmas, the energy confinement time increases with the current and is approximately independent of the density. The similarity in the confinement between the ohmic and ICRF regimes opens the possibility that the thermal transport in Alcator C-Mod may be described by one transport coefficient for both regimes. We introduce a modified form of a transport coefficient previously used to describe ohmic plasmas in Alcator C-Mod. The coefficient is inspired by the properties of the so-called open-quotes ubiquitousclose quotes mode that can be excited in the presence of a significant fraction of trapped electrons and also includes the constraint of profile consistency. A detailed series of transport simulations are used to show that the proposed coefficient can reproduce the observed temperature profiles, loop voltage and energy confinement time for both ohmic and ICRF discharges. A total of nearly two dozen ohmic and ICRF Alcator C-Mod discharges have been fit over the range of parameter space available using this transport coefficient

  10. Mechanisms for ITB formation and control in Alcator C-Mod identified through gyrokinetic simulations of TEM turbulence

    International Nuclear Information System (INIS)

    Ernst, D.R.; Basse, N.; Bonoli, P.T.; Catto, P.J.; Fiore, C.L.; Greenwald, M.; Hubbard, A.E.; Marmar, E.S.; Porkolab, M.; Rice, J.E.; Zeller, K.; Zhurovich, K.; Dorland, W.

    2005-01-01

    Internal particle and thermal energy transport barriers are produced in Alcator C-Mod with off-axis ICRF heating, with core densities exceeding 10 21 m -3 , without core fueling, and with little change in the temperature profile. Applying on-axis ICRF heating controls the core density gradient and rate of rise. The present study employs linear and nonlinear gyrokinetic simulations of trapped electron mode (TEM) turbulence to explore mechanisms for ITB formation and control in Alcator C-Mod ITB experiments. Anomalous pinches are found to be negligible in our simulations; further, the collisional Ware pinch is sufficient to account for the slow density rise, lasting many energy confinement times. The simulations have revealed new nonlinear physics of TEM turbulence. The critical density gradient for onset of TEM turbulent transport is nonlinearly up-shifted by zonal flows. As the density profile peaks, during ITB formation, this nonlinear critical gradient is eventually exceeded, and the turbulent particle diffusivity from GS2 gyrokinetic simulations matches the particle diffusivity from transport analysis, within experimental errors. A stable equilibrium is then established when the TEM turbulent diffusion balances the Ware pinch in the ITB. This equilibrium is sensitive to temperature through gyroBohm scaling of the TEM turbulent transport, and the collisionality dependence of the neoclassical pinch, providing for control of the density rate of rise with on-axis RF heating. With no core particle fueling, and ∼1 mm between density spatial channels, the C-Mod experiments provide a nearly ideal test bed for particle transport studies. The pure TEM is the only unstable drift mode in the ITB, producing particle transport driven by the density gradient. (author)

  11. A new interferometry-based electron density fluctuation diagnostic on Alcator C-Moda)

    Science.gov (United States)

    Kasten, C. P.; Irby, J. H.; Murray, R.; White, A. E.; Pace, D. C.

    2012-10-01

    The two-color interferometry diagnostic on the Alcator C-Mod tokamak has been upgraded to measure fluctuations in the electron density and density gradient for turbulence and transport studies. Diagnostic features and capabilities are described. In differential mode, fast phase demodulation electronics detect the relative phase change between ten adjacent, radially-separated (ΔR = 1.2 cm, adjustable), vertical-viewing chords, which allows for measurement of the line-integrated electron density gradient. The system can be configured to detect the absolute phase shift of each chord by comparison to a local oscillator, measuring the line-integrated density. Each chord is sensitive to density fluctuations with kR < 20.3 cm-1 and is digitized at up to 10 MS/s, resolving aspects of ion temperature gradient-driven modes and other long-wavelength turbulence. Data from C-Mod discharges is presented, including observations of the quasi-coherent mode in enhanced D-alpha H-mode plasmas and the weakly coherent mode in I-mode.

  12. 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)

  13. Diamagnetic measurements on the Alcator C tokamak

    International Nuclear Information System (INIS)

    Shepard, T.D.

    1985-07-01

    A procedure for determining the total thermal energy content of a magnetically confined plasma from a measurement of the plasma magnetization has been successfully implemented on the Alcator C tokamak. When a plasma is confined by a magnetic field, the kinetic pressure of the plasma is supported by an interaction between the confining magnetic field and drift currents which flow in the plasma. These drift currents induce an additional magnetic field which can be measured by means of appropriately positioned pickup coils. From a measurement of this magnetic field and of the confining magnetic field, one can calculate the spatially averaged plasma pressure, which is related to the thermal energy content of the plasma by the equation of state of the plasma. The theory on which this measurement is based is described in detail. The fields and currents which flow in the plasma are related to the confining magnetic field and the plasma pressure by requiring that the plasma be in equilibrium, i.e., by balancing the forces due to pressure gradients against those due to magnetic interactions. The apparatus used to make this measurement is described and some example data analyses are carried out

  14. Testing Gyrokinetics on C-Mod and NSTX

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  15. Helium experiments on Alcator C-Mod in support of ITER early operations

    Science.gov (United States)

    Kessel, C. E.; Wolfe, S. M.; Reinke, M. L.; Hughes, J. W.; Lin, Y.; Wukitch, S. J.; Baek, S. G.; Bonoli, P. T.; Chilenski, M.; Diallo, A.; the Alcator C-Mod Team

    2018-05-01

    Helium majority experiments on Alcator C-Mod were performed to compare with deuterium discharges, and inform ITER early operations. ELMy H-modes were produced with a special plasma shape at B T  =  5.3 T, I P  =  0.9 MA, at q 95 ~ 3.8. The He fraction ranged over, n He,L/n L  =  0.2-0.4, with n D,L/n L  =  0.15-0.26, compared to D plasmas with n D,L/n L  =  0.85-0.97. The power to enter the H-mode in He was found to be greater than ~2 times that for D discharges, in the low density region  operation in ITER.

  16. Overview of Recent Alcator C-Mod Highlights

    Science.gov (United States)

    Marmar, Earl; C-Mod Team

    2013-10-01

    Analysis and modeling of recent C-Mod experiments has yielded significant results across multiple research topics. I-mode provides routine access to high confinement plasma (H98 up to 1.2) in quasi-steady state, without large ELMs; pedestal pressure and impurity transport are regulated by short-wavelength EM waves, and core turbulence is reduced. Multi-channel transport is being investigated in Ohmic and RF-heated plasmas, using advanced diagnostics to validate non-linear gyrokinetic simulations. Results from the new field-aligned ICRF antenna, including significantly reduced high-Z metal impurity contamination, and greatly improved load-tolerance, are being understood through antenna-plasma modeling. Reduced LHCD efficiency at high density correlates with parametric decay and enhanced edge absorption. Strong flow drive and edge turbulence suppression are seen from LHRF, providing new approaches for plasma control. Plasma density profiles directly in front of the LH coupler show non-linear modifications, with important consequences for wave coupling. Disruption-mitigation experiments using massive gas injection at multiple toroidal locations show unexpected results, with potentially significant implications for ITER. First results from a novel accelerator-based PMI diagnostic are presented. What would be the world's first actively-heated high-temperature advanced tungsten divertor is designed and ready for construction. Conceptual designs are being developed for an ultra-advanced divertor facility, Alcator DX, to attack key FNSF and DEMO heat-flux challenges integrated with a high-performance core. Supported by USDOE.

  17. 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

  18. Numerical investigation of edge plasma phenomena in an enhanced D-alpha discharge at Alcator C-Mod: Parallel heat flux and quasi-coherent edge oscillations

    International Nuclear Information System (INIS)

    Russell, D. A.; D’Ippolito, D. A.; Myra, J. R.; LaBombard, B.; Terry, J. L.; Zweben, S. J.

    2012-01-01

    Reduced-model scrape-off layer turbulence (SOLT) simulations of an enhanced D-alpha (EDA) H-mode shot observed in the Alcator C-Mod tokamak were conducted to compare with observed variations in the scrape-off-layer (SOL) width of the parallel heat flux profile. In particular, the role of the competition between sheath- and conduction-limited parallel heat fluxes in determining that width was studied for the turbulent SOL plasma that emerged from the simulations. The SOL width decreases with increasing input power and with increasing separatrix temperature in both the experiment and the simulation, consistent with the strong temperature dependence of the parallel heat flux in balance with the perpendicular transport by turbulence and blobs. The particularly strong temperature dependence observed in the case analyzed is attributed to the fact that these simulations produce SOL plasmas which are in the conduction-limited regime for the parallel heat flux. A persistent quasi-coherent (QC) mode dominates the SOLT simulations and bears considerable resemblance to the QC mode observed in C-Mod EDA operation. The SOLT QC mode consists of nonlinearly saturated wave-fronts located just inside the separatrix that are convected poloidally by the mean flow, continuously transporting particles and energy and intermittently emitting blobs into the SOL.

  19. Simulated plasma facing component measurements for an in situ surface diagnostic on Alcator C-Moda)

    Science.gov (United States)

    Hartwig, Z. S.; Whyte, D. G.

    2010-10-01

    The ideal in situ plasma facing component (PFC) diagnostic for magnetic fusion devices would perform surface element and isotope composition measurements on a shot-to-shot (˜10 min) time scale with ˜1 μm depth and ˜1 cm spatial resolution over large areas of PFCs. To this end, the experimental adaptation of the customary laboratory surface diagnostic—nuclear scattering of MeV ions—to the Alcator C-Mod tokamak is being guided by ACRONYM, a Geant4 synthetic diagnostic. The diagnostic technique and ACRONYM are described, and synthetic measurements of film thickness for boron-coated PFCs are presented.

  20. Intermittent fluctuations in the Alcator C-Mod scrape-off layer for ohmic and high confinement mode plasmas

    Science.gov (United States)

    Garcia, O. E.; Kube, R.; Theodorsen, A.; LaBombard, B.; Terry, J. L.

    2018-05-01

    Plasma fluctuations in the scrape-off layer of the Alcator C-Mod tokamak in ohmic and high confinement modes have been analyzed using gas puff imaging data. In all cases investigated, the time series of emission from a single spatially resolved view into the gas puff are dominated by large-amplitude bursts, attributed to blob-like filament structures moving radially outwards and poloidally. There is a remarkable similarity of the fluctuation statistics in ohmic plasmas and in edge localized mode-free and enhanced D-alpha high confinement mode plasmas. Conditionally averaged waveforms have a two-sided exponential shape with comparable temporal scales and asymmetry, while the burst amplitudes and the waiting times between them are exponentially distributed. The probability density functions and the frequency power spectral densities are similar for all these confinement modes. These results provide strong evidence in support of a stochastic model describing the plasma fluctuations in the scrape-off layer as a super-position of uncorrelated exponential pulses. Predictions of this model are in excellent agreement with experimental measurements in both ohmic and high confinement mode plasmas. The stochastic model thus provides a valuable tool for predicting fluctuation-induced plasma-wall interactions in magnetically confined fusion plasmas.

  1. Multispecies density peaking in gyrokinetic turbulence simulations of low collisionality Alcator C-Mod plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D. R., E-mail: dmikkelsen@pppl.gov; Bitter, M.; Delgado-Aparicio, L.; Hill, K. W. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543 (United States); Greenwald, M.; Howard, N. T.; Hughes, J. W.; Rice, J. E. [MIT Plasma Science and Fusion Center, 175 Albany St., Cambridge, Massachusetts 02139 (United States); Reinke, M. L. [MIT Plasma Science and Fusion Center, 175 Albany St., Cambridge, Massachusetts 02139 (United States); York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Podpaly, Y. [MIT Plasma Science and Fusion Center, 175 Albany St., Cambridge, Massachusetts 02139 (United States); AAAS S and T Fellow placed in the Directorate for Engineering, NSF, 4201 Wilson Blvd., Arlington, Virginia 22230 (United States); Ma, Y. [MIT Plasma Science and Fusion Center, 175 Albany St., Cambridge, Massachusetts 02139 (United States); ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Candy, J.; Waltz, R. E. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

    2015-06-15

    Peaked density profiles in low-collisionality AUG and JET H-mode plasmas are probably caused by a turbulently driven particle pinch, and Alcator C-Mod experiments confirmed that collisionality is a critical parameter. Density peaking in reactors could produce a number of important effects, some beneficial, such as enhanced fusion power and transport of fuel ions from the edge to the core, while others are undesirable, such as lower beta limits, reduced radiation from the plasma edge, and consequently higher divertor heat loads. Fundamental understanding of the pinch will enable planning to optimize these impacts. We show that density peaking is predicted by nonlinear gyrokinetic turbulence simulations based on measured profile data from low collisionality H-mode plasma in Alcator C-Mod. Multiple ion species are included to determine whether hydrogenic density peaking has an isotope dependence or is influenced by typical levels of low-Z impurities, and whether impurity density peaking depends on the species. We find that the deuterium density profile is slightly more peaked than that of hydrogen, and that experimentally relevant levels of boron have no appreciable effect on hydrogenic density peaking. The ratio of density at r/a = 0.44 to that at r/a = 0.74 is 1.2 for the majority D and minority H ions (and for electrons), and increases with impurity Z: 1.1 for helium, 1.15 for boron, 1.3 for neon, 1.4 for argon, and 1.5 for molybdenum. The ion temperature profile is varied to match better the predicted heat flux with the experimental transport analysis, but the resulting factor of two change in heat transport has only a weak effect on the predicted density peaking.

  2. Local gas injection as a scrape-off layer diagnostic on the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Jablonski, D.F.

    1996-05-01

    A capillary puffing array has been installed on Alcator C-Mod which allows localized introduction of gaseous species in the scrape-off layer. This system has been utilized in experiments to elucidate both global and local properties of edge transport. Deuterium fueling and recycling impurity screening are observed to be characterized by non-dimensional screening efficiencies which are independent of the location of introduction. In contrast, the behavior of non-recycling impurities is seen to be characterized by a screening time which is dependent on puff location. The work of this thesis has focused on the use of the capillary array with a camera system which can view impurity line emission plumes formed in the region of an injection location. The ionic plumes observed extend along the magnetic field line with a comet-like asymmetry, indicative of background plasma ion flow. The flow is observed to be towards the nearest strike-point, independent of x-point location, magnetic field direction, and other plasma parameters. While the axes of the plumes are generally along the field line, deviations are seen which indicate cross-field ion drifts. A quasi-two dimensional fluid model has been constructed to use the plume shapes of the first charge state impurity ions to extract information about the local background plasma, specifically the temperature, parallel flow velocity, and radial electric field. Through comparisons of model results with those of a three dimensional Monte Carlo code, and comparisons of plume extracted parameters with scanning probe measurements, the efficacy of the model is demonstrated. Plume analysis not only leads to understandings of local edge impurity transport, but also presents a novel diagnostic technique

  3. Local gas injection as a scrape-off layer diagnostic on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jablonski, David F. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1996-05-01

    A capillary puffing array has been installed on Alcator C-Mod which allows localized introduction of gaseous species in the scrape-off layer. This system has been utilized in experiments to elucidate both global and local properties of edge transport. Deuterium fueling and recycling impurity screening are observed to be characterized by non-dimensional screening efficiencies which are independent of the location of introduction. In contrast, the behavior of non-recycling impurities is seen to be characterized by a screening time which is dependent on puff location. The work of this thesis has focused on the use of the capillary array with a camera system which can view impurity line emission plumes formed in the region of an injection location. The ionic plumes observed extend along the magnetic field line with a comet-like asymmetry, indicative of background plasma ion flow. The flow is observed to be towards the nearest strike-point, independent of x-point location, magnetic field direction, and other plasma parameters. While the axes of the plumes are generally along the field line, deviations are seen which indicate cross-field ion drifts. A quasi-two dimensional fluid model has been constructed to use the plume shapes of the first charge state impurity ions to extract information about the local background plasma, specifically the temperature, parallel flow velocity, and radial electric field. Through comparisons of model results with those of a three dimensional Monte Carlo code, and comparisons of plume extracted parameters with scanning probe measurements, the efficacy of the model is demonstrated. Plume analysis not only leads to understandings of local edge impurity transport, but also presents a novel diagnostic technique.

  4. Density limit and cross-field edge transport scaling in Alcator C-Mod

    International Nuclear Information System (INIS)

    LaBombard, B.; Greenwald, M.; Hughes, J.W.; Lipschultz, B.; Mossessian, D.; Terry, J.L.; Boivin, R.L.; Carreras, B.A.; Pitcher, C.S.; Zweben, S.J.

    2003-01-01

    Recent experiments in Alcator C-Mod have uncovered a direct link between the character and scaling of cross-field particle transport in the edge plasma and the density limit, n G . As n-bar e /n G is increased from low values to values approaching ∼1, an ordered progression in the cross-field edge transport physics occurs: first benign cross-field heat convection, then cross-field heat convection impacting the scrape-off layer (SOL) power loss channels and reducing the separatrix electron temperature, and finally 'bursty' transport (normally associated with the far SOL) invading into closed flux surface regions and carrying a convective power loss that impacts the power balance of the discharge. These observations suggest that SOL transport and its scaling with plasma conditions plays a key role in setting the empirically observed density limit scaling law. (author)

  5. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  6. Influence of boronization on operation with high-Z plasma facing components in Alcator C-Mod

    International Nuclear Information System (INIS)

    Lipschultz, B.; Lin, Y.; Marmar, E.S.; Whyte, D.G.; Wukitch, S.; Hutchinson, I.H.; Irby, J.; LaBombard, B.; Reinke, M.L.; Terry, J.L.; Wright, G.

    2007-01-01

    We report the results of operation of Alcator C-Mod with all high-Z molybdenum plasma facing component (PFC) surfaces. Without boron-coated PFCs energy confinement was poor (H ITER,89 ∼ 1) due to high core molybdenum (n Mo /n e ≤ 0.1%) and radiation. After applying boron coatings, n Mo /n e was reduced by a factor of 10-20 with H ITER,89 approaching 2. Results of between-discharge boronization, localized at various major radii, point towards important molybdenum source regions being small, outside the divertor, and due to RF-sheath-rectification. Boronization also has a significant effect on the plasma startup phase lowering Z eff , radiation, and lowering the runaway electron damage. The requirement of low-Z coatings over at least a fraction of the Mo PFCs in C-Mod for best performance together with the larger than expected D retention in Mo, give impetus for further high-Z PFC investigations to better predict the performance of un-coated tungsten surfaces in ITER and beyond

  7. Discrepancies between soft x-ray emissivity contours and magnetic flux surfaces in Alcator C-Mod

    International Nuclear Information System (INIS)

    Borras, M.C.; Granetz, R.S.

    1996-01-01

    The soft x-ray diagnostic system of Alcator C-Mod, equipped with 152 detectors distributed in four arrays, is used to obtain iso-emissivity surfaces. These surfaces have been characterized by giving their elongation and relative shift from the centre of the tokamak as functions of plasma radius. Flux surfaces, provided by magnetic diagnostics, have also been described with elongation and shift. Results from the comparison of the two sets of geometric parameters obtained from magnetic and x-ray diagnostics are presented. We find that, whereas the shifts obtained from these two diagnostic methods are always in good agreement, the corresponding elongation curves show different patterns. An agreement between elongations better than 2% is only found in a range of about 2 cm in minor radius. On the other hand, the elongations can differ by 10% towards the plasma edge and the plasma centre. Error bars for the x-ray diagnostic are obtained by propagating the effect of ± 1% random errors at the detector signals, and can amount to ± 1-2% of the estimated values near the edge and the centre of the plasma. The estimated uncertainties in the determination of elongation from magnetic flux surfaces are of the order of 4%. A series of tests and simulations performed to verify the accuracy of the X-ray diagnostic system is presented. The discrepancies found could imply the existence of asymmetries in impurity concentration. (Author)

  8. Control of internal transport barriers on Alcator C-Mod

    International Nuclear Information System (INIS)

    Fiore, C.L.; Bonoli, P.T.; Ernst, D.R.; Hubbard, A.E.; Greenwald, M.J.; Lynn, A.; Marmar, E.S.; Phillips, P.; Redi, M.H.; Rice, J.E.; Wolfe, S.M.; Wukitch, S.J.; Zhurovich, K.

    2004-01-01

    Recent studies of internal transport and double transport barrier regimes in the Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] have explored the limits for forming, maintaining, and controlling these plasmas. The C-Mod provides a unique platform for studying such discharges: the ions and electrons are tightly coupled by collisions and the plasma has no internal particle or momentum sources. The double-barrier mode comprised of an edge barrier with an internal transport barrier (ITB) can be induced at will using off-axis ion cyclotron range of frequency (ICRF) injection on either the low or high field side of the plasma with either of the available ICRF frequencies (70 or 80 MHz). When an enhanced D α high confinement mode (EDA H-mode) is accessed in Ohmic plasmas, the double barrier ITB forms spontaneously if the H-mode is sustained for ∼2 energy confinement times. The ITBs formed in both Ohmic and ICRF heated plasmas are quite similar regardless of the trigger method. They are characterized by strong central peaking of the electron density, and a reduction of the core particle and energy transport. The control of impurity influx and heating of the core plasma in the presence of the ITB have been achieved with the addition of central ICRF power in both the Ohmic H-mode and ICRF induced ITBs. The radial location of the particle transport barrier is dependent on the toroidal magnetic field but not on the location of the ICRF resonance. A narrow region of decreased electron thermal transport, as determined by sawtooth heat pulse analysis, is found in these plasmas as well. Transport analysis indicates that a reduction of the particle diffusivity in the barrier region allows the neoclassical pinch to drive the density and impurity accumulation in the plasma center. An examination of the gyrokinetic stability at the trigger time for the ITB suggests that the density and temperature profiles are inherently stable to ion temperature gradient and

  9. Poloidal asymmetries in the limiter shadow plasma of the Alcator C tokamak. Volume 1

    International Nuclear Information System (INIS)

    LaBombard, B.

    1986-05-01

    This thesis investigates conditions which exist in the limiter shadow plasma of the Alcator C tokamak. The understanding of this edge plasma region is approached from both experimental and theoretical points of view. First, a general overview of edge plasma physical processes is presented. Simple edge plasma models and conditions which can theoretically result in a poloidally asymmetric edge plasma are discussed. A review of data obtained from previous diagnostics in the Alcator C edge plasma is then used to motivate the development of a new edge plasma diagnostic system (DENSEPACK) to experimentally investigate poloidal asymmetries in this region. The bulk of this thesis focuses on the marked poloidal asymmetries detected by this poloidal probe array and possible mechanisms which might support such asymmetries on a magnetic flux surface. In processing the probe data, some important considerations on fitting Langmuir probe characteristics are identified. The remainder of this thesis catalogues edge versus central plasma parameter dependences. Regression analysis techniques are applied to characterize edge density for various central plasma parameters. Edge plasma conditions during lower hybrid radio frequency heating and pellet injection are also discussed

  10. Two dimensional radiated power diagnostics on Alcator C-Moda)

    Science.gov (United States)

    Reinke, M. L.; Hutchinson, I. H.

    2008-10-01

    The radiated power diagnostics for the Alcator C-Mod tokamak have been upgraded to measure two dimensional structure of the photon emissivity profile in order to investigate poloidal asymmetries in the core radiation. Commonly utilized unbiased absolute extreme ultraviolet (AXUV) diode arrays view the plasma along five different horizontal planes. The layout of the diagnostic set is shown and the results from calibrations and recent experiments are discussed. Data showing a significant, 30%-40%, inboard/outboard emissivity asymmetry during ELM-free H-mode are presented. The ability to use AXUV diode arrays to measure absolute radiated power is explored by comparing diode and resistive bolometer-based emissivity profiles for highly radiative L-mode plasmas seeded with argon. Emissivity profiles match in the core but disagree radially outward resulting in an underprediction of Prad of nearly 50% by the diodes compared to Prad determined using resistive bolometers.

  11. The design and performance of a twenty barrel hydrogen pellet injector for Alcator C-Mod

    International Nuclear Information System (INIS)

    Urbahn, J.A.

    1994-05-01

    A twenty barrel hydrogen pellet injector has been designed, built and tested both in the laboratory and on the Alcator C-Mod Tokamak at MIT. The injector functions by firing pellets of frozen hydrogen or deuterium deep into the plasma discharge for the purpose of fueling the plasma, modifying the density profile and increasing the global energy confinement time. The design goals of the injector are: (1) Operational flexibility, (2) High reliability, (3) Remote operation with minimal maintenance. These requirements have lead to a single stage, pipe gun design with twenty barrels. Pellets are formed by in- situ condensation of the fuel gas, thus avoiding moving parts at cryogenic temperatures. The injector is the first to dispense with the need for cryogenic fluids and instead uses a closed cycle refrigerator to cool the thermal system components. The twenty barrels of the injector produce pellets of four different size groups and allow for a high degree of flexibility in fueling experiments. Operation of the injector is under PLC control allowing for remote operation, interlocked safety features and automated pellet manufacturing. The injector has been extrusively tested and shown to produce pellets reliably with velocities up to 1400 m/sec. During the period from September to November of 1993, the injector was successfully used to fire pellets into over fifty plasma discharges. Experimental results include data on the pellet penetration into the plasma using an advanced pellet tracking diagnostic with improved time and spatial response. Data from the tracker indicates pellet penetrations were between 30 and 86 percent of the plasma minor radius

  12. Impurity toroidal rotation and transport in Alcator C-Mod ohmic high confinement mode plasmas

    International Nuclear Information System (INIS)

    Rice, J. E.; Goetz, J. A.; Granetz, R. S.; Greenwald, M. J.; Hubbard, A. E.; Hutchinson, I. H.; Marmar, E. S.; Mossessian, D.; Pedersen, T. Sunn; Snipes, J. A.

    2000-01-01

    Central toroidal rotation and impurity transport coefficients have been determined in Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] Ohmic high confinement mode (H-mode) plasmas from observations of x-ray emission following impurity injection. Rotation velocities up to 3x10 4 m/sec in the co-current direction have been observed in the center of the best Ohmic H-mode plasmas. Purely ohmic H-mode plasmas display many characteristics similar to ion cyclotron range of frequencies (ICRF) heated H-mode plasmas, including the scaling of the rotation velocity with plasma parameters and the formation of edge pedestals in the electron density and temperature profiles. Very long impurity confinement times (∼1 sec) are seen in edge localized mode-free (ELM-free) Ohmic H-modes and the inward impurity convection velocity profile has been determined to be close to the calculated neoclassical profile. (c) 2000 American Institute of Physics

  13. The flush-mounted rail Langmuir probe array designed for the Alcator C-Mod vertical target plate divertor

    Science.gov (United States)

    Kuang, A. Q.; Brunner, D.; LaBombard, B.; Leccacorvi, R.; Vieira, R.

    2018-04-01

    An array of flush-mounted and toroidally elongated Langmuir probes (henceforth called rail probes) have been specifically designed for the Alcator C-Mod's vertical target plate divertor and operated over multiple campaigns. The "flush" geometry enables the tungsten electrodes to survive high heat flux conditions in which traditional "proud" tungsten electrodes suffer damage from melting. The toroidally elongated rail-like geometry reduces the influence of sheath expansion, which is an important effect to consider in the design and interpretation of flush-mounted Langmuir probes. The new rail probes successfully operated during C-Mod's FY2015 and FY2016 experimental campaigns with no evidence of damage, despite being regularly subjected to heat flux densities parallel to the magnetic field exceeding ˜1 GW m-2 for short periods of time. A comparison between rail and proud probe data indicates that sheath expansion effects were successfully mitigated by the rail design, extending the use of these Langmuir probes to incident magnetic field line angles as low as 0.5°.

  14. Internal transport barrier production and control in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Fiore, C L [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Bonoli, P T [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Ernst, D R [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Greenwald, M J [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Marmar, E S [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Redi, M H [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Rice, J E [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Wukitch, S J [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Zhurovich, K [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2004-12-01

    Internal transport barriers (ITBs), marked by a steep density profile, even stronger peaking in the pressure profile and reduction of core transport are obtained in Alcator C-Mod. They are induced by the use of off-axis D(H) ICRF (ion cyclotron range of frequencies) power deposition. They also arise spontaneously in Ohmic H-mode plasmas once the H-mode lasts for several energy confinement times. Recent studies have explored the limits for forming, maintaining and controlling these plasmas. The C-Mod provides a unique platform for studying such discharges: the high density (up to 8 x 10{sup 20} m{sup -3}) causes the ions and electrons to be tightly coupled by collisions with T{sub i}/T{sub e} = 1, and the plasma has no internal particle or momentum sources. The ITBs formed in both Ohmic and ICRF heated plasmas are quite similar regardless of the trigger method. Control of impurity influx and heating of the core plasma in the presence of the ITB have been achieved with the addition of central ICRF power, in both Ohmic H-mode and ICRF induced ITBs. Control of the radial location of the transport barrier is achieved through manipulation of the toroidal magnetic field and plasma current. A narrow region of decreased electron thermal transport, as determined by sawtooth heat pulse analysis, is found in these plasmas as well. Transport analysis indicates that reduction of the particle diffusivity in the barrier region allows the neoclassical pinch to drive the density and impurity accumulation in the plasma centre. Examination of the gyro-kinetic stability indicates that the density and temperature profiles of the plasma core are inherently stable to long-wavelength drift mode driven turbulence at the onset time of the ITB, but that the increasing density gradients cause the trapped electron mode to play a role in providing a control mechanism to ultimately limit the density and impurity rise in the plasma centre.

  15. Radiative and three-body recombination in the Alcator C-Mod divertor

    International Nuclear Information System (INIS)

    Lumma, D.; Terry, J.L.; Lipschultz, B.

    1997-01-01

    Significant recombination of the majority ion species has been observed in the divertor region of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under detached conditions. This determination is made by analysis of the visible spectrum from the divertor, in particular the Balmer series line emission and the observed recombination continuum, including an apparent recombination edge at ∼375 nm. The analysis shows that the electron temperature in the recombining plasma is 0.8 endash 1.5 eV. The measured volume recombination rate is comparable to the rate of ion collection at the divertor plates. The dominant recombination mechanism is three-body recombination into excited states (e+e+D + Right-arrow D 0 +e), although radiative recombination (e+D + Right-arrow D 0 +hν) contributes ∼5% to the total rate. Analysis of the Balmer series line intensities (from n upper =3 through 10) shows that the upper levels of these transitions are populated primarily by recombination. Thus the brightnesses of the Balmer series (and Lyman series) are directly related to the recombination rate. copyright 1997 American Institute of Physics

  16. Experimental and gyrokinetic investigation of core impurity transport in Alcator C-mod

    Science.gov (United States)

    Howard, N.; Greenwald, M.; Podpaly, Y.; Reinke, M. L.; Rice, J. E.; White, A. E.; Mikkelsen, D. R.; Puetterich, T.

    2010-11-01

    A new multiple pulse laser blow-off system coupled with an upgraded high resolution x-ray spectrometer with spatial resolution allow for the most detailed studies of impurity transport on Alcator C-mod to date. Trace impurity injections created by the laser blow-off technique were introduced into plasmas with a wide range of parameters and time evolving profiles of He-like calcium were measured. The unique measurement of a single charge state profile and line integrated emission measurements from spectroscopic diagnostics were compared with the simulated emission from the impurity transport code STRAHL. A nonlinear least squares fitting routine was coupled with STRAHL, allowing for core impurity transport coefficients with errors to be determined. With this method, experimental data from trace calcium injections were analyzed and radially dependent, core values (< r/a ˜.6) of the diffusive and convective components of the impurity flux were obtained. The STRAHL results are compared with linear and global, nonlinear simulations from the gyrokinetic code GYRO. Results of this comparison and an investigation of the underlying physics associated with turbulent impurity transport will be presented.

  17. Time-resolving electron temperature diagnostic for ALCATOR C

    International Nuclear Information System (INIS)

    Fairfax, S.A.

    1984-05-01

    A diagnostic that provides time-resolved central electron temperatures has been designed, built, and tested on the ALCATOR C Tokamak. The diagnostic uses an array of fixed-wavelength x-ray crystal monochromators to sample the x-ray continuum and determine the absolute electron temperature. The resolution and central energy of each channel were chosen to exclude any contributions from impurity line radiation. This document describes the need for such a diagnostic, the design methodology, and the results with typical ALCATOR C plasmas. Sawtooth (m = 1) temperature oscillations were observed after pellet fueling of the plasma. This is the first time that such oscillations have been observed with an x-ray temperature diagnostic

  18. Observations of core toroidal rotation reversals in Alcator C-Mod ohmic L-mode plasmas

    International Nuclear Information System (INIS)

    Rice, J.E.; Reinke, M.L.; Podpaly, Y.A.; Churchill, R.M.; Cziegler, I.; Dominguez, A.; Ennever, P.C.; Fiore, C.L.; Granetz, R.S.; Greenwald, M.J.; Hubbard, A.E.; Hughes, J.W.; Irby, J.H.; Ma, Y.; Marmar, E.S.; McDermott, R.M.; Porkolab, M.; Duval, B.P.; Bortolon, A.; Diamond, P.H.

    2011-01-01

    Direction reversals of intrinsic toroidal rotation have been observed in Alcator C-Mod ohmic L-mode plasmas following modest electron density or toroidal magnetic field ramps. The reversal process occurs in the plasma interior, inside of the q = 3/2 surface. For low density plasmas, the rotation is in the co-current direction, and can reverse to the counter-current direction following an increase in the electron density above a certain threshold. Reversals from the co- to counter-current direction are correlated with a sharp decrease in density fluctuations with k R ≥ 2 cm -1 and with frequencies above 70 kHz. The density at which the rotation reverses increases linearly with plasma current, and decreases with increasing magnetic field. There is a strong correlation between the reversal density and the density at which the global ohmic L-mode energy confinement changes from the linear to the saturated regime.

  19. Experimental/theoretical comparisons of the turbulence in the scrape-off-layers of Alcator C-Mod, DIII-D, and NSTX

    International Nuclear Information System (INIS)

    Terry, J.L. . E-mail : terry@psfc.mit.edu; Zweben, S.J.; Rudakov, D.L.

    2003-01-01

    The intermittent turbulent transport in the scrape-off-layers of Alcator C-Mod, DIII-D, and NSTX is studied experimentally. On DIII-D the fluctuations of both density and temperature have strongly non-Gaussian statistics, and events with amplitudes above 10 times the mean level are responsible for large fractions of the net particle and heat transport, indicating the importance of turbulence on the transport. In C-Mod and NSTX the turbulence is imaged with a very high density of spatial measurements. The 2-D structure and dynamics of emission from a localized gas puff are observed, and intermittent features (also sometimes called 'blobs') are typically seen. On DIII-D the turbulence is imaged using BES and similar intermittent features are seen. The dynamics of these intermittent features are discussed. The experimental observations are compared with numerical simulations of edge turbulence. The electromagnetic turbulence in a 3-D geometry is computed using non-linear plasma fluid equations. The wavenumber spectra in the poloidal dimension of the simulations are in reasonable agreement with those of the C-Mod experimental images once the response of the optical system is accounted for. The resistive ballooning mode is the dominant linear instability in the simulations. (author)

  20. Studies of photonuclear neutron emission during the start-up phase of the Alcator C tokamak

    International Nuclear Information System (INIS)

    Pappas, D.S.; Furnstahl, R.; Kochanski, G.P.

    1981-05-01

    Alcator C operations commenced with discharge cleaning and tokamak operation using hydrogen filling gas. Prior to and during these experiments no deuterium gas was allowed into the device. The earliest operation resulted in dosimeter readings of a few Roentgen per shot in the vicinity of the limiter and a localized source of neutron emission of up to 10 9 neutrons/shot. A strong correlation of the neutron emissions with hard x-ray emissions from the limiter and nonthermal features on the synchrotron emissions was observed during these discharges. Gamma energy spectroscopy of the activated limiter after removal from Alcator allowed identification of 16 radioisotopes which were consistent with photonuclear processes (γ,n , γ,p , γ,α reactions) arising in the limiter. After seven months of hydrogen operation conditions were achieved that resulted in substantially less non-thermal activity. Typical neutron emission rates of equal to or less than 10 6 n/sec were observed, i.e., about four orders of magnitude less than the expected D-D thermonuclear neutron emission rates for the same type of discharge if D 2 was used as the filling gas

  1. Active control system upgrade design for lower hybrid current drive system on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Kanojia, A.D., E-mail: akanojia@mit.edu [Massachusetts Institute of Technology Plasma Science and Fusion Center, Cambridge, MA (United States); Wallace, G.M.; Terry, D.R.; Stillerman, J.A.; Burke, W.M.; MacGibbon, P.A.; Johnson, D.K. [Massachusetts Institute of Technology Plasma Science and Fusion Center, Cambridge, MA (United States)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Initial tests of the Hittite microwave components show good or better control of phase and amplitude when compared to the vector modulators used in the current system. Black-Right-Pointing-Pointer With an analog based control component system the system complexity is dramatically reduced. Black-Right-Pointing-Pointer Historically, D-tAcq hardware/software has performed more reliably on DPCS and FFT controllers than the current lower hybrid control system Black-Right-Pointing-Pointer Cost and lead time of the Hittite microwave components is significantly small compared to vector modulators. - Abstract: As a part of the scheduled expansion of the Alcator C-Mod lower hybrid current drive (LHCD) system from 12 to 16 klystrons to accommodate installation of a second LH antenna, the active control system (ACS) is being redesigned to accommodate the additional klystrons. Digitizers and output modules will be cPCI modules provided by D-tAcq Solutions. The real-time application will run on a standard PC server running Linux. Initially, the new ACS system will be designed to control 8 klystrons on the second LH antenna and the existing ACS will control the remaining 8 klystrons on the existing LH antenna. Experience gained operating the existing LHCD system has given us insight into the design of a more robust, compact, efficient and simple system for the new ACS. The design upgrade will be patterned on the digital plasma control system (DPCS [1]) in use on C-Mod.

  2. Measured MHD equilibrium in Alcator C

    International Nuclear Information System (INIS)

    Pribyl, P.A.

    1986-03-01

    A method of processing data from a set of partial Rogowski loops is developed to study the MHD equilibrium in Alcator C. Time dependent poloidal fields in the vicinity of the plasma are calculated from measured currents, with field penetration effects being accounted for. Fields from eddy currents induced by the plasma in the tokamak structure are estimated as well. Each of the set of twelve B/sub θ/ measurements can then be separated into a component from the plasma current and a component from currents external to the pickup loops. Harmonic solutions to Maxwell's equations in toroidal coordinates are fit to these measurements in order to infer the fields everywhere in the vacuum region surrounding the plasma. Using this diagnostic, plasma current, position, shape, and the Shafranov term Λ = β/sub p/ + l/sub i//2 - 1 may be computed, and systematic studies of these plasma parameters are undertaken for Alcator C plasmas

  3. ICRF-enhanced plasma potentials in the SOL of Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Ochoukov, R.; Whyte, D. G.; Brunner, D.; LaBombard, B.; Lipschultz, B.; Terry, J. L.; Wukitch, S. J. [PSFC MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); D' Ippolito, D. A.; Myra, J. R. [Lodestar Research Corporation, 2400 Central Avenue, Boulder, Colorado 80301 (United States)

    2014-02-12

    We performed an extensive survey of the plasma potential in the scrape-off layer (SOL) of Ion Cyclotron Range-of Frequencies (ICRF)-heated discharges on Alcator C-Mod. Our results show that plasma potentials are enhanced in the presence of ICRF power and plasma potential values of >100 V are often observed. Such potentials are high enough to induce sputtering of high-Z molybdenum (Mo) plasma facing components by deuterium ions on C-Mod. For comparison, the plasma potential in Ohmic discharges is typically less than 10 V, well below the threshold needed to induce Mo sputtering by deuterium ions. ICRF-enhanced plasma potentials are observed in the SOL regions that both magnetically map and do not map to active ICRF antennas. Regions that magnetically map to active ICRF antennas are accessible to slow waves directly launched by the antennas and these regions experience plasma potential enhancement that is partially consistent with the slow wave rectification mechanism. One of the most defining features of the slow wave rectification is a threshold appearance of significant plasma potentials (>100 V) when the dimensionless rectification parameter Λ{sub −o} is above unity and this trend is observed experimentally. We also observe ICRF-enhanced plasma potentials >100 V in regions that do not magnetically map to the active antennas and, hence, are not accessible for slow waves launched directly by the active antennas. However, unabsorbed fast waves can reach these regions. The general trend that we observe in these 'un-mapped' regions is that the plasma potential scales with the strength of the local RF wave fields with the fast wave polarization and the highest plasma potentials are observed in discharges with the highest levels of unabsorbed ICRF power. Similarly, we find that core Mo levels scale with the level of unabsorbed ICRF power suggesting a link between plasma potentials in the SOL and the strength of the impurity source.

  4. High-confinement-mode edge stability of Alcator C-mod plasmas

    International Nuclear Information System (INIS)

    Mossessian, D.A.; Snyder, P.; Hubbard, A.; Hughes, J.W.; Greenwald, M.; La Bombard, B.; Snipes, J.A.; Wolfe, S.; Wilson, H.

    2003-01-01

    For steady state high-confinement-mode (H-mode) operation, a relaxation mechanism is required to limit build-up of the edge gradient and impurity content. Alcator C-Mod [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] sees two such mechanisms--EDA (enhanced D-alpha H mode) and grassy ELMs (edge localized modes), but not large type I ELMs. In EDA the edge relaxation is provided by an edge localized quasicoherent (QC) electromagnetic mode that exists at moderate pedestal temperature T 95 >3.5, and does not limit the buildup of the edge pressure gradient. The q boundary of the operational space of the mode depends on plasma shape, with the q 95 limit moving down with increasing plasma triangularity. At high edge pressure gradients and temperatures the mode is replaced by broadband fluctuations ( f<50 kHz) and small irregular ELMs are observed. Ideal MHD (magnetohydrodynamic) stability analysis that includes both pressure and current driven edge modes shows that the discharges where the QC mode is observed are stable. The ELMs are identified as medium n (10< n<50) coupled peeling/ballooning modes. The predicted stability boundary of the modes as a function of pedestal current and pressure gradient is reproduced in experimental observations. The measured dependence of the ELMs' threshold and amplitude on plasma triangularity is consistent with the results of ideal MHD analysis performed with the linear stability code ELITE [Wilson et al., Phys. Plasmas 9, 1277 (2002)

  5. Measurement of particle transport coefficients on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Luke, T.C.T.

    1994-10-01

    The goal of this thesis was to study the behavior of the plasma transport during the divertor detachment in order to explain the central electron density rise. The measurement of particle transport coefficients requires sophisticated diagnostic tools. A two color interferometer system was developed and installed on Alcator C-Mod to measure the electron density with high spatial ({approx} 2 cm) and high temporal ({le} 1.0 ms) resolution. The system consists of 10 CO{sub 2} (10.6 {mu}m) and 4 HeNe (.6328 {mu}m) chords that are used to measure the line integrated density to within 0.08 CO{sub 2} degrees or 2.3 {times} 10{sup 16}m{sup {minus}2} theoretically. Using the two color interferometer, a series of gas puffing experiments were conducted. The density was varied above and below the threshold density for detachment at a constant magnetic field and plasma current. Using a gas modulation technique, the particle diffusion, D, and the convective velocity, V, were determined. Profiles were inverted using a SVD inversion and the transport coefficients were extracted with a time regression analysis and a transport simulation analysis. Results from each analysis were in good agreement. Measured profiles of the coefficients increased with the radius and the values were consistent with measurements from other experiments. The values exceeded neoclassical predictions by a factor of 10. The profiles also exhibited an inverse dependence with plasma density. The scaling of both attached and detached plasmas agreed well with this inverse scaling. This result and the lack of change in the energy and impurity transport indicate that there was no change in the underlying transport processes after detachment.

  6. Measurement of particle transport coefficients on Alcator C-Mod

    International Nuclear Information System (INIS)

    Luke, T.C.T.

    1994-10-01

    The goal of this thesis was to study the behavior of the plasma transport during the divertor detachment in order to explain the central electron density rise. The measurement of particle transport coefficients requires sophisticated diagnostic tools. A two color interferometer system was developed and installed on Alcator C-Mod to measure the electron density with high spatial (∼ 2 cm) and high temporal (≤ 1.0 ms) resolution. The system consists of 10 CO 2 (10.6 μm) and 4 HeNe (.6328 μm) chords that are used to measure the line integrated density to within 0.08 CO 2 degrees or 2.3 x 10 16 m -2 theoretically. Using the two color interferometer, a series of gas puffing experiments were conducted. The density was varied above and below the threshold density for detachment at a constant magnetic field and plasma current. Using a gas modulation technique, the particle diffusion, D, and the convective velocity, V, were determined. Profiles were inverted using a SVD inversion and the transport coefficients were extracted with a time regression analysis and a transport simulation analysis. Results from each analysis were in good agreement. Measured profiles of the coefficients increased with the radius and the values were consistent with measurements from other experiments. The values exceeded neoclassical predictions by a factor of 10. The profiles also exhibited an inverse dependence with plasma density. The scaling of both attached and detached plasmas agreed well with this inverse scaling. This result and the lack of change in the energy and impurity transport indicate that there was no change in the underlying transport processes after detachment

  7. Divertor IR thermography on Alcator C-Moda)

    Science.gov (United States)

    Terry, J. L.; LaBombard, B.; Brunner, D.; Payne, J.; Wurden, G. A.

    2010-10-01

    Alcator C-Mod is a particularly challenging environment for thermography. It presents issues that will similarly face ITER, including low-emissivity metal targets, low-Z surface films, and closed divertor geometry. In order to make measurements of the incident divertor heat flux using IR thermography, the C-Mod divertor has been modified and instrumented. A 6° toroidal sector has been given a 2° toroidal ramp in order to eliminate magnetic field-line shadowing by imperfectly aligned divertor tiles. This sector is viewed from above by a toroidally displaced IR camera and is instrumented with thermocouples and calorimeters. The camera provides time histories of surface temperatures that are used to compute incident heat-flux profiles. The camera sensitivity is calibrated in situ using the embedded thermocouples, thus correcting for changes and nonuniformities in surface emissivity due to surface coatings.

  8. EMC3-EIRENE modeling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lore, J.D., E-mail: lorejd@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Reinke, M.L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); LaBombard, B. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Lipschultz, B. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Churchill, R.M. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Feng, Y. [Max Planck Institute for Plasma Physics, Greifswald (Germany)

    2015-08-15

    Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ∼50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modeling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.

  9. Density limit and cross-field edge transport scaling in Alcator C-Mod

    International Nuclear Information System (INIS)

    LaBombard, B.

    2002-01-01

    Experiments in Alcator C-Mod have uncovered a direct link between the character and scaling of edge transport and the empirical Greenwald density limit (n G ). In low to moderate density discharges, the scrape-off layer (SOL) exhibits a two-layer structure: a near SOL (∼5 mm zone) with steep density and temperature gradients and a far SOL with flatter profiles. In the far SOL, the transport fluxes exhibit large transport events ('bursts' which carry particles to main-chamber structures. In the near SOL, transport fluxes appear to be less 'bursty' particle diffusivities in this region is found to increase strongly with local plasma collisionality. As n/n G (or collisionality) is raised, cross-field heat convection begins to compete with parallel conduction to the divertor. At N/n G ∼0.5, T E at the separatrix is reduced. As n/n G approaches ∼1, regions inside the separatrix exhibit flatter profiles with 'bursty' transport behavior; cross-field heat convection to main-chamber structures becomes comparable to the radiated power. Thus as n/n G is increased, cross-field edge transport physics progressively changes, ultimately impacting the power balance of the discharge near N/n G ∼1. (author)

  10. High-resolution disruption halo current measurements using Langmuir probes in Alcator C-Mod

    Science.gov (United States)

    Tinguely, R. A.; Granetz, R. S.; Berg, A.; Kuang, A. Q.; Brunner, D.; LaBombard, B.

    2018-01-01

    Halo currents generated during disruptions on Alcator C-Mod have been measured with Langmuir ‘rail’ probes. These rail probes are embedded in a lower outboard divertor module in a closely-spaced vertical (poloidal) array. The dense array provides detailed resolution of the spatial dependence (~1 cm spacing) of the halo current distribution in the plasma scrape-off region with high time resolution (400 kHz digitization rate). As the plasma limits on the outboard divertor plate, the contact point is clearly discernible in the halo current data (as an inversion of current) and moves vertically down the divertor plate on many disruptions. These data are consistent with filament reconstructions of the plasma boundary, from which the edge safety factor of the disrupting plasma can be calculated. Additionally, the halo current ‘footprint’ on the divertor plate is obtained and related to the halo flux width. The voltage driving halo current and the effective resistance of the plasma region through which the halo current flows to reach the probes are also investigated. Estimations of the sheath resistance and halo region resistivity and temperature are given. This information could prove useful for modeling halo current dynamics.

  11. Novel energy resolving x-ray pinhole camera on Alcator C-Moda)

    Science.gov (United States)

    Pablant, N. A.; Delgado-Aparicio, L.; Bitter, M.; Brandstetter, S.; Eikenberry, E.; Ellis, R.; Hill, K. W.; Hofer, P.; Schneebeli, M.

    2012-10-01

    A new energy resolving x-ray pinhole camera has been recently installed on Alcator C-Mod. This diagnostic is capable of 1D or 2D imaging with a spatial resolution of ≈1 cm, an energy resolution of ≈1 keV in the range of 3.5-15 keV and a maximum time resolution of 5 ms. A novel use of a Pilatus 2 hybrid-pixel x-ray detector [P. Kraft et al., J. Synchrotron Rad. 16, 368 (2009), 10.1107/S0909049509009911] is employed in which the lower energy threshold of individual pixels is adjusted, allowing regions of a single detector to be sensitive to different x-ray energy ranges. Development of this new detector calibration technique was done as a collaboration between PPPL and Dectris Ltd. The calibration procedure is described, and the energy resolution of the detector is characterized. Initial data from this installation on Alcator C-Mod is presented. This diagnostic provides line-integrated measurements of impurity emission which can be used to determine impurity concentrations as well as the electron energy distribution.

  12. Transport phenomena in the edge of Alcator C-Mod plasmas

    International Nuclear Information System (INIS)

    Terry, J.L.; Basse, N.P.; Cziegler, I.; Greenwald, M.; LaBombard, B.; Edlund, E.M.; Hughes, J.W.; Lin, L.; Lin, Y.; Porkolab, M.; Veto, B.; Wukitch, S.J.; Grulke, O.; Zweben, S.J.; Sampsell, M.

    2005-01-01

    Two aspects of edge turbulence and transport in Alcator C-Mod are explored. The quasi-coherent mode, an edge fluctuation present in Enhanced Da H-mode plasmas, is examined with regard to its role in the enhanced particle transport found in these plasmas, its in/out asymmetry, its poloidal wave number, and its radial width and location. It is shown to play a dominant role in the perpendicular particle transport. The QCM is not observed at the inboard midplane, indicating that its amplitude there is significantly smaller than on the outboard side. The peak amplitude of the QCM is found just inside the separatrix, with a radial width ≥5 mm, leading to a non-zero amplitude outside the separatrix and qualitatively consistent with its transport enhancement. Also examined are the characteristics of the intermittent convective transport, associated with 'blobs' and typically occurring in the scrape-off-layer. The blobs are qualitatively similar in L- and H-mode. When their sizes, occurrence frequencies, and magnitudes are compared, it is found that the blob size may be somewhat smaller in ELMfree H-Mode, and blob frequency is similar. A clear difference is seen in the blob magnitude in the far SOL, with ELMfree H-mode showing a smaller perturbation there than L-mode. As the Greenwald density limit is approached (n/n GW ≥0.7), blobs are seen inside the separatrix, consistent with the observation that the high cross-field transport region, normally found in the far scrape-off, penetrates the closed flux surfaces at high n/n GW . (author)

  13. SOLPS-ITER Study of neutral leakage and drift effects on the alcator C-Mod divertor plasma

    Directory of Open Access Journals (Sweden)

    W. Dekeyser

    2017-08-01

    Full Text Available As part of an effort to validate the edge plasma model in the SOLPS-ITER code suite under ITER-relevant divertor plasma and neutral conditions, we report on progress in the modeling of the Alcator C-Mod divertor plasma with the new code. We perform simulations with a complete drifts model and kinetic neutrals, including effects of neutral viscosity, ion-molecule collisions and Lyα-opaque conditions, but assuming a pure deuterium plasma. Through a series of simulations with varying divertor geometries, we show the importance of including neutal leakage paths through the divertor substructure on the divertor plasma solution. Moreover, the impact of drifts on inner-outer target asymmetries is assessed. Including both effects, we achieve excellent agreement between simulations and upstream and outer target Langmuir Probe data. In absence of strong volumetric losses due to e.g. impurity radiation in our simulations, the strong inner target detachment observed experimentally remains elusive in our modeling at present.

  14. Study of volume recombination and radiation opacity effects in Alcator C-Mod

    International Nuclear Information System (INIS)

    Terry, J.L.; Lipschultz, B.; Pigarov, A.Y.; Boswell, C.; Krasheninnikov, S.I.; LaBombard, B.; Pappas, D.A.

    1998-01-01

    Observations of significant volume recombination within the Alcator C-Mod divertor plasma and in the edge plasma (MARFE) are described. The recombination occurs in regions where T e approx-lt 1 eV and n e approx-gt 1x10 21 m -3 . The determinations of the recombination rates are made by measuring the D 0 Lyman and/or Balmer spectra and by using a collisional radiative model describing the level populations, ionization and recombination of D 0 . In regions of strong recombination the upper levels (n approx-gt 4) populations are close to those determined by Saha-Boltzmann distribution and are independent of the ground state density. Thus the intensities of lines from these levels are related to the recombination rate, and curves determining the number of open-quote recombinations per photon close-quote are calculated. Ly β line emission is shown to be trapped in some cases, meaning that Ly α can be strongly trapped. Since opacity affects the recombination rates, the effects of the trapping of Ly α,β photons on the open-quote recombinations per photon close-quote curves are calculated and considered in the recombination rate determinations. Total recombination rates in the detached divertor plasma and in MARFEs located at the periphery of the main plasma are determined. Recombination can be a significant sink for ions. copyright 1998 American Institute of Physics

  15. Integrated numerical design of an innovative Lower Hybrid launcher for Alcator C-Mod

    International Nuclear Information System (INIS)

    Meneghini, O.; Shiraiwa, S.; Beck, W.; Irby, J.; Koert, P.; Parker, R. R.; Viera, R.; Wukitch, S.; Wilson, J.

    2009-01-01

    The new Alcator C-Mod LHCD system (LH2) is based on the concept of a four way splitter [1] which evenly splits the RF power among the four waveguides that compose one of the 16 columns of the LH grill. In this work several simulation tools have been used to study the LH2 coupling performance and the launched spectra when facing a plasma, numerically verifying the effectiveness of the four way splitter concept and further improving its design. The TOPLHA code has been used for modeling reflections at the antenna/plasma interface. TOPLHA results have been then coupled to the commercial code CST Microwave Studio to efficiently optimize the four way splitter geometry for several plasma scenarios. Subsequently, the COMSOL Multiphysics code has been used to self consistently take into account the electromagnetic-thermal-structural interactions. This comprehensive and predictive analysis has proven to be very valuable for understanding the behavior of the system when facing the plasma and has profoundly influenced several design choices of the LH2. According to the simulations, the final design ensures even poloidal power splitting for a wide range of plasma parameters, which ultimately results in an improvement of the wave coupling and an increased maximum operating power.

  16. Fluctuating zonal flows in the I-mode regime in Alcator C-Moda)

    Science.gov (United States)

    Cziegler, I.; Diamond, P. H.; Fedorczak, N.; Manz, P.; Tynan, G. R.; Xu, M.; Churchill, R. M.; Hubbard, A. E.; Lipschultz, B.; Sierchio, J. M.; Terry, J. L.; Theiler, C.

    2013-05-01

    Velocity fields and density fluctuations of edge turbulence are studied in I-mode [F. Ryter et al., Plasma Phys. Controlled Fusion 40, 725 (1998)] plasmas of the Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] tokamak, which are characterized by a strong thermal transport barrier in the edge while providing little or no barrier to the transport of both bulk and impurity particles. Although previous work showed no clear geodesic-acoustic modes (GAM) on C-Mod, using a newly implemented, gas-puff-imaging based time-delay-estimate velocity inference algorithm, GAM are now shown to be ubiquitous in all I-mode discharges examined to date, with the time histories of the GAM and the I-mode specific [D. Whyte et al., Nucl. Fusion 50, 105005 (2010)] Weakly Coherent Mode (WCM, f = 100-300 kHz, Δf/f≈0.5, and kθ≈1.3 cm-1) closely following each other through the entire duration of the regime. Thus, the I-mode presents an example of a plasma state in which zero frequency zonal flows and GAM continuously coexist. Using two-field (density-velocity and radial-poloidal velocity) bispectral methods, the GAM are shown to be coupled to the WCM and to be responsible for its broad frequency structure. The effective nonlinear growth rate of the GAM is estimated, and its comparison to the collisional damping rate seems to suggest a new view on I-mode threshold physics.

  17. Transport Studies in Alcator C-Mod ITB Plasmas

    Science.gov (United States)

    Fiore, C. L.; Bonoli, P. T.; Ernst, D.; Greenwald, M. J.; Ince-Cushman, A.; Lin, L.; Marmar, E. S.; Porkolab, M.; Rice, J. E.; Wukitch, S.; Rowan, W.; Bespamyatnov, I.; Phillips, P.

    2008-11-01

    Internal transport barriers occur in C-Mod plasmas that have off-axis ICRF heating and also in Ohmic H-mode plasmas. These ITBs are marked by highly peaked density and pressure profiles, as they rely on a reduction of particle and thermal flux in the barrier region which allows the neoclassical pinch to peak the central density without reducing the central temperature. Enhancement of several core diagnostics has resulted in increased understanding of C-Mod ITBs. Ion temperature profile measurements have been obtained using an innovative design for x-ray crystal spectrometry and clearly show a barrier forming in the ion temperature profile. The phase contrast imaging (PCI) provides limited localization of the ITB related fluctuations that increase in strength as the central density increases. Simulation of triggering conditions, integrated simulations with fluctuation measurements, parametric studies, and transport implications of fully ionized boron impurity profiles in the plasma are under study. A summary of these results will be presented.

  18. Parallel transport studies of high-Z impurities in the core of Alcator C-Mod plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Reinke, M. L.; Hutchinson, I. H.; Rice, J. E.; Greenwald, M.; Howard, N. T.; Hubbard, A.; Hughes, J. W.; Terry, J. L.; Wolfe, S. M. [MIT-Plasma Science and Fusion Center Cambridge, Massachusetts 02139 (United States)

    2013-05-15

    Measurements of poloidal variation, ñ{sub z}/, in high-Z impurity density have been made using photodiode arrays sensitive to vacuum ultraviolet and soft x-ray emission in Alcator C-Mod plasmas. In/out asymmetries in the range of −0.2<0.3 are observed for r/a<0.8, and accumulation on both the high-field side, n{sub z,cos}<0, and low-field side, n{sub z,cos}>0, of a flux surface is found to be well described by a combination of centrifugal, poloidal electric field, and ion-impurity friction effects. Up/down asymmetries, −0.05<0.10, are observed over 0.50 corresponding to accumulation opposite the ion ∇B drift direction. Measurements of the up/down asymmetry of molybdenum are found to disagree with predictions from recent neoclassical theory in the trace limit, n{sub z}Z{sup 2}/n{sub i}≪1. Non-trace levels of impurities are expected to modify the main-ion poloidal flow and thus change friction-driven impurity density asymmetries and impurity poloidal rotation, v{sub θ,z}. Artificially modifying main-ion flow in parallel transport simulations is shown to impact both ñ{sub z}/ and v{sub θ,z}, but simultaneous agreement between measured and predicted up/down and in/out asymmetry as well as impurity poloidal rotation is not possible for these C-Mod data. This link between poloidal flow and poloidal impurity density variation outlines a more stringent test for parallel neoclassical transport theory than has previously been performed. Measurement and computational techniques specific to the study of poloidal impurity asymmetry physics are discussed as well.

  19. Experimental studies of edge turbulence and confinement in Alcator C-Moda)

    Science.gov (United States)

    Cziegler, I.; Terry, J. L.; Hughes, J. W.; LaBombard, B.

    2010-05-01

    The steep gradient edge region and scrape-off-layer (SOL) on the low-field-side of Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] tokamak plasmas are studied using gas-puff-imaging diagnostics. In L-mode plasmas, the region extending ˜2 cm inside the magnetic separatrix has fluctuations showing a broad, turbulent spectrum, propagating in the electron diamagnetic drift direction, whereas features in the open field line region propagate in the ion diamagnetic drift direction. This structure is robust against toroidal field strength, poloidal null-point geometry, plasma current, and plasma density. Global parameter dependence of spectral and spatial structure of the turbulence inside the separatrix is explored and characterized, and both the intensity and spectral distributions are found to depend strongly on the plasma density normalized to the tokamak density limit. In H-mode discharges the fluctuations at and inside the magnetic separatrix show fundamentally different trends compared to L-mode, with the electron diamagnetic direction propagating turbulence greatly reduced in ELM-free [F. Wagner et al., Proceedings of the Thirteenth Conference on Plasma Physics and Controlled Nuclear Fusion Research (IAEA, Vienna, 1982), Vol. I, p. 277], and completely dominated by the modelike structure of the quasicoherent mode in enhanced D-alpha regimes [A. E. Hubbard, R. L. Boivin, R. S. Granetz et al., Phys. Plasmas 8, 2033 (2001)], while the normalized SOL turbulence is largely unaffected.

  20. FIR laser scattering and heterodyne receiver measurements on Alcator C

    International Nuclear Information System (INIS)

    Woskoboinikow, P.; Praddaude, H.C.; Mulligan, W.J.; Cohn, D.R.; Lax, B.

    1982-01-01

    The MIT program to develop high power collective Thomson scattering diagnostics is presented. The D 2 O laser Thomson scattering system is operational on Alcator C tokamak. The major components include a 0.5 MW, 150 ns D 2 O laser, a heterodyne receiver mixer, a 25 MW, 381 μ DCOOD laser local oscillator and X-band I.F. electronics including a 32 channel multiplexer filter centered at 9.4 GHz with 80 MHz wide channels. Initial scattering measurement showed high level of stray D 2 O laser power. The spectrum was obtained by operating the Thomson scattering diagnostics with no plasma in the tokamak. An X-band notch filter was placed after the Schottky diode mixer to reject a 240 MHz band centered at 9.4 GHz. The stray light level was reduced by 16 to 20 db. Other sources of background noise such as strong non-thermal scattering and ECE did not appear to be a problem. A gas filled cell was placed on the Alcator C scattering system to reduce the level of stray light. Work is underway to improve the transverse mode quality of the laser and receiver to improve matching to the beam and viewing dumps. (Kato, T.)

  1. A Study of Electron Modes in Off-axis Heated Alcator C-Mod Plasmas

    Science.gov (United States)

    Fiore, C. L.; Ernst, D. R.; Mikkelsen, D.; Ennever, P. C.; Howard, N. T.; Gao, C.; Reinke, M. L.; Rice, J. E.; Hughes, J. W.; Walk, J. R.

    2013-10-01

    Understanding the underlying physics and stability of the peaked density internal transport barriers (ITB) that have been observed during off-axis ICRF heating of Alcator C-Mod plasmas is the goal of recent gyro-kinetic simulations. Two scenarios are examined: an ITB plasma formed with maximal (4.5 MW) off-axis heating power; also the use of off-axis heating in an I-mode plasma as a target in the hopes of establishing an ITB. In the former, it is expected that evidence of trapped electron mode instabilities could be found if a sufficiently high electron temperature is achieved in the core. Linear simulations show unstable modes are present across the plasma core from r/a = 0.2 and greater. In the latter case, despite establishing similar conditions to those in which ITBS were formed, none developed in the I-mode plasmas. Linear gyrokinetic analyses show no unstable ion modes at r/a < 0.55 in these I-mode plasmas, with both ITG and ETG modes present beyond r/a = 0.65. The details of the experimental results will be presented. Linear and non-linear simulations of both of these cases will attempt to explore the underlying role of electron and ion gradient driven instabilities to explain the observations. This work was supported by US-DoE DE-FC02-99ER54512 and DE-AC02-09CH11466.

  2. Measurement of LHCD edge power deposition through modulation techniques on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Faust, I. C.; Brunner, D.; LaBombard, B.; Parker, R. R.; Baek, S. G.; Chilenksi, M. A.; Hubbard, A.; Hughes, J. W.; Terry, J. L.; Shiraiwa, S.; Walk, J. R.; Wallace, G. M.; Whyte, D. G. [MIT Plasma Science and Fusion Center, Cambridge, MA USA (United States); Edlund, E. [Princeton Plasma Physics Laboratory, Princeton, NJ USA (United States)

    2015-12-10

    The efficiency of LHCD on Alcator C-Mod drops exponentially with line average density. At reactor relevant densities (> 1 · 1020 [m{sup −3}]) no measurable current is driven. While a number of causes have been suggested, no specific mechanism has been shown to be responsible for the loss of current drive at high density. Fast modulation of the LH power was used to isolate and quantify the LHCD deposition within the plasma. Measurements from these plasmas provide unique evidence for determining a root cause. Modulation of LH power in steady plasmas exhibited no correlated change in the core temperature. A correlated, prompt response in the edge suggests that the loss in efficiency is related to a edge absorption mechanism. This follows previous results which found the generation of n{sub ||}-independent SOL currents. Multiple Langmuir probe array measurements of the conducted heat conclude that the lost power is deposited near the last closed flux surface. The heat flux induced by LH waves onto the outer divertor is calculated. Changes in the neutral pressure, ionization and hard X-ray emission at high density highlight the importance of the active divertor in the loss of efficiency. Results of this study implicate a mechanism which may occur over multiple passes, leading to power absorption near the LCFS.

  3. C-Mod Collaboration Informal Technical Progress Report

    International Nuclear Information System (INIS)

    Kenneth W. Gentle

    2007-01-01

    The aims of the collaboration have not changed. A specific list of tasks was agreed upon during the Fall of 2006 in preparation for the 2007 C-Mod campaign by Earl Marmar, Head of the Alcator Project, Kenneth Gentle, Principal Investigator, and William Rowan, Collaboration Coordinator with the facilitation of Adam Rosenberg (DOE grant monitor for the collaboration). The activities follow the list of tasks and are discussed in this progress report

  4. Detached divertor plasmas in Alcator C-Mod: A study of the role of atomic physics

    International Nuclear Information System (INIS)

    Lipschultz, B.; Boswell, C.; Goetz, J.A.

    1999-01-01

    Detailed profiles of the volumetric recombination occurring in Alcator C-Mod plasmas are presented. During detachment the recombination sink is compared to the divertor plate sink as well as the divertor ion source. Depending on plasma conditions, volume recombination removes between 10 and 75% of the ions before they reach the plates. A second, equally important process that leads to a drop in plate ion current is inferred to be a reduction in divertor ion source, which is correlated with a drop in power flowing into the ionization region and the pressure loss of detachment. For high n e the divertor recombination can cross the separatrix near the x-point, cool the core and lead to a disruption. Experimental measurements show a difference in ion and neutral velocities for H-mode detached plasmas. The resulting ion-neutral collisions are found to be more efficacious than recombination in removing momentum from the ions. The neutral component of volumetric power emission from the divertor has been measured by means of a novel filtering technique to be substantial (∼ 20% of the total divertor volumetric emission). (author)

  5. Diagnosis of mildly relativistic electron velocity distributions by electron cyclotron emission in the Alcator C tokamak

    International Nuclear Information System (INIS)

    Kato, K.

    1986-09-01

    Mildly relativistic electron velocity distributions are diagnosed from measurements of the first few electron cyclotron emission harmonics in the Alcator C tokamak. The approach employs a vertical viewing chord through the center of the tokamak plasma terminating at a compact, high-performance viewing dump. The cyclotron emission spectra obtained in this way are dominated by frequency downshifts due to the relativistic mass increase, which discriminates the electrons by their total energy. In this way a one-to-one correspondence between the energy and the emission frequency is accomplished in the absence of harmonic superpositions. The distribution, described by f/sub p/, the line-averaged phase space density, and Λ, the anisotropy factor, is determined from the ratio of the optically thin harmonics or polarizations. Diagnosis of spectra in the second and the third harmonic range of frequencies obtained during lower hybrid heating, current drive, and low density ohmic discharges are carried out, using different methods depending on the degree of harmonic superposition present in the spectrum and the availability of more than one ratio measurement. Discussions of transient phenomena, the radiation temperature measurement from the optically thick first harmonic, and the measurements compared to the angular hard x-ray diagnostic results illuminate the capabilities of the vertically viewing electron cyclotron emission diagnostic

  6. Ohmic ignition of Neo-Alcator tokamak with adiabatic compression

    International Nuclear Information System (INIS)

    Inoue, Nobuyuki; Ogawa, Yuichi

    1992-01-01

    Ohmic ignition condition on axis of the DT tokamak plasma heated by minor radius and major radius adiabatic compression is studied assuming parabolic profiles for plasma parameters, elliptic plasma cross section, and Neo-Alcator confinement scaling. It is noticeable that magnetic compression reduces the necessary total plasma current for Ohmic ignition device. Typically in compact ignition tokamak of the minor radius of 0.47 m, major radius of 1.5 m and on-axis toroidal field of 20 T, the plasma current of 6.8 MA is sufficient for compression plasma, while that of 11.7 MA is for no compression plasma. Another example with larger major radius is also described. In such a device the large flux swing of Ohmic transformer is available for long burn. Application of magnetic compression saves the flux swing and thereby extends the burn time. (author)

  7. Ion Bernstein wave experiments on the Alcator C tokamak

    International Nuclear Information System (INIS)

    Moody, J.D.

    1988-09-01

    Ion Bernstein wave experiments are carried out on the Alcator C tokamak to study wave excitation, propagation, absorption, and plasma heating due to wave power absorption. It is shown that ion Bernstein wave power is coupled into the plasma and follows the expected dispersion relation. The antenna loading is maximized when the hydrogen second harmonic layer is positioned just behind the antenna. Plasma heating results at three values of the toroidal magnetic field are presented. Central ion temperature increases of ΔT/sub i//Ti /approx lt/ 0.1 and density increases Δn/n 6 s/sup /minus/1/ for plasmas within the density range 0.6 /times/ 10 20 m/sup /minus/3/ ≤ /bar n//sub e/ ≤ 4 /times/ 10 20 m/sup /minus/3/ and magnetic fields 2.4 ≥ ω/Ω/sub H/ ≥ 1.1. The density increases is usually accompanied by an improvement in the global particle confinement time relative to the Ohmic value. The ion heating rate is measured to be ΔT/sub i//P/sub rf/ ≅ 2-4.5 eV/kW at low densities. At higher densities /bar n//sub e/ ≤ 1.5 /times/ 10 20 m/sup /minus/3/ the ion heating rate dramatically decreases. It is shown that the decrease in the ion heating rate can be explained by the combined effects of wave scattering through the edge turbulence and the decreasing on energy confinement of these discharges with density. The effect of observed edge turbulence is shown to cause a broadening of the rf power deposition profile with increasing density. It is shown that the inferred value of the Ohmic ion thermal conduction, when compared to the Chang-Hinton neoclassical prediction, exhibits an increasing anomaly with increasing plasma density

  8. Nonaxisymmetric field effects on Alcator C-Moda)

    Science.gov (United States)

    Wolfe, S. M.; Hutchinson, I. H.; Granetz, R. S.; Rice, J.; Hubbard, A.; Lynn, A.; Phillips, P.; Hender, T. C.; Howell, D. F.; La Haye, R. J.; Scoville, J. T.

    2005-05-01

    A set of external coils (A-coils) capable of producing nonaxisymmetric, predominantly n =1, fields with different toroidal phase and a range of poloidal mode m spectra has been used to determine the threshold amplitude for mode locking over a range of plasma parameters in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda, P. Bonoli, S. Fairfax, C. Fiore, J. Goetz, S. Golovato, R. Granetz, M. Greenwald et al., Phys. Plasmas 1, 1511 (1994)]. The threshold perturbations and parametric scalings, expressed in terms of (B21/BT), are similar to those observed on larger, lower field devices. The threshold is roughly linear in density, with typical magnitudes of order 10-4. This result implies that locked modes should not be significantly more problematic for the International Thermonuclear Experimental Reactor [I. P. B. Editors, Nucl. Fusion 39, 2286 (1999)] than for existing devices. Coordinated nondimensional identity experiments on the Joint European Torus [Fusion Technol. 11, 13 (1987)], DIII-D [Fusion Technol. 8, 441 (1985)], and C-Mod, with matching applied mode spectra, have been carried out to determine more definitively the field and size scalings. Locked modes on C-Mod are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement, frequently leading to disruptions. Intrinsic error fields inferred from the threshold studies are found to be consistent in amplitude and phase with a comprehensive model of the sources of field errors based on "as-built" coil and bus-work details and coil imperfections inferred from measurements using in situ magnetic diagnostics on dedicated test pulses. Use of the A-coils to largely cancel the 2/1 component of the intrinsic nonaxisymmetric field has led to expansion of the accessible operating space in C-Mod, including operation up to 2 MA plasma current at 8 T.

  9. Measurements of Mode Converted Ion Cyclotron Wave with Phase Contrast Imaging in Alcator C-Mod and Comparisons with Synthetic PCI Simulations in TORIC

    International Nuclear Information System (INIS)

    Tsujii, N.; Porkolab, M.; Edlund, E. M.; Lin, L.; Lin, Y.; Wright, J. C.; Wukitch, S. J.

    2009-01-01

    Mode converted ion cyclotron wave (ICW) has been observed with phase contrast imaging (PCI) in D- 3 He plasmas in Alcator C-Mod. The measurements were carried out with the optical heterodyne technique using acousto-optic modulators which modulate the CO2 laser beam intensity near the ion cyclotron frequency. With recently improved calibration of the PCI system using a calibrated sound wave source, the measurements have been compared with the full-wave code TORIC, as interpreted by a synthetic diagnostic. Because of the line-integrated nature of the PCI signal, the predictions are sensitive to the exact wave field pattern. The simulations are found to be in qualitative agreement with the measurements.

  10. Plasma profiles and flows in the high-field side scrape-off layer in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Smick, N. [MIT Plasma Science and Fusion Center, NW17-170, 175 Albany St., Cambridge, MA 02139 (United States)]. E-mail: nsmick@mit.edu; LaBombard, B. [MIT Plasma Science and Fusion Center, NW17-170, 175 Albany St., Cambridge, MA 02139 (United States); Pitcher, C.S. [132 Bowood Ave., Toronto, M4N1Y5 (Canada)

    2005-03-01

    A novel, magnetically-driven swing probe was recently installed near the midplane on the high-field side SOL in Alcator C-Mod. The probe collects plasma from co- and counter-current directions during its respective 0-90 deg and 90-180 deg of motion, thus providing profiles of density, electron temperature and plasma flow parallel to magnetic field lines (Mach number, M{sub parallel}) up to the separatrix. Results are reported from discharges with different magnetic topologies: lower single-null, upper single-null, and double-null. In single-null, a strong parallel flow (vertical bar M{sub parallel} vertical bar {approx} 1) is detected, which is always directed from the low- to high-field SOL. In double-null discharges, e-folding lengths in the high-field SOL are a factor of {approx}4 shorter than the low-field SOL. Thus, plasma appears to 'fill-in' the high-field SOL in single-null plasmas, not by cross-field transport but by parallel flow from the low-field SOL - a picture consistent with a very strong ballooning-like component to the cross-field transport.

  11. Massachusetts Institute of Technology, Plasma Fusion Center FY97--FY98 work proposal

    International Nuclear Information System (INIS)

    1996-03-01

    Alcator C-Mod is the high-field, high-density divertor tokamak in the world fusion program. It is one of five divertor experiments capable of plasma currents exceeding one megamp. Because of its compact dimensions, Alcator C-Mod investigates an essential area in parameter space, which complements the world's larger experiments, in establishing the tokamak physics database. Three key areas of investigation have been called out in which Alcator C-Mod has a vital role to play: (1) divertor research on C-Mod takes advantage of the advanced divertor shaping, the very high scrap-off-layer power density, unique abilities in impurity diagnosis, and the High-Z metal wall, to advance the physics understanding of this critical topic; (2) in transport studies, C-Mod is making critical tests of both empirical scalings and theoretically based interpretations of tokamak transport, at dimensional parameters that are unique but dimensionless parameters often comparable to those in much larger experiments; (3) in the area of Advanced Tokamak research, so important to concept optimization, the high-field design of the device also provides long pulse length, compared to resistive skin time, which provides an outstanding opportunity to investigate the extent to which enhanced confinement and stability can be sustained in steady-state, using active profile control. In addition to these main programmatic emphasis, important enabling research is being performed in MHD stability and control, which has great significance for the immediate design of ITER, and in the physics and engineering of ICRF, which is the main auxiliary heating method on C-Mod

  12. EUV impurity study of the Alcator tokamak

    International Nuclear Information System (INIS)

    Terry, J.L.; Chen, K.I.; Moos, H.W.; Marmar, E.S.

    1978-01-01

    The intensity of resonance line radiation from oxygen, nitrogen, carbon and molybdenum impurities has been measured in the high-field (80kG), high-density (6x10 14 cm -3 ) discharges of the Alcator Tokamak, using a 0.4-m normal-incidence monochromator (300-1300A) with its line of sight fixed along a major radius. Total light-impurity concentrations of a few tenths of a percent have been estimated by using both a simple model and a computer code which included Pfirsch-Schlueter impurity diffusion. The resulting values of Zsub(eff), including the contributions due to both the light impurities and molybdenum, were close to one. The power lost through the impurity line radiation from the lower ionization states accounted for approximately 10% of the total Ohmic input power at high densities. (author)

  13. Poloidal asymmetries in the scrape-off layer plasma of the Alcator C tokamak

    International Nuclear Information System (INIS)

    LaBombard, B.; Lipschultz, B.

    1987-01-01

    Large poloidal asymmetries in density, electron temperature, radial density e-folding length and floating potential have been measured in the plasma existing between the limiter radius and the wall of the Alcator C tokamak. Typically, variations in density by factors of about 4-20 and variations in radial density e-folding length by factors of about 3-8 are recorded in discharges which are bounded by poloidally symmetric ring limiters. These poloidal asymmetries show that pressure is a function of poloidal angle on open magnetic flux surfaces in this region of the plasma. Observations of toroidally symmetric MARFE (multifaceted asymmetric radiation from the edge) phenomena further imply that density and perhaps pressure are also a function of poloidal angle on closed flux surfaces existing just inside the limiter radius. The magnitude of these poloidal asymmetries and their dependence on poloidal angle persists independent of machine parameters (central plasma density, plasma current, toroidal field, MARFE versus non-MARFE discharges). Analysis of the data indicates that these asymmetries are caused by poloidal variations in perpendicular particle and heat transport in both the main plasma and the scrape-off layer. A number of possible asymmetric perpendicular transport processes in the scrape-off layer plasma are examined, including diffusion and E-vectorxB-vector plasma convection. (author)

  14. 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

  15. Viewgraphs presented at the ASDEX/DOE workshop on disruptions in divertor tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Granetz, R.; Gruber, O.; Zohm, H. [and others

    1994-09-01

    The emphasis of this year`s ASDEX/DOE workshop was on disruptions in diverted tokamaks. The meeting was held here at MIT on 14--15 March. It is particularly appropriate that MIT hosted the workshop this year, since Alcator C-Mod had just recently completed its very first run campaign, and disruptions are one of the key areas of research in our program. There were a total of 14 speakers, with participants from IPP (Garching), CRPP (Lausanne), Culham, General Atomics, PPPL, Sandia, ORNL, the ITER JCT, and MIT. The subjects addressed included statistical analysis of disruption probabilities in ASDEX, modelling of the vertical axisymmetric plasma motion in DIII-D, impact of disruptions on the design of the ITER divertors, modelling of runaway electrons, and TSC calculations of disruption-induced currents and forces in TPX, etc. One item of particular interest to us was the experimental correlation of halo current magnitude with plasma current on ASDEX-Upgrade. The data indicates at least a linear, and possibly even a quadractic dependence. This has important implications for Alcator C-Mod, since it would predict halo currents of order 1 MA or more at full performance. At the conclusion of the talks, an informal discussion of disruption databases was held, primarily for the purpose of helping us develop a useful one for C-Mod.

  16. Viewgraphs presented at the ASDEX/DOE workshop on disruptions in divertor tokamaks

    International Nuclear Information System (INIS)

    Granetz, R.; Gruber, O.; Zohm, H.

    1994-01-01

    The emphasis of this year's ASDEX/DOE workshop was on disruptions in diverted tokamaks. The meeting was held here at MIT on 14--15 March. It is particularly appropriate that MIT hosted the workshop this year, since Alcator C-Mod had just recently completed its very first run campaign, and disruptions are one of the key areas of research in our program. There were a total of 14 speakers, with participants from IPP (Garching), CRPP (Lausanne), Culham, General Atomics, PPPL, Sandia, ORNL, the ITER JCT, and MIT. The subjects addressed included statistical analysis of disruption probabilities in ASDEX, modelling of the vertical axisymmetric plasma motion in DIII-D, impact of disruptions on the design of the ITER divertors, modelling of runaway electrons, and TSC calculations of disruption-induced currents and forces in TPX, etc. One item of particular interest to us was the experimental correlation of halo current magnitude with plasma current on ASDEX-Upgrade. The data indicates at least a linear, and possibly even a quadractic dependence. This has important implications for Alcator C-Mod, since it would predict halo currents of order 1 MA or more at full performance. At the conclusion of the talks, an informal discussion of disruption databases was held, primarily for the purpose of helping us develop a useful one for C-Mod

  17. Study of the L-mode tokamak plasma “shortfall” with local and global nonlinear gyrokinetic δf particle-in-cell simulation

    Energy Technology Data Exchange (ETDEWEB)

    Chowdhury, J.; Wan, Weigang; Chen, Yang; Parker, Scott E. [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); Holland, C. [University of California at San Diego, La Jolla, California 92093 (United States); Howard, N. T. [Oak Ridge Institute for Science and Education (ORISE), Oak Ridge, Tennessee 37831 (United States)

    2014-11-15

    The δ f particle-in-cell code GEM is used to study the transport “shortfall” problem of gyrokinetic simulations. In local simulations, the GEM results confirm the previously reported simulation results of DIII-D [Holland et al., Phys. Plasmas 16, 052301 (2009)] and Alcator C-Mod [Howard et al., Nucl. Fusion 53, 123011 (2013)] tokamaks with the continuum code GYRO. Namely, for DIII-D the simulations closely predict the ion heat flux at the core, while substantially underpredict transport towards the edge; while for Alcator C-Mod, the simulations show agreement with the experimental values of ion heat flux, at least within the range of experimental error. Global simulations are carried out for DIII-D L-mode plasmas to study the effect of edge turbulence on the outer core ion heat transport. The edge turbulence enhances the outer core ion heat transport through turbulence spreading. However, this edge turbulence spreading effect is not enough to explain the transport underprediction.

  18. Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

    Science.gov (United States)

    Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; Grierson, B. A.; Staebler, G. M.; Rice, J. E.; Yuan, X.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Sciortino, F.

    2018-02-01

    A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. This Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and time scales of cold-pulse experiments in tokamak plasmas.

  19. Spectral measurements of fluctuating ω/sub pe/ radiation from Alcator C tokamak

    International Nuclear Information System (INIS)

    Gandy, R.F.; Yates, D.

    1984-01-01

    High resolution spectral measurements have been made of the fluctuating electron plasma frequency (ω/sub pe/) radiation from Alcator C. Three techniques have been used in making the measurements. Features as narrow as 350 kHz have been observed (Δf/f approx. = 6 x 10 -6 ), impling that a highly coherent process is responsible for the emission

  20. Changes in core electron temperature fluctuations across the ohmic energy confinement transition in Alcator C-Mod plasmas

    International Nuclear Information System (INIS)

    Sung, C.; White, A.E.; Howard, N.T.; Oi, C.Y.; Rice, J.E.; Gao, C.; Ennever, P.; Porkolab, M.; Parra, F.; Ernst, D.; Walk, J.; Hughes, J.W.; Irby, J.; Kasten, C.; Hubbard, A.E.; Greenwald, M.J.; Mikkelsen, D.

    2013-01-01

    The first measurements of long wavelength (k y ρ s < 0.3) electron temperature fluctuations in Alcator C-Mod made with a new correlation electron cyclotron emission diagnostic support a long-standing hypothesis regarding the confinement transition from linear ohmic confinement (LOC) to saturated ohmic confinement (SOC). Electron temperature fluctuations decrease significantly (∼40%) crossing from LOC to SOC, consistent with a change from trapped electron mode (TEM) turbulence domination to ion temperature gradient (ITG) turbulence as the density is increased. Linear stability analysis performed with the GYRO code (Candy and Waltz 2003 J. Comput. Phys. 186 545) shows that TEMs are dominant for long wavelength turbulence in the LOC regime and ITG modes are dominant in the SOC regime at the radial location (ρ ∼ 0.8) where the changes in electron temperature fluctuations are measured. In contrast, deeper in the core (ρ < 0.8), linear stability analysis indicates that ITG modes remain dominant across the LOC/SOC transition. This radial variation suggests that the robust global changes in confinement of energy and momentum occurring across the LOC/SOC transition are correlated to local changes in the dominant turbulent mode near the edge. (paper)

  1. The Fusion Science Research Plan for the Major U.S. Tokamaks. Advisory report

    International Nuclear Information System (INIS)

    1996-01-01

    In summary, the community has developed a research plan for the major tokamak facilities that will produce impressive scientific benefits over the next two years. The plan is well aligned with the new mission and goals of the restructured fusion energy sciences program recommended by FEAC. Budget increases for all three facilities will allow their programs to move forward in FY 1997, increasing their rate of scientific progress. With a shutdown deadline now established, the TFTR will forego all but a few critical upgrades and maximize operation to achieve a set of high-priority scientific objectives with deuterium-tritium plasmas. The DIII-D and Alcator C-Mod facilities will still fall well short of full utilization. Increasing the run time in vii DIII-D is recommended to increase the scientific output using its existing capabilities, even if scheduled upgrades must be further delayed. An increase in the Alcator C-Mod budget is recommended, at the expense of equal and modest reductions (~1%) in the other two facilities if necessary, to develop its capabilities for the long-term and increase its near-term scientific output.

  2. An experimental assessment of methods used to compute secondary electron emission yield for tungsten and molybdenum electrodes based on exposure to Alcator C-Mod scrape-off layer plasmas

    Science.gov (United States)

    McCarthy, W.; LaBombard, B.; Brunner, D.; Kuang, A. Q.

    2018-03-01

    Plasma potentials computed from Langmuir probe data rely on a method to account for secondary electron emission (SEE) from the electrodes. However, significant variations exist among published models for SEE and the reported experimental parameters used to evaluate them. As a means to critically assess SEE computation methods, two of four tungsten electrodes on a Langmuir-Mach probe head were replaced with molybdenum and exposed to Alcator C-Mod boundary plasmas where electron temperatures exceed 50 eV and SEE becomes significant. In this situation, plasma potentials computed for either material should be identical—the SEE evaluation method should properly account for the differences in SEE yields. Of the six methods used to compute SEE, two are found to produce consistent results (Sternglass model with Bronstein experimental parameters and Young-Dekker model with Bronstein experimental parameters). In contrast, the method previously used for C-Mod data analysis (Sternglass model with Kollath parameters) was found to be inconsistent. We have since adopted Young-Dekker-Bronstein as the preferred method.

  3. Visible continuum measurements on the Alcator C Tokamak: Changes in particle transport during pellet fuelled discharges

    International Nuclear Information System (INIS)

    Foord, M.E.

    1986-12-01

    A spatially resolving visible light detector system is used to measure continuum radiation near 5360A on the Alcator C Tokamak. For the typically hot plasmas studied, the continuum emission is found to be dominated by bremsstrahlung radiation near this wavelength region. Accurate determinations of Z/sub eff/ are obtained from continuum measurements using independently determined temperature and density measurements. Density profiles during high density, clean pellet fueled discharges, are also determined and are used to study the changes in particle transport after injection. For discharges with sufficiently large pellet density increases, density profiles are found to become more peaked following the injection. In these cases, the profiles are found to remain peaked for the remainder of the discharge, or until a ''giant'' sawtooth or minor disruption abruptly returns the profiles to a flatter pre-pellet condition. Analysis of density profiles after pellet injection yields information about the radial diffusion and convection velocity of the plasma particles. The peakedness in the density profiles, observed after pellet injection, is attributable mostly to increases in inward convection. It is concluded that neoclassical fluxes are too small to account for these changes. 70 refs., 55 figs

  4. 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)

  5. First results of the SOL reflectometer on Alcator C-Moda)

    Science.gov (United States)

    Lau, C.; Hanson, G.; Lin, Y.; Wilgen, J.; Wukitch, S.; Labombard, B.; Wallace, G.

    2012-10-01

    A swept-frequency X-mode reflectometer has been built on Alcator C-Mod to measure the scrape-off layer (SOL) density profiles adjacent to the lower hybrid launcher. The reflectometer system operates between 100 and 146 GHz at sweep rates from 10 μs to 1 ms and covers a density range of ˜1016-1020 m-3 at B0 = 5-5.4 T. This paper discusses the analysis of reflectometer density profiles and presents first experimental results of SOL density profile modifications due to the application of lower hybrid range-of-frequencies power to L-mode discharges. Comparison between density profiles measured by the X-mode reflectometer and scanning Langmuir probes is also shown.

  6. 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

  7. Workshop on High Power ICH Antenna Designs for High Density Tokamaks

    Science.gov (United States)

    Aamodt, R. E.

    1990-02-01

    A workshop in high power ICH antenna designs for high density tokamaks was held to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of RF auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made.

  8. Intermittent electron density and temperature fluctuations and associated fluxes in the Alcator C-Mod scrape-off layer

    Science.gov (United States)

    Kube, R.; Garcia, O. E.; Theodorsen, A.; Brunner, D.; Kuang, A. Q.; LaBombard, B.; Terry, J. L.

    2018-06-01

    The Alcator C-Mod mirror Langmuir probe system has been used to sample data time series of fluctuating plasma parameters in the outboard mid-plane far scrape-off layer. We present a statistical analysis of one second long time series of electron density, temperature, radial electric drift velocity and the corresponding particle and electron heat fluxes. These are sampled during stationary plasma conditions in an ohmically heated, lower single null diverted discharge. The electron density and temperature are strongly correlated and feature fluctuation statistics similar to the ion saturation current. Both electron density and temperature time series are dominated by intermittent, large-amplitude burst with an exponential distribution of both burst amplitudes and waiting times between them. The characteristic time scale of the large-amplitude bursts is approximately 15 μ {{s}}. Large-amplitude velocity fluctuations feature a slightly faster characteristic time scale and appear at a faster rate than electron density and temperature fluctuations. Describing these time series as a superposition of uncorrelated exponential pulses, we find that probability distribution functions, power spectral densities as well as auto-correlation functions of the data time series agree well with predictions from the stochastic model. The electron particle and heat fluxes present large-amplitude fluctuations. For this low-density plasma, the radial electron heat flux is dominated by convection, that is, correlations of fluctuations in the electron density and radial velocity. Hot and dense blobs contribute only a minute fraction of the total fluctuation driven heat flux.

  9. Design and operation of a high-heat flux, flush-mounted ‘rail’ Langmuir probe array on Alcator C-Mod

    Directory of Open Access Journals (Sweden)

    A.Q. Kuang

    2017-08-01

    Full Text Available A poloidal array of toroidally-extended, flush-mounted ‘rail’ Langmuir probes was recently installed on Alcator C-Mod's vertical target plate divertor. The aim was to investigate if a Langmuir probe array could be designed to survive reactor-level heat fluxes and have the ability to make measurements that could be reliably interpreted under reactor-level plasma densities, neutral densities and magnetic fields. Langmuir probes are typically built to have incident field-line angles >10° to avoid interpretation issues associated with sheath expansion. However, at the high parallel heat fluxes experienced in reactor-relevant conditions such a probe would quickly overheat and melt. To mitigate both the issues of extreme heat flux and sheath expansion, each probe was designed to be flush with the divertor surface, toroidally-extended and field-aligned, giving it a ‘rail’ geometry. The flush mounted probes have proven to be exceptionally robust surviving the 2015–2016 campaign – a first for a C-Mod probe system. Examination of the probe current-voltage (I-V characteristics reveals that they are immune to sheath expansion at incident field angles down to ∼0.5°. Comparison of the flush probes to traditional proud probes shows that both measure the same electron pressure across the divertor plate. However, there are significant and systematic differences in the density, temperature and floating potential. This suggests that there is important physics, perhaps unique to conditions in a vertical-target plate divertor with small field-line attack angles, that affects the I-V characteristics and is not currently included in probe data analyses. Finally, the probe response is examined in the ‘death-ray’ regime, just near detachment. Previous work using proud probes has suggested that the ‘death-ray’ is an artefact of the probe bias. However, on flush mounted probes the ‘death-ray’ manifests itself under different conditions, which

  10. Hydrogen pellet injection into Alcator C

    International Nuclear Information System (INIS)

    Greenwald, M.

    1983-09-01

    A four-shot pneumatic pellet injector, based on an ORNL design, has been built and operated on the Alcator C tokamak at MIT. The injector fires four independently-timed frozen hydrogen pellets with velocities in the range 8 x 10 4 - 1 x 10 5 cm/sec. Each contains 6 x 10 19 particles which corresponds to = 2 x 10 14 /cm 3 . The objectives of this experiment are to study pellet fueling and penetration, particle confinement, dependence of energy confinement on density profile and fueling mode, and edge physics and recycling as a function of fueling mode. Typical pre-injection plasmas have had anti n/sub e/ = 2 - 3 x 10 14 , Bt = 80 - 100 kG, Ip = 400 - 500 kA, T/sub e/(0) = 1200 - 1500 ev. A single pellet injected into this plasma will roughly double the electron density. Record plasma densities have been obtained by multiple injections. Line average densities in excess of 8 x 10 14 have been achieved, with highly peaked profiles. Central densities of 1.5 - 2 x 10 15 have been measured

  11. Fast Wave Transmission Measurements on Alcator C-Mod

    Science.gov (United States)

    Reardon, J.; Bonoli, P. T.; Porkolab, M.; Takase, Y.; Wukitch, S. J.

    1997-11-01

    Data are presented from an array of single-turn loop probes newly installed on the inner wall of C-Mod, directly opposite one of the two fast-wave antennas. The 8-loop array extends 32^circ in the toroidal direction at the midplane and can distinguish electromagnetic from electrostatic modes. Data are acquired by 1GHz digitizer, spectrum analyzer, and RF detector circuit. Phase measurements during different heating scenarios show evidence of both standing and travelling waves. The measurement of toroidal mode number N_tor (conserved under the assumption of axisymmetry) is used to guide the toroidal full-wave code TORIC(Brambilla, M., IPP Report 5/66, February 1996). Amplitude measurements show modulation both by Type III ELMs and sawteeth; the observed sawtooth modulation may be interpreted as due to changes in central absorption. The amplitude of tildeB_tor measured at the inner wall is compared to the prediction of TORIC.

  12. Design of a correlation electron cyclotron emission diagnostic for Alcator C-Moda)

    Science.gov (United States)

    Sung, C.; White, A. E.; Irby, J. H.; Leccacorvi, R.; Vieira, R.; Oi, C. Y.; Peebles, W. A.; Nguyen, X.

    2012-10-01

    A correlation electron cyclotron emission (CECE) diagnostic has been installed in Alcator C-Mod. In order to measure electron temperature fluctuations, this diagnostic uses a spectral decorrelation technique. Constraints obtained with nonlinear gyrokinetic simulations guided the design of the optical system and receiver. The CECE diagnostic is designed to measure temperature fluctuations which have kθ ≤ 4.8 cm-1 (kθρs < 0.5) using a well-focused beam pattern. Because the CECE diagnostic is a dedicated turbulence diagnostic, the optical system is also flexible, which allows for various collimating lenses and antenna to be used. The system overview and the demonstration of its operability as designed are presented in this paper.

  13. Investigation of lower hybrid physics through power modulation experiments on Alcator C-Moda)

    Science.gov (United States)

    Schmidt, A.; Bonoli, P. T.; Meneghini, O.; Parker, R. R.; Porkolab, M.; Shiraiwa, S.; Wallace, G.; Wright, J. C.; Harvey, R. W.; Wilson, J. R.

    2011-05-01

    Lower hybrid current drive (LHCD) is an attractive tool for off-axis current profile control in magnetically confined tokamak plasmas and burning plasmas (ITER), because of its high current drive efficiency. The LHCD system on Alcator C-Mod operates at 4.6 GHz, with ~ 1 MW of coupled power, and can produce a wide range of launched parallel refractive index (n||) spectra. A 32 chord, perpendicularly viewing hard x-ray camera has been used to measure the spatial and energy distribution of fast electrons generated by lower hybrid (LH) waves. Square-wave modulation of LH power on a time scale much faster than the current relaxation time does not significantly alter the poloidal magnetic field inside the plasma and thus allows for realistic modeling and consistent plasma conditions for different n|| spectra. Inverted hard x-ray profiles show clear changes in LH-driven fast electron location with differing n||. Boxcar binning of hard x-rays during LH power modulation allows for ~ 1 ms time resolution which is sufficient to resolve the build-up, steady-state, and slowing-down phases of fast electrons. Ray-tracing/Fokker-Planck modeling in combination with a synthetic hard x-ray diagnostic shows quantitative agreement with the x-ray data for high n|| cases. The time histories of hollow x-ray profiles have been used to measure off-axis fast electron transport in the outer half of the plasma, which is found to be small on a slowing down time scale.

  14. 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.

  15. Verification of GENE and GYRO with L-mode and I-mode plasmas in Alcator C-Mod

    Science.gov (United States)

    Mikkelsen, D. R.; Howard, N. T.; White, A. E.; Creely, A. J.

    2018-04-01

    Verification comparisons are carried out for L-mode and I-mode plasma conditions in Alcator C-Mod. We compare linear and nonlinear ion-scale calculations by the gyrokinetic codes GENE and GYRO to each other and to the experimental power balance analysis. The two gyrokinetic codes' linear growth rates and real frequencies are in good agreement throughout all the ion temperature gradient mode branches and most of the trapped electron mode branches of the kyρs spectra at r/a = 0.65, 0.7, and 0.8. The shapes of the toroidal mode spectra of heat fluxes in nonlinear simulations are very similar for kyρs ≤ 0.5, but in most cases GENE has a relatively higher heat flux than GYRO at higher mode numbers. The ratio of ion to electron heat flux is similar in the two codes' simulations, but the heat fluxes themselves do not agree in almost all cases. In the I-mode regime, GENE's heat fluxes are ˜3 times those from GYRO, and they are ˜60%-100% higher than GYRO in the L-mode conditions. The GYRO under-prediction of Qe is much reduced in GENE's L-mode simulations, and it is eliminated in the I-mode simulations. This largely improved agreement with the experimental electron heat flux is offset, however, by the large overshoot of GENE's ion heat fluxes, which are 2-3 times the experimental level, and its electron heat flux overshoot at r/a = 0.80 in the I-mode. Rotation effects can explain part of the difference between the two codes' predictions, but very significant differences remain in simulations without any rotation effects.

  16. The Role of Plasma Rotation in C-Mod Internal Transport Barriers

    Science.gov (United States)

    Fiore, C. L.; Ernst, D. R.; Rice, J. E.; Podpaly, Y.; Reinke, M. L.; Greenwald, M. J.; Hughes, J. W.; Ma, Y.; Bespamyatnov, I. O.; Rowan, W. L.

    2010-11-01

    ITBs in Alcator C-Mod featuring highly peaked density and pressure profiles are induced by injecting ICRF power with the second harmonic of the resonant frequency for minority hydrogen off-axis at the plasma half radius. These ITBs are formed in the absence of particle or momentum injection, and with monotonic q profiles with qmin ITB forms, this rotation decreases in the center of the plasma and forms a well, and often reverses direction in the core. This indicates that there is a strong EXB shearing rate in the region where the foot in the ITB density profile is observed. Preliminary gyrokinetic analyses indicate that this shearing rate is comparable to the ion temperature gradient mode (ITG) growth rate at this location and may be responsible for stabilizing the turbulence. Gyrokinetic analyses of recent experimental data obtained from a complete scan of the ICRF resonance position across the entire C-Mod plasma will be presented.

  17. Fast wave ion cyclotron resonance heating experiments on the Alcator C tokamak

    International Nuclear Information System (INIS)

    Shepard, T.D.

    1988-09-01

    Minority regime fast wave ICRF heating experiments have been conducted on the Alcator C tokamak at rf power levels sufficient to produce significant changes in plasma properties, and in particular to investigate the scaling to high density of the rf heating efficiency. Up to 450 kW of rf power at frequency f = 180 MHz, was injected into plasmas composed of deuterium majority and hydrogen minority ion species at magnetic field B 0 = 12 T, density 0.8 ≤ /bar n/sub e// ≤ 5 /times/ 10 20 m -3 , ion temperature T/sub D/(0) /approximately/ 1 keV, electron temperature T/sub e/(0) /approximately/ 1.5--2.5 keV, and minority concentration 0.25 /approx lt/ /eta/sub H// ≤ 8%. Deuterium heating ΔT/sub D/(0) = 400 eV was observed at /bar n/sub e// = 1 /times/ 10 20 m -3 , with smaller temperature increases at higher density. However, there was no significant change in electron temperature and the minority temperatures were insufficient to account for the launched rf power. Minority concentration scans indicated most efficient deuterium heating at the lowest possible concentration, in apparent contradiction with theory. Incremental heating /tau/sub inc// /equivalent to/ ΔW/ΔP up to 5 ms was independent of density, in spite of theoretical predictions of favorable density scaling of rf absorption and in stark contrast to Ohmic confinement times /tau/sub E// /equivalent to/ W/P. After accounting for mode conversion and minority losses due to toroidal field ripple, unconfined orbits, asymmetric drag, neoclassical and sawtooth transport, and charge-exchange, it was found that the losses as well as the net power deposition on deuterium do scale very favorably with density. Nevertheless, when the net rf and Ohmic powers deposited on deuterium are compared, they are found to be equally efficient at heating the deuterium. 139 refs

  18. Workshop on high power ICH antenna designs for high density tokamaks

    International Nuclear Information System (INIS)

    Aamodt, R.E.

    1990-01-01

    A workshop in high power ICH antenna designs for high density tokamaks was held in Boulder, Colorado on January 31 through February 2, 1990. The purposes of the workshop were to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of rf auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made

  19. EUV impurity study of the Alcator tokamak

    International Nuclear Information System (INIS)

    Terry, J.L.; Chen, K.I.; Moos, H.W.; Marmar, E.S.

    1977-06-01

    The intensity of resonance line radiation from oxygen, nitrogen, carbon and molybdenum impurities has been measured in the high field (80 kG), high density (6 x 10 14 cm -3 ) discharges of the Alcator tokamak, using a 0.4 m normal incidence monochromator (300 to 1300 A) with its line of sight fixed along a major radius. The total light impurity concentrations were 2 x 10 -3 , 7 x 10 -4 , and 3 x 10 -3 at central electron densities of 4.5 x 10 13 cm -3 (burnout), 4.0 x 10 13 (low density plateau) and 6.0 x 10 14 (high density plateau). Both a simple model and a computer code which included Pfirsch-Schluter impurity diffusion were used to estimate oxygen influxes of 1.6 x 10 13 cm -2 sec -1 and 1.5 x 10 14 cm -2 sec -1 at the plasma edge in the low and high density emission plateaus. The resulting values of Z/sub eff/, including the contributions due to both the light impurities and molybdenum, were close to one. The power lost through the impurity line radiation accounted for approximately equal to 7 percent of the total ohmic input power at high densities

  20. Scrape-off layer reflectometer for Alcator C-Moda)

    Science.gov (United States)

    Lau, Cornwall; Hanson, Greg; Wilgen, John; Lin, Yijun; Wukitch, Steve

    2010-10-01

    A swept-frequency X-mode reflectometer is being built for Alcator C-Mod to measure the scrape-off layer density profiles at the top, middle, and bottom locations in front of both the new lower hybrid launcher and the new ion cyclotron range of frequencies antenna. The system is planned to operate between 100 and 146 GHz at sweep rates from 10 μs to 1 ms, and will cover a density range of approximately 1016-1020 m-3 at B0=5-5.4 T. To minimize the effects of density fluctuations, both differential phase and full phase reflectometry will be employed. Design, test data, and calibration results of this electronics system will be discussed. To reduce attenuation losses, tallguide (TE01) will be used for most of the transmission line system. Simulations of high mode conversion in tallguide components, such as e-plane hyperbolic secant radius of curvature bends, tapers, and horn antennas will be shown. Experimental measurements of the total attenuation losses of these components in the lower hybrid waveguide run will also be presented.

  1. Small ELM regimes with good confinement on JET and comparison to those on ASDEX Upgrade, Alcator C-mod, and JT-60U

    International Nuclear Information System (INIS)

    Stober, J.; Lomas, P.; Saibene, G.

    2005-01-01

    Since it is uncertain if ITER operation is compatible with type-I ELMs, the study of alternative H-mode pedestals is an urgent issue. This paper reports on experiments on JET aiming to find scenarios with small ELMs and good confinement, such as the type-II ELMs in ASDEX Upgrade, the enhanced D-alpha H-mode in Alcator C-mod or the grassy ELMs in JT-60U. The study includes shape variations, especially the closeness to a double-null configuration, variations of q 95 , density and beta poloidal. H-mode pedestals without type-I ELMs have been observed only at the lowest currents (≤ 1.2 MA), showing similarities to the observations in the devices mentioned above. These are discussed in detail on the basis of edge fluctuation analysis. For higher currents, only the mixed type-I/II scenario is observed. Although the increased inter-ELM transport reduces the type-I ELM frequency, a single type-I ELM is not significantly reduced in size. Obviously, these results do question the accessibility of such small ELM scenarios on ITER, except perhaps the high beta-poloidal scenario at higher q 95 , which could not be tested at higher currents at JET due to limitations in heating power. (author)

  2. Low q operations in Alcator

    International Nuclear Information System (INIS)

    Overskei, D.O.; Gondhalekar, A.; Hutchinson, I.; Pappas, D.; Parker, R.; Rice, J.; Scaturro, L.; Wolfe, S.

    1979-01-01

    The Alcator Tokamak has successfully achieved stable and routine operation in the low q[q(a) <= 2.6] high density regime. Significant increases in the total β are observed with no degradation in the energy confinement times. Strong MHD activity is suppressed by the programmed increase of the plasma current. (Auth.)

  3. Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

    Science.gov (United States)

    Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

    2017-10-01

    The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

  4. Development of a High Resolution X-Ray Imaging Crystal Spectrometer for Measurement of Ion-Temperature and Rotation-Velocity Profiles in Fusion Energy Research Plasmas

    International Nuclear Information System (INIS)

    Hill, K.W.; Bitter, M.L.; Broennimann, Ch.; Eikenberry, E.F.; Ince-Cushman, A.; Lee, S.G.; Rice, J.E.; Scott, S.; Barnsley, R.

    2008-01-01

    A new imaging high resolution x-ray crystal spectrometer (XCS) has been developed to measure continuous profiles of ion temperature and rotation velocity in fusion plasmas. Following proof-of-principle tests on the Alcator C-Mod tokamak and the NSTX spherical tokamak, and successful testing of a new silicon, pixilated detector with 1MHz count rate capability per pixel, an imaging XCS is being designed to measure full profiles of T i and ν φ on C-Mod. The imaging XCS design has also been adopted for ITER. Ion-temperature uncertainty and minimum measurable rotation velocity are calculated for the C-Mod spectrometer. The affects of x-ray and nuclear-radiation background on the measurement uncertainties are calculated to predict performance on ITER

  5. An in situ accelerator-based diagnostic for plasma-material interactions science on magnetic fusion devices.

    Science.gov (United States)

    Hartwig, Zachary S; Barnard, Harold S; Lanza, Richard C; Sorbom, Brandon N; Stahle, Peter W; Whyte, Dennis G

    2013-12-01

    This paper presents a novel particle accelerator-based diagnostic that nondestructively measures the evolution of material surface compositions inside magnetic fusion devices. The diagnostic's purpose is to contribute to an integrated understanding of plasma-material interactions in magnetic fusion, which is severely hindered by a dearth of in situ material surface diagnosis. The diagnostic aims to remotely generate isotopic concentration maps on a plasma shot-to-shot timescale that cover a large fraction of the plasma-facing surface inside of a magnetic fusion device without the need for vacuum breaks or physical access to the material surfaces. Our instrument uses a compact (~1 m), high-current (~1 milliamp) radio-frequency quadrupole accelerator to inject 0.9 MeV deuterons into the Alcator C-Mod tokamak at MIT. We control the tokamak magnetic fields--in between plasma shots--to steer the deuterons to material surfaces where the deuterons cause high-Q nuclear reactions with low-Z isotopes ~5 μm into the material. The induced neutrons and gamma rays are measured with scintillation detectors; energy spectra analysis provides quantitative reconstruction of surface compositions. An overview of the diagnostic technique, known as accelerator-based in situ materials surveillance (AIMS), and the first AIMS diagnostic on the Alcator C-Mod tokamak is given. Experimental validation is shown to demonstrate that an optimized deuteron beam is injected into the tokamak, that low-Z isotopes such as deuterium and boron can be quantified on the material surfaces, and that magnetic steering provides access to different measurement locations. The first AIMS analysis, which measures the relative change in deuterium at a single surface location at the end of the Alcator C-Mod FY2012 plasma campaign, is also presented.

  6. Plasma position control on Alcator C

    International Nuclear Information System (INIS)

    Pribyl, P.A.

    1981-05-01

    The Alcator C MHD equilibrium is investigated from the standpoint of determining the plasma position. A review of equilibrium theory is presented, indicating that the central flux surfaces of the plasma should be displaced about 1 to 2 cm from the outermost. Further, the plasma should have a slightly noncircular cross-section. A comparison is made between the observed and predicted profiles. Flux loops sensitive to plasma position generate the error signal for the feedback control circuit. This measurement agrees with other position-sensitive diagnostics, such as limiter heating, and centroids of density, soft x-ray, and electron cyclotron emission. A linear model is developed for the position control feedback system, including the vertical field SCR supply, plasma, and feedback electronics. Operation of the control system agrees well with that predicted by the model, with acceptable plasma position being maintained for the duration of the discharge. The feedback control system is in daily use for Alcator C runs

  7. 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)

  8. 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

  9. Effect of magnetic and density fluctuations on the propagation of lower hybrid waves in tokamaks

    Science.gov (United States)

    Vahala, George; Vahala, Linda; Bonoli, Paul T.

    1992-12-01

    Lower hybrid waves have been used extensively for plasma heating, current drive, and ramp-up as well as sawteeth stabilization. The wave kinetic equation for lower hybrid wave propagation is extended to include the effects of both magnetic and density fluctuations. This integral equation is then solved by Monte Carlo procedures for a toroidal plasma. It is shown that even for magnetic/density fluctuation levels on the order of 10-4, there are significant magnetic fluctuation effects on the wave power deposition into the plasma. This effect is quite pronounced if the magnetic fluctuation spectrum is peaked within the plasma. For Alcator-C-Mod [I. H. Hutchinson and the Alcator Group, Proceedings of the IEEE 13th Symposium on Fusion Engineering (IEEE, New York, 1990), Cat. No. 89CH 2820-9, p. 13] parameters, it seems possible to be able to infer information on internal magnetic fluctuations from hard x-ray data—especially since the effects of fluctuations on electron power density can explain the hard x-ray data from the JT-60 tokamak [H. Kishimoto and JT-60 Team, in Plasma Physics and Controlled Fusion (International Atomic Energy Agency, Vienna, 1989), Vol. I, p. 67].

  10. Shear flows at the tokamak edge and their interaction with edge-localized modes

    International Nuclear Information System (INIS)

    Aydemir, A. Y.

    2007-01-01

    Shear flows in the scrape-off layer (SOL) and the edge pedestal region of tokamaks are shown to arise naturally out of transport processes in a magnetohydrodynamic model. In quasi-steady-state conditions, collisional resistivity coupled with a simple bootstrap current model necessarily leads to poloidal and toroidal flows, mainly localized to the edge and SOL. The role of these flows in the grad-B drift direction dependence of the power threshold for the L (low) to H (high) transition, and their effect on core rotation, are discussed. Theoretical predictions based on symmetries of the underlying equations, coupled with computational results, are found to be in agreement with observations in Alcator C-Mod [Phys. Plasmas 12, 056111 (2005)]. The effects of these self-consistent flows on linear peeling/ballooning modes and their nonlinear consequences are also examined

  11. Phase contrast imaging measurements of reversed shear Alfvén eigenmodes during sawteeth in Alcator C-Moda)

    Science.gov (United States)

    Edlund, E. M.; Porkolab, M.; Kramer, G. J.; Lin, L.; Lin, Y.; Wukitch, S. J.

    2009-05-01

    Reversed shear Alfvén eigenmodes (RSAEs) have been observed with the phase contrast imaging diagnostic and Mirnov coils during the sawtooth cycle in Alcator C-mod [M. Greenwald et al., Nucl. Fusion 45, S109 (2005)] plasmas with minority ion-cyclotron resonance heating. Both down-chirping RSAEs and up-chirping RSAEs have been observed during the sawtooth cycle. Experimental measurements of the spatial structure of the RSAEs are compared to theoretical models based on the code NOVA [C. Z. Cheng and M. S. Chance, J. Comput. Phys. 71, 124 (1987)] and used to derive constraints on the q profile. It is shown that the observed RSAEs can be understood by assuming a reversed shear q profile (up chirping) or a q profile with a local maximum (down chirping) with q ≈1.

  12. 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.

  13. Edge energy transport barrier and turbulence in the I-mode regime on Alcator C-Moda)

    Science.gov (United States)

    Hubbard, A. E.; Whyte, D. G.; Churchill, R. M.; Cziegler, I.; Dominguez, A.; Golfinopoulos, T.; Hughes, J. W.; Rice, J. E.; Bespamyatnov, I.; Greenwald, M. J.; Howard, N.; Lipschultz, B.; Marmar, E. S.; Reinke, M. L.; Rowan, W. L.; Terry, J. L.

    2011-05-01

    We report extended studies of the I-mode regime [Whyte et al., Nucl. Fusion 50, 105005 (2010)] obtained in the Alcator C-Mod tokamak [Marmar et al., Fusion Sci. Technol. 51(3), 3261 (2007)]. This regime, usually accessed with unfavorable ion B × ∇B drift, features an edge thermal transport barrier without a strong particle transport barrier. Steady I-modes have now been obtained with favorable B × ∇B drift, by using specific plasma shapes, as well as with unfavorable drift over a wider range of shapes and plasma parameters. With favorable drift, power thresholds are close to the standard scaling for L-H transitions, while with unfavorable drift they are ˜ 1.5-3 times higher, increasing with Ip. Global energy confinement in both drift configurations is comparable to H-mode scalings, while density profiles and impurity confinement are close to those in L-mode. Transport analysis of the edge region shows a decrease in edge χeff, by typically a factor of 3, between L- and I-mode. The decrease correlates with a drop in mid-frequency fluctuations (f ˜ 50-150 kHz) observed on both density and magnetics diagnostics. Edge fluctuations at higher frequencies often increase above L-mode levels, peaking at f ˜ 250 kHz. This weakly coherent mode is clearest and has narrowest width (Δf/f ˜ 0.45) at low q95 and high Tped, up to 1 keV. The Er well in I-mode is intermediate between L- and H-mode and is dominated by the diamagnetic contribution in the impurity radial force balance, without the Vpol shear typical of H-modes.

  14. Modification of ordinary-mode reflectometry system to detect lower-hybrid waves in Alcator C-Moda)

    Science.gov (United States)

    Baek, S. G.; Shiraiwa, S.; Parker, R. R.; Dominguez, A.; Kramer, G. J.; Marmar, E. S.

    2012-10-01

    Backscattering experiments to detect lower-hybrid (LH) waves have been performed in Alcator C-Mod, using the two modified channels (60 GHz and 75 GHz) of an ordinary-mode reflectometry system with newly developed spectral recorders that can continuously monitor spectral power at a target frequency. The change in the baseline of the spectral recorder during the LH wave injection is highly correlated to the strength of the X-mode non-thermal electron cyclotron emission. In high density plasmas where an anomalous drop in the lower hybrid current drive efficiency is observed, the observed backscattered signals are expected to be generated near the last closed flux surface, demonstrating the presence of LH waves within the plasma. This experimental technique can be useful in identifying spatially localized LH electric fields in the periphery of high-density plasmas.

  15. UCSD Performance in the Edge Plasma Simulation (EPSI) Project. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Tynan, George Robert [Univ. of California, San Diego, CA (United States). Center for Energy Research

    2017-12-12

    This report contains a final report on the activities of UC San Diego PI G.R. Tynan and his collaborators as part of the EPSI Project, that was led by Dr. C.S. Chang, from PPPL. As a part of our work, we carried out several experiments on the ALCATOR C-­MOD tokamak device, aimed at unraveling the “trigger” or cause of the spontaneous transition from low-­mode confinement (L-­mode) to high confinement (H-­mode) that is universally observed in tokamak devices, and is planned for use in ITER.

  16. High Resolution Spectroscopy in the Divertor and Edge Regions of Alcator-C Mode and Measurement of Radiative Transfer in Vacuum-UV Line Emission from Magnetic Fusion Devices

    International Nuclear Information System (INIS)

    Griem, Hans R.

    2005-01-01

    Spectroscopic diagnostics were carried out both at MIT and at the University of Maryland. At MIT, measurements were made of toroidal flow velocities in the mid-plane of the inner and outer scrape-off layers (SOL) of Alcator C-Mod plasmas, using a high-resolution spectrograph. Subsequently, the MIT/Alcator procedures based upon visible spectroscopy were transferred to the new Maryland centrifugal experiment (MCX). In a further effort towards data refinement, we expanded the hydrogen measurements from the n approaches to 2 Balmer series in the visible to the n approaches to 1 Lyman series in the vacuum-ultraviolet (vuv) spectral region. Recent results were presented at APS Division of Plasma Physics meetings and published in Physics of Plasmas in 2004 and 2005. Further details can be found in the annual progress reports to the Department of Energy

  17. Analysis of C-MOD molybdenum divertor erosion and code/data comparison

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, J.N., E-mail: brooksjn@purdue.edu [Purdue University, West Lafayette, IN (United States); Allain, J.P. [Purdue University, West Lafayette, IN (United States); Whyte, D.G.; Ochoukov, R.; Lipschultz, B. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    2011-08-01

    We analyze an important 15 year old Alcator C-MOD study of campaign-integrated molybdenum divertor erosion in which the measured net erosion was significantly higher ({approx}X3) than originally predicted by a simple model . We perform full process sputtering erosion/redeposition computational analysis including the effect of a possible RF induced sheath. The simulations show that most sputtered Mo atoms are ionized close to the surface and strongly redeposited, via Lorentz force motion and collisional friction with the high density incoming plasma. The predicted gross erosion profile is a reasonable match to MoI influx data, however, the critically important net erosion comparison with post-exposure Mo tile analysis is poor, with {approx}X10 higher peak erosion measured than computed. An RF sheath increases predicted erosion by {approx}45%, thus being significant but not fundamental. We plan future analysis.

  18. Measurements of the parallel wavenumber of lower hybrid waves in the scrape-off layer of a high-density tokamak

    International Nuclear Information System (INIS)

    Baek, S. G.; Wallace, G. M.; Parker, R. R.; Shiraiwa, S.; Bonoli, P. T.; Brunner, D.; Faust, I.; LaBombard, B. L.; Wukitch, S.; Shinya, T.; Takase, Y.

    2016-01-01

    In lower hybrid current drive (LHCD) experiments on tokamaks, the parallel wavenumber of lower hybrid waves is an important physics parameter that governs the wave propagation and absorption physics. However, this parameter has not been experimentally well-characterized in the present-day high density tokamaks, despite the advances in the wave physics modeling. In this paper, we present the first measurement of the dominant parallel wavenumber of lower hybrid waves in the scrape-off layer (SOL) of the Alcator C-Mod tokamak with an array of magnetic loop probes. The electric field strength measured with the probe in typical C-Mod plasmas is about one-fifth of that of the electric field at the mouth of the grill antenna. The amplitude and phase responses of the measured signals on the applied power spectrum are consistent with the expected wave energy propagation. At higher density, the observed k || increases for the fixed launched k || , and the wave amplitude decreases rapidly. This decrease is correlated with the loss of LHCD efficiency at high density, suggesting the presence of loss mechanisms. Evidence of the spectral broadening mechanisms is observed in the frequency spectra. However, no clear modifications in the dominant k || are observed in the spectrally broadened wave components, as compared to the measured k || at the applied frequency. It could be due to (1) the probe being in the SOL and (2) the limited k || resolution of the diagnostic. Future experiments are planned to investigate the roles of the observed spectral broadening mechanisms on the LH density limit problem in the strong single pass damping regime.

  19. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    1982-03-01

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  20. Design of a microwave calorimeter for the microwave tokamak experiment

    International Nuclear Information System (INIS)

    Marinak, M.

    1988-01-01

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs

  1. CSPM (Center for the Study of Plasma Microturbulence)

    Energy Technology Data Exchange (ETDEWEB)

    Candy, Jeff [General Atomics, San Diego, CA (United States)

    2017-11-13

    This final report for CSPM documents the contributions from General Atomics over the period of the award (15 Feb. 2011 to 14 Feb. 2017). This project focused on the development of gyrokinetic simulation capabilities relevant for the description of particle and energy transport in tokamak plasmas. This included significant comparisons with results on the DIII-D and Alcator C-Mod tokamaks. In addition, advances in a number of theoretical areas (collision operators, discretization methods, transport model recalibration, etc) were also documented and are described herein.

  2. Comparison of Alcator C data with the Rebut-Lallia-Watkins critical gradient scaling

    International Nuclear Information System (INIS)

    Hutchinson, I.H.

    1992-12-01

    The critical temperature gradient model of Rebut, Lallia and Watkins is compared with data from Alcator C. The predicted central electron temperature is derived from the model, and a simple analytic formula is given. It is found to be in quite good agreement with the observed temperatures on Alcator C under ohmic heating conditions. However, the thermal diffusivity postulated in the model for gradients that exceed the critical is not consistent with the observed electron heating by Lower Hybrid waves

  3. Improved Controls for Fusion RF Systems. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Casey, Jeffrey A. [Rockfield Research Inc., Las Vegas, NV (United States)

    2011-11-08

    We have addressed the specific requirements for the integrated systems controlling an array of klystrons used for Lower Hybrid Current Drive (LHCD). The immediate goal for our design was to modernize the transmitter protection system (TPS) for LHCD on the Alcator C-Mod tokamak at the MIT Plasma Science and Fusion Center (MIT-PSFC). Working with the Alcator C-Mod team, we have upgraded the design of these controls to retrofit for improvements in performance and safety, as well as to facilitate the upcoming expansion from 12 to 16 klystrons. The longer range goals to generalize the designs in such a way that they will be of benefit to other programs within the international fusion effort was met by designing a system which was flexible enough to address all the MIT system requirements, and modular enough to adapt to a large variety of other requirements with minimal reconfiguration.

  4. Improved Controls for Fusion RF Systems. Final technical report

    International Nuclear Information System (INIS)

    Casey, Jeffrey A.

    2011-01-01

    We have addressed the specific requirements for the integrated systems controlling an array of klystrons used for Lower Hybrid Current Drive (LHCD). The immediate goal for our design was to modernize the transmitter protection system (TPS) for LHCD on the Alcator C-Mod tokamak at the MIT Plasma Science and Fusion Center (MIT-PSFC). Working with the Alcator C-Mod team, we have upgraded the design of these controls to retrofit for improvements in performance and safety, as well as to facilitate the upcoming expansion from 12 to 16 klystrons. The longer range goals to generalize the designs in such a way that they will be of benefit to other programs within the international fusion effort was met by designing a system which was flexible enough to address all the MIT system requirements, and modular enough to adapt to a large variety of other requirements with minimal reconfiguration

  5. Spectroscopic diagnostics of high temperature plasmas

    International Nuclear Information System (INIS)

    Moos, W.

    1990-01-01

    A three-year research program for the development of novel XUV spectroscopic diagnostics for magnetically confined fusion plasmas is proposed. The new diagnostic system will use layered synthetic microstructures (LSM) coated, flat and curved surfaces as dispersive elements in spectrometers and narrow band XUV filter arrays. In the framework of the proposed program we will develop impurity monitors for poloidal and toroidal resolved measurements on PBX-M and Alcator C-Mod, imaging XUV spectrometers for electron density and temperature fluctuation measurements in the hot plasma core in TEXT or other similar tokamaks and plasma imaging devices in soft x-ray light for impurity behavior studies during RF heating on Phaedrus T and carbon pellet ablation in Alcator C-Mod. Recent results related to use of multilayer in XUV plasma spectroscopy are presented. We also discuss the latest results reviewed to q o and local poloidal field measurements using Zeeman polarimetry

  6. Characterization of axisymmetric disruption dynamics toward VDE avoidance in tokamaks

    International Nuclear Information System (INIS)

    Nakamura, Y.

    2002-01-01

    Disruption experiments on Alcator C-Mod and ASDEX-Upgrade tokamaks and axisymmetric MHD simulations using the TSC have explicated the underlying mechanisms of Vertical Displacement Events (VDEs) and a diversity of disruption dynamics. First, the neutral point, which is known as an initial vertical plasma position advantageous to VDE avoidance, is shown to be fairly insensitive to plasma shape and current profile parameters, while the VDE rate significantly depends on those parameters. Secondly, it is clarified that a rapid flattening of the plasma current profile frequently seen at the thermal quench drags a single null-diverted, up-down asymmetric plasma vertically toward divertor, whereas the dragging effect is absent in up-down symmetric limiter discharges. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges, being consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the current quench and vanishes at the neutral point, the dragging effect explains many details of the VDE dynamics over the whole period of disruptive termination. (author)

  7. Active MHD Spectroscopy of Alfvén Eigenmodes on Alcator C-Mod

    Science.gov (United States)

    Sears, J.; Snipes, J.; Burke, W.; Parker, R.; Fasoli, A.

    2004-11-01

    Alfvén eigenmode resonances are excited in a variety of plasma conditions in C-Mod with two moderate-n antennas positioned above and below the outboard midplane. Power amplifiers (≈ 3 kW) sweep the driving frequency over the audio range (< 30 kHz) or over a selected ± 50 kHz range from 100 kHz to 1 MHz. Logic circuitry that calculates the center frequency of the Toroidal Alfven Eigenmode gap, f_TAE=v_A/4π qR, in real-time from BT and e measurements is being developed to enable the antennas to track f_TAE. Simultaneous in-vessel phase calibration of the pick-up coils will be used to better identify toroidal mode numbers. Shot-to-shot elongation scans do not show the dependence of damping on edge shear that was seen in results at JET. Inner wall limited plasmas with moderate outer gaps show higher damping rates than diverted plasmas with low outer gaps. Low frequency experiments below 20kHz will also be presented.

  8. Extraordinary mode absorption at the electron cyclotron harmonic frequencies as a Tokamak plasma diagnostic

    International Nuclear Information System (INIS)

    Pachtman, A.

    1986-09-01

    Measurements of Extraordinary mode absorption at the electron cyclotron harmonic frequencies are of unique value in high temperature, high density Tokamak plasma diagnostic applications. An experimental study of Extraordinary mode absorption at the semi-opaque second and third harmonics has been performed on the ALCATOR C Tokamak. A narrow beam of submillimeter laser radiation was used to illuminate the plasma in a horizontal plane, providing a continuous measurement of the one-pass, quasi-perpendicular transmission

  9. Testing of low Z coated limiters in tokamak fusion devices

    International Nuclear Information System (INIS)

    Whitely, J.B.; Mullendore, A.W.; Langley, R.A.

    1980-01-01

    Extensive testing on a laboratory scale has been used to select those coatings most suitable for this environment. From this testing which included pulsed electron beam heating, low energy ion bombardment and arcing, chemical vapor deposited coating of TiB 2 and TiC on Poco graphite substrates have been selected and tested as limiters in ISX. Both limiter materials gave clean, stable, reproducible tokamak discharges the first day of operation. After one weeks exposure, the TiC limiter showed only superficial damage with no coating failure. The TiB 2 limiter had some small areas of coating failure. TiC coated graphite limiters have also been briefly tested in the tokamaks Alcator and PDX with favorable results

  10. A Third Generation Lower Hybrid Coupler

    International Nuclear Information System (INIS)

    Bernabei, S.; Hosea, J.; Kung, C.; Loesser, D.; Rushinski, J.; Wilson, J.R.; Parker, R.

    2001-01-01

    The Princeton Plasma Physics Laboratory (PPPL) and the Massachusetts Institute of Technology (MIT) are preparing an experiment of current profile control using lower-hybrid waves in order to produce and sustain advanced tokamak regimes in steady-state conditions in Alcator C-Mod. Unlike JET's, ToreSupra's and JT60's couplers, the C-Mod lower-hybrid coupler does not employ the now conventional multijunction design, but will have similar characteristics, compactness, and internal power division while retaining full control of the antenna element phasing. This is achieved by using 3 dB vertical power splitters and a stack of laminated plates with the waveguides milled in them. Construction is simplified and allows easy control and maintenance of all parts. Many precautions are taken to avoid arcing. Special care is also taken to avoid the recycling of reflected power which could affect the coupling and the launched n(subscript ||) spectrum. The results from C-Mod should allow further simplification in the designs of the coupler planned for KSTAR (Korea Superconducting Tokamak Advanced Research) and ITER (International Thermonuclear Experimental Reactor)

  11. Examining Innovative Divertor and Main Chamber Options for a National Divertor Test Tokamak

    Science.gov (United States)

    Labombard, B.; Umansky, M.; Brunner, D.; Kuang, A. Q.; Marmar, E.; Wallace, G.; Whyte, D.; Wukitch, S.

    2016-10-01

    The US fusion community has identified a compelling need for a National Divertor Test Tokamak. The 2015 Community Planning Workshop on PMI called for a national working group to develop options. Important elements of a NDTT, adopted from the ADX concept, include the ability to explore long-leg divertor `solutions for power exhaust and particle control' (Priority Research Direction B) and to employ inside-launch RF actuators combined with double-null topologies as `plasma solution for main chamber wall components, including tools for controllable sustained operation' (PRD-C). Here we examine new information on these ideas. The projected performance of super-X and X-point target long-leg divertors is looking very promising; a stable fully-detached divertor condition handling an order-of-magnitude increase in power handling over conventional divertors may be possible. New experiments on Alcator C-Mod are addressing issues of high-field side versus low-field side heat flux sharing in double-null topologies and the screening of impurities that might originate from RF actuators placed in the high-field side - both with favorable results. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC52-07NA27344.

  12. Radiofrequency-heated enhanced confinement modes in the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Takase, Y.; Boivin, R.L.; Bombarda, F.; Bonoli, P.T.; Christensen, C.; Fiore, C.; Garnier, D.; Goetz, J.A.; Golovato, S.N.; Granetz, R.; Greenwald, M.; Horne, S.F.; Hubbard, A.; Hutchinson, I.H.; Irby, J.; LaBombard, B.; Lipschultz, B.; Marmar, E.; May, M.; Mazurenko, A.; McCracken, G.; OShea, P.; Porkolab, M.; Reardon, J.; Rice, J.; Rost, C.; Schachter, J.; Snipes, J.A.; Stek, P.; Terry, J.; Watterson, R.; Welch, B.; Wolfe, S.

    1997-01-01

    Enhanced confinement modes up to a toroidal field of B T =8T have been studied with up to 3.5 MW of radiofrequency (rf) heating power in the ion cyclotron range of frequencies (ICRF) at 80 MHz. H-mode is observed when the edge temperature exceeds a threshold value. The high confinement mode (H-mode) with higher confinement enhancement factors (H) and longer duration became possible after boronization by reducing the radiated power from the main plasma. A quasi-steady state with high confinement (H=2.0), high normalized beta (β N =1.5), low radiated power fraction (P rad main /P loss =0.3), and low effective charge (Z eff =1.5) has been obtained in Enhanced D α H-mode. This type of H-mode has enhanced levels of continuous D α emission and very little or no edge localized mode (ELM) activity, and reduced core particle confinement time relative to ELM-free H-mode. The pellet enhanced performance (PEP) mode is obtained by combining core fueling with pellet injection and core heating. A highly peaked pressure profile with a central value of 8 atmospheres was observed. The steep pressure gradient drives off-axis bootstrap current, resulting in a shear reversed safety factor (q) profile. Suppression of sawteeth appears to be important in maintaining the highly peaked pressure profile. Lithium pellets were found to be more effective than deuterium pellets in raising q 0 . copyright 1997 American Institute of Physics

  13. Study of mode-converted and directly-excited ion Bernstein waves by CO2 laser scattering in Alcator C

    International Nuclear Information System (INIS)

    Takase, Y.; Fiore, C.L.; McDermott, F.S.; Moody, J.D.; Porkolab, M.; Shepard, T.; Squire, J.

    1987-01-01

    Mode-converted and directly excited ion Bernstein waves (IBW) were studied using CO 2 laser scattering in the Alcator C tokamak. During the ICRF fast wave heating experiments, mode-converted IBW was observed on the high-field side of the resonance in both second harmonic and minority heating regimes. By comparing the relative scattered powers from the two antennas separated by 180 0 toroidally, an increased toroidal wave damping with increasing density was inferred. In the IBW heating experiments, optimum direct excitation is obtained when an ion-cyclotron harmonic layer is located just behind the antenna. Wave absorption at the ω = 3Ω/sub D/ = 1.5Ω/sub H/ layer was directly observed. Edge ion heating was inferred from the IBW dispersion when this absorption layer was located in the plasma periphery, which may be responsible for the observed improvement in particle confinement

  14. Comparative study of fundamental and second-harmonic ICRF wave propagation and damping at high density in the Alcator tokamak

    International Nuclear Information System (INIS)

    Gaudreau, M.P.J.

    1981-09-01

    Due to the versatility of the high power apparatus, the fast magnetosonic branch is used with ω 0 = 1,2,3,4 ω/sub ci/, unlike most other ICRF experiments. Unusually high magnetic field (B 0 = 40 to 80 kG), plasma density (n/sub e/ = 10 13 - 5 x 10 14 /cm 3 ), generator frequency (f 0 = 90 to 200 MHz) and transmitter power, with shielded and unshielded antennas, are the key parameters of the experiment. This wide parameter range allows a direct comparison between fundamental and second harmonic regimes, and shielded and unshielded antennas, our prime goals. The real and imaginary parts of the parallel and perpendicular wave numbers are measured with extensive magnetic probe diagnostics for a spectrum of plasma parameters and compared with theory. Qualitative and quantitative evaluations of the wave structure and scaling laws are derived analytically in simple geometries and computed numerically for realistic plasma parameters and profiles. General figures of merit, such as radiation resistance and quality factor, are also derived and compared with the experiment. Secondary effects of the high power wave launching, such as changes in plasma current, density, Z/sub eff/, energetic neutral flux, soft x-rays, neutron flux, and impurities are also discussed. Most important, a general synthesis of the many engineering, physics, and experimental problems and conclusions of the Alcator A ICRF program are inspected in detail. Finally, the derived and experimentally determined scaling laws and engineering constraints are used to estimate the ICRF requrements, advantages, and potential pitfalls of the next generations of experiments on the Alcator tokamaks

  15. Impact and effects of simultaneous MeV-ion irradiation and helium plasma exposure to the formation of tungsten nano-tendrils

    Science.gov (United States)

    Wright, Graham; Kesler, Leigh Ann; Whyte, Dennis

    2013-10-01

    The extrusion of nano-tendrils from high temperature (>1000 K) tungsten (W) targets exposed to helium (He) plasma ions remains a concern for future fusion reactors. Previous work on the Alcator C-Mod tokamak has demonstrated it is possible to form these structures in a tokamak environment. However, one area where Alcator C-Mod and a fusion reactor differ is total neutron flux at the wall and the displacement damage these neutrons produce in the plasma-facing materials. This dsiplacement damage may affect the size and number He bubbles precipitating in the W target, which is a key factor in the formation and growth of the nano-tendrils. The DIONISOS experiment directly measures the impact of the displacement damage by simultaneously bombarding high temperature W targets with MeV-range ions (to simulate the displacement damage caused by neutron flux) and high flux of He plasma ions. Different combinations of irradiating ion species and W target temperatures are used to vary the different processes and rates that are involved such as He trapping rate, vacancy production and annealing rates, and nano-tendril growth rate. The nano-tendril growth is characterized by SEM imaging and focused ion beam (FIB) cross-sectioning and compared to nano-tendril formation without the presence of the irradiating ion beam. This work is supported by US DOE award DE-SC00-02060.

  16. Analysis of Rotation and Transport Data in C-Mod ITB Plasmas

    Science.gov (United States)

    Fiore, C. L.; Rice, J. E.; Reinke, M. L.; Podpaly, Y.; Bespamyatnov, I. O.; Rowan, W. L.

    2009-11-01

    Internal transport barriers (ITBs) spontaneously form near the half radius of Alcator C-Mod plasmas when the EDA H-mode is sustained for several energy confinement times in either off-axis ICRF heated discharges or in purely ohmic heated plasmas. These plasmas exhibit strongly peaked density and pressure profiles, static or peaking temperature profiles, peaking impurity density profiles, and thermal transport coefficients that approach neoclassical values in the core. It has long been observed that the intrinsic central plasma rotation that is strongly co-current following the H-mode transition slows and often reverses as the density peaks as the ITB forms. Recent spatial measurements demonstrate that the rotation profile develops a well in the core region that decreases continuously as central density rises while the value outside of the core remains strongly co-current. This results in the formation of a steep potential gradient/strong electric field at the location of the foot of the ITB density profile. The resulting E X B shearing rate is also quite significant at the foot. These analyses and the implications for plasma transport and stability will be presented.

  17. Physics of the Tokamak Pedestal, and Implications for Magnetic Fusion Energy

    Science.gov (United States)

    Snyder, Philip

    2017-10-01

    High performance in tokamaks is achieved via the spontaneous formation of a transport barrier in the outer few percent of the confined plasma. This narrow insulating layer, referred to as a ``pedestal,'' typically results in a >30x increase in pressure across a 0.4-5cm layer. Predicted fusion power scales with the square of the pedestal top pressure (or ``pedestal height''), hence a fusion reactor strongly benefits from a high pedestal, provided this can be attained without large Edge Localized Modes (ELMs), which may erode plasma facing materials. The overlap of drift orbit, turbulence, and equilibrium scales across this narrow layer leads to rich and complex physics, and challenges traditional analytic and computational approaches. We review studies employing gyrokinetic, neoclassical, MHD, and other methods, which have explored how a range of instabilities, influenced by complex geometry, and strong ExB flows and bootstrap current, drive transport across the pedestal and guide its structure and dynamics. Development of high resolution diagnostics, and coordinated experiments on several tokamaks, have validated understanding of important aspects of the physics, while highlighting open issues. A predictive model (EPED) has proven capable of predicting the pedestal height and width to 20-25% accuracy in large statistical studies. This model was used to predict a new, high pedestal ``Super H-Mode'' regime, which was subsequently discovered on DIII-D, and motivated experiments on Alcator C-Mod which achieved world record, reactor relevant pedestal pressure. We review open issues including improved formalism, particle and momentum transport, the role of neutrals and impurities, ELM control, and pedestal formation. Finally we discuss coupling pedestal and core predictive models to enable more comprehensive optimization of the tokamak fusion concept. Supported by the US DOE under DE-FG02-95ER54309, FC02-06ER54873, DE-FC02-04ER54698, DE-FC02-99ER54512.

  18. Free-boundary toroidal Alfvén eigenmodes

    Science.gov (United States)

    Chen, Eugene Y.; Berk, H. L.; Breizman, B.; Zheng, L. J.

    2011-05-01

    A numerical study is presented for the n = 1 free-boundary toroidal Alfvén eigenmodes (TAE) in tokamaks, which shows that there is considerable sensitivity of n = 1 modes to the position of the conducting wall. An additional branch of the TAE is shown to emerge from the upper continuum as the ratio of conducting wall radius to plasma radius increases. Such phenomena arise in plasma equilibria with both circular and shaped cross sections, where the shaped profile studied here is similar to that found in Alcator C-Mod.

  19. D-D neutron energy-spectra measurements in Alcator C

    International Nuclear Information System (INIS)

    Pappas, D.S.; Wysocki, F.J.; Furnstahl, R.J.

    1982-08-01

    Measurements of energy spectra of neutrons produced during high density (anti n/sub e/ > 2 x 10 14 cm -3 ) deuterium discharges have been performed using a proton-recoil (NE 213) spectrometer. A two foot section of light pipe (coupling the scintillator and photomultiplier) was used to extend the scintillator into a diagnostic viewing port to maximize the neutron detection efficiency while not imposing excessive magnetic shielding requirements. A derivative unfolding technique was used to deduce the energy spectra. The results showed a well defined peak at 2.5 MeV which was consistent with earlier neutron flux measurements on Alcator C that indicated the neutrons were of thermonuclear origin

  20. Transport-driven scrape-off layer flows and the x-point dependence of the L-H power threshold in Alcator C-Moda)

    Science.gov (United States)

    LaBombard, B.; Rice, J. E.; Hubbard, A. E.; Hughes, J. W.; Greenwald, M.; Granetz, R. S.; Irby, J. H.; Lin, Y.; Lipschultz, B.; Marmar, E. S.; Marr, K.; Mossessian, D.; Parker, R.; Rowan, W.; Smick, N.; Snipes, J. A.; Terry, J. L.; Wolfe, S. M.; Wukitch, S. J.

    2005-05-01

    Factor of ˜2 higher power thresholds for low- to high-confinement mode transitions (L-H) with unfavorable x-point topologies in Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are linked to flow boundary conditions imposed by the scrape-off layer (SOL). Ballooning-like transport drives flow along magnetic field lines from low- to high-field regions with toroidal direction dependent on upper/lower x-point balance; the toroidal rotation of the confined plasma responds, exhibiting a strong counter-current rotation when B ×∇B points away from the x point. Increased auxiliary heating power (rf, no momentum input) leads to an L-H transition at approximately twice the edge electron pressure gradient when B ×∇B points away. As gradients rise prior to the transition, toroidal rotation ramps toward the co-current direction; the H mode is seen when the counter-current rotation imposed by the SOL flow becomes compensated. Remarkably, L-H thresholds in lower-limited discharges are identical to lower x-point discharges; SOL flows are also found similar, suggesting a connection.

  1. Model Development for VDE Computations in NIMROD

    Science.gov (United States)

    Bunkers, K. J.; Sovinec, C. R.

    2017-10-01

    Vertical displacement events (VDEs) and the disruptions associated with them have potential for causing considerable physical damage to ITER and other tokamak experiments. We report on simulations of generic axisymmetric VDEs and a vertically unstable case from Alcator C-MOD using the NIMROD code. Previous calculations have been done with closures for heat flux and viscous stress. Initial calculations show that halo current width is dependent on temperature boundary conditions, and so transport together with plasma-surface interaction may play a role in determining halo currents in experiments. The behavior of VDEs with Braginskii thermal conductivity and viscosity closures and Spitzer-like resistivity are investigated for both the generic axisymmetric VDE case and the C-MOD case. This effort is supported by the U.S. Dept. of Energy, Award Numbers DE-FG02-06ER54850 and DE-FC02-08ER54975.

  2. Response to comment on 'Magnetic topology effects on Alcator C-Mod scrape-off layer flow'

    International Nuclear Information System (INIS)

    Simakov, Andrei N; Catto, Peter J

    2009-01-01

    In his comment to our recent work (Simakov et al 2008 Plasma Phys. Control. Fusion 50 105010), Aydemir has asserted that poloidal plasma flow reversal is not a valid response to toroidal magnetic field reversal in an up-down symmetric tokamak, and that the toroidal plasma flow must reverse instead. We show that this assertion is wrong due to his misunderstanding of the corresponding symmetry transformation. (reply)

  3. Intrinsic torque reversals induced by magnetic shear effects on the turbulence spectrum in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Z. X.; Tynan, G. [Center for Energy Research and Department of Mechanical and Aerospace Engineering, University of California at San Diego, San Diego, California 92093 (United States); Center for Momentum Transport and Flow Organization and Center for Astrophysics and Space Science, University of California, San Diego, California 92093 (United States); Wang, W. X.; Ethier, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Diamond, P. H. [Center for Momentum Transport and Flow Organization and Center for Astrophysics and Space Science, University of California, San Diego, California 92093 (United States); Gao, C.; Rice, J. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    2015-05-15

    Intrinsic torque, which can be generated by turbulent stresses, can induce toroidal rotation in a tokamak plasma at rest without direct momentum injection. Reversals in intrinsic torque have been inferred from the observation of toroidal velocity changes in recent lower hybrid current drive (LHCD) experiments. This work focuses on understanding the cause of LHCD-induced intrinsic torque reversal using gyrokinetic simulations and theoretical analyses. A new mechanism for the intrinsic torque reversal linked to magnetic shear (s{sup ^}) effects on the turbulence spectrum is identified. This reversal is a consequence of the ballooning structure at weak s{sup ^}. Based on realistic profiles from the Alcator C-Mod LHCD experiments, simulations demonstrate that the intrinsic torque reverses for weak s{sup ^} discharges and that the value of s{sup ^}{sub crit} is consistent with the experimental results s{sup ^}{sub crit}{sup exp}≈0.2∼0.3 [Rice et al., Phys. Rev. Lett. 111, 125003 (2013)]. The consideration of this intrinsic torque feature in our work is important for the understanding of rotation profile generation at weak s{sup ^} and its consequent impact on macro-instability stabilization and micro-turbulence reduction, which is crucial for ITER. It is also relevant to internal transport barrier formation at negative or weakly positive s{sup ^}.

  4. An alcator-like confinement time scaling law derived from buckingham's PI theorem

    International Nuclear Information System (INIS)

    Roth, J.R.

    1983-01-01

    The unsatisfactory state of understanding of particle transport and confinement in tokamaks is well known. The best available theory, neoclassical transport, predicts a confinement time which scales as the square of the magnetic field, and inversely as the number density. Until recently, the best available phenomenological scaling law was the Alcator scaling law. This scaling law has recently been supplanted by the neoAlcator scaling law. Both of these expressions are unsatisfactory, because they not only are unsupported by any physical theory, but also their numerical constants are dimensional, suggesting that additional physical parameters need to be accounted for. A more firmly based scaling law can be derived from Buckingham's pi theorem. We adopt the particle confinement time as the dependent variable (derived dimension), and as independent variables (fundamental dimensions) we use the plasma volume, the average ion charge density, the ion current on the limiter, and the magnetic induction. From Buckingham's pi theorem, we obtain an equation which correctly predicts the absence of magnetic induction dependence, and the direct dependence on the ion density. The dependence on the product of the major radius and the plasma radius is intermediate between the original and neoAlcator scaling laws, and may be consistent with the data if the ion kinetic temperature and limiter area were accounted for

  5. ALCBEAM - Neutral beam formation and propagation code for beam-based plasma diagnostics

    Science.gov (United States)

    Bespamyatnov, I. O.; Rowan, W. L.; Liao, K. T.

    2012-03-01

    ALCBEAM is a new three-dimensional neutral beam formation and propagation code. It was developed to support the beam-based diagnostics installed on the Alcator C-Mod tokamak. The purpose of the code is to provide reliable estimates of the local beam equilibrium parameters: such as beam energy fractions, density profiles and excitation populations. The code effectively unifies the ion beam formation, extraction and neutralization processes with beam attenuation and excitation in plasma and neutral gas and beam stopping by the beam apertures. This paper describes the physical processes interpreted and utilized by the code, along with exploited computational methods. The description is concluded by an example simulation of beam penetration into plasma of Alcator C-Mod. The code is successfully being used in Alcator C-Mod tokamak and expected to be valuable in the support of beam-based diagnostics in most other tokamak environments. Program summaryProgram title: ALCBEAM Catalogue identifier: AEKU_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKU_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 66 459 No. of bytes in distributed program, including test data, etc.: 7 841 051 Distribution format: tar.gz Programming language: IDL Computer: Workstation, PC Operating system: Linux RAM: 1 GB Classification: 19.2 Nature of problem: Neutral beams are commonly used to heat and/or diagnose high-temperature magnetically-confined laboratory plasmas. An accurate neutral beam characterization is required for beam-based measurements of plasma properties. Beam parameters such as density distribution, energy composition, and atomic excited populations of the beam atoms need to be known. Solution method: A neutral beam is initially formed as an ion beam which is extracted from

  6. Detection of lower hybrid waves in the scrape-off layer of tokamak plasmas with microwave backscattering

    International Nuclear Information System (INIS)

    Baek, S. G.; Shiraiwa, S.; Parker, R. R.; Bonoli, P. T.; Marmar, E. S.; Wallace, G. M.; Lau, C.; Dominguez, A.; Kramer, G. J.

    2014-01-01

    Microwave backscattering experiments have been performed on the Alcator C-Mod tokamak in order to investigate the propagation of lower hybrid (LH) waves in reactor-relevant, high-density plasmas. When the line-averaged density is raised above 1 × 10 20 m –3 , lower hybrid current drive efficiency is found to be lower than expected [Wallace et al., Phys. Plasmas 19, 062505 (2012)] and LH power is thought to be dissipated at the plasma edge. Using a single channel (60 GHz) ordinary-mode (O-mode) reflectometer system, we demonstrate radially localized LH wave measurements in the scrape-off layer of high density plasmas (n ¯ e  ≳ 0.9×10 20  m −3 ). Measured backscattered O-mode power varies depending on the magnetic field line mapping, suggesting the resonance cone propagation of LH waves. Backscattered power is also sensitive to variations in plasma density and the launched parallel refractive index of the LH waves. LH ray-tracing simulations have been carried out to interpret the observed variations. To understand the measured LH waves in regions not magnetically connected to the launcher, two hypotheses are examined. One is the weak single pass absorption and the other is scattering of LH waves by non-linear effects

  7. Quantitative comparison of electron temperature fluctuations to nonlinear gyrokinetic simulations in C-Mod Ohmic L-mode discharges

    Energy Technology Data Exchange (ETDEWEB)

    Sung, C., E-mail: csung@physics.ucla.edu [University of California, Los Angeles, Los Angeles, California 90095 (United States); White, A. E.; Greenwald, M.; Howard, N. T. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Mikkelsen, D. R.; Churchill, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Holland, C. [University of California, San Diego, La Jolla, California 92093 (United States); Theiler, C. [Ecole Polytechnique Fédérale de Lausanne, SPC, Lausanne 1015 (Switzerland)

    2016-04-15

    Long wavelength turbulent electron temperature fluctuations (k{sub y}ρ{sub s} < 0.3) are measured in the outer core region (r/a > 0.8) of Ohmic L-mode plasmas at Alcator C-Mod [E. S. Marmar et al., Nucl. Fusion 49, 104014 (2009)] with a correlation electron cyclotron emission diagnostic. The relative amplitude and frequency spectrum of the fluctuations are compared quantitatively with nonlinear gyrokinetic simulations using the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] in two different confinement regimes: linear Ohmic confinement (LOC) regime and saturated Ohmic confinement (SOC) regime. When comparing experiment with nonlinear simulations, it is found that local, electrostatic ion-scale simulations (k{sub y}ρ{sub s} ≲ 1.7) performed at r/a ∼ 0.85 reproduce the experimental ion heat flux levels, electron temperature fluctuation levels, and frequency spectra within experimental error bars. In contrast, the electron heat flux is robustly under-predicted and cannot be recovered by using scans of the simulation inputs within error bars or by using global simulations. If both the ion heat flux and the measured temperature fluctuations are attributed predominantly to long-wavelength turbulence, then under-prediction of electron heat flux strongly suggests that electron scale turbulence is important for transport in C-Mod Ohmic L-mode discharges. In addition, no evidence is found from linear or nonlinear simulations for a clear transition from trapped electron mode to ion temperature gradient turbulence across the LOC/SOC transition, and also there is no evidence in these Ohmic L-mode plasmas of the “Transport Shortfall” [C. Holland et al., Phys. Plasmas 16, 052301 (2009)].

  8. RELAP5/MOD2 code assessment for the Semiscale Mod-2C Test S-LH-1

    International Nuclear Information System (INIS)

    Fineman, C.P.

    1986-01-01

    RELAP5/MOD2, Cycle 36.02, was assessed using data from Semiscale Mod-2C experiment S-LH-1. The major phenomena that occurred during the experiment were calculated by RELAP5/MOD2, although the duration and the magnitude of their effect on the transient were not always well calculated. Areas defined where further work was needed to improve the RELAP5 calculation include: (1) the system energy balance, (2) core interfacial drag, and 3) the heat transfer logic rod dryout criterion

  9. Study of parametric instabilities during the Alcator C lower hybrid wave heating experiments

    International Nuclear Information System (INIS)

    Takase, Y.

    1983-10-01

    Parametric excitation of ion-cyclotron quasi-modes (ω/sub R/ approx. = nω/sub ci/) and ion-sound quasi-modes (ω/sub R/ approx. = k/sub parallel to/v/sub ti/) during lower hybrid wave heating of tokamak plasmas have been studied in detail. Such instabilities may significantly modify the incident wavenumber spectrum near the plasma edge. Convective losses for these instabilities are high if well-defined resonance cones exist, but they are significantly reduced if the resonance cones spread and fill the plasma volume (or some region of it). These instabilities preferentially excite lower hybrid waves with larger values of n/sub parallel to/ than themselves possess, and the new waves tend to be absorbed near the outer layers of the plasma. Parametric instabilities during lower hybrid heating of Alcator C plasmas have been investigated using rf probes (to study tilde phi and tilde n/sub i/) and CO 2 scattering technique (to study tilde n/sub e/). At lower densities (anti n/sub e/ less than or equal to 0.5 x 10 14 cm -3 ) where waves observed in the plasma interior using CO 2 scattering appear to be localized, parametric decay is very weak. Both ion-sound and ion-cyclotron parametric decay processes have been observed at higher densities (anti n greater than or equal to 1.5 x 10 14 cm -3 ) where waves appear to be unlocalized. Finally, at still higher densities (anti n /sub e/ greater than or equal to 2 x 10 4 cm -3 ) pump depletion has been observed. Above these densities heating and current drive efficiencies are expected to degrade significantly

  10. A transport model for Alcator scaling in tokamaks

    International Nuclear Information System (INIS)

    Ohkawa, T.

    1978-01-01

    A theoretical model is proposed to explain the tokamak energy confinement time. With no adjustable numerical coefficients, the model predicts experimentally observed values to within a level of uncertainty consistent with the intrinsic spread of the experimental data and the necessity of calculating the confinement time without precise knowledge of the temperature profile. (Auth.)

  11. Papers presented at the eleventh topical conference on high-temperature plasma diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    This report contains the following eleven papers presented at the conference: Neutral Beam Diagnostics for Alcator C-Mod; A Study for the Installation of the TEXT HIBP on DIII-D; Time-domain Triple-probe Measurement of Edge Plasma Turbulence on TEXT-U; A Langmuir/Mach Probe Array for Edge Plasma Turbulence and Flow; Determination of Field Line Location and Safety Factor in TEXT-U; Hybrid ECE Imaging Array System for TEXT-U; First Results from the Phase Contrast Imaging System on TEXT-U; A Fast Tokamak Plasma Flux and Electron Density Reconstruction Technique; Time-series Analysis of Nonstationary Plasma Fluctuations Using Wavelet Transforms; Quantitative Modeling of 3-D Camera Views for Tokamak Divertors; and Variable-frequency Complex Demodulation Technique for Extracting Amplitude and Phase Information. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  12. Papers presented at the eleventh topical conference on high-temperature plasma diagnostics

    International Nuclear Information System (INIS)

    1996-01-01

    This report contains the following eleven papers presented at the conference: Neutral Beam Diagnostics for Alcator C-Mod; A Study for the Installation of the TEXT HIBP on DIII-D; Time-domain Triple-probe Measurement of Edge Plasma Turbulence on TEXT-U; A Langmuir/Mach Probe Array for Edge Plasma Turbulence and Flow; Determination of Field Line Location and Safety Factor in TEXT-U; Hybrid ECE Imaging Array System for TEXT-U; First Results from the Phase Contrast Imaging System on TEXT-U; A Fast Tokamak Plasma Flux and Electron Density Reconstruction Technique; Time-series Analysis of Nonstationary Plasma Fluctuations Using Wavelet Transforms; Quantitative Modeling of 3-D Camera Views for Tokamak Divertors; and Variable-frequency Complex Demodulation Technique for Extracting Amplitude and Phase Information. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  13. Investigation of fluctuations in the HDH and H* regime of Wendelstein 7-AS

    International Nuclear Information System (INIS)

    Baeumel, S.; Werner, A.; McCormick, K.

    2003-01-01

    The High Density H-Mode Regime was discovered in Island Divertor operation of the Wendelstein 7-AS (W7-AS)stellerator. This regime is characterized by low impurity, high-energy confinement times - up to twice the value of the International Stellarator Scaling ISS95 - and edge radiated power fractions of up to 90% in detached state. Regarding the enhanced impurity transport at good energy confinement there are similarities to the enhanced D α H-mode found on the Alcator C-Mod tokamak. In W7-AS studies were performed in order to compare the HDH regime with the classical ELM-free discharges (H * ). Although both regimes are similar in collisionality and have almost the same n e (r)- and T e (r)-profile shapes, the H * regime suffers a radiation collapse due to impurity accumulation. The short impurity confinement times in HDH discharges requires enhanced transport at the plasma edge. The cause is not clear and this contribution looks for similarities to the enhanced D α H-mode of Alcator C-Mod, that is, whether quasicoherent modes exist in W7-AS causing enhanced edge transport. Discharges with a variation of magnetic configurations, densities (up to 4.10 20 m -3 ) and powers (up to 3.2 MW absorbed) will be discussed with respect to the different behaviour of fluctuations. (orig.)

  14. Studies of Turbulence and Transport in Alcator C-Mod H-Mode Plasmas with Phase Contrast Imaging and Comparisons with GYRO

    Science.gov (United States)

    Porkolab, M.; Lin, L.; Edlund, E. M.; Rost, J. C.; Fiore, C. L.; Greenwald, M.; Mikkelsen, D.

    2008-11-01

    We present recent experimental measurements of turbulence and transport in C-Mod H-Mode plasmas with and without internal transport barriers (ITB) using the phase contrast imaging (PCI) diagnostic and compare the results with GYRO predictions. In plasmas without ITB, the fluctuation above 300 kHz observed by PCI agrees with ITG in GYRO simulation, including the direction of propagation, wavenumber spectrum, and absolute intensity within experimental uncertainly (+/-75%). After transition to ITBs, the observed overall fluctuation intensity increases. GYRO simulation in the core shows that ITG dominates in ITBs but its intensity is lower than the overall experimental measurements which may also include contributions from the plasma edge. These results, as well as the impact of varying ∇Ti, ∇n, and ExB shear on turbulence will be discussed. C.L. Fiore et al., Fusion Sci. Technol., 51, 303 (2007). M. Porkolab et al., IEEE Trans. Plasma Sci. 34, 229 (2006). J. Candy et al., Phys. Rev. Lett., 91, 045001 (2003).

  15. High-energy-ion depletion in the charge exchange spectrum of Alcator C

    International Nuclear Information System (INIS)

    Schissel, D.P.

    1982-01-01

    A three-dimensional, guiding center, Monte Carlo code is developed to study ion orbits in Alcator C. The highly peaked ripple of the magnetic field of Alcator is represented by an analytical expression for the vector potential. The analytical ripple field is compared to the resulting magnetic field generated by a current model of the toroidal plates; agreement is excellent. Ion-Ion scattering is simulated by a pitch angle and an energy scattering operator. The equations of motion are integrated with a variable time step, extrapolating integrator. The code produces collisionless banana and ripple trapped loss cones which agree well with present theory. Global energy distributions have been calculated and show a slight depletion above 8.5 keV. Particles which are ripple trapped and lost are at energies below where depletion is observed. It is found that ions pitch angle scatter less as energy is increased. The result is that, when viewed in velocity space, ions form probability lobes the shape of mouse ears which are fat near the thermal energy. Therefore, particles enter the loss cone at low energies near the bottom of the core. Recommendations for future work include improving the analytic model of the ripple field, testing the effect of del . B not equal to 0 on ion orbits, and improving the efficiency of the code by either using a spline fit for the magnetic fields or by creating a vectorized Monte Carlo code

  16. Some introductory notes on the problem of nuclear energy by controlled fusion reactions

    International Nuclear Information System (INIS)

    Pedretti, E.

    1988-01-01

    Written for scientists and technologist interested in, but unfamiliar with nuclear energy by controlled fusion reactions, this ''sui generis'' review paper attempts to provide the reader, as shortly as possible, with a general idea of the main issues at stake in nuclear fusion research. With the purpose of keeping this paper within a reasonable length, the various subjects are only outlined in their essence, basic features, underlying principles, etc., without entering into details, which are left to the quoted literature. Due to the particular readership of this journal, vacuum problems and/or aspects of fusion research anyhow related with vacuum science and technology are evidentiated. After reviewing fusion reactions' cross sections, fusion by accelerators and muon catalyzed fusion are described, followed by mention of Lawson's criteria and of plasma confinement features. Then, inertial confinement fusion is dealt with, also including one example of laser system (Nova), one of accelerator facility (PBFA-II) and some guesses on the classified Centurion-Halite program. Magnetic confinement fusion research is also reviewed, in particulary reporting one example of linear machine (MFTF-B), two examples of toroidal machines other than Tokamak (ATF and Eta-Beta-II) and various examples of Tokamaks, including PBX and PBX-M; TFTR, JET, JT-60, T-15 and Tore-Supra (large machines); Alcator A, FT, Alcator C/MTX, Alcator C-Mod and T-14 (compact high field machines). Tokamaks under design for ignition experiments (Ignitor, CIT, Ignitex and NET) are also illustrated. Thermal conversion of fusion power and direct generation of electricity are mentioned; conceptual design of fusion power plants are considered and illustrated by four examples (STARFIRE, WILDCAT, MARS and CASCADE). The D 3 He fuel cycle is discussed as an alternative more acceptable than Deuterium-Tritium, and thw Candor proposal is reported. After recalling past experience of the fission power development, some

  17. Microwave Tokamak Experiment: An overview of the construction and checkout phase

    International Nuclear Information System (INIS)

    Lang, L.L.; Bell, H.H.

    1989-01-01

    At Lawrence Livermore National Laboratory (LLNL) we constructed and presently operate the Microwave Tokamak Experiment (MTX) to demonstrate the feasibility of using microwave pulses produced from a free electron laser (FEL) to provide electron cyclotron heating (ECH) for use in tokamaks, particularly high-field machines. The MTX consists primarily of the ALCATOR C tokamak and power supplies that were documented and disassembled at the Massachusetts Institute of Technology (MIT) and shipped to LLNL in April 1987. We made many additions, including a new primary power system from the magnetic Fusion Test Facility (MFTF) substation, a new commutation system, substantially upgraded seismic support system for earthquake loading, a fast controls system for use with the FEL, a new data-acquisition system, and a new vault facility. We checked out these systems and put them into operation in October 1988; we achieved the first plasma in November 1988. We have also constructed and installed the microwave transmission system and the local microwave system to be used with the FEL. These systems transmit the microwaves to MTX quasi-optically through an evacuated tube. The ongoing plasma operations, both with and without FEL heating, are described in a companion paper. 12 refs., 2 figs., 2 tabs

  18. X-ray imaging crystal spectroscopy for use in plasma transport research

    Science.gov (United States)

    Reinke, M. L.; Podpaly, Y. A.; Bitter, M.; Hutchinson, I. H.; Rice, J. E.; Delgado-Aparicio, L.; Gao, C.; Greenwald, M.; Hill, K.; Howard, N. T.; Hubbard, A.; Hughes, J. W.; Pablant, N.; White, A. E.; Wolfe, S. M.

    2012-11-01

    This research describes advancements in the spectral analysis and error propagation techniques associated with x-ray imaging crystal spectroscopy (XICS) that have enabled this diagnostic to be used to accurately constrain particle, momentum, and heat transport studies in a tokamak for the first time. Doppler tomography techniques have been extended to include propagation of statistical uncertainty due to photon noise, the effect of non-uniform instrumental broadening as well as flux surface variations in impurity density. These methods have been deployed as a suite of modeling and analysis tools, written in interactive data language (IDL) and designed for general use on tokamaks. Its application to the Alcator C-Mod XICS is discussed, along with novel spectral and spatial calibration techniques. Example ion temperature and radial electric field profiles from recent I-mode plasmas are shown, and the impact of poloidally asymmetric impurity density and natural line broadening is discussed in the context of the planned ITER x-ray crystal spectrometer.

  19. Development of the gas puff charge exchange recombination spectroscopy (GP-CXRS) technique for ion measurements in the plasma edge

    International Nuclear Information System (INIS)

    Churchill, R. M.; Theiler, C.; Lipschultz, B.; Dux, R.; Pütterich, T.; Viezzer, E.

    2013-01-01

    A novel charge-exchange recombination spectroscopy (CXRS) diagnostic method is presented, which uses a simple thermal gas puff for its donor neutral source, instead of the typical high-energy neutral beam. This diagnostic, named gas puff CXRS (GP-CXRS), is used to measure ion density, velocity, and temperature in the tokamak edge/pedestal region with excellent signal-background ratios, and has a number of advantages to conventional beam-based CXRS systems. Here we develop the physics basis for GP-CXRS, including the neutral transport, the charge-exchange process at low energies, and effects of energy-dependent rate coefficients on the measurements. The GP-CXRS hardware setup is described on two separate tokamaks, Alcator C-Mod and ASDEX Upgrade. Measured spectra and profiles are also presented. Profile comparisons of GP-CXRS and a beam based CXRS system show good agreement. Emphasis is given throughout to describing guiding principles for users interested in applying the GP-CXRS diagnostic technique

  20. Is X-ray emissivity constant on magnetic flux surfaces?

    International Nuclear Information System (INIS)

    Granetz, R.S.; Borras, M.C.

    1997-01-01

    Knowledge of the elongations and shifts of internal magnetic flux surfaces can be used to determine the q profile in elongated tokamak plasmas. X-ray tomography is thought to be a reasonable technique for independently measuring internal flux surface shapes, because it is widely believed that X-ray emissivity should be constant on a magnetic flux surface. In the Alcator C-Mod tokamak, the X-ray tomography diagnostic system consists of four arrays of 38 chords each. A comparison of reconstructed X-ray contours with magnetic flux surfaces shows a small but consistent discrepancy in the radial profile of elongation. Numerous computational tests have been performed to verify these findings, including tests of the sensitivity to calibration and viewing geometry errors, the accuracy of the tomography reconstruction algorithms, and other subtler effects. We conclude that the discrepancy between the X-ray contours and the magnetic flux surfaces is real, leading to the conclusion that X-ray emissivity is not exactly constant on a flux surface. (orig.)

  1. Development of the gas puff charge exchange recombination spectroscopy (GP-CXRS) technique for ion measurements in the plasma edge

    Energy Technology Data Exchange (ETDEWEB)

    Churchill, R. M.; Theiler, C.; Lipschultz, B. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Dux, R.; Pütterich, T.; Viezzer, E. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, D-85748 Garching (Germany); Collaboration: Alcator C-Mod Team; ASDEX Upgrade Team

    2013-09-15

    A novel charge-exchange recombination spectroscopy (CXRS) diagnostic method is presented, which uses a simple thermal gas puff for its donor neutral source, instead of the typical high-energy neutral beam. This diagnostic, named gas puff CXRS (GP-CXRS), is used to measure ion density, velocity, and temperature in the tokamak edge/pedestal region with excellent signal-background ratios, and has a number of advantages to conventional beam-based CXRS systems. Here we develop the physics basis for GP-CXRS, including the neutral transport, the charge-exchange process at low energies, and effects of energy-dependent rate coefficients on the measurements. The GP-CXRS hardware setup is described on two separate tokamaks, Alcator C-Mod and ASDEX Upgrade. Measured spectra and profiles are also presented. Profile comparisons of GP-CXRS and a beam based CXRS system show good agreement. Emphasis is given throughout to describing guiding principles for users interested in applying the GP-CXRS diagnostic technique.

  2. Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

    Science.gov (United States)

    Chen, Bin

    2017-10-01

    QCMs (quasi-coherent modes) are well characterized in the edge of Alcator C-Mod, when operating in the Enhanced Dα (EDA) H-mode, a promising alternative regime for ELM (edge localized modes) suppressed operation. To improve the understanding of the physics behind the QCMs, three typical C-Mod EDA H-Mode discharges are simulated by BOUT + + using a six-field two-fluid model (based on the Braginskii equations). The simulated characteristics of the frequency versus wave number spectra of the modes is in reasonable agreement with phase contrast imaging data. The key simulation results are: 1) Linear spectrum analysis and the nonlinear phase relationship indicate the dominance of resistive-ballooning modes and drift-Alfven wave instabilities; 2) QCMs originate inside the separatrix; (3) magnetic flutter causes the mode spreading into the SOL; 4) the boundary electric field Er changes the turbulent characteristics of the QCMs and controls edge transport and the divertor heat flux width; 5) the magnitude of the divertor heat flux depends on the physics models, such as sources and sinks, sheath boundary conditions, and parallel heat flux limiting coefficient. The BOUT + + simulations have also been performed for inter-ELM periods of DIII-D and EAST discharges, and similar quasi-coherent modes have been found. The parallel electron heat fluxes projected onto the target from these BOUT + + simulations follow the experimental heat flux width scaling, in particular the inverse dependence of the width on the poloidal magnetic field with an outlier. Further turbulence statistics analysis shows that the blobs are generated near the pedestal peak gradient region inside the separatrix and contribute to the transport of the particle and heat in the SOL region. To understand the Goldston heuristic drift-based model, results will also be presented from self-consistent transport simulations with the electric and magnetic drifts in BOUT + + and with the sheath potential included in the

  3. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

    Chan, V.S.

    1994-09-01

    In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

  4. Observation of ion cyclotron range of frequencies mode conversion plasma flow drive on Alcator C-Moda)

    Science.gov (United States)

    Lin, Y.; Rice, J. E.; Wukitch, S. J.; Greenwald, M. J.; Hubbard, A. E.; Ince-Cushman, A.; Lin, L.; Marmar, E. S.; Porkolab, M.; Reinke, M. L.; Tsujii, N.; Wright, J. C.; Alcator C-Mod Team

    2009-05-01

    At modest H3e levels (n3He/ne˜8%-12%), in relatively low density D(H3e) plasmas, n¯e≤1.3×1020 m-3, heated with 50 MHz rf power at Bt0˜5.1 T, strong (up to 90 km/s) toroidal rotation (Vϕ) in the cocurrent direction has been observed by high-resolution x-ray spectroscopy on Alcator C-Mod. The change in central Vϕ scales with the applied rf power (≤30 km s-1 MW-1), and is generally at least a factor of 2 higher than the empirically determined intrinsic plasma rotation scaling. The rotation in the inner plasma (r /a≤0.3) responds to the rf power more quickly than that of the outer region (r /a≥0.7), and the rotation profile is broadly peaked for r /a≤0.5. Localized poloidal rotation (0.3≤r/a≤0.6) in the ion diamagnetic drift direction (˜2 km/s at 3 MW) is also observed, and similarly increases with rf power. Changing the toroidal phase of the antenna does not affect the rotation direction, and it only weakly affects the rotation magnitude. The mode converted ion cyclotron wave (MC ICW) has been detected by a phase contrast imaging system and the MC process is confirmed by two-dimensional full wave TORIC simulations. The simulations also show that the MC ICW is strongly damped on H3e ions in the vicinity of the MC layer, approximately on the same flux surfaces where the rf driven flow is observed. The flow shear in our experiment is marginally sufficient for plasma confinement enhancement based on the comparison of the E ×B shearing rate and gyrokinetic linear stability analysis.

  5. One-dimensional fluid model for transport in divertor and limiter tokamak scrape-off layers

    International Nuclear Information System (INIS)

    Lipschultz, B.

    1983-11-01

    Single-fluid transport in the plasma scrape-off layer is modeled for poloidal divertor and mechanically limited discharges. This numerical model is one-dimensional along a field line and time-independent. Conductive and convective transport, as well as impurity and neutral source (sink) terms are included. A simple shooting method technique is used for obtaining solutions. Results are shown for the case of the proposed Alcator DCT tokamak

  6. The Escherichia coli modE gene: effect of modE mutations on molybdate dependent modA expression.

    Science.gov (United States)

    McNicholas, P M; Chiang, R C; Gunsalus, R P

    1996-11-15

    The Escherichia coli modABCD operon, which encodes a high-affinity molybdate uptake system, is transcriptionally regulated in response to molybdate availability by ModE. Here we describe a highly effective enrichment protocol, applicable to any gene with a repressor role, and establish its application in the isolation of transposon mutations in modE. In addition we show that disruption of the ModE C-terminus abolishes derepression in the absence of molybdate, implying this region of ModE controls the repressor activity. Finally, a mutational analysis of a proposed molybdate binding motif indicates that this motif does not function in regulating the repressor activity of ModE.

  7. On the density limit of Tokamaks

    International Nuclear Information System (INIS)

    Lehnert, B.

    1982-12-01

    Under the conditions of so far performed quasi-steady tokamak experiments near the density limit, the plasma pressure gradient in the outer layers of the plasma body becomes mainly determined by the plasma-neutral gas balance. An earlier analysis of ballooning instabilities driven by this gradient in regions of bad curvature has been extended to deduce an explicit stability criterion which determines the density limit. This criterion is closely related to the empirical Murakami limit. At relevant tokamak data, the deduced limit becomes proportional to J(sub)zR(sup)1/2 where J(sub)z is the average current density and R the major plasma radius. It is further found to be independent of the toroidal magnetic field strength and anomalous transport, as well as to be a slow function of the outer layer temperature and the mass number. The deduced stability criterion is consistent with so far performed experiments. Provided that the present analysis can be extrapolated to a wider range of parameter data and be combined with Alcator scaling, conditions near ignition appear to become realizable in small tokamaks by ohmic heating alone. These conditions can be satisfied at relevant magnetic field strengths and plasma currents, by imposing a high plasma current density. (author)

  8. Numerical investigation of ICRF antenna designs and heating schemes for Alcator C

    International Nuclear Information System (INIS)

    Blackfield, D.T.; Blackwell, B.D.

    1983-02-01

    Initially, approximately 500 kW of rf power will be launched through one port of Alcator C at a frequency of 180 to 220 MHz. We use a hot-plasma slab model to examine antenna coupling and power deposition for second harmonic (H, B 0 = 6.7 T) heating regime. The plasma medium is assumed uniform in y and z and is stratified in x. Three-dimensional, four-mode field solutions are obtained by solution of the boundary condition equations in the x direction, and by Fourier transform methods in y and z. We study the dependence of radiation resistance upon coupling parameters such as antenna width and radial position. This is interpreted in terms of the spectral (k/sub y/, k/sub z/) distribution of power, the effect of the evanescent region at the edge, and the attenuation length of the waves

  9. Study of electron temperature evolution during sawtoothing and pellet injection using thermal electron cyclotron emission in the Alcator C tokamak

    International Nuclear Information System (INIS)

    Gomez, C.C.

    1986-05-01

    A study of the electron temperature evolution has been performed using thermal electron cyclotron emission. A six channel far infrared polychromator was used to monitor the radiation eminating from six radial locations. The time resolution was <3 μs. Three events were studied, the sawtooth disruption, propagation of the sawtooth generated heatpulse and the electron temperature response to pellet injection. The sawtooth disruption in Alcator takes place in 20 to 50 μs, the energy mixing radius is approx. 8 cm or a/2. It is shown that this is inconsistent with single resonant surface Kadomtsev reconnection. Various forms of scalings for the sawtooth period and amplitude were compared. The electron heatpulse propagation has been used to estimate chi e(the electron thermal diffusivity). The fast temperature relaxation observed during pellet injection has also been studied. Electron temperature profile reconstructions have shown that the profile shape can recover to its pre-injection form in a time scale of 200 μs to 3 ms depending on pellet size

  10. A minimum-size tokamak concept for conditions near ignition

    International Nuclear Information System (INIS)

    Lehnert, B.

    1983-01-01

    Based on a combination of Alcator scaling and a recent theory on the Murakami density limit, a minimum-size tokamak concept (Minitor) is proposed. Even if this concept does not aim at alpha particle containment, it has the important goal of reaching plasma core temperatures and Lawson parameter values required for ignition, by ohmic heating alone and under macroscopically stable conditions. The minimized size, and the associated enhancement of the plasma current density, are found to favour high plasma temperatues, average densities, and beta values. The goal of this concept appears to be realizable by relatively modest technical means. (author)

  11. Infrared laser diagnostics for ITER

    International Nuclear Information System (INIS)

    Hutchinson, D.P.; Richards, R.K.; Ma, C.H.

    1995-01-01

    Two infrared laser-based diagnostics are under development at ORNL for measurements on burning plasmas such as ITER. The primary effort is the development of a CO 2 laser Thomson scattering diagnostic for the measurement of the velocity distribution of confined fusion-product alpha particles. Key components of the system include a high-power, single-mode CO 2 pulsed laser, an efficient optics system for beam transport and a multichannel low-noise infrared heterodyne receiver. A successful proof-of-principle experiment has been performed on the Advanced Toroidal Facility (ATF) stellerator at ORNL utilizing scattering from electron plasma frequency satellites. The diagnostic system is currently being installed on Alcator C-Mod at MIT for measurements of the fast ion tail produced by ICRH heating. A second diagnostic under development at ORNL is an infrared polarimeter for Faraday rotation measurements in future fusion experiments. A preliminary feasibility study of a CO 2 laser tangential viewing polarimeter for measuring electron density profiles in ITER has been completed. For ITER plasma parameters and a polarimeter wavelength of 10.6 microm, a Faraday rotation of up to 26 degree is predicted. An electro-optic polarization modulation technique has been developed at ORNL. Laboratory tests of this polarimeter demonstrated a sensitivity of ≤ 0.01 degree. Because of the similarity in the expected Faraday rotation in ITER and Alcator C-Mod, a collaboration between ORNL and the MIT Plasma Fusion Center has been undertaken to test this polarimeter system on Alcator C-Mod. A 10.6 microm polarimeter for this measurement has been constructed and integrated into the existing C-Mod multichannel two-color interferometer. With present experimental parameters for C-Mod, the predicted Faraday rotation was on the order of 0.1 degree. Significant output signals were observed during preliminary tests. Further experiment and detailed analyses are under way

  12. Atomic physics for fusion plasma spectroscopy; a soft x-ray study of molybdenum ions

    International Nuclear Information System (INIS)

    Fournier, K.B.

    1996-01-01

    Understanding the radiative patterns of the ions of heavy atoms (Z approx-gt 18) is crucial to fusion experiments. The present thesis applies ab initio, relativistic calculations of atomic data to modeling the emission of molybdenum (Z = 42) ions in magnetically confined fusion plasmas. The models are compared to observations made in the Alcator C-Mod tokamak (Plasma Fusion Center, Massachusetts Institute of Technology), and the Frascati Tokamak Upgrade. Experimental confirmation of these models allows confidence in calculations of the total molybdenum concentration and quantitative estimates of the total power lost from the plasmas due to molybdenum line radiation. Charge states in the plasma core (Mo 33+ to Mo 29+ ) emit strong x-ray and XUV spectra which allow benchmarking of models for the spatial distribution of highly stripped molybdenum ions; the models only achieve agreement with observations when the rates of indirect ionization and recombination processes are included in the calculation of the charge state distribution of the central molybdenum ions. The total concentration of molybdenum in the core of the plasma is found, and the total power radiated from the plasma core is computed. Observations of line emission from more highly charged molybdenum ions (Mo 36+ to Mo 34+ ) are presented. open-quotes Bulkclose quotes molybdenum charge states (Mo 25+ to Mo 23+ ) emit complicated XUV spectra from a position in the plasma near C-Mod's half radius; spatial profiles of these ions' emission are analyzed. Models for the line-emission spectra of adjacent ions (Mo 28+ to Mo 26+ ) are offered, and the accuracy and limits of ab initio energy level calculations are discussed. open-quotes Edgeclose quotes charge states (Mo 22+ to Mo 15 ) extend to the last closed magnetic flux surface of the C-Mod plasma. The strongest features from these charge states are emitted in a narrow band from ∼70 Angstrom

  13. Low-temperature operating regime of the tokamak evacuating limiter

    International Nuclear Information System (INIS)

    Tokar', M.Z.

    1987-01-01

    The conditions for realizing the regime of strong recycling of a cold dense plasma of an evacuating limiter were determined based on a previously proposed model for describing the limiter layer of a tokamak. The scaling for the dependence of the gas pressure in the evacuation system on the average plasma density in the limiter layer was found, and agreed quantitatively with the results of measurements on the Alcator and ISX-B tokamaks. For the tokamak reactor of the INTOR scale the calculations show that the low-temperature operating regime of the evacuating limiter can be realized with a quite low pumping rate. It has the advantages of reduced erosion of the limiter and small fluxes of impurities into the working volume of the reactor. In addition, the relative concentration of the helium ash in the limiter layer does not exceed 2-3%, but the density of the main plasma is comparable to the proposed average density in the reactor. The concept of a stochastic limiter is of interest for lowering the plasma density in the limiter layer and lowering the thermal loads on the limiter

  14. User's manual for DSTAR MOD1: A comprehensive tokamak disruption code

    International Nuclear Information System (INIS)

    Merrill, B.J.; Jardin, S.J.

    1986-01-01

    A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces that can occur during major tokamak plasma disruptions. The DSTAR code development effort has been accomplished by coupling a recently developed free boundary tokamak plasma transport computational model with other models developed to predict impurity transport and radiation, and the electromagnetic and thermal dynamic response of vacuum vessel components. The combined model, DSTAR, is a unique tool for predicting the consequences of tokamak disruptions. This informal report discusses the sequence of events of a resistive disruption, models developed to predict plasma transport and electromagnetic field evolution, the growth of the stochastic region of the plasma, the transport and nonequilibrium ionization/emitted radiation of the ablated vacuum vessel material, the vacuum vessel thermal and magnetic response, and user input and code output

  15. FY-2013 FES (Fusion Energy Sciences) Joint Research Target Report

    Energy Technology Data Exchange (ETDEWEB)

    Fenstermacher, M. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Garofalo, A. M. [General Atomics, San Diego, CA (United States); Gerhardt, S. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Hubbard, A. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Maingi, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Whyte, D. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2013-09-30

    investigated. The research will strengthen the basis for extrapolation of stationary regimes which combine high energy confinement with good particle and impurity control, to ITER and other future fusion facilities for which avoidance of large ELMs is a critical issue. Data from the Alcator C-Mod tokamak (MIT), DIII-D tokamak (General Atomics), and NSTX spherical tokamak (PPPL) contribute to this report. Experiments specifically motivated by this research target were conducted on DIII-D, with a national team of researchers from GA, LLNL, PPPL, MIT and ORNL contributing. Both the Alcator C-Mod and NSTX-U teams contributed analysis of previously collected data, as those two facilities did not operate in FY2013. Within each of the three research groups, members from both the host institutions and collaborating institutions made critical contributions. Highlights from these research activities are provided, with additional details.

  16. Lower hybrid current drive in tokamak plasmas

    International Nuclear Information System (INIS)

    Ushigusa, Kenkichi

    1999-03-01

    Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar e - 10 20 m -3 , ALCATOR-C) and the highest current drive efficiency (η CD = 3.5x10 19 m -2 A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)

  17. Observation of Self-Generated Flows in Tokamak Plasmas with Lower-Hybrid-Driven Current

    International Nuclear Information System (INIS)

    Ince-Cushman, A.; Rice, J. E.; Reinke, M.; Greenwald, M.; Wallace, G.; Parker, R.; Fiore, C.; Hughes, J. W.; Bonoli, P.; Shiraiwa, S.; Hubbard, A.; Wolfe, S.; Hutchinson, I. H.; Marmar, E.; Bitter, M.; Wilson, J.; Hill, K.

    2009-01-01

    In Alcator C-Mod discharges lower hybrid waves have been shown to induce a countercurrent change in toroidal rotation of up to 60 km/s in the central region of the plasma (r/a∼<0.4). This modification of the toroidal rotation profile develops on a time scale comparable to the current redistribution time (∼100 ms) but longer than the energy and momentum confinement times (∼20 ms). A comparison of the co- and countercurrent injected waves indicates that current drive (as opposed to heating) is responsible for the rotation profile modifications. Furthermore, the changes in central rotation velocity induced by lower hybrid current drive (LHCD) are well correlated with changes in normalized internal inductance. The application of LHCD has been shown to generate sheared rotation profiles and a negative increment in the radial electric field profile consistent with a fast electron pinch

  18. Confinement of ohmically heated plasmas and turbulent heating in high-magnetic field tokamak TRIAM-1

    Energy Technology Data Exchange (ETDEWEB)

    Hiraki, N; Itoh, S; Kawai, Y; Toi, K; Nakamura, K [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

    1979-12-01

    TRIAM-1, the tokamak device with high toroidal magnetic field, has been constructed to establish the scaling laws of advanced tokamak devices such as Alcator, and to study the possibility of the turbulent heating as a further economical heating method of the fusion oriented plasmas. The plasma parameters obtained by ohmic heating alone are as follows; central electron temperature T sub(e0) = 640 eV, central ion temperature T sub(i0) = 280 eV and line-average electron density n average sub(e) = 2.2 x 10/sup 14/ cm/sup -3/. The empirical scaling laws are investigated concerning T sub(e0), T sub(i0) and n average sub(e). The turbulent heating has been carried out by applying the high electric field in the toroidal direction to the typical tokamak discharge with T sub(i0) asymptotically equals 200 eV. The efficient ion heating is observed and T sub(i0) attains to about 600 eV.

  19. Electron diffusion in tokamaks due to electromagnetic fluctuations

    International Nuclear Information System (INIS)

    Horton, W.; Choi, D.I.; Yushmanov, P.N.; Parail, V.V.

    1986-05-01

    Calculations for the stochastic diffusion of electrons in tokamaks due to a spectrum of electromagnetic drift fluctuations are presented. The parametric dependence of the diffusion coefficient on the amplitude and phase velocity of the spectrum, and the bounce frequency for the electrons is studied. The wavenumber spectrum is taken to be a low order (5 x 5) randomly-phased, isotropic, Monotonic spectrum extending from k /sub perpendicular min/ approx. = ω/sub ci//c/sub s/ to k/sub perpendicular max/ approx. = 3ω/sub pe//c with different power laws of decrease phi k approx. = phi 1/k/sup m/, 1 less than or equal to m less than or equal to 3. A nonlinear Ohm's law is derived for the self-consistent relation between the electrostatic and parallel vector potentials. The parallel structure of the fluctuations is taken to be such that k parallel/sup nl/upsilon/sub e/ < w/sub k/ due to the nonlinear perpendicular motion of the electrons described in the nonlinear Ohm's law. The diffusion coefficient scales approximately as the neo-Alcator and Merezhkin-Mukhovatoc empirical formulas for plasma densities above a critical density

  20. The rate of plasma heating by harmonic ion cyclotron waves in tokamaks

    International Nuclear Information System (INIS)

    Moslehi-Fard, M.; Sobhanian, S.; Solati-Kia, F.

    2002-01-01

    In tokamaks, the toroidal magnetic field, B φ , is due to the current in coils around plasma, and the poloidal magnetic field B p results from the plasma itself. Usually B φ p , and the combination of these two fields forms a nested set of toroidal magnetic surfaces. The equilibrium Grad-Shafranov equation is investigated and it is shown that the particle products of fusion with different pitch angles on these surfaces have different orbital shapes. In the JET tokamak, the α particles with pitch angle θ smaller than 54.8 deg are passing, those with θ between 54.8 deg and 65.1 deg have trapping-passing orbits but for θ greater than 65.1 deg the orbit has a banana form. Other tokamaks such as Alcator and ITER are also considered. The passing, trapping-passing and banana orbits in these tokamaks are traced. The results obtained from this calculation are analyzed. The wave damping has been investigated produced from interaction with particles, particularly α particles, and the rate of heating for l = 1 to 8 harmonics is plotted. The results of calculation show that heating at the fourth harmonic reaches a maximum. For higher harmonics, the heating does not change much from the fourth harmonic. (author)

  1. Electron diffusion in tokamaks due to electromagnetic fluctuations

    International Nuclear Information System (INIS)

    Horton, W.; Choi, D.-I.; Yushmanov, P.N.; Parail, V.V.

    1987-01-01

    Calculations for the stochastic diffusion of electrons in Tokamaks due to a spectrum of electromagnetic drift fluctuations are presented. The parametric dependence of the diffusion coefficient on the amplitude and phase velocity of the spectrum, and the bounce frequency for the electrons is studied. The wavenumber spectrum is taken to be a low order (5 x 5) randomly-phased, isotropic, monotonic spectrum extending from k sub(perpendicular to min) ≅ ωsub(ci)/Csub(s) to k sub(perpendicular to max) ≅ 3ωsub(pe)/C with different power laws of decrease φsub(k) ≅ φ 1 /ksup(m), 1 ≤ m ≤ 3. A nonlinear Ohm's law is derived for the self-consistent relation between the electrostatic and parallel vector potentials. The parallel structure of the fluctuations is taken to be such that ksup(nl)sub(parallel to)Vsub(e) < ωsub(k) due to the nonlinear perpendicular motion of the electrons described in the nonlinear Ohm's law. The diffusion coefficient scales approximately as the neo-Alcator and Merezhkin-Mukhovatov empirical formulas for plasma densities below a critical density. (author)

  2. Design and operation of the RFX-mod plasma shape control system

    Energy Technology Data Exchange (ETDEWEB)

    Marchiori, G., E-mail: giuseppe.marchiori@igi.cnr.it [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Finotti, C. [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Kudlacek, O. [Università di Padova, Padova (Italy); Villone, F. [Dipartimento di Ingegneria Elettrica e dell’Informazione (DIEI), Università di Cassino (Italy); Zanca, P. [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Abate, D. [Dipartimento di Ingegneria Elettrica e dell’Informazione (DIEI), Università di Cassino (Italy); Cavazzana, R. [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Jackson, G.L.; Luce, T.C. [General Atomics, San Diego, CA (United States); Marrelli, L. [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy)

    2016-10-15

    Highlights: • Linearized plasma response model of RFX-mod Tokamak Double/Single Null discharges. • Model based design of a vertical stability control system. • Model based design of a plasma shape LQG control system with Kalman state estimator. • Real time plasma boundary reconstruction algorithm. • Tracking and disturbance rejection experimental tests. - Abstract: The aim of executing Single Null discharges in RFX-mod operating as a Tokamak led to the design and implementation of a plasma shape feedback control system. A fully model-based approach was followed which allowed dealing with critical issues such as the presence of a conducting shell, the strong coupling of the poloidal field coils and the voltage limits of the power supplies. A Linear Quadratic regulator and a Kalman state estimator were designed and implemented in the real time MARTe framework together with an algorithm for the real-time plasma boundary reconstruction. The problem of a number of sensors along the poloidal direction adequate only for circular discharges was also successfully tackled. The development of the system and its performances in terms of tracking and disturbance rejection capability are presented in the paper.

  3. Final Report: Spectral Analysis of L-shell Data in the Extreme Ultraviolet from Tokamak Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lepson, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jernigan, J. Garrett [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beiersdorfer, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-02-05

    We performed detailed analyses of extreme ultraviolet spectra taken by Lawrence Livermore National Laboratory on the National Spherical Torus Experiment at Princeton Plasma Physics Laboratory and on the Alcator CKmod tokamak at the M.I.T. Plasma Science and Fusion Center. We focused on the emission of iron, carbon, and other elements in several spectral band pass regions covered by the Atmospheric Imaging Assembly on the Solar Dynamics Observatory. We documented emission lines of carbon not found in currently used solar databases and demonstrated that this emission was due to charge exchange.

  4. Intra-shot MSE Calibration Technique For LHCD Experiments

    International Nuclear Information System (INIS)

    Ko, Jinseok; Scott, Steve; Shiraiwa, Syun'ichi; Greenwald, Martin; Parker, Ronald; Wallace, Gregory

    2009-01-01

    The spurious drift in pitch angle of order several degrees measured by the Motional Stark Effect (MSE) diagnostic in the Alcator C-Mod tokamak1 over the course of an experimental run day has precluded direct utilization of independent absolute calibrations. Recently, the underlying cause of the drift has been identified as thermal stress-induced birefringence in a set of in-vessel lenses. The shot-to-shot drift can be avoided by using MSE to measure only the change in pitch angle between a reference phase and a phase of physical interest within a single plasma discharge. This intra-shot calibration technique has been applied to the Lower Hybrid Current Drive (LHCD) experiments and the measured current profiles qualitatively demonstrate several predictions of LHCD theory such as an inverse dependence of current drive efficiency on the parallel refractive index and the presence of off-axis current drive.

  5. Laboratory calibration of density-dependent lines in the extreme ultraviolet spectral region

    Science.gov (United States)

    Lepson, J. K.; Beiersdorfer, P.; Gu, M. F.; Desai, P.; Bitter, M.; Roquemore, L.; Reinke, M. L.

    2012-05-01

    We have been making spectral measurements in the extreme ultraviolet (EUV) from different laboratory sources in order to investigate the electron density dependence of various astrophysically important emission lines and to test the atomic models underlying the diagnostic line ratios. The measurement are being performed at the Livermore EBIT-I electron beam ion trap, the National Spherical Torus Experiment (NSTX) at Princeton, and the Alcator C-Mod tokamak at the Massachusetts Institute of Technology, which together span an electron density of four orders of magnitude and which allow us to test the various models at high and low density limits. Here we present measurements of Fe XXII and Ar XIV, which include new data from an ultra high resolution (λ/Δλ >4000) spectrometer at the EBIT-I facility. We found good agreement between the measurements and modeling calculations for Fe XXII, but poorer agreement for Ar XIV.

  6. Experiences with remote collaborations in fusion research

    International Nuclear Information System (INIS)

    Wurden, G.A.; Davis, S.; Barnes, D.

    1998-03-01

    The magnetic fusion research community has considerable experience in placing remote collaboration tools in the hands of real user. The ability to remotely view operations and to control selected instrumentation and analysis tasks has been demonstrated. University of Wisconsin scientists making turbulence measurements on TFTR: (1) were provided with a remote control room from which they could operate their diagnostic, while keeping in close contact with their colleagues in Princeton. LLNL has assembled a remote control room in Livermore in support of a large, long term collaboration on the DIII-D tokamak in San Diego. (2) From the same control room, a joint team of MIT and LLNL scientists has conducted full functional operation of the Alcator C-Mod tokamak located 3,000 miles away in Cambridge Massachusetts. (3) These early efforts have been highly successful, but are only the first steps needed to demonstrate the technical feasibility of a complete facilities on line environment. These efforts have provided a proof of principle for the collaboratory concept and they have also pointed out shortcomings in current generation tools and approaches. Current experiences and future directions will be discussed

  7. Lower hybrid current drive in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1999-03-01

    Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar{sub e} - 10{sup 20}m{sup -3}, ALCATOR-C) and the highest current drive efficiency ({eta}{sub CD} = 3.5x10{sup 19} m{sup -2}A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)

  8. The design of a second harmonic tangential array interferometer for C-Mod

    International Nuclear Information System (INIS)

    Bretz, N.; Jobes, F.; Irby, J.

    1997-01-01

    A design for a tangential array interferometer for C-Mod operating at 1.06 and 0.53 μm is presented. This is a special type of two color interferometer in which a Nd:YAG laser is frequency doubled in a nonlinear crystal. Because the doubling efficiency is imperfect, two frequencies propagate collinearly through the plasma after which the 1.06 μm ray is doubled again mixing in the optical domain with the undoubled ray. The resulting interference is insensitive to path length but is affected by plasma dispersion in the usual way. A typical central fringe shift in C-Mod is expected to be 0.1 endash 1.0, but the absolute and relative accuracy in n e l measurements can be as high as in a conventional interferometer. This design uses a repetitively pulsed laser which is converted to a fan beam crossing the horizontal midplane. The chordal array is defined by internal retroreflectors on the C-Mod midplane which return the beam to the second doubler and a detector array. This interferometer design has beam diameters of a few millimeters and element spacings of a few centimeters, uses a repetitively pulsed, TEM 00 Nd:YAG laser, fiber optic beam transport, commercial components, and a compact optical design which minimizes port space requirements. An optical system design is presented which is based on the performance of a tabletop prototype at Princeton Plasma Physics Laboratory. copyright 1997 American Institute of Physics

  9. Avoidance of Tearing Mode Locking and Disruption with Electro-Magnetic Torque Introduced by Feedback-based Mode Rotation Control in DIII-D and RFX-mod

    Energy Technology Data Exchange (ETDEWEB)

    Okabayashi, M. [PPPL; Zanca, P. [Euratom-ENEA; Strait, E. J. [General Atomics

    2014-09-01

    Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. Here we have demonstrated a very promising scheme to avoid such disruptions by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque to the modes is created by a toroidal phase shift between the externally-applied field and the excited TM fields, compensating for the mode momentum loss due to the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance is provided by a feedback scheme. We have explored this approach in two vastly different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high βN plasmas in a non-circular divertor tokamak. In RFX-mod, the plasma was ohmically-heated plasma with ultralow safety factor in a circular limiter discharge of active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. The internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited to this purpose.

  10. Self-organized 3D equilibrium formation and its feedback control in RFX-mod

    International Nuclear Information System (INIS)

    Piovesan, P.; Bonfiglio, D.; Marrelli, L.; Soppelsa, A.; Spolaore, M.; Terranova, D.

    2014-01-01

    The reversed-field pinch exhibits a strong tendency to self-organize into a helical equilibrium as the plasma current is increased. The helical reversed-field pinch is characterized by reduced magnetic stochasticity and by the formation of electron internal transport barriers. The paper gives an update on recent experimental and modelling work on helical states in RFX-mod (Sonato et al 2003 Fusion Eng. Des. 66 161), also discussing similarities with 3D equilibria in tokamaks. The helical equilibrium is modelled with 3D codes developed for stellarators, such as VMEC/V3FIT. The reconstructed safety factor profile has low or reversed magnetic shear in the core, which may be related to transport barrier formation. A significant extension of the RFX-mod database to high current and density confirms the dependence observed before of various helical state properties on macroscopic quantities. Even under conditions where it does not form spontaneously, such as at low current or high density, the 3D magnetic equilibrium can be stimulated and robustly controlled with external fields applied by an extensive set of non axi-symmetric coils. An advanced magnetic feedback algorithm that compensates for error fields induced by eddy currents in the 3D wall structures has been developed. This work stimulated similar experiments in RFX-mod run as a tokamak, where external 3D fields are applied to control a m = 1/n = 1 helical equilibrium. (paper)

  11. Three-dimensional simulations of plasma turbulence in the RFX-mod scrape-off layer and comparison with experimental measurements

    Science.gov (United States)

    Riva, Fabio; Vianello, Nicola; Spolaore, Monica; Ricci, Paolo; Cavazzana, Roberto; Marrelli, Lionello; Spagnolo, Silvia

    2018-02-01

    The tokamak scrape-off layer (SOL) plasma dynamics is investigated in a circular limiter configuration with a low edge safety factor. Focusing on the experimental parameters of two ohmic tokamak inner-wall limited plasma discharges in RFX-mod [Sonato et al., Fusion Eng. Des. 74, 97 (2005)], nonlinear SOL plasma simulations are performed with the GBS code [Ricci et al., Plasma Phys. Controlled Fusion 54, 124047 (2012)]. The numerical results are compared with the experimental measurements, assessing the reliability of the GBS model in describing the RFX-mod SOL plasma dynamics. It is found that the simulations are able to quantitatively reproduce the RFX-mod experimental measurements of the electron plasma density, electron temperature, and ion saturation current density (jsat) equilibrium profiles. Moreover, there are indications that the turbulent transport is driven by the same instability in the simulations and in the experiment, with coherent structures having similar statistical properties. On the other hand, it is found that the simulation results are not able to correctly reproduce the floating potential equilibrium profile and the jsat fluctuation level. It is likely that these discrepancies are, at least in part, related to simulating only the tokamak SOL region, without including the plasma dynamics inside the last close flux surface, and to the limits of applicability of the drift approximation. The turbulence drive is then identified from the nonlinear simulations and with the linear theory. It results that the inertial drift wave is the instability driving most of the turbulent transport in the considered discharges.

  12. Measurements of ion cyclotron range of frequencies mode converted wave intensity with phase contrast imaging in Alcator C-Mod and comparison with full-wave simulations

    International Nuclear Information System (INIS)

    Tsujii, N.; Porkolab, M.; Bonoli, P. T.; Lin, Y.; Wright, J. C.; Wukitch, S. J.; Jaeger, E. F.; Green, D. L.; Harvey, R. W.

    2012-01-01

    Radio frequency waves in the ion cyclotron range of frequencies (ICRF) are widely used to heat tokamak plasmas. In ICRF heating schemes involving multiple ion species, the launched fast waves convert to ion cyclotron waves or ion Bernstein waves at the two-ion hybrid resonances. Mode converted waves are of interest as actuators to optimise plasma performance through current drive and flow drive. In order to describe these processes accurately in a realistic tokamak geometry, numerical simulations are essential, and it is important that these codes be validated against experiment. In this study, the mode converted waves were measured using a phase contrast imaging technique in D-H and D- 3 He plasmas. The measured mode converted wave intensity in the D- 3 He mode conversion regime was found to be a factor of ∼50 weaker than the full-wave predictions. The discrepancy was reduced in the hydrogen minority heating regime, where mode conversion is weaker.

  13. Transport and Stability in C-Mod ITBs in Diverse Regimes

    Science.gov (United States)

    Fiore, C. L.; Ernst, D. R.; Howard, N. T.; Kasten, C. P.; Mikkelsen, D.; Reinke, M. L.; Rice, J. E.; White, A. E.; Rowan, W. L.; Bespamyatnov, I.

    2012-10-01

    Internal Transport Barriers (ITBs) in C-Mod feature highly peaked density and pressure profiles and are typically induced by the introduction of radio frequency power in the ion cyclotron range of frequencies (ICRF) with the second harmonic of the resonance for minority hydrogen ions positioned off-axis at the plasma half radius on either the low or high field side of the plasma. These ITBs are formed in the absence of particle or momentum injection, and with monotonic q profiles with qminITB dynamics in a reactor relevant regime. Recently, linear and non-linear gyrokinetic simulations have demonstrated that changes in the ion temperature and plasma rotation profiles, coincident with the application of off-axis ICRF heating, contribute to greater stability to ion temperature gradient driven fluctuation in the plasma. This results in reduced turbulent driven outgoing heat flux. To date, ITB formation in C-Mod has only been observed in EDA H-mode plasmas with moderate (2-3 MW) ICRF power. Experiments to explore the formation of ITBs in other operating regimes such as I-mode and also with high ICRF power are being undertaken to understand further the process of ITB formation and sustainment, especially with regard to turbulent driven transport.

  14. The physics of an ignited tokamak

    International Nuclear Information System (INIS)

    Troyon, F.

    1990-10-01

    There appears to be a consensus that time has come to embark on the design and construction of the next generation of tokamaks which is at the origin of the ITER initiative. Different proposals have been made based on different appreciation as to the size of the step which can be taken, related to considerations of cost, risk and duration of construction. A class of devices which may be considered the last the very high-field, high density ALCATOR-Frascati line of tokamaks have been proposed for some years specifically for this purpose. Today there remain three such projects: Ignitor, Ignitex and CIT. The technology chosen limits the pulse length to a few seconds. These devices have evolved through the years becoming larger and much more expensive than originally anticipated, increasing the pressure to do more than just a simple demonstration of ignition. There is another class of more ambitious devices which aim at creating long burning plasmas in conditions as close as possible to those of a tokamak reactor in order to address all the plasma physics problems associated with long burn. Three such projects, NET, the european next step after JET, ITER and JIT are good examples of this approach. The ideal would be to design a device with sufficient margin to study burning plasmas over a wide range of parameters. The object of this didactic presentation is to describe the common physics basis of all these projects, compare their expected performance using present knowledge and list the physics problems associated with a burning plasma experiment. The comparison is not meant to be a judgement since the important parameter is the cost/benefit ratio which is a matter of appreciation at this stage. 6 refs., 3 figs., 1 tab

  15. An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

    International Nuclear Information System (INIS)

    Hall, D.G.; Watkins, J.C.

    1987-01-01

    This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

  16. 13C-Tracer Experiments in DIII-D Preliminary to Thermal Oxidation Experiments to Understand Tritium Recovery in DIII-D, JET, C-Mod, and MAST

    International Nuclear Information System (INIS)

    Stangeby, P.; Allen, S.; Bekris, N.; Brooks, N.; Christie, K.; Chrobak, C.; Coad, J.; Counsell, G.; Davis, J.; Elder, J.; Fenstermacher, M.; Groth, M.; Haasz, A.; Likonen, J.; Lipschultz, B.; McLean, A.; Philipps, V.; Porter, G.; Rudakov, D.; Shea, J.; Wampler, W.; Watkins, J.; West, W.; Whyte, D.

    2006-01-01

    Retention of tritium in carbon co-deposits is a serious concern for ITER. Developing a reliable in-situ removal method of the co-deposited tritium would allow the use of carbon plasma-facing components which have proven reliable in high heat flux conditions and compatible with high performance plasmas. Thermal oxidation is a potential solution, capable of reaching even hidden locations. It is necessary to establish the least severe conditions to achieve adequate tritium recovery, minimizing damage and reconditioning time. The first step in this multi-machine project is 13 C-tracer experiments in DIII-D, JET, C-Mod and MAST. In DIII-D and JET, 13 CH 4 has been (and in C-Mod and MAST, will be) injected toroidally symmetrically, facilitating quantification and interpretation of the results. Tiles have been removed, analyzed for 13 C content and will next be evaluated in a thermal oxidation test facility in Toronto with regard to the ability of different severities of oxidation exposure to remove the different types of (known and measured) 13 C co-deposit. Removal of D/T from B on Mo tiles from C-Mod will also be tested. OEDGE interpretive code analysis of the 13 C deposition patterns is used to generate the understanding needed to apply findings to ITER. First results are reported here for the 13 C injection experiments IN DIII-D

  17. Laudation Laudation

    Science.gov (United States)

    Bychkov, A.

    2015-01-01

    During the opening session of the 2014 IAEA Fusion Energy Conference in St Petersburg, Russian Federation, I was pleased to present the 2013 and 2014 Nuclear Fusion journal prizes. The IAEA's monthly journal, Nuclear Fusion, has been publishing preeminent research in controlled thermonuclear fusion for over 50 years. In 2006, we inaugurated an annual journal prize to recognise outstanding contributions. This prize is presented to the lead author of the paper judged by the journal's Board of Editors to have made the greatest scientific contribution in the two years following its publication. The 2013 prize was awarded to Dennis Whyte, Massachusetts Institute of Technology, for the 2010 paper: 'I-mode: an H-mode energy confinement regime with L-mode particle transport in Alcator C-Mod' [1] This ground breaking paper, presenting results from Alcator C-Mod, enhances our understanding of the formation of energy transport barriers and temperature pedestals, without particle barriers, through the I-mode regime. The discovery of a stationary ELM-free improved confinement regime with no impurity accumulation in a metallic high field tokamak, like ITER, has implications that will stimulate much future research. The 2014 prize was awarded to Philip Snyder, General Atomics, USA, for the 2011 paper: 'A first-principles predictive model of the pedestal height and width: development, testing and ITER optimization with the EPED model' [2]. Pedestal height will have a dramatic impact on overall fusion performance in next-step devices. This exceptional paper presents a compelling model for the edge pedestal width and height based on coupling peeling-ballooning theory for stability and kinetic ballooning transport theory. Comparison is made to experimental observations across a range of devices and convincing agreement is demonstrated. This model, therefore, has the potential to significantly focus the predictions of performance in future devices. I congratulate the prize winners and

  18. Effect of the scrape-off layer in AORSA full wave simulations of fast wave minority, mid/high harmonic, and helicon heating regimes

    Energy Technology Data Exchange (ETDEWEB)

    Bertelli, N., E-mail: nbertell@pppl.gov; Gerhardt, S.; Hosea, J. C.; LeBlanc, B.; Perkins, R. J.; Phillips, C. K.; Taylor, G.; Valeo, E. J.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaeger, E. F. [XCEL Engineering Inc., Oak Ridge, TN 37830 (United States); Lau, C.; Blazevski, D.; Green, D. L.; Berry, L.; Ryan, P. M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Bonoli, P. T.; Wright, J. C. [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Pinsker, R. I.; Prater, R. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Qin, C. M. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); and others

    2015-12-10

    Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves, have found strong interactions between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 3D AORSA results for the National Spherical Torus eXperiment (NSTX), where a full antenna spectrum is reconstructed, are shown, confirming the same behavior found for a single toroidal mode results in Bertelli et al, Nucl. Fusion, 54 083004, 2014, namely, a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is moved away from in front of the antenna by increasing the edge density. Additionally, full wave simulations have been extended to “conventional” tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for Alcator C-Mod and EAST, which operate in the minority heating regime unlike NSTX/NSTX-U and DIII-D, which operate in the mid/high harmonic regime. A substantial discussion of some of the main aspects, such as (i) the pitch angle of the magnetic field; (ii) minority heating vs. mid/high harmonic regimes is presented showing the different behavior of the RF field in the SOL region for NSTX-U scenarios with different plasma current. Finally, the preliminary results of the impact of the SOL region on the evaluation of the helicon current drive efficiency in DIII-D is presented for the first time and briefly compared with the different regimes

  19. Momentum transport studies from multi-machine comparisons

    International Nuclear Information System (INIS)

    Yoshida, M.; Kamada, Y.; Sakamoto, Y.; Kaye, S.; Solomon, W.; Bell, R.E.; Rice, J.; Podpaly, Y.; Reinke, M.L.; Tala, T.; Salmi, A.; Burrell, K.H.; Ferreira, J.; McDonald, D.; Mantica, P.

    2012-01-01

    A database of toroidal momentum transport on five tokamaks, Alcator C-Mod, DIII-D, JET, NSTX and JT-60U, has been constructed under a wide range of conditions in order to understand the characteristics of toroidal momentum transport coefficients, namely the toroidal momentum diffusivity (χ φ ) and the pinch velocity (V pinch ). Through an inter-machine comparison, the similarities and differences in the properties of χ φ and V pinch among the machines have been clarified. Parametric dependences of these momentum transport coefficients have been investigated over a wide range of plasma parameters taking advantage of the different operation regimes in machines. The approach offers insights into the parametric dependences as follows. The toroidal momentum diffusivity (χ φ ) generally increases with increasing heat diffusivity (χ i ). The correlation is observed over a wide range of χ φ , covering roughly two orders of magnitude, and within each of the machines over the whole radius. Through the inter-machine comparison, it is found that χ φ becomes larger in the outer region of the plasma. Also observed is a general trend for V pinch in tokamaks; the inward pinch velocity (−V pinch ) increases with increasing χ φ . The results that are commonly observed in machines will support a toroidal rotation prediction in future devices. On the other hand, differences among machines have been observed. The toroidal momentum diffusivity, χ φ , is larger than or equal to χ i in JET and JT-60U; on the other hand, χ φ is smaller than or equal to χ i in NSTX, DIII-D and Alcator C-Mod. In DIII-D, the ratio −RV pinch /χ φ at r/a = 0.5–0.6 is about 2, which is small compared with that in other tokamaks (−RV pinch /χ φ ≈ 5). Based on these different observations, parametric dependences of χ φ /χ i , RV pinch /χ φ and χ φ have been investigated in H-mode plasmas. Across the dataset from all machines, the ratio χ φ /χ i tends to be larger in low

  20. Validation of TGLF in C-Mod and DIII-D using machine learning and integrated modeling tools

    Science.gov (United States)

    Rodriguez-Fernandez, P.; White, Ae; Cao, Nm; Creely, Aj; Greenwald, Mj; Grierson, Ba; Howard, Nt; Meneghini, O.; Petty, Cc; Rice, Je; Sciortino, F.; Yuan, X.

    2017-10-01

    Predictive models for steady-state and perturbative transport are necessary to support burning plasma operations. A combination of machine learning algorithms and integrated modeling tools is used to validate TGLF in C-Mod and DIII-D. First, a new code suite, VITALS, is used to compare SAT1 and SAT0 models in C-Mod. VITALS exploits machine learning and optimization algorithms for the validation of transport codes. Unlike SAT0, the SAT1 saturation rule contains a model to capture cross-scale turbulence coupling. Results show that SAT1 agrees better with experiments, further confirming that multi-scale effects are needed to model heat transport in C-Mod L-modes. VITALS will next be used to analyze past data from DIII-D: L-mode ``Shortfall'' plasma and ECH swing experiments. A second code suite, PRIMA, allows for integrated modeling of the plasma response to Laser Blow-Off cold pulses. Preliminary results show that SAT1 qualitatively reproduces the propagation of cold pulses after LBO injections and SAT0 does not, indicating that cross-scale coupling effects play a role in the plasma response. PRIMA will be used to ``predict-first'' cold pulse experiments using the new LBO system at DIII-D, and analyze existing ECH heat pulse data. Work supported by DE-FC02-99ER54512, DE-FC02-04ER54698.

  1. Bringing the mountain to Mohammed

    International Nuclear Information System (INIS)

    Hibbs, S.M.; Bell, H.H.; Bowman, F.M.; Hitchin, C.; Jackson, M.

    1987-01-01

    New free electron laser (FEL) technology at Lawrence Livermore National Laboratory (LLNL) promises electron-cyclotron plasma heating at power levels, cost efficiency, and tunable frequency range far beyond the capabilities of existing technology. LLNL has the high-current induction linear accelerators needed to drive such an FEL. Thus, the first stage of the Microwave Tokamak Experiment (MTX), designed to test this new technology, was to bring the Alcator-C tokamak across the United States from the Massachusetts Institute of Technology (MIT) to LLNL in California. The authors discuss why the tokamak was moved across the country and described the move

  2. An analysis of the binding of repressor protein ModE to modABCD (molybdate transport) operator/promoter DNA of Escherichia coli.

    Science.gov (United States)

    Grunden, A M; Self, W T; Villain, M; Blalock, J E; Shanmugam, K T

    1999-08-20

    Expression of the modABCD operon in Escherichia coli, which codes for a molybdate-specific transporter, is repressed by ModE in vivo in a molybdate-dependent fashion. In vitro DNase I-footprinting experiments identified three distinct regions of protection by ModE-molybdate on the modA operator/promoter DNA, GTTATATT (-15 to -8; region 1), GCCTACAT (-4 to +4; region 2), and GTTACAT (+8 to +14; region 3). Within the three regions of the protected DNA, a pentamer sequence, TAYAT (Y = C or T), can be identified. DNA-electrophoretic mobility experiments showed that the protected regions 1 and 2 are essential for binding of ModE-molybdate to DNA, whereas the protected region 3 increases the affinity of the DNA to the repressor. The stoichiometry of this interaction was found to be two ModE-molybdate per modA operator DNA. ModE-molybdate at 5 nM completely protected the modABCD operator/promoter DNA from DNase I-catalyzed hydrolysis, whereas ModE alone failed to protect the DNA even at 100 nM. The apparent K(d) for the interaction between the modA operator DNA and ModE-molybdate was 0.3 nM, and the K(d) increased to 8 nM in the absence of molybdate. Among the various oxyanions tested, only tungstate replaced molybdate in the repression of modA by ModE, but the affinity of ModE-tungstate for modABCD operator DNA was 6 times lower than with ModE-molybdate. A mutant ModE(T125I) protein, which repressed modA-lac even in the absence of molybdate, protected the same region of modA operator DNA in the absence of molybdate. The apparent K(d) for the interaction between modA operator DNA and ModE(T125I) was 3 nM in the presence of molybdate and 4 nM without molybdate. The binding of molybdate to ModE resulted in a decrease in fluorescence emission, indicating a conformational change of the protein upon molybdate binding. The fluorescence emission spectra of mutant ModE proteins, ModE(T125I) and ModE(Q216*), were unaffected by molybdate. The molybdate-independent mutant Mod

  3. Critical need for MFE: the Alcator DX advanced divertor test facility

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

  4. Vessel coolant mass depletion during a 5% SBLOCA in the Semiscale Mod-2C facility

    International Nuclear Information System (INIS)

    Shaw, R.A.; Loomis, G.G.

    1985-01-01

    Experimental results are presented from two 5% small-break loss-of-coolant accident (SBLOCA) simulations in the Semiscale Mod-2C facility. In performing the simulated 5% SBLOCAs, boundary conditions scaled from a pressurized water reactor (PWR) were used. The experiment was run with initial conditions typical of a PWR (15.6 MPa pressure and 35 K core differential temperature). The Mod-2C facility represents the state-of-the-art in small facilities scaled from PWRs. Phenomena which occurred during the transient included: primary fluid saturation (change from subcooled to saturated blowdown), break uncovery (a centerline break was simulated), condensation-induced liquid hold-up in the steam generator primary tubes, pump suction liquid seal formation and core level depression with resulting core rod temperature excursion, pump suction liquid seal clearance, loop fluid mass redistribution, and gradual core rewet. The influence of core bypass flow is also discussed. 11 refs., 13 figs

  5. Comparison and Evaluation of Annual NDVI Time Series in China Derived from the NOAA AVHRR LTDR and Terra MODIS MOD13C1 Products.

    Science.gov (United States)

    Guo, Xiaoyi; Zhang, Hongyan; Wu, Zhengfang; Zhao, Jianjun; Zhang, Zhengxiang

    2017-06-06

    Time series of Normalized Difference Vegetation Index (NDVI) derived from multiple satellite sensors are crucial data to study vegetation dynamics. The Land Long Term Data Record Version 4 (LTDR V4) NDVI dataset was recently released at a 0.05 × 0.05° spatial resolution and daily temporal resolution. In this study, annual NDVI time series that are composited by the LTDR V4 and Moderate Resolution Imaging Spectroradiometer (MODIS) NDVI datasets (MOD13C1) are compared and evaluated for the period from 2001 to 2014 in China. The spatial patterns of the NDVI generally match between the LTDR V4 and MOD13C1 datasets. The transitional zone between high and low NDVI values generally matches the boundary of semi-arid and sub-humid regions. A significant and high coefficient of determination is found between the two datasets according to a pixel-based correlation analysis. The spatially averaged NDVI of LTDR V4 is characterized by a much weaker positive regression slope relative to that of the spatially averaged NDVI of the MOD13C1 dataset because of changes in NOAA AVHRR sensors between 2005 and 2006. The measured NDVI values of LTDR V4 were always higher than that of MOD13C1 in western China due to the relatively lower atmospheric water vapor content in western China, and opposite observation appeared in eastern China. In total, 18.54% of the LTDR V4 NDVI pixels exhibit significant trends, whereas 35.79% of the MOD13C1 NDVI pixels show significant trends. Good agreement is observed between the significant trends of the two datasets in the Northeast Plain, Bohai Economic Rim, Loess Plateau, and Yangtze River Delta. By contrast, the datasets contrasted in northwestern desert regions and southern China. A trend analysis of the regression slope values according to the vegetation type shows good agreement between the LTDR V4 and MOD13C1 datasets. This study demonstrates the spatial and temporal consistencies and discrepancies between the AVHRR LTDR and MODIS MOD13C1 NDVI

  6. Pumped-limiter study for Alcator DCT

    International Nuclear Information System (INIS)

    Brooks, J.N.; Mattas, R.F.; Cha, Y.S.; Hassanein, A.M.; Majumdar, S.

    1983-06-01

    A study was performed for a pumped-limiter design for the proposed Alcator DCT device. The study focused on reactor-relevant issues. The main issues examined were configuration, surface erosion, thermal hydraulics, and the choice of structural and surface materials. A bottom, flat limiter, with a copper-alloy substrate, seems to be a reasonable design and should provide an opportunity to test high power and particle loadings. Carbon is recommended as a surface material if acceptable redeposition properties can be demonstrated

  7. Pumped-limiter study for Alcator DCT

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, J.N.; Mattas, R.F.; Cha, Y.S.; Hassanein, A.M.; Majumdar, S.

    1983-06-01

    A study was performed for a pumped-limiter design for the proposed Alcator DCT device. The study focused on reactor-relevant issues. The main issues examined were configuration, surface erosion, thermal hydraulics, and the choice of structural and surface materials. A bottom, flat limiter, with a copper-alloy substrate, seems to be a reasonable design and should provide an opportunity to test high power and particle loadings. Carbon is recommended as a surface material if acceptable redeposition properties can be demonstrated.

  8. Experimental operation of the KT-5C tokamak in USTC

    International Nuclear Information System (INIS)

    Wen Yizhi; Wan Shude; Zhai Kan; Liu Wandong

    1995-01-01

    Experimental operation of the KT-5C tokamak was started in early 1991. More than 3 x 10 4 shots of discharges have been performed so far. The authors deals with the major features of operation control of the KT-5C device. As the machine is usually controlled by an experimenter himself without duty operator, an audible indicator makes it much easier to control and operate. Accident and interference prevention systems prove to be reliable, and the operation system works successfully

  9. Simulations of the L-H transition on experimental advanced superconducting Tokamak

    International Nuclear Information System (INIS)

    Weiland, Jan

    2014-01-01

    We have simulated the L-H transition on the EAST tokamak [Baonian Wan, EAST and HT-7 Teams, and International Collaborators, “Recent experiments in the EAST and HT-7 superconducting tokamaks,” Nucl. Fusion 49, 104011 (2009)] using a predictive transport code where ion and electron temperatures, electron density, and poloidal and toroidal momenta are simulated self consistently. This is, as far as we know, the first theory based simulation of an L-H transition including the whole radius and not making any assumptions about where the barrier should be formed. Another remarkable feature is that we get H-mode gradients in agreement with the α – α d diagram of Rogers et al. [Phys. Rev. Lett. 81, 4396 (1998)]. Then, the feedback loop emerging from the simulations means that the L-H power threshold increases with the temperature at the separatrix. This is a main feature of the C-mod experiments [Hubbard et al., Phys. Plasmas 14, 056109 (2007)]. This is also why the power threshold depends on the direction of the grad B drift in the scrape off layer and also why the power threshold increases with the magnetic field. A further significant general H-mode feature is that the density is much flatter in H-mode than in L-mode

  10. Mod i ledelse

    DEFF Research Database (Denmark)

    Mellon, Karsten

    2016-01-01

    Mod i ledelse er en efterspurgt vare i offentligt regi som modsvar på stigende kompleksitet og pres. Men hvad er ’mod i ledelse’ – og er du selv en modig leder?......Mod i ledelse er en efterspurgt vare i offentligt regi som modsvar på stigende kompleksitet og pres. Men hvad er ’mod i ledelse’ – og er du selv en modig leder?...

  11. Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

    International Nuclear Information System (INIS)

    Nishio, S.; Ueda, S.; Kurihara, R.; Kuroda, T.; Miura, H.; Sako, K.; Takase, H.; Seki, Y.; Adachi, J.; Yamazaki, S.; Hashimoto, T.; Mori, S.; Shinya, K.; Murakami, Y.; Senda, I.; Okano, K.; Asaoka, Y.; Yoshida, T.

    2000-01-01

    If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features of the prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900 deg. C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal

  12. Algebraic Structures on MOD Planes

    OpenAIRE

    Kandasamy, Vasantha; Ilanthenral, K.; Smarandache, Florentin

    2015-01-01

    Study of MOD planes happens to a very recent one. In this book, systematically algebraic structures on MOD planes like, MOD semigroups, MOD groups and MOD rings of different types are defined and studied. Such study is innovative for a large four quadrant planes are made into a small MOD planes. Several distinct features enjoyed by these MOD planes are defined, developed and described.

  13. π Scope: python based scientific workbench with visualization tool for MDSplus data

    Science.gov (United States)

    Shiraiwa, S.

    2014-10-01

    π Scope is a python based scientific data analysis and visualization tool constructed on wxPython and Matplotlib. Although it is designed to be a generic tool, the primary motivation for developing the new software is 1) to provide an updated tool to browse MDSplus data, with functionalities beyond dwscope and jScope, and 2) to provide a universal foundation to construct interface tools to perform computer simulation and modeling for Alcator C-Mod. It provides many features to visualize MDSplus data during tokamak experiments including overplotting different signals and discharges, various plot types (line, contour, image, etc.), in-panel data analysis using python scripts, and publication quality graphics generation. Additionally, the logic to produce multi-panel plots is designed to be backward compatible with dwscope, enabling smooth migration for dwscope users. πScope uses multi-threading to reduce data transfer latency, and its object-oriented design makes it easy to modify and expand while the open source nature allows portability. A built-in tree data browser allows a user to approach the data structure both from a GUI and a script, enabling relatively complex data analysis workflow to be built quickly. As an example, an IDL-based interface to perform GENRAY/CQL3D simulations was ported on πScope, thus allowing LHCD simulation to be run between-shot using C-Mod experimental profiles. This workflow is being used to generate a large database to develop a LHCD actuator model for the plasma control system. Supported by USDoE Award DE-FC02-99ER54512.

  14. Transport in the tokamak plasma edge

    International Nuclear Information System (INIS)

    Vold, E.L.

    1989-01-01

    Experimental observations characterize the edge plasma or boundary layer in magnetically confined plasmas as a region of great complexity. Evidence suggests the edge physics plays a key role in plasma confinement although the mechanism remains unresolved. This study focuses on issues in two areas: observed poloidal asymmetries in the Scrape Off Layer (SOL) edge plasma and the physical nature of the plasma-neutral recycling. A computational model solves the coupled two dimensional partial differential equations governing the plasma fluid density, parallel and radial velocities, electron and ion temperatures and neutral density under assumptions of toroidal symmetry, ambipolarity, anomalous diffusive radial flux, and neutral-ion thermal equilibrium. Drift flow and plasma potential are calculated as dependent quantities. Computational results are compared to experimental data for the CCT and TEXTOR:ALT-II tokamak limiter cases. Comparisons show drift flux is a major component of the poloidal flow in the SOL along the tangency/separatrix. Plasma-neutral recycling is characterized in several tokamak divertors, including the C-MOD device using magnetic flux surface coordinates. Recycling is characterized by time constant, τ rc , on the order of tens of milliseconds. Heat flux transients from the core into the edge on shorter time scales significantly increase the plasma temperatures at the target and may increase sputtering. Recycling conditions in divertors vary considerably depending on recycled flux to the core. The high density, low temperature solution requires that the neutral mean free path be small compared to the divertor target to x-point distance. The simulations and analysis support H-mode confinement and transition models based on the recycling divertor solution bifurcation

  15. Distribution of the type III DNA methyltransferases modA, modB and modD among Neisseria meningitidis genotypes: implications for gene regulation and virulence.

    Science.gov (United States)

    Tan, Aimee; Hill, Dorothea M C; Harrison, Odile B; Srikhanta, Yogitha N; Jennings, Michael P; Maiden, Martin C J; Seib, Kate L

    2016-02-12

    Neisseria meningitidis is a human-specific bacterium that varies in invasive potential. All meningococci are carried in the nasopharynx, and most genotypes are very infrequently associated with invasive meningococcal disease; however, those belonging to the 'hyperinvasive lineages' are more frequently associated with sepsis or meningitis. Genome content is highly conserved between carriage and disease isolates, and differential gene expression has been proposed as a major determinant of the hyperinvasive phenotype. Three phase variable DNA methyltransferases (ModA, ModB and ModD), which mediate epigenetic regulation of distinct phase variable regulons (phasevarions), have been identified in N. meningitidis. Each mod gene has distinct alleles, defined by their Mod DNA recognition domain, and these target and methylate different DNA sequences, thereby regulating distinct gene sets. Here 211 meningococcal carriage and >1,400 disease isolates were surveyed for the distribution of meningococcal mod alleles. While modA11-12 and modB1-2 were found in most isolates, rarer alleles (e.g., modA15, modB4, modD1-6) were specific to particular genotypes as defined by clonal complex. This suggests that phase variable Mod proteins may be associated with distinct phenotypes and hence invasive potential of N. meningitidis strains.

  16. Fluctuation measurements at c/ωpe spatial scales in a tokamak

    International Nuclear Information System (INIS)

    Haines, E.J.; Tan, I.H.; Prager, S.C.

    1994-07-01

    Magnetic and electrostatic fluctuations have been measured at the short scale length of the collisionless skin depth (c/ω pe using small magnetic and electrostatic probes in the Tokapole II tokamak. For certain conditions and at high frequency (MHz range) the amplitude is observed to increase as wavelength is decreased toward the c/ω pe scale. Wavelength dependence is inferred from measurements with probes of varying sizes. The amplitude of the turbulence at the c/ω pe scale is smaller than the dominant low frequency turbulence, and is thus not relevant to transport in Tokapole II. Comparison with theoretical treatments of c/ω pe turbulence is discussed

  17. Fast IR diodes thermometer for tokamak

    International Nuclear Information System (INIS)

    Chen Xiangbo

    2001-01-01

    A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices

  18. Plasma confinement using biased electrode in the TCABR tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Machida, M.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.

    2005-01-01

    Experimental data obtained on the TCABR tokamak (R = 0.61 m, a = 0.18 m) with an electrically polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1) and hydrogen gas injection adjusted for keeping the electron density at 1.0 x 10 19 m -3 without bias. Time evolution and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the central line-averaged density, up to a maximum factor of 2.6, while H α hydrogen spectral line intensity decreases and the C III impurity stays on the same level. The analysis of temporal behaviour and radial profiles of plasma parameters indicates that the confined plasma enters the H-mode regime. The data analysis shows a maximum enhanced energy confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the Scrape-Off-Layer, using measured floating potentials and ion saturation currents, show a strong decrease in the power spectra and transport. Bifurcation was not observed and the decrease in the saturation current occurs in 50 μs

  19. Analyzing the Radiation Properties of High-Z Impurities in High-Temperature Plasmas

    International Nuclear Information System (INIS)

    Reinke, M. L.; Ince-Cushman, A.; Podpaly, Y.; Rice, J. E.; Bitter, M.; Hill, K. W.; Fournier, K. B.; Gu, M. F.

    2009-01-01

    Most tokamak-based reactor concepts require the use of noble gases to form either a radiative mantle or divertor to reduce conductive heat exhaust to tolerable levels for plasma facing components. Predicting the power loss necessary from impurity radiation is done using electron temperature-dependent 'cooling-curves' derived from ab initio atomic physics models. We present here a technique to verify such modeling using highly radiative, argon infused discharges on Alcator C-Mod. A novel x-ray crystal imaging spectrometer is used to measure spatially resolved profiles of line-emissivity, constraining impurity transport simulations. Experimental data from soft x-ray diodes, bare AXUV diodes and foil bolometers are used to determine the local emissivity in three overlapping spectral bands, which are quantitatively compared to models. Comparison of broadband measurements show agreement between experiment and modeling in the core, but not over the entire profile, with the differences likely due to errors in the assumed radial impurity transport outside of the core. Comparison of Ar 16+ x-ray line emission modeling to measurements suggests an additional problem with the collisional-radiative modeling of that charge state.

  20. Feedback-assisted extension of the tokamak operating space to low safety factor

    International Nuclear Information System (INIS)

    Hanson, J. M.; Bialek, J. M.; Navratil, G. A.; Olofsson, K. E. J.; Shiraki, D.; Turco, F.; Baruzzo, M.; Bolzonella, T.; Marrelli, L.; Martin, P.; Piovesan, P.; Piron, C.; Piron, L.; Terranova, D.; Zanca, P.; Hyatt, A. W.; Jackson, G. L.; La Haye, R. J.; Lanctot, M. J.; Strait, E. J.

    2014-01-01

    Recent DIII-D and RFX-mod experiments have demonstrated stable tokamak operation at very low values of the edge safety factor q(a) near and below 2. The onset of n = 1 resistive wall mode (RWM) kink instabilities leads to a disruptive stability limit, encountered at q(a) = 2 (limiter plasmas) and q 95  = 2 (divertor plasmas). However, passively stable operation can be attained for q(a) and q 95 values as low as 2.2. RWM damping in the q(a) = 2 regime was measured using active MHD spectroscopy. Although consistent with theoretical predictions, the amplitude of the damped response does not increase significantly as the q(a) = 2 limit is approached, in contrast with damping measurements made approaching the pressure-driven RWM limit. Applying proportional gain magnetic feedback control of the n = 1 modes has resulted in stabilized operation with q 95 values reaching as low as 1.9 in DIII-D and q(a) reaching 1.55 in RFX-mod. In addition to being consistent with the q(a) = 2 external kink mode stability limit, the unstable modes have growth rates on the order of the characteristic wall eddy-current decay timescale in both devices, and a dominant m = 2 poloidal structure that is consistent with ideal MHD predictions. The experiments contribute to validating MHD stability theory and demonstrate that a key tokamak stability limit can be overcome with feedback

  1. Magnetic fluctuations in the plasma of KT-5C tokamak

    International Nuclear Information System (INIS)

    Lu Ronghua; Pan Gesheng; Wang Zhijiang; Wen Yizhi; Yu Changxuan; Wan Shude; Liu Wandong; Wang Jun; Xu Min; Xiao Delong; Yu Yi

    2004-01-01

    A newly developed moveable magnetic probe array was installed on KT-5C tokamak. The profiles of radial and poloidal magnetic fluctuations of the plasma have been measured for (0.5r/a1.1). The experimental results indicate that there is a radial gradient which is greater than relative electrostatic fluctuations and the magnetic fluctuations contribute a little to losses. A strong coherence between fluctuations of 4 mm nearby two points suggests that the magnetic fluctuations have quite a long correlation length

  2. Modélisation et automatisation des procédés d’écriture et de production de supports de formation numérisés - Le modèle M.A.Ï.HEU.T.I.C. de la CCI de Paris

    Directory of Open Access Journals (Sweden)

    José Martin

    2005-01-01

    Full Text Available Cet article expose les conclusions d’un projet de recherche de la Direction de l’Enseignement de la Chambre de Commerce et d’Industrie de Paris (CCIP, mené de 2003 à 2005, dont l’aboutissement est un modèle didactique de production de contenus de cours numérisés, baptisé M.A.Ï.HEU.T.I.C.[[ M.A.Ï.HEU.T.I.C. : Modèle Appliqué d’Interprétation Heuristique des Technologies de l’Information et de la Communication.

  3. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-03-01

    This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

  4. TC-13 Mod 0 and Mod 2 Steam Catapult Test Site

    Data.gov (United States)

    Federal Laboratory Consortium — Located on 11,000 feet of test runway, the TC-13 Mod 0 and Mod 2 Steam Catapult Test Site has in-ground catapults identical to those aboard carriers. This test site...

  5. Three-dimensional equilibria and transport in RFX-mod: A description using stellarator tools

    International Nuclear Information System (INIS)

    Gobbin, M.; Bonfiglio, D.; Lorenzini, R.; Marrelli, L.; Martin, P.; Martines, E.; Momo, B.; Predebon, I.; Puiatti, M. E.; Spizzo, G.; Terranova, D.; Boozer, A. H.; Cooper, A. W.; Escande, D. F.; Hirshman, S. P.; Lore, J.; Sanchez, R.; Spong, D. A.; Pomphrey, N.

    2011-01-01

    RFX-mod self-organized single helical axis (SHAx) states provide a unique opportunity to advance 3D fusion physics and establish a common knowledge basis in a parameter region not covered by stellarators and tokamaks. The VMEC code has been adapted to the reversed-field pinch (RFP) to model SHAx equilibria in fixed boundary mode with experimental measurements as constraint. The averaged particle diffusivity over the helical volume, estimated with the Monte Carlo code ORBIT, has a neoclassical-like dependence on collisionality and does not show the 1/ν trend of un-optimized stellarators. In particular, the helical region boundary, corresponding to an electron transport barrier with zero magnetic shear and improved confinement, has been investigated using numerical codes common to the stellarator community. In fact, the DKES/PENTA codes have been applied to RFP for local neoclassical transport computations, including radial electric field, to estimate thermal diffusion coefficients in the barrier region for typical RFX-mod temperature and density profiles. A comparison with power balance estimates shows that residual chaos due to secondary tearing modes and small-scale turbulence still contribute to drive anomalous transport in the barrier region.

  6. Feedback-assisted extension of the tokamak operating space to low safety factor

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, J. M., E-mail: jmh2130@columbia.edu; Bialek, J. M.; Navratil, G. A.; Olofsson, K. E. J.; Shiraki, D.; Turco, F. [Department of Applied Mathematics and Applied Physics, Columbia University, New York, New York 10027-6900 (United States); Baruzzo, M.; Bolzonella, T.; Marrelli, L.; Martin, P.; Piovesan, P.; Piron, C.; Piron, L.; Terranova, D.; Zanca, P. [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Hyatt, A. W.; Jackson, G. L.; La Haye, R. J.; Lanctot, M. J.; Strait, E. J. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); and others

    2014-07-15

    Recent DIII-D and RFX-mod experiments have demonstrated stable tokamak operation at very low values of the edge safety factor q(a) near and below 2. The onset of n = 1 resistive wall mode (RWM) kink instabilities leads to a disruptive stability limit, encountered at q(a) = 2 (limiter plasmas) and q{sub 95} = 2 (divertor plasmas). However, passively stable operation can be attained for q(a) and q{sub 95} values as low as 2.2. RWM damping in the q(a) = 2 regime was measured using active MHD spectroscopy. Although consistent with theoretical predictions, the amplitude of the damped response does not increase significantly as the q(a) = 2 limit is approached, in contrast with damping measurements made approaching the pressure-driven RWM limit. Applying proportional gain magnetic feedback control of the n = 1 modes has resulted in stabilized operation with q{sub 95} values reaching as low as 1.9 in DIII-D and q(a) reaching 1.55 in RFX-mod. In addition to being consistent with the q(a) = 2 external kink mode stability limit, the unstable modes have growth rates on the order of the characteristic wall eddy-current decay timescale in both devices, and a dominant m = 2 poloidal structure that is consistent with ideal MHD predictions. The experiments contribute to validating MHD stability theory and demonstrate that a key tokamak stability limit can be overcome with feedback.

  7. Mercier criterion for high-β tokamaks

    International Nuclear Information System (INIS)

    Galvao, R.M.O.

    1984-01-01

    An expression, for the application of the Mercier criterion to numerical studies of diffuse high-β tokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the high-β tokamak approximation is derived. (L.C.) [pt

  8. Apo and ligand-bound structures of ModA from the archaeon Methanosarcina acetivorans.

    Science.gov (United States)

    Chan, Sum; Giuroiu, Iulia; Chernishof, Irina; Sawaya, Michael R; Chiang, Janet; Gunsalus, Robert P; Arbing, Mark A; Perry, L Jeanne

    2010-03-01

    The trace-element oxyanion molybdate, which is required for the growth of many bacterial and archaeal species, is transported into the cell by an ATP-binding cassette (ABC) transporter superfamily uptake system called ModABC. ModABC consists of the ModA periplasmic solute-binding protein, the integral membrane-transport protein ModB and the ATP-binding and hydrolysis cassette protein ModC. In this study, X-ray crystal structures of ModA from the archaeon Methanosarcina acetivorans (MaModA) have been determined in the apoprotein conformation at 1.95 and 1.69 A resolution and in the molybdate-bound conformation at 2.25 and 2.45 A resolution. The overall domain structure of MaModA is similar to other ModA proteins in that it has a bilobal structure in which two mixed alpha/beta domains are linked by a hinge region. The apo MaModA is the first unliganded archaeal ModA structure to be determined: it exhibits a deep cleft between the two domains and confirms that upon binding ligand one domain is rotated towards the other by a hinge-bending motion, which is consistent with the 'Venus flytrap' model seen for bacterial-type periplasmic binding proteins. In contrast to the bacterial ModA structures, which have tetrahedral coordination of their metal substrates, molybdate-bound MaModA employs octahedral coordination of its substrate like other archaeal ModA proteins.

  9. Apo and ligand-bound structures of ModA from the archaeon Methanosarcina acetivorans

    International Nuclear Information System (INIS)

    Chan, Sum; Giuroiu, Iulia; Chernishof, Irina; Sawaya, Michael R.; Chiang, Janet; Gunsalus, Robert P.; Arbing, Mark A.; Perry, L. Jeanne

    2010-01-01

    Crystal structures of ModA from M. acetivorans in the apo and ligand-bound conformations confirm domain rotation upon ligand binding. The trace-element oxyanion molybdate, which is required for the growth of many bacterial and archaeal species, is transported into the cell by an ATP-binding cassette (ABC) transporter superfamily uptake system called ModABC. ModABC consists of the ModA periplasmic solute-binding protein, the integral membrane-transport protein ModB and the ATP-binding and hydrolysis cassette protein ModC. In this study, X-ray crystal structures of ModA from the archaeon Methanosarcina acetivorans (MaModA) have been determined in the apoprotein conformation at 1.95 and 1.69 Å resolution and in the molybdate-bound conformation at 2.25 and 2.45 Å resolution. The overall domain structure of MaModA is similar to other ModA proteins in that it has a bilobal structure in which two mixed α/β domains are linked by a hinge region. The apo MaModA is the first unliganded archaeal ModA structure to be determined: it exhibits a deep cleft between the two domains and confirms that upon binding ligand one domain is rotated towards the other by a hinge-bending motion, which is consistent with the ‘Venus flytrap’ model seen for bacterial-type periplasmic binding proteins. In contrast to the bacterial ModA structures, which have tetrahedral coordination of their metal substrates, molybdate-bound MaModA employs octahedral coordination of its substrate like other archaeal ModA proteins

  10. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.; Saramad, S.

    2012-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

  11. Functional characterization of the Bradyrhizobium japonicum modA and modB genes involved in molybdenum transport.

    Science.gov (United States)

    Delgado, María J; Tresierra-Ayala, Alvaro; Talbi, Chouhra; Bedmar, Eulogio J

    2006-01-01

    A modABC gene cluster that encodes an ABC-type, high-affinity molybdate transporter from Bradyrhizobium japonicum has been isolated and characterized. B. japonicum modA and modB mutant strains were unable to grow aerobically or anaerobically with nitrate as nitrogen source or as respiratory substrate, respectively, and lacked nitrate reductase activity. The nitrogen-fixing ability of the mod mutants in symbiotic association with soybean plants grown in a Mo-deficient mineral solution was severely impaired. Addition of molybdate to the bacterial growth medium or to the plant mineral solution fully restored the wild-type phenotype. Because the amount of molybdate required for suppression of the mutant phenotype either under free-living or under symbiotic conditions was dependent on sulphate concentration, it is likely that a sulphate transporter is also involved in Mo uptake in B. japonicum. The promoter region of the modABC genes has been characterized by primer extension. Reverse transcription and expression of a transcriptional fusion, P(modA)-lacZ, was detected only in a B. japonicum modA mutant grown in a medium without molybdate supplementation. These findings indicate that transcription of the B. japonicum modABC genes is repressed by molybdate.

  12. Who owns the mods?

    OpenAIRE

    Kow, Yong Ming; Nardi, Bonnie

    2010-01-01

    Modding, the development of end user software extensions to commercial products, is popular among video gamers. Modders form communities to help each other. Mods can shape software products by weaving in contributions from users themselves based on their own experience of a product. The purpose of this paper is to investigate a conflict between a modding community and a gaming company which reveals contested issues of ownership and governance. We studied an online game, World of Warcraft, a l...

  13. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

  14. A distributed high speed data acquisition system for KT5C tokamak

    International Nuclear Information System (INIS)

    Sun Xiang; Wang Zhijiang; Lu Ronghua; Wang Jun; Yu Yi; Zhu Zhenghua; Wen Yizhi; Wan Shude; Liu Wandong; Yu Changxuan

    2005-01-01

    The development of a distributed data acquisition system with low cost to implement high speed data collection through the campus networks for a small tokamak, KT5C, is presented. Data of 512 k bytes at 5 MHz from 5 channels for each can be collected during about 10s after three researchers at different positions demand this system for acquisitions. This system realizes long distance multiuser operations; virtually efficiency of the data acquisition is enhanced. (authors)

  15. Neutral beam energy and power requirements for expanding radius and full bore startup of tokamak reactors

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Mense, A.T.; Attenberger, S.E.

    1979-09-01

    Natural beam power and energy requirements are compared for full density full bore and expanding radius startup scenarios in an elongated plasma, The Next Step (TNS), as a function of beam pulse time and plasma density. Because of the similarity of parameters, the results should also be applicable to Engineering Test Facility (ETF) and International Tokamak Reactor (INTOR) studies. A transport model consisting of neoclassical ion conduction and anomalous electron conduction and diffusion based on ALCATOR scaling leads to average densities in the range approx. 0.8 to 1.2 x 10 14 cm -3 being sufficient for ignition. Neutral deuterium beam energies in the range 120 to 180 keV are adequate for penetration, with the required power injected into the plasma decreasing with increasing beam energy. The neutral beam power decreases strongly with increasing beam pulse length b/sub b/ until t/sub b/ exceeds a few total energy confinement times, yielding b/sub b/ approx. = 4 to 6 s for the TNS plasma

  16. Performance of a compact four-strap fast wave antenna

    International Nuclear Information System (INIS)

    Wukitch, S.J.

    2002-01-01

    Ion cyclotron range of frequency (ICRF) is expected to be a primary auxiliary heating source in future experiments and fusion reactors. Compact antennas with high power density able to withstand large disruption forces present significant challenges to ICRF antenna design. A compact four-strap antenna has been installed in Alcator C-Mod and its performance has been compared with a pair of two-strap antennas. The key design features are the long vacuum strip line feeds, folded current strap configuration, use of ceramic insulators in the Faraday screen, and open Faraday screen. The heating efficiency and impurity generation are nearly identical to the other antennas while the loading is ∼2.5 higher. The power handling of the antenna was limited by arcing at relatively low maximum voltage. The strip line and antenna strap had arc damage localized to regions where the RF E-field was parallel to the tokamak B-field. For E parallel B, the breakdown voltage was determined to be ∼15 kV/cm. Redesign of the strip line has resulted in an increase in the maximum voltage from 17 kV to 25 kV. Finally, the current strap is being modified to increase the maximum voltage to ∼35 kV. (author)

  17. Full wave simulations of lower hybrid wave propagation in tokamaks

    International Nuclear Information System (INIS)

    Wright, J. C.; Bonoli, P. T.; Phillips, C. K.; Valeo, E.; Harvey, R. W.

    2009-01-01

    Lower hybrid (LH) waves have the attractive property of damping strongly via electron Landau resonance on relatively fast tail electrons at (2.5-3)xv te , where v te ≡ (2T e /m e ) 1/2 is the electron thermal speed. Consequently these waves are well-suited to driving current in the plasma periphery where the electron temperature is lower, making LH current drive (LHCD) a promising technique for off-axis (r/a≥0.60) current profile control in reactor grade plasmas. Established techniques for computing wave propagation and absorption use WKB expansions with non-Maxwellian self-consistent distributions.In typical plasma conditions with electron densities of several 10 19 m -3 and toroidal magnetic fields strengths of 4 Telsa, the perpendicular wavelength is of the order of 1 mm and the parallel wavelength is of the order of 1 cm. Even in a relatively small device such as Alcator C-Mod with a minor radius of 22 cm, the number of wavelengths that must be resolved requires large amounts of computational resources for the full wave treatment. These requirements are met with a massively parallel version of the TORIC full wave code that has been adapted specifically for the simulation of LH waves [J. C. Wright, et al., Commun. Comput. Phys., 4, 545 (2008), J. C. Wright, et al., Phys. Plasmas 16 July (2009)]. This model accurately represents the effects of focusing and diffraction that occur in LH propagation. It is also coupled with a Fokker-Planck solver, CQL3D, to provide self-consistent distribution functions for the plasma dielectric as well as a synthetic hard X-ray (HXR) diagnostic for direct comparisons with experimental measurements of LH waves.The wave solutions from the TORIC-LH zero FLR model will be compared to the results from ray tracing from the GENRAY/CQL3D code via the synthetic HXR diagnostic and power deposition.

  18. Massachusetts Institute of Technology, Plasma Fusion Center, technical research programs

    International Nuclear Information System (INIS)

    1982-02-01

    Research programs have produced significant results on four fronts: (1) the basic physics of high-temperature fusion plasmas (plasma theory, RF heating, development of advanced diagnostics and small-scale experiments on the Versator tokamak and Constance mirror devices); (2) major confinement results on the Alcator A and C tokamaks, including pioneering investigations of the equilibrium, stability, transport and radiation properties of fusion plasmas at high densities, temperatures and magnetic fields; (3) development of a new and innovative design for axisymmetric tandem mirrors with inboard thermal barriers, with initial operation of the TARA tandem mirror experimental facility scheduled for 1983; and (4) a broadly based program of fusion technology and engineering development that addresses problems in several critical subsystem areas

  19. Évolution des contraintes résiduelles dans la couche de diffusion d’un acier modèle Fe-Cr-C nitruré

    DEFF Research Database (Denmark)

    Jegou, Sébastien; Barrallier, Laurent; Somers, Marcel A. J.

    2011-01-01

    Limiter la fatigue et la corrosion des pièces est possible grâce à une nitruration. Des contraintes résiduelles en découlent. Le rôle de la diffusion du carbone sur le développement de ces contraintes a été étudié sur un acier modèle Fe-3%m.Cr-0.35%m.C.......Limiter la fatigue et la corrosion des pièces est possible grâce à une nitruration. Des contraintes résiduelles en découlent. Le rôle de la diffusion du carbone sur le développement de ces contraintes a été étudié sur un acier modèle Fe-3%m.Cr-0.35%m.C....

  20. Molybdate binding by ModA, the periplasmic component of the Escherichia coli mod molybdate transport system.

    Science.gov (United States)

    Imperial, J; Hadi, M; Amy, N K

    1998-03-13

    ModA, the periplasmic-binding protein of the Escherichia coli mod transport system was overexpressed and purified. Binding of molybdate and tungstate to ModA was found to modify the UV absorption and fluorescence emission spectra of the protein. Titration of these changes showed that ModA binds molybdate and tungstate in a 1:1 molar ratio. ModA showed an intrinsic fluorescence emission spectrum attributable to its three tryptophanyl residues. Molybdate binding caused a conformational change in the protein characterized by: (i) a shift of tryptophanyl groups to a more hydrophobic environment; (ii) a quenching (at pH 5.0) or enhancement (at pH 7.8) of fluorescence; and (iii) a higher availability of tryptophanyl groups to the polar quencher acrylamide. The tight binding of molybdate did not allow an accurate estimation of the binding constants by these indirect methods. An isotopic binding method with 99MoO42- was used for accurate determination of KD (20 nM) and stoichiometry (1:1 molar ratio). ModA bound tungstate with approximately the same affinity, but did not bind sulfate or phosphate. These KDs are 150- to 250-fold lower than those previously reported, and compatible with the high molybdate transport affinity of the mod system. The affinity of ModA for molybdate was also determined in vivo and found to be similar to that determined in vitro. Copyright 1998 Elsevier Science B.V.

  1. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    McWilliams, R.

    1988-01-01

    The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  2. SCDAP/RELAP5/MOD3 code development

    International Nuclear Information System (INIS)

    Allison, C.M.; Siefken, J.L.; Coryell, E.W.

    1992-01-01

    The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding

  3. Tokamak physics studies using x-ray diagnostic methods

    International Nuclear Information System (INIS)

    Hill, K.W.; Bitter, M.; von Goeler, S.

    1987-03-01

    X-ray diagnostic measurements have been used in a number of experiments to improve our understanding of important tokamak physics issues. The impurity content in TFTR plasmas, its sources and control have been clarified through soft x-ray pulse-height analysis (PHA) measurements. The dependence of intrinsic impurity concentrations and Z/sub eff/ on electron density, plasma current, limiter material and conditioning, and neutral-beam power have shown that the limiter is an important source of metal impurities. Neoclassical-like impurity peaking following hydrogen pellet injection into Alcator C and a strong effect of impurities on sawtooth behavior were demonstrated by x-ray imaging (XIS) measurements. Rapid inward motion of impurities and continuation of m = 1 activity following an internal disruption were demonstrated with XIS measurements on PLT using injected aluminum to enhance the signals. Ion temperatures up to 12 keV and a toroidal plasma rotation velocity up to 6 x 10 5 m/s have been measured by an x-ray crystal spectrometer (XCS) with up to 13 MW of 85-keV neutral-beam injection in TFTR. Precise wavelengths and relative intensities of x-ray lines in several helium-like ions and neon-like ions of silver have been measured in TFTR and PLT by the XCS. The data help to identify the important excitation processes predicted in atomic physics. Wavelengths of n = 3 to 2 silver lines of interest for x-ray lasers were measured, and precise instrument calibration techniques were developed. Electron thermal conductivity and sawtooth dynamics have been studied through XIS measurements on TFTR of heat-pulse propagation and compound sawteeth. A non-Maxwellian electron distribution function has been measured, and evidence of the Parail-Pogutse instability identified by hard x-ray PHA measurements on PLT during lower-hybrid current-drive experiments

  4. Overexpression, purification, and partial characterization of ADP-ribosyltransferases modA and modB of bacteriophage T4.

    Science.gov (United States)

    Tiemann, B; Depping, R; Rüger, W

    1999-01-01

    There is increasing experimental evidence that ADP-ribosylation of host proteins is an important means to regulate gene expression of bacteriophage T4. Surprisingly, this phage codes for three different ADP-ribosyltransferases, gene products Alt, ModA, and ModB, modifying partially overlapping sets of host proteins. While gene product Alt already has been isolated as a recombinant protein and its action on host RNA polymerases and transcription regulation have been studied, the nucleotide sequences of the two mod genes was published only recently. Their mode of action in the course of the infection cycle and the consequences of the ADP-ribosylations catalyzed by these enzymes remain to be investigated. Here we describe the cloning of the genes, the overexpression, purification, and partial characterization of ADP-ribosyltransferases ModA and ModB. Both proteins seem to act independently, and the ADP-ribosyl moieties are transferred to different sets of host proteins. While gene product ModA, similarly to the Alt protein, acts also on the alpha-subunit of host RNA polymerase, the ModB activity serves another set of proteins, one of which was identified as the S1 protein associated with the 30S subunit of the E. coli ribosomes.

  5. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  6. Quick look report for semiscale MOD-2C Test S-FS-11

    International Nuclear Information System (INIS)

    Plessinger, M.P.

    1985-11-01

    Results of a preliminary analysis of the fifth test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-11 simulated a pressurized water reactor transient initiated by a 50% break in a steam generator bottom feedwater line downstream of the check valve. With the exception of primary pressure, the initial conditions represented the initial conditions used for the C-E System 80 Final Safety Analysis Report (FSAR) Appendix 15B calculations. The transient included an initial 600 s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant followed by break isolation and affected loop steam generator refill with auxiliary feedwater. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overpressurization and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overpressurization and primary-to-secondary heat transfer. 64 figs

  7. SCDAP/RELAP5/MOD3 code development and assessment

    International Nuclear Information System (INIS)

    Allison, C.M.; Heath, C.H.; Siefken, L.J.; Hohorst, J.K.

    1991-01-01

    The SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the Nuclear Regulatory Commission (NRC). SCDAP/RELAP5/MOD3, created in January, 1991, is the result of merging RELAP5/MOD3 with SCDAP and TRAP-MELT models from SCDAP/RELAP5/MOD2.5. The RELAP5 models calculate the overall RCS thermal-hydraulics, control system interactions, reactor kinetics, and the transport of noncondensible gases, fission products, and aerosols. The SCDAP models calculate the damage progression in the core structures, the formation, heatup, and melting of debris, and the creep rupture failure of the lower head and other RCS structures. The TRAP-MELT models calculate the deposition of fission products upon aerosols or structural surfaces; the formation, growth, or deposition of aerosols; and the evaporation of species from surfaces. The systematic assessment of modeling uncertainties in SCDAP/RELAP5 code is currently underway. This assessment includes (a) the evaluation of code-to-data comparisons using stand-alone SCDAP and SCDAP/RELAP5/MOD3, (b) the estimation of modeling and experimental uncertainties, and (c) the determination of the influence of those uncertainties on predicted severe accident behavior

  8. Assessment of RELAP5/MOD2 and RELAP5/MOD1-EUR codes on the basis of LOBI-MOD2 test results

    International Nuclear Information System (INIS)

    D'Auria, F.; Mazzini, M.; Oriolo, F.; Galassi, G.M.

    1989-10-01

    The present report deals with an overview of the application of RELAP5/MOD2 and RELAP5/MOD1-EUR codes to tests performed in the LOBI/MOD2 facility. The work has been carried out in the frame of a contract between Dipartimento di Costruzioni Meccaniche e Nucleari (DCMN) of Pisa University and CEC. The Universities of Roma, Pisa, Bologna and Palermo and the Polytechnic of Torino performed the post-test analysis of the LOBI experiment under the supervision of DCMN. In the report the main outcomes from the analysis of the LOBI experiments are given with the attempt to identify deficiencies in the modelling capabilities of the used codes

  9. Cloud-based uniform ChIP-Seq processing tools for modENCODE and ENCODE.

    Science.gov (United States)

    Trinh, Quang M; Jen, Fei-Yang Arthur; Zhou, Ziru; Chu, Kar Ming; Perry, Marc D; Kephart, Ellen T; Contrino, Sergio; Ruzanov, Peter; Stein, Lincoln D

    2013-07-22

    Funded by the National Institutes of Health (NIH), the aim of the Model Organism ENCyclopedia of DNA Elements (modENCODE) project is to provide the biological research community with a comprehensive encyclopedia of functional genomic elements for both model organisms C. elegans (worm) and D. melanogaster (fly). With a total size of just under 10 terabytes of data collected and released to the public, one of the challenges faced by researchers is to extract biologically meaningful knowledge from this large data set. While the basic quality control, pre-processing, and analysis of the data has already been performed by members of the modENCODE consortium, many researchers will wish to reinterpret the data set using modifications and enhancements of the original protocols, or combine modENCODE data with other data sets. Unfortunately this can be a time consuming and logistically challenging proposition. In recognition of this challenge, the modENCODE DCC has released uniform computing resources for analyzing modENCODE data on Galaxy (https://github.com/modENCODE-DCC/Galaxy), on the public Amazon Cloud (http://aws.amazon.com), and on the private Bionimbus Cloud for genomic research (http://www.bionimbus.org). In particular, we have released Galaxy workflows for interpreting ChIP-seq data which use the same quality control (QC) and peak calling standards adopted by the modENCODE and ENCODE communities. For convenience of use, we have created Amazon and Bionimbus Cloud machine images containing Galaxy along with all the modENCODE data, software and other dependencies. Using these resources provides a framework for running consistent and reproducible analyses on modENCODE data, ultimately allowing researchers to use more of their time using modENCODE data, and less time moving it around.

  10. Integrated uncertainty analysis using RELAP/SCDAPSIM/MOD4.0

    International Nuclear Information System (INIS)

    Perez, M.; Reventos, F.; Wagner, R.; Allison, C.

    2009-01-01

    The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis package being developed jointly by the Technical University of Catalunya (UPC) and Innovative Systems Software (ISS). The integrated uncertainty analysis approach used in the package uses the following steps: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. The first four steps are performed by the user prior to the RELAP/SCDAPSIM/MOD4.0 analysis. The remaining steps are included with the MOD4.0 integrated uncertainty analysis (IUA) package. This paper briefly describes the integrated uncertainty analysis package including (a) the features of the package, (b) the implementation of the package into RELAP/SCDAPSIM/MOD4.0, and

  11. Quick Look Report for Semiscale MOD-2C Test S-FS-2

    International Nuclear Information System (INIS)

    Boucher, T.J.; Chen, T.H.

    1985-01-01

    Results of a preliminary analysis of the first test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-2 simulated a pressurized water reactor transient initiated by a double-ended offset shear of a steam generator main steam line upstream of the flow restrictor. Initial conditions represented normal ''hot-standby'' operation. The transient included an initial 600-s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant at conditions required to allow a natural circulation cooldown. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overcooling and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overcooling and primary-to-secondary heat transfer. 57 figs., 3 tabs

  12. Behavior of oxygen impurities in tokamak. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    El-Sharif, R N; Beket, A H [Plasma and Nuclear Fusion Department, Nuclear Research Center, Atomic Energy Aurhority, Cairo (Egypt)

    1996-03-01

    Impurity transport in tokamak plasma is a subject of great importance in present day tokamak experiments. The transport of oxygen as an impurity element in small tokamak was studied theoretically. The viscosity coefficient of oxygen has been calculated in different approximation 13 and 21 moment approximation, taking into consideration {chi}>>1,{chi}{omega}{sub c} {tau}. It was found that in 21 moment approximation additional terms added to the perturbation from equilibrium leads to increase in viscosity coefficients than in 13 moments approximation. 9 figs.

  13. Transient electromagnetic analysis in tokamaks using TYPHOON code

    International Nuclear Information System (INIS)

    Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

    1996-01-01

    The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

  14. Plasma confinement using biased electrode in the TCABR tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.; Machida, M.

    2005-01-01

    Experimental data obtained on the TCABR tokamak (R = 0.61 m, r = 0.18 m) with an electrally polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1), and hydrogen gas injection adjusted for keeping the electron density at 1.0x10(19) m(-3) without bias. Temporal and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the density, up to a maximum factor of 2.6, while H-alpha hydrogen spectral line intensity decreases, and the CIII impurity stays on the same level. The analysis of temporal and radial profiles of plasma parameters indicates that the confined plasma entered in the H-mode regime. The data analysis shows a maximum enhanced confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the plasma edge, using measured floating potentials and ion saturation currents, show strong decrease in the power spectra and transport. Bifurcation was not observed, and the decrease in the saturation current occurs in 50 microseconds. (author)

  15. Expression, purification and DNA-binding activities of two putative ModE proteins of Herbaspirillum seropedicae (Burkholderiales, Oxalobacteraceae

    Directory of Open Access Journals (Sweden)

    André L.F. Souza

    2008-01-01

    Full Text Available In prokaryotes molybdenum is taken up by a high-affinity ABC-type transporter system encoded by the modABC genes. The endophyte β-Proteobacterium Herbaspirillum seropedicae has two modABC gene clusters and two genes encoding putative Mo-dependent regulator proteins (ModE1 and ModE2. Analysis of the amino acid sequence of the ModE1 protein of H. seropedicae revealed the presence of an N-terminal domain containing a DNA-binding helix-turn-helix motif (HTH and a C-terminal domain with a molybdate-binding motif. The second putative regulator protein, ModE2, contains only the helix-turn-helix motif, similar to that observed in some sequenced genomes. We cloned the modE1 (810 bp and modE2 (372 bp genes and expressed them in Escherichia coli as His-tagged fusion proteins, which we subsequently purified. The over-expressed recombinant His-ModE1 was insoluble and was purified after solubilization with urea and then on-column refolded during affinity chromatography. The His-ModE2 was expressed as a soluble protein and purified by affinity chromatography. These purified proteins were analyzed by DNA band-shift assays using the modA2 promoter region as probe. Our results indicate that His-ModE1 and His-ModE2 are able to bind to the modA2 promoter region, suggesting that both proteins may play a role in the regulation of molybdenum uptake and metabolism in H. seropedicae.

  16. The Tokamak IST-TOK

    International Nuclear Information System (INIS)

    Varandas, C.A.F.; Cabral, J.A.C.; Manso, M.E.

    1991-01-01

    A small tokamak is under construction at the Portuguese Technical Superior Institute. The main objective is to create a home based laboratory in which an independent scientific program might be developed. (L.C.J.A.). 14 refs, 6 figs

  17. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  18. R-HyMOD: an R-package for the hydrological model HyMOD

    Science.gov (United States)

    Baratti, Emanuele; Montanari, Alberto

    2015-04-01

    A software code for the implementation of the HyMOD hydrological model [1] is presented. HyMOD is a conceptual lumped rainfall-runoff model that is based on the probability-distributed soil storage capacity principle introduced by R. J. Moore 1985 [2]. The general idea behind this model is to describe the spatial variability of some process parameters as, for instance, the soil structure or the water storage capacities, through probability distribution functions. In HyMOD, the rainfall-runoff process is represented through a nonlinear tank connected with three identical linear tanks in parallel representing the surface flow and a slow-flow tank representing groundwater flow. The model requires the optimization of five parameters: Cmax (the maximum storage capacity within the watershed), β (the degree of spatial variability of the soil moisture capacity within the watershed), α (a factor for partitioning the flow between two series of tanks) and the two residence time parameters of quick-flow and slow-flow tanks, kquick and kslow respectively. Given its relatively simplicity but robustness, the model is widely used in the literature. The input data consist of precipitation and potential evapotranspiration at the given time scale. The R-HyMOD package is composed by a 'canonical' R-function of HyMOD and a fast FORTRAN implementation. The first one can be easily modified and can be used, for instance, for educational purposes; the second part combines the R user friendly interface with a fast processing unit. [1] Boyle D.P. (2000), Multicriteria calibration of hydrological models, Ph.D. dissertation, Dep. of Hydrol. and Water Resour., Univ of Arizona, Tucson. [2] Moore, R.J., (1985), The probability-distributed principle and runoff production at point and basin scale, Hydrol. Sci. J., 30(2), 273-297.

  19. Perspectives d’avenir du modèle autrichien

    OpenAIRE

    Neisser, Heinrich

    2018-01-01

    Plusieurs de ceux qui m’ont précédé ont déjà décrit des éléments de ce modèle autrichien qui suscite beaucoup d’intérêt à l’étranger, mais qui – et cela aussi a été signalé un certain nombre de fois – ne peut pas être transposé dans sa totalité à d’autres pays. De tels modèles concrets pour la solution de conflits sociaux naissent à partir d’un certain contexte historique. Ils se développent de manière pragmatique, c’est-à-dire pour résoudre, le mieux possible, des problèmes qui se posent à u...

  20. A study of quasi-mode parametric excitations in lower-hybrid heating of tokamak plasmas

    International Nuclear Information System (INIS)

    Villalon, E.; Bers, A.

    1980-01-01

    A detailed linear and non-linear analysis of quasi-mode parametric excitations relevant to experiments in supplementary heating of tokamak plasmas is presented. The linear analysis includes the full ion-cyclotron harmonic quasi-mode spectrum. The non-linear analysis, considering depletion of the pump electric field, is applied to the recent Alcator A heating experiment. Because of the very different characteristics of a tokamak plasma near the wall (in the shadow of the limiter) and inside, the quasi-mode excitations are studied independently for the plasma edge and the main bulk of the plasma, and for two typical regimes in overall density, the low (peak in density, n 0 =1.5x10 14 cm -3 ) and high (n 0 =5x10 14 cm -3 ) density regimes. At the edge of the plasma and for the low-density regime, it is found that higher nsub(z)(nsub(z)=cksub(z)/ω) than those predicted by the linear theory are strongly excited. Inside the plasma, the excitation of higher wave numbers is also significant. These results indicate that a large amount of the RF-power may not penetrate to the plasma centre, but will rather be either Landau-damped on the electrons or mode-converted into thermal modes, close to the plasma edge. Moreover, for sufficiently high peaks in density, it is found that all the RF-power is mode-converted before reaching the plasma centre. Inside the plasma, the power density of the excited sideband fields is shown to be always very small in comparison with their excitation at the plasma edge. (author)

  1. ICRF Mode Conversion Studies with Phase Contrast Imaging and Comparisons with Full-Wave Simulations

    International Nuclear Information System (INIS)

    Tsujii, N.; Bonoli, P. T.; Lin, Y.; Wright, J. C.; Wukitch, S. J.; Porkolab, M.; Jaeger, E. F.; Harvey, R. W.

    2011-01-01

    Waves in the ion cyclotron range of frequencies (ICRF) are widely used to heat toka-mak plasmas. In a multi-ion-species plasma, the FW converts to ion cyclotron waves (ICW) and ion Bernstein waves (IBW) around the ion-ion hybrid resonance (mode conversion). The mode converted wave is of interest as an actuator to optimise plasma performance through flow drive and current drive. Numerical simulations are essential to describe these processes accurately, and it is important that these simulation codes be validated. On Alcator C-Mod, direct measurements of the mode converted waves have been performed using Phase Contrast Imaging (PCI), which measures the line-integrated electron density fluctuations. The results were compared to full-wave simulations AORSA and TORIC. AORSA is coupled to a Fokker-Planck code CQL3D for self-consistent simulation of the wave electric field and the minority distribution function. The simulation results are compared to PCI measurements using synthetic diagnostic. The experiments were performed in D-H and D- 3 He plasmas over a wide range of ion species concentrations. The simulations agreed well with the measurements in the strong absorption regime. However, the measured fluctuation intensity was smaller by 1-2 orders of magnitudes in the weakly abosorbing regime, and a realistic description of the plasma edge including dissipation and antenna geometry may be required in these cases.

  2. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

    With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

  3. Plasma density measurements on COMPASS-C tokamak from electron cyclotron emission cutoffs

    International Nuclear Information System (INIS)

    Chenna Reddy, D.; Edlington, T.

    1996-01-01

    Electron cyclotron emission (ECE) is a standard diagnostic in present day tokamak devices for temperature measurement. When the plasma density is high enough the emission at some frequencies is cut off. Of these cutoff frequencies, the first frequency to cut off depends on the shape of the density profile. If the density profile can be described by a few parameters, in some circumstances, this first cutoff frequency can be used to obtain two of these parameters. If more than two parameters are needed to describe the density profile, then additional independent measurements are required to find all the parameters. We describe a technique by which it is possible to obtain an analytical relation between the radius at which the first cutoff occurs and the profile parameters. Assuming that the shape of the profile does not change as the average density rises after the first cutoff, one can use the cutoffs at other frequencies to obtain the average density at the time of these cutoffs. The plasma densities obtained with this technique using the data from a 14 channel ECE diagnostic on COMPASS-C tokamak are in good agreement with those measured by a standard 2 mm interferometer. The density measurement using the ECE cutoffs is an independent measurement and requires only a frequency calibration of the ECE diagnostic. copyright 1996 American Institute of Physics

  4. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  5. On mod 2 and higher elliptic genera

    International Nuclear Information System (INIS)

    Liu Kefeng

    1992-01-01

    In the first part of this paper, we construct mod 2 elliptic genera on manifolds of dimensions 8k+1, 8k+2 by mod 2 index formulas of Dirac operators. They are given by mod 2 modular forms or mod 2 automorphic functions. We also obtain an integral formula for the mod 2 index of the Dirac operator. As a by-product we find topological obstructions to group actions. In the second part, we construct higher elliptic genera and prove some of their rigidity properties under group actions. In the third part we write down characteristic series for all Witten genera by Jacobi theta-functions. The modular property and transformation formulas of elliptic genera then follow easily. We shall also prove that Krichever's genera, which come from integrable systems, can be written as indices of twisted Dirac operators for SU-manifolds. Some general discussions about elliptic genera are given. (orig.)

  6. Neutronic analysis of fusion tokamak devices by PHITS

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

    2011-01-01

    A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

  7. Design and operation of Alvand II C tokamak

    International Nuclear Information System (INIS)

    Avakian, M.; Azodi, H.; Naraghi, M.; Tabatabaie, H.; Rezvani, B.; Shahnazi, Sh.; Behrozinia, J.; Solaimani, M.

    1984-01-01

    ALVAND IIC is a research tokamak having a circular cross section with major radius of 45.5 cm and minor radius of 12.6 cm, which has been designed and constructed in the plasma physics division of NRC of AEOI. It is used for developing various diagnostics for low beta plasmas and study of the physical characteristics of these plasmas. This paper discusses the design of ALVAND IIC. (Author)

  8. The role of turbulent suppression in the triggering ITBs on C-Mod

    Science.gov (United States)

    Zhurovich, K.; Fiore, C. L.; Ernst, D. R.; Bonoli, P. T.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Marmar, E. S.; Mikkelsen, D. R.; Phillips, P.; Rice, J. E.

    2007-11-01

    Internal transport barriers can be routinely produced in C-Mod steady EDA H-mode plasmas by applying ICRF at |r/a|>= 0.5. Access to the off-axis ICRF heated ITBs may be understood within the paradigm of marginal stability. Analysis of the Te profiles shows a decrease of R/LTe in the ITB region as the RF resonance is moved off axis. Ti profiles broaden as the ICRF power deposition changes from on-axis to off-axis. TRANSP calculations of the Ti profiles support this trend. Linear GS2 calculations do not reveal any difference in ETG growth rate profiles for ITB vs. non-ITB discharges. However, they do show that the region of stability to ITG modes widens as the ICRF resonance is moved outward. Non-linear simulations show that the outward turbulent particle flux exceeds the Ware pinch by factor of 2 in the outer plasma region. Reducing the temperature gradient significantly decreases the diffusive flux and allows the Ware pinch to peak the density profile. Details of these experiments and simulations will be presented.

  9. A new formulation of the law of octic reciprocity for primes ≡±3(mod8 and its consequences

    Directory of Open Access Journals (Sweden)

    Richard H. Hudson

    1982-01-01

    Full Text Available Let p and q be odd primes with q≡±3(mod8, p≡1(mod8=a2+b2=c2+d2 and with the signs of a and c chosen so that a≡c≡1(mod4. In this paper we show step-by-step how to easily obtain for large q necessary and sufficient criteria to have (−1(q−1/2q(p−1/8≡(a−bd/acj(modp for j=1,…,8 (the cases with j odd have been treated only recently [3] in connection with the sign ambiguity in Jacobsthal sums of order 4. This is accomplished by breaking the formula of A.E. Western into three distinct parts involving two polynomials and a Legendre symbol; the latter condition restricts the validity of the method presented in section 2 to primes q≡3(mod8 and significant modification is needed to obtain similar results for q≡±1(mod8. Only recently the author has completely resolved the case q≡5(mod8, j=1,…,8 and a sketch of the method appears in the closing section of this paper.

  10. The physics of magnetic confinement configurations : Tokamak theory and experiment

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1982-01-01

    Several aspects, both theoretical and experimental, in plasma physics are discussed. The problem of magnetic confinement in Tokamak devices is treated. A discussion on the history of the development and on the future problems to be solved in Tokamaks is made. (L.C.) [pt

  11. Particle injection into the Castor tokamak by electric arcs

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.

    1989-01-01

    The influence of arcing on the tokamak discharge was investigated in the Castor tokamak. A special calibrated gun which emitted tantalum by artificially ignited electric arcs, was used to study the transport of the injected tantalum ions, neutrals and droplets. The injection of tantalum led to an increase in electron density and to a change of plasma position only if the transported charge was higher than 0.01 C. As the naturally occurring arcs are well below this limit, the arcing in tokamaks is rather the consequence than the reason of instabilities. (J.U.)

  12. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  13. ModA and ModB, two ADP-ribosyltransferases encoded by bacteriophage T4: catalytic properties and mutation analysis.

    Science.gov (United States)

    Tiemann, Bernd; Depping, Reinhard; Gineikiene, Egle; Kaliniene, Laura; Nivinskas, Rimas; Rüger, Wolfgang

    2004-11-01

    Bacteriophage T4 encodes three ADP-ribosyltransferases, Alt, ModA, and ModB. These enzymes participate in the regulation of the T4 replication cycle by ADP-ribosylating a defined set of host proteins. In order to obtain a better understanding of the phage-host interactions and their consequences for regulating the T4 replication cycle, we studied cloning, overexpression, and characterization of purified ModA and ModB enzymes. Site-directed mutagenesis confirmed that amino acids, as deduced from secondary structure alignments, are indeed decisive for the activity of the enzymes, implying that the transfer reaction follows the Sn1-type reaction scheme proposed for this class of enzymes. In vitro transcription assays performed with Alt- and ModA-modified RNA polymerases demonstrated that the Alt-ribosylated polymerase enhances transcription from T4 early promoters on a T4 DNA template, whereas the transcriptional activity of ModA-modified polymerase, without the participation of T4-encoded auxiliary proteins for middle mode or late transcription, is reduced. The results presented here support the conclusion that ADP-ribosylation of RNA polymerase and of other host proteins allows initial phage-directed mRNA synthesis reactions to escape from host control. In contrast, subsequent modification of the other cellular target proteins limits transcription from phage early genes and participates in redirecting transcription to phage middle and late genes.

  14. Tokamak concept innovations

    International Nuclear Information System (INIS)

    1986-04-01

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  15. Development of TiC and TiN coated molybdenum limiter system and initial results of the thermal testing in neutral beam heated JFT-2 tokamak

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Sengoku, Seio; Maeno, Masaki; Yamamoto, Shin; Seki, Masahiro; Kazawa, Minoru

    1982-06-01

    This paper describes the limiter drive system for TiC and TiN coated molybdenum limiters and the thermal testing results of the TiC coated limiter in the JFT-2 tokamak using neutral beam injection (0.7 MW). To investigate the influence of TiC coated limiter on plasma behavior and adhesion property under tokamak plasma, a full scale limiter test has been performed in the JFT-2. Reproducible plasma was obtained after the plasma conditioning. Maximum heat flux to the limiter, measured by IR camera, was 1.5 -- 6.5 kW/cm 2 in 25 msec. Cracking, exfoliation and melting on TiC coated limiter were not observed, except for a number of arc tracks. Finally, the permissible heat fluxes of TiC coated molybdenum first wall are discussed. (author)

  16. RELAP5/MOD2 code assessment

    International Nuclear Information System (INIS)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

    1985-01-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G

  17. RELAP5/MOD2 code assessment

    Energy Technology Data Exchange (ETDEWEB)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

    1985-11-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G.

  18. Orbit Representations from Linear mod 1 Transformations

    Directory of Open Access Journals (Sweden)

    Carlos Correia Ramos

    2012-05-01

    Full Text Available We show that every point $x_0in [0,1]$ carries a representationof a $C^*$-algebra that encodes the orbit structure of thelinear mod 1 interval map $f_{eta,alpha}(x=eta x +alpha$. Such $C^*$-algebra is generated by partial isometries arising from the subintervals of monotonicity of the underlying map $f_{eta,alpha}$. Then we prove that such representation is irreducible. Moreover two such of representations are unitarily equivalent if and only if the points belong to the same generalized orbit, for every $alphain [0,1[$ and $etageq 1$.

  19. Steady-state simulations of a 30-tube once-through steam generator with the RELAP5/MOD3 and RELAP5/MOD2 computer codes

    International Nuclear Information System (INIS)

    Hassan, Y.A.; Salim, P.

    1991-01-01

    This paper reports on a steady-state analysis of a 30-tube once-through steam generator that has been performed on the RELAPS/MOD3 and RELAPS/MOD2 computer codes for 100, 75, and 65% loads. Results obtained are compared with experimental data. The RELAP5/MOD3 results for the test facility generally agree reasonably well with the data for the primary-side temperature profiles. The secondary-side temperature profile predicted by RELAP5/MOD3 at 75 and 65% loads agrees fairly well with the data and is better than the RELAP5/MOD2 results. However, the RELAP5/MOD3 calculated secondary-side temperature profile does not compare well with the 100% load data

  20. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  1. Vectorization, parallelization and implementation of nuclear codes [MVP/GMVP, QMDRELP, EQMD, HSABC, CURBAL, STREAM V3.1, TOSCA, EDDYCAL, RELAP5/MOD2/C36-05, RELAP5/MOD3] on the VPP500 computer system. Progress report 1995 fiscal year

    International Nuclear Information System (INIS)

    Nemoto, Toshiyuki; Watanabe, Hideo; Fujita, Toyozo; Kawai, Wataru; Harada, Hiroo; Gorai, Kazuo; Yamasaki, Kazuhiko; Shoji, Makoto; Fujii, Minoru.

    1996-07-01

    At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)

  2. Vectorization, parallelization and implementation of nuclear codes =MVP/GMVP, QMDRELP, EQMD, HSABC, CURBAL, STREAM V3.1, TOSCA, EDDYCAL, RELAP5/MOD2/C36-05, RELAP5/MOD3= on the VPP500 computer system. Progress report 1995 fiscal year

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Toshiyuki; Watanabe, Hideo; Fujita, Toyozo [Fujitsu Ltd., Tokyo (Japan); Kawai, Wataru; Harada, Hiroo; Gorai, Kazuo; Yamasaki, Kazuhiko; Shoji, Makoto; Fujii, Minoru

    1996-06-01

    At Center for Promotion of Computational Science and Engineering, time consuming eight nuclear codes suggested by users have been vectorized, parallelized on the VPP500 computer system. In addition, two nuclear codes used on the VP2600 computer system were implemented on the VPP500 computer system. Neutron and photon transport calculation code MVP/GMVP and relativistic quantum molecular dynamics code QMDRELP have been parallelized. Extended quantum molecular dynamics code EQMD and adiabatic base calculation code HSABC have been parallelized and vectorized. Ballooning turbulence simulation code CURBAL, 3-D non-stationary compressible fluid dynamics code STREAM V3.1, operating plasma analysis code TOSCA and eddy current analysis code EDDYCAL have been vectorized. Reactor safety analysis code RELAP5/MOD2/C36-05 and RELAP5/MOD3 were implemented on the VPP500 computer system. (author)

  3. Power plant design study of a high aspect ratio Tokamak using a SiC composite structure

    International Nuclear Information System (INIS)

    Murakami, Y.; Takase, H.; Shinya, K.

    1998-01-01

    The DREAM (drastically easy maintenance) tokamak is a fusion power plant which is designed from the viewpoint of maintenance feasibility. For this purpose, the DREAM reactor uses a plasma with a very high aspect ratio (A) and adopts SiC as a structural material. The choice of SiC affects the design of the core plasma, i.e. large inboard shield thickness, low synchrotron radiation reflectivity, and small plasma elongation for positional stability. The objectives of this study are to explore the feasibility of a high-A device, such as a power plant, and to clarify the technological impact of SiC material on the plasma design. Plasma size is optimized by the physics guidelines similar to ITER. The plasma major and minor radii of DREAM are 16 m and 2 m, respectively, and the average neutron wall load is 2.5 MW m -2 , the maximum toroidal field is 20 T, and the fusion power is 5.5 GW. Steady-state operation is obtained with 50 MW of external current-drive power and 90% bootstrap current. The divertor heat load is estimated to be about 10 MW m -2 . A radiative divertor concept is adopted to achieve a low divertor plasma temperature. The DREAM tokamak concept is found to be a possible candidate for a future power plant with more than 5 GW of fusion power and an acceptable divertor condition. (orig.)

  4. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  5. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

    Wesson, John; Campbell, D.J.; Connor, J.W.

    1997-01-01

    It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

  6. Tokamak devices: towards controlled fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

  7. Post-test analysis of LOBI BT-01 using RELAP5/MOD2 and RELAP5/MOD3

    International Nuclear Information System (INIS)

    Holmes, B.J.

    1991-08-01

    LOBI is a high pressure, electrically heated integral system test facility simulating a KWU 1300 MW PWR scaled 1:712 by volume, although full scale has been maintained in the vertical direction. This report describes the results of an analysis of test BT-01, which simulates a 10% steam line break. The bulk of the analysis was performed using the Project Version of RELAP5/MOD2, with additional calculations using RELAP5/MOD3 for comparison. The codes provided generally good agreement with data. In particular, the break flows were well modelled, although the mass flow data proved to be unreliable, and this conclusion had to be derived from interpreting other signals. RELAP over-predicted primary/secondary heat transfer in the broken loop, however, leading to a more rapid cool-down of the primary circuit. Furthermore, the primary side pressure response was critically dependent upon the pressuriser behaviour, and the correct timing of the uncovery of the surge line. Inter-phase drag was not well predicted in the broken loop steam generator intermals, although some improvement was seen in the RELAP5/MOD3 predictions. MOD3 gave a reduction in primary/secondary heat transfer during the test pre-conditioning phase, resulting in a lower secondary side pressure at the start of the transient compared with MOD2. (author)

  8. Analysis of the OECD-LOFT International Standard Problem 31 using SCDAP/RELAP5/MOD3

    International Nuclear Information System (INIS)

    Hohorst, J.K.; Allison, C.M.

    1992-01-01

    The CORA-13 bundle heating and melting experiment performed at the Kernforechungszentrum, Karlaruhe, (KfK) was analyzed at the Idaho National Engineering Laboratory (INEL) using SCDAP/RELAP5/MOD3. This analysis was part of a systematic assessment of SCDAP/RELAP5/MOD3 for the US Nuclear Regulatory Commission to (a) evaluate the variances between calculated and observed behavior, (b) identify outstanding modeling deficiencies, and (c) to evaluate the impact of ongoing modeling improvements. A brief discussion of the CORA-13 experiment including a description of the facility, important test conditions, and comparisons with other CORA experimental conditions and results is provided in this report. This report describes the results of the SCDAP/RELAPS/MOD3 analysis including a description of the SCDAP/RELAPS model of the facility, base case results, sensitivity results, and a comparison with other SCDAP/RELAP5/MOD3 code-to-data comparisons

  9. Irradiation hardening of Mod.9Cr-1Mo steel

    International Nuclear Information System (INIS)

    Ryu, Woo-Seog; Kim, Sung-Ho; Choo, Kee-Nam; Kim, Do-Sik

    2009-01-01

    An irradiation test of Mod.9Cr-1Mo steel was carried out in the OR5 test hole of HANARO of a 30 MW thermal power at 390±10degC up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E > 1.0 MeV). The dpa of the irradiated specimens was evaluated to be 0.034 - 0.07. Tensile and impact tests of the irradiated Mod.9Cr-1Mo were done in the hot cell of the IMEF. The change of the tensile strength by irradiation was similar to the change of the yield strength. The increase of the yield and tensile strengths was up to 18% and 10% respectively. The elongation reduction of the weldment was up to 65%. (author)

  10. CORCON-MOD1 modelling improvements

    International Nuclear Information System (INIS)

    Corradini, M.L.; Gonzales, F.G.; Vandervort, C.L.

    1986-01-01

    Given the unlikely occurrence of a severe accident in a light water reactor (LWR), the core may melt and slump into the reactor cavity below the reactor vessel. The interaction of the molten core with exposed concrete (a molten-core-concrete-interaction, MCCI) causes copious gas production which influences further heat transfer and concrete attack and may threaten containment integrity. In this paper the authors focus on the low-temperature phase of the MCCI where the molten pool is partially solidified, but is still capable of attacking concrete. The authors have developed some improved phenomenological models for pool freezing and molten core-coolant heat transfer and have incorporated them into the CORCON-MOD1 computer program. In the paper the authors compare the UW-CORCON/MOD1 calculations to CORCON/MOD2 and WECHSL results as well as the BETA experiments which are being conducted in Germany

  11. TRAC-PF1/MOD2 status and plans

    International Nuclear Information System (INIS)

    Spore, J.W.; Steinke, R.G.; Nelson, R.A.; Cappiello, M.W.; Jenks, R.

    1989-01-01

    The development of the TRAC-PF1/MOD1 code was completed in July 1988 with the release of Version 14.4. A TRAC-PF1/MOD2 code development plan addresses code deficiencies identified in the MOD1 code in order to provide an accurate and defensible tool that can be used to simulate large-break loss-of-coolant accidents (LOCAs), small-break LOCAs, and operational transients. The MOD2 code development plan is an international cooperative effort that includes contributions from Los Alamos National Laboratory, Idaho National Engineering Laboratory (INEL), Japanese Atomic Energy Research Institute (JAERI), Cray Research, Central Electricity Generating Board (CEGB), and United Kingdom Atomic Energy Authority (UKAEA)

  12. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    Caldas, I.L.; Vannucci, A.

    1985-01-01

    A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.) [pt

  13. Prise en compte des ``courants de London'' dans la modélisation des supraconducteurs

    Science.gov (United States)

    Bossavit, Alain

    1997-10-01

    A model is given, in variational form, in which volumic “Bean currents”, ruled by Bean's law, and surface “London currents” coexist. This macroscopic model generalizes Bean's one, by appending to the critical density j_c a second parameter, with the dimension of a length, similar to London's depth λ. The one-dimensional version of the model is investigated, in order to link this parameter with the standard observable H-M characteristics On propose un modèle, sous forme variationnelle, associant des “courants de Bean” volumiques, décrits par la loi de Bean, et des “courants de London”, surfaciques. Ce modèle macroscopique généralise celui de Bean, caractérisé par le courant critique j_c, et fait intervenir un second paramètre, homogène à une longueur, analogue au λ de London. La version unidimensionnelle du modèle est étudiée en détail de manière à relier ce paramètre à l'observation des caractéristiques H-M usuelles.

  14. Flux pinning characteristics of Sn-doped YBCO film by the MOD process

    International Nuclear Information System (INIS)

    Choi, S.M.; Shin, G.M.; Yoo, S.I.

    2013-01-01

    Highlights: ► The pinning effects of undoped and Sn-doped YBCO films by MOD were characterized. ► Sn-containing nanoparticles were trapped in Sn-doped YBCO films by MOD. ► Sn-containing nanoparticles were identified as the YBa 2 SnO 5.5 (YBSO) phase by TEM. ► The YBSO nanoparticles are responsible for improved flux pinning effect. ► We report the orientation relationship between YBSO nanoparticles and YBCO matrix. -- Abstract: Compared with the undoped YBa 2 Cu 3 O 7−δ (YBCO) film, 10 mol% Sn-doped YBCO film exhibited significantly enhanced critical current densities (J c ) in magnetic fields up to 5 T at 65 and 77 K for H//c, indicating that the Sn-doped YBCO film possesses more effective flux pinning centers. Both samples were grown on the SrTiO 3 (STO) (1 0 0) single crystal substrates by the metal-organic deposition (MOD) process. Larger J c (77 K, 1 T) values of Sn-doped YBCO film are observed over a wide field-orientation angle (θ) except the field-orientations close to the ab-plane of YBCO (85° c values for 85° 2 SnO 5.5 (YBSO) phase by STEM (scanning transmission electron microscopy)-EDS (energy dispersive X-ray spectroscopy) analysis. Further analyses by HR-TEM (high resolution-transmission electron microscopy) revealed that YBSO nanoparticles completely surrounded by the YBCO matrix had random orientation with YBCO while those located at the interface of YBCO/STO substrate had epitaxial relationship with YBCO

  15. Investigation of dynamics of ELM crashes and their mitigation techniques

    Energy Technology Data Exchange (ETDEWEB)

    Pankin, Alexei Y. [Tech-X Corporation, Boulder, CO (United States)

    2015-08-14

    The accurate prediction of H-mode pedestal dynamics is critical for planning experiments in existing tokamaks and in the design of future tokamaks such as ITER and DEMO. The main objective of the proposed research is to advance the understanding of the physics of H-mode pedestal. Through advances in coupled kinetic-MHD simulations, a new model for H-mode pedestal and ELM crashes as well as an improved model for the bootstrap current will be developed. ELMmitigation techniques will also be investigated. The proposed research will help design efficient confinement scenarios and reduce transient heat loads on the divertor and plasma facing components. During the last two years, the principal investigator (PI) of this proposal actively participated in physics studies related to the DOE Joint Research Targets. These studies include the modeling of divertor heat load in the DIII-D, Alcator C-Mod, and NSTX tokamaks in 2010, and the modeling of H-mode pedestal structure in the DIII-D tokamak in 2011. It is proposed that this close collaboration with experimentalists from major US tokamaks continue during the next funding period. Verification and validation will be a strong component of the proposed research. During the course of the project, advances will be made in the following areas; Dynamics of the H-mode pedestal buildup and recovery after ELM crashes – The effects of neutral fueling, particle and thermal pinches will be explored; Dynamics of ELM crashes in realistic tokamak geometries – Heat loads associated with ELM crashes will be validated against experimental measurements. An improved model for ELM crashes will be developed; ELM mitigation – The effect of resonant magnetic perturbations on ELMs stability and their evolution will be investigated; Development of a new bootstrap current model – A reduced model for will be developed through careful verification of existing models for bootstrap current against first-principle kinetic neoclassical simulations

  16. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

    For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

  17. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

    The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

  18. NetMOD Version 2.0 User?s Manual.

    Energy Technology Data Exchange (ETDEWEB)

    Merchant, Bion J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-10-01

    NetMOD ( Net work M onitoring for O ptimal D etection) is a Java-based software package for conducting simulation of seismic, hydracoustic, and infrasonic networks. Specifically, NetMOD simulates the detection capabilities of monitoring networks. Network simulations have long been used to study network resilience to station outages and to determine where additional stations are needed to reduce monitoring thresholds. NetMOD makes use of geophysical models to determine the source characteristics, signal attenuation along the path between the source and station, and the performance and noise properties of the station. These geophysical models are combined to simulate the relative amplitudes of signal and noise that are observed at each of the stations. From these signal-to-noise ratios (SNR), the probability of detection can be computed given a detection threshold. This manual describes how to configure and operate NetMOD to perform detection simulations. In addition, NetMOD is distributed with simulation datasets for the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) International Monitoring System (IMS) seismic, hydroacoustic, and infrasonic networks for the purpose of demonstrating NetMOD's capabilities and providing user training. The tutorial sections of this manual use this dataset when describing how to perform the steps involved when running a simulation. ACKNOWLEDGEMENTS We would like to thank the reviewers of this document for their contributions.

  19. Comparison of SCDAP/RELAP5/MOD3 to TRAC-PF1/MOD1 for timing analysis of PWR fuel pin failures

    International Nuclear Information System (INIS)

    Jones, K.R.; Katsma, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.

    1991-01-01

    A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calculate the thermal-hydraulic boundary conditions for a complete, double-ended, offset-shear break of a cold leg in a Westinghouse 4-loop pressurized water reactor. Both calculations used the FRAPCON-2 code to calculate the steady-state fuel rod behavior and the FRAP-T6 code to calculate the transient fuel rod behavior. The analysis was performed for 16 combinations of fuel burnups and power peaking factors extending up to the Technical Specifications limits. While all calculations were made on a best-estimate basis, the SCDAP/RELAP5/MOD3 code has not yet been fully assessed for large-break LOCA analysis. The results indicate that SCDAP/RELAP5/MOD3 yields conservative fuel pin failure timing results in comparison to those generated using TRAC-PF1/MOD1. 7 refs., 5 figs

  20. Integrated, Reactor Relevant Solutions for Lower Hybrid Range of Frequencies Actuators

    Science.gov (United States)

    Shiraiwa, S.; Bonoli, P. T.; Lin, Y.; Wallace, G. M.; Wukitch, S. J.

    2017-10-01

    RF (radiofrequency) actuators with high system efficiency (wall-plug to plasma) and ability for continuous operation have long be recognized as essential tools for realizing a steady state tokamak. A number of physics and technological challenges to utilization remain including current drive efficiency and location, efficient coupling, and impurity contamination. In a reactor environment, plasma material interaction (PMI) issues associated with coupling structures are similar to the first wall and have been identified as a potential show-stopper. High field side (HFS) launch of LHRF power represents an integrated solution that both improves core wave physics and mitigates PMI/coupling issues. For HFS LHRF, wave penetration is vastly improves because wave accessibility scales as 1/B allowing for launching the wave at lower n|| (parallel refractive index). The lower n|| penetrate to higher electron temperature resulting in higher current drive efficiency (1/n||2). HFS RF launch also provides for a means to dramatically improve launcher robustness in a reactor environment. On the HFS, the SOL is quiescent; local density profile is steep and controlled through magnetic shape; fast particle, neutron, turbulent heat and particle fluxes are eliminated or minim Work supported by the U.S. DoE, Office of Science, Office of Fusion Energy Sciences, User Facility Alcator C-Mod under DE-FC02-99ER54512 and US DoE Contract No. DE-FC02-01ER54648 under a Scientific Discovery through Advanced Computing Initiative.

  1. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  2. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    Hendel, H.W.; Jassby, D.L.

    1989-11-01

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  3. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  4. Bascule d'un modèle poutre à un modèle 3D en dynamique des machines tournantes

    OpenAIRE

    Tannous , Mikhael; Cartraud , Patrice; Dureisseix , David; Torkhani , Mohamed

    2013-01-01

    National audience; Les problèmes de machines tournantes incluant un contact rotor-stator, nécessitent un maillage 3D de la zone de contact. Cependant, un modèle 3D pour toute la durée de simulation conduit à des temps de calcul rédhibitoires. Or un modèle poutre est suffisant pour décrire la dynamique de la machine tournante hors contact. Une stratégie qui permet d'utiliser un modèle poutre et un autre 3D, pendant deux phases différentes durant la même simulation, permet donc de gagner en tem...

  5. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  6. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  7. Modelling of transient electromagnetics in Tokamaks during off-normal conditions

    Science.gov (United States)

    Schneider, J. H.; Crutzen, Y. R.; Papadopoulos, S.; Richard, N.

    1993-03-01

    During plasma disruption events in Tokamaks, a considerable amount of magnetic and thermal energy is associated to the transfer of plasma current into eddy and Halo currents. In this paper, a predictive numerical modelling is described concerning plasma-wall interactions during disruptive instabilities. Preliminary results are presented, giving an estimation of heat transfer interaction with electromagnetic phenomena. Eddy currents and heat deposition increase significantly with decreasing disruption time. An estimation of the order of magnitude of Halo currents and associated forces on plasma-facing conducting components is also presented. Durant une disruption de plasma dans les Tokamaks, une quantité considérable d'énergie magnétique et thermique est associée au transfert de courant de plasma sous forme de courants de Foucault et de “Halo”. Cette contribution décrit un modèle numérique capable de prédire les interactions plasma-première paroi durant ces instabilités disruptives. La présentation de résultats préliminaires comprend une estimation de l'interaction entre le phénomène électromagnétique et le transfert de chaleur. Les courants de Foucault et les pertes Joule augmentent fortement quand la durée de disruption devient plus brève. Une évaluation de l'ordre de grandeur des courants de “Halo” et des forces résultantes, appliqués aux composants métalliques placés face au plasma, est également illustrée.

  8. An Optimization Scheme for ProdMod

    International Nuclear Information System (INIS)

    Gregory, M.V.

    1999-01-01

    A general purpose dynamic optimization scheme has been devised in conjunction with the ProdMod simulator. The optimization scheme is suitable for the Savannah River Site (SRS) High Level Waste (HLW) complex operations, and able to handle different types of optimizations such as linear, nonlinear, etc. The optimization is performed in the stand-alone FORTRAN based optimization deliver, while the optimizer is interfaced with the ProdMod simulator for flow of information between the two

  9. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  10. Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

    International Nuclear Information System (INIS)

    Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka

    2014-03-01

    The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

  11. Conceptual design of the 7 megawatt Mod-5B wind turbine generator

    Science.gov (United States)

    Douglas, R. R.

    1982-01-01

    Similar to MOD-2, the MOD-5B wind turbine generator system is designed for the sole purpose of providing electrical power for distribution by a major utility network. The objectives of the MOD-2 and MOD-5B programs are essentially identical with one important exception; the cost-of-electricity (COE) target is reduced from 4 cent/Kwhr on MOD-2 to 3 cent/Kwhr on MOD-5B, based on mid 1977 dollars and large quantity production. The MOD-5B concept studies and eventual concept selection confirmed that the program COE targets could not only be achieved but substantially bettered. Starting from the established MOD-2 technology as a base, this achievement resulted from a combination of concept changes, size changes, and design refinements. The result of this effort is a wind turbine system that can compete with conventional power generation over significant geographical areas, increasing commercial market potential by an order of magnitude.

  12. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

    The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

  13. NetMOD version 1.0 user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Merchant, Bion John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-01-01

    NetMOD (Network Monitoring for Optimal Detection) is a Java-based software package for conducting simulation of seismic networks. Specifically, NetMOD simulates the detection capabilities of seismic monitoring networks. Network simulations have long been used to study network resilience to station outages and to determine where additional stations are needed to reduce monitoring thresholds. NetMOD makes use of geophysical models to determine the source characteristics, signal attenuation along the path between the source and station, and the performance and noise properties of the station. These geophysical models are combined to simulate the relative amplitudes of signal and noise that are observed at each of the stations. From these signal-to-noise ratios (SNR), the probability of detection can be computed given a detection threshold. This manual describes how to configure and operate NetMOD to perform seismic detection simulations. In addition, NetMOD is distributed with a simulation dataset for the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) International Monitoring System (IMS) seismic network for the purpose of demonstrating NetMOD's capabilities and providing user training. The tutorial sections of this manual use this dataset when describing how to perform the steps involved when running a simulation.

  14. MOD silver metallization for photovoltaics

    Science.gov (United States)

    Vest, G. M.; Vest, R. W.

    1984-01-01

    The development of flat plate solar arrays is reported. Photovoltaic cells require back side metallization and a collector grid system on the front surface. Metallo-organic decomposition (MOD) silver films can eliminate most of the present problems with silver conductors. The objectives are to: (1) identify and characterize suitable MO compounds; (2) develop generic synthesis procedures for the MO compounds; (3) develop generic fabrication procedures to screen printable MOD silver inks; (4) optimize processing conditions to produce grid patterns and photovoltaic cells; and (5) develop a model which describes the adhesion between the fired silver film and the silicon surface.

  15. NetMOD Version 2.0 Parameters

    Energy Technology Data Exchange (ETDEWEB)

    Merchant, Bion J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    NetMOD ( Net work M onitoring for O ptimal D etection) is a Java-based software package for conducting simulation of seismic, hydroacoustic and infrasonic networks. Network simulations have long been used to study network resilience to station outages and to determine where additional stations are needed to reduce monitoring thresholds. NetMOD makes use of geophysical models to determine the source characteristics, signal attenuation along the path between the source and station, and the performance and noise properties of the station. These geophysical models are combined to simulate the relative amplitudes of signal and noise that are observed at each of the stations. From these signal-to-noise ratios (SNR), the probability of detection can be computed given a detection threshold. This document describes the parameters that are used to configure the NetMOD tool and the input and output parameters that make up the simulation definitions.

  16. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

    1980-06-01

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  17. Numerical modeling of a fast-neutron collimator for the Alcator A fusion device

    International Nuclear Information System (INIS)

    Fisher, W.A.

    1982-12-01

    A numerical procedure is developed to analyze neutron collimators used for spatial neutron measurements of plasma neutrons. The procedure is based upon Monte-Carlo methods and uses a standard Monte-Carlo code. The specific developments described herein involve a new approach to represent complex spatial details in a method that is conservative of computer time, retains accuracy and required only modest changes in already-developed Monte-Carlo procedures. The procedure was used to model the Alcator A collimator. The collimator consists of 448 cells and has a measured spatial point source response of 0.7 cm. The numerical procedure successfully predicts this response

  18. Feedback-Assisted Extension of the Tokamak Operating Space to Low Safety Factor

    Science.gov (United States)

    Hanson, J. M.

    2013-10-01

    Recent DIII-D experiments have demonstrated stable operation at very low edge safety factor, q95 instability. The performance of tokamak fusion devices may benefit from increased plasma current, and thus, decreased q. However, disruptive stability limits are commonly encountered in experiments at qedge ~ 2 (limited plasmas) and q95 ~ 2 (diverted plasmas), limiting exploration of low q regimes. In the recent DIII-D experiments, the impact and control of key disruptive instabilities was studied. Locked n = 1 modes with exponential growth times on the order of the wall eddy current decay timescale τw preceded disruptions at q95 = 2 . The instabilities have a poloidal structure that is consistent with VALEN simulations of the RWM mode structure at q95 = 2 . Applying proportional gain magnetic feedback control of the n = 1 mode resulted in stabilized operation with q95 reaching 1.9, and an extension of the discharge lifetime for > 100τw . Loss of feedback control was accompanied by power supply saturation, followed by a rapidly growing n = 1 mode and disruption. Comparisons of the feedback dynamics with VALEN simulations will be presented. The DIII-D results complement and will be discussed alongside recent RFX-MOD demonstrations of RWM control using magnetic feedback in limited tokamak discharges with qedge economical fusion power production. Supported by the US Department of Energy under DE-FG02-04ER54761 and DE-FC02-04ER54698.

  19. RELAP5/MOD3 code coupling model

    International Nuclear Information System (INIS)

    Martin, R.P.; Johnsen, G.W.

    1994-01-01

    A new capability has been incorporated into RELAP5/MOD3 that enables the coupling of RELAP5/MOD3 to other computer codes. The new capability has been designed to support analysis of the new advanced reactor concepts. Its user features rely solely on new RELAP5 open-quotes styledclose quotes input and the Parallel Virtual Machine (PVM) software, which facilitates process management and distributed communication of multiprocess problems. RELAP5/MOD3 manages the input processing, communication instruction, process synchronization, and its own send and receive data processing. The flexible capability requires that an explicit coupling be established, which updates boundary conditions at discrete time intervals. Two test cases are presented that demonstrate the functionality, applicability, and issues involving use of this capability

  20. ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

    Science.gov (United States)

    Whyte, Dennis; ADX Team

    2015-11-01

    The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

  1. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  2. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  3. Modeling of EAST ICRF antenna performance using the full-wave code TORIC

    Energy Technology Data Exchange (ETDEWEB)

    Edlund, E. M., E-mail: eedlund@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Bonoli, P. T.; Porkolab, M.; Wukitch, S. J. [MIT Plasma Science and Fusion Center, Cambridge, MA (United States)

    2015-12-10

    Access to advanced operating regimes in the EAST tokamak will require a combination of electron-cyclotron resonance heating (ECRH), neutral beam injection (NBI) and ion cyclotron range frequency heating (ICRF), with the addition of lower-hybrid current drive (LHCD) for current profile control. Prior experiments at the EAST tokamak facility have shown relatively weak response of the plasma temperature to application of ICRF heating, with typical coupled power about 2 MW out of 12 MW source. The launched spectrum, at n{sub φ} = 34 for 0-π -0-π phasing and 27 MHz, is largely inaccessible at line-averaged densities of approximately 2 × 10{sup 19} m{sup −3}. However, with variable antenna phasing and frequency, this system has considerable latitude to explore different heating schemes. To develop an ICRF actuator control model, we have used the full-wave code TORIC to explore the physics of ICRF wave propagation in EAST. The results presented from this study use a spectrum analysis using a superposition of n{sub φ} spanning −50 to +50. The low density regime typical of EAST plasmas results in a perpendicular wavelength comparable to the minor radius which results in global cavity resonance effects and eigenmode formation when the single-pass absorption is low. This behavior indicates that improved performance can be attained by lowering the peak of the k{sub ||} spectrum by using π/3 phasing of the 4-strap antenna. Based on prior studies conducted at Alcator C-Mod, this phasing is also expected to have the advantage of nearly divergence-free box currents, which should result in reduced levels of impurity production. Significant enhancements of the loading resistance may be achieved by using low k{sub ||} phasing and a combination of magnetic field and frequency to vary the location of the resonance and mode conversion regions. TORIC calculations indicate that the significant power may be channeled to the electrons and deuterium majority. We expect that

  4. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  5. Advanced Tokamak Stability Theory

    Science.gov (United States)

    Zheng, Linjin

    2015-03-01

    The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

  6. VARSKIN MOD 2 and SADDE MOD2: Computer codes for assessing skin dose from skin contamination

    International Nuclear Information System (INIS)

    Durham, J.S.

    1992-12-01

    The computer code VARSKIN has been modified to calculate dose to skin from three-dimensional sources, sources separated from the skin by layers of protective clothing, and gamma dose from certain radionuclides correction for backscatter has also been incorporated for certain geometries. This document describes the new code, VARSKIN Mod 2, including installation and operation instructions, provides detailed descriptions of the models used, and suggests methods for avoiding misuse of the code. The input data file for VARSKIN Mod 2 has been modified to reflect current physical data, to include the contribution to dose from internal conversion and Auger electrons, and to reflect a correction for low-energy electrons. In addition, the computer code SADDE: Scaled Absorbed Dose Distribution Evaluator has been modified to allow the generation of scaled absorbed dose distributions for mixtures of radionuclides and intereat conversion and Auger electrons. This new code, SADDE Mod 2, is also described in this document. Instructions for installation and operation of the code and detailed descriptions of the models used in the code are provided

  7. Properties of the periplasmic ModA molybdate-binding protein of Escherichia coli.

    Science.gov (United States)

    Rech, S; Wolin, C; Gunsalus, R P

    1996-02-02

    The modABCD operon, located at 17 min on the Escherichia coli chromosome, encodes the protein components of a high affinity molybdate uptake system. Sequence analysis of the modA gene (GenBank L34009) predicts that it encodes a periplasmic binding protein based on the presence of a leader-like sequence at its N terminus. To examine the properties of the ModA protein, the modA structural gene was overexpressed, and its product was purified. The ModA protein was localized to the periplasmic space of the cell, and it was released following a gentle osmotic shock. The N-terminal sequence of ModA confirmed that a leader region of 24 amino acids was removed upon export from the cell. The apparent size of ModA is 31.6 kDa as determined by gel sieve chromatography, whereas it is 22.5 kDa when examined by SDS-polyacrylamide gel electrophoresis. A ligand-dependent protein mobility shift assay was devised using a native polyacrylamide gel electrophoresis protocol to examine binding of molybdate and other anions to the ModA periplasmic protein. Whereas molybdate and tungstate were bound with high affinity (approximately 5 microM), sulfate, chromate, selenate, phosphate, and chlorate did not bind even when tested at 2 mM. A UV spectral assay revealed apparent Kd values of binding for molybdate and tungstate of 3 and 7 microM, respectively. Strains defective in the modA gene were unable to transport molybdate unless high levels of the anion were supplied in the medium. Therefore the modA gene product is essential for high affinity molybdate uptake by the cell. Tungstate interference of molybdate acquisition by the cell is apparently due in part to the high affinity of the ModA protein for this anion.

  8. High-β steady-state advanced tokamak regimes for ITER and FIRE

    International Nuclear Information System (INIS)

    Meade, D.M.; Sauthoff, N.R.; Kessel, C.E.; Budny, R.V.; Gorelenkov, N.; Jardin, S.C.; Schmidt, J.A.; Navratil, G.A.; Bialek, J.; Ulrickson, M.A.; Rognlein, T.; Mandrekas, J.

    2005-01-01

    An attractive tokamak-based fusion power plant will require the development of high-β steady-state advanced tokamak regimes to produce a high-gain burning plasma with a large fraction of self-driven current and high fusion-power density. Both ITER and FIRE are being designed with the objective to address these issues by exploring and understanding burning plasma physics both in the conventional H-mode regime, and in advanced tokamak regimes with β N ∼ 3 - 4, and f bs ∼50-80%. ITER has employed conservative scenarios, as appropriate for its nuclear technology mission, while FIRE has employed more aggressive assumptions aimed at exploring the scenarios envisioned in the ARIES power-plant studies. The main characteristics of the advanced scenarios presently under study for ITER and FIRE are compared with advanced tokamak regimes envisioned for the European Power Plant Conceptual Study (PPCS-C), the US ARIES-RS Power Plant Study and the Japanese Advanced Steady-State Tokamak Reactor (ASSTR). The goal of the present work is to develop advanced tokamak scenarios that would fully exploit the capability of ITER and FIRE. This paper will summarize the status of the work and indicate critical areas where further R and D is needed. (author)

  9. Implementation of the thermal-hydraulic transient analysis code RELAP4/MOD5 and MOD6 on the FACOM 230/75 computer system

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Ishigai, Takahiro; Kumakura, Toshimasa; Naraoka, Ken-itsu

    1979-03-01

    Development efforts have continued on the extensively used LOCA analysis code RELAP-4, as seen in its history; that is, from the prototype version MOD2 to the latest one MOD6 which is capable of one-through calculations from blowdown to reflood phase of PWR-LOCA. Many improvements and refinements of the models have enlarged the scopes and extents of phenomena to treat. Correspondingly the size of program has increased version to version, and special programming techniques have continuously been introduced to manage the program within limited capacity of core memory. For example, the Dynamic Storage Allocation of MOD5 and the PRELOAD Preprocessor newly incorporated in MOD6 are those designed for the CDC computer with relatively small core size. Described are these programming techniques in detail and experiences on implementation of the codes on FACOM 230/75, together with some results of confirmatory calculations. (author)

  10. Sustainment of spherical tokamak by means of repetitive injection of compact torus plasma

    International Nuclear Information System (INIS)

    Shimamura, Shin; Matsura, Ken; Takahashi, Tsutomu; Nogi, Yasuyuki

    2000-01-01

    Sustainment of spherical tokamak (S.T.) has been studied. A compact torus (C.T.) plasma was injected into confinement region by magnetized coaxial gun. For start-up and sustainment of large main spherical tokamak, single pulsed injection of small C.T. is not sufficient in many cases. C.T.plasma injection of high repetition rate is required. For this purpose magnetized coaxial gun was driven with high repetition rate current. The first injected C.T. plasma could start-up S.T. without other help. The repetitive C.T. injection grew and sustained the S.T. plasma. A CCD camera with fast gated image intensifier took a cross sectional view of S.T. during the repetitive C.T. injection. (author)

  11. Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

    International Nuclear Information System (INIS)

    Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

    2009-01-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

  12. Tokamak control simulator

    International Nuclear Information System (INIS)

    Edelbaum, T.N.; Serben, S.; Var, R.E.

    1976-01-01

    A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks

  13. Device description and general physics requirements

    International Nuclear Information System (INIS)

    Neilson, G.H. Jr.

    1992-01-01

    To accomplish the dual goals set forth in the mission statement - determination of burning plasma physics and demonstration of fusion power production - requires a tokamak with special characteristics. A conceptual design for such a facility has been developed by the CIT/BPX Project team over a period of about 5 years. The process has drawn extensively upon the world tokamak physics data base as well as engineering experience gained on actual machines like Tokamak Fusion Test Reactor (TFTR) and the Alcators and on design studies for machines like the Long-Pulse Ignited Test experiment (LITE) and the Tokamak Fusion Core Experiment (TFCX). The Joint European Torus (JET) and Doublet III-D (DIII-D) experiments in particular have had a significant influence on the physics base. The resulting design incorporates features that have proven successful in tokamak experiments, combined with new features that are needed in the regime of high-power-density deuterium-tritium (D-T) fusion plasmas

  14. NetMOD Version 2.0 Mathematical Framework

    Energy Technology Data Exchange (ETDEWEB)

    Merchant, Bion J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Young, Christopher J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Chael, Eric P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    NetMOD ( Net work M onitoring for O ptimal D etection) is a Java-based software package for conducting simulation of seismic, hydroacoustic and infrasonic networks. Network simulations have long been used to study network resilience to station outages and to determine where additional stations are needed to reduce monitoring thresholds. NetMOD makes use of geophysical models to determine the source characteristics, signal attenuation along the path between the source and station, and the performance and noise properties of the station. These geophysical models are combined to simulate the relative amplitudes of signal and noise that are observed at each of the stations. From these signal-to-noise ratios (SNR), the probabilities of signal detection at each station and event detection across the network of stations can be computed given a detection threshold. The purpose of this document is to clearly and comprehensively present the mathematical framework used by NetMOD, the software package developed by Sandia National Laboratories to assess the monitoring capability of ground-based sensor networks. Many of the NetMOD equations used for simulations are inherited from the NetSim network capability assessment package developed in the late 1980s by SAIC (Sereno et al., 1990).

  15. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1977-01-01

    The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

  16. Mod 1 ICS TI Report: ICS Conversion of a 140% HPGe Detector

    Energy Technology Data Exchange (ETDEWEB)

    Bounds, John Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-07-05

    This report evaluates the Mod 1 ICS, an electrically cooled 140% HPGe detector. It is a custom version of the ORTEC Integrated Cooling System (ICS) modified to make it more practical for us to use in the field. Performance and operating characteristics of the Mod 1 ICS are documented, noting both pros and cons. The Mod 1 ICS is deemed a success. Recommendations for a Mod 2 ICS, a true field prototype, are provided.

  17. Tokamak building-design considerations for a large tokamak device

    International Nuclear Information System (INIS)

    Barrett, R.J.; Thomson, S.L.

    1981-01-01

    Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

  18. Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

    International Nuclear Information System (INIS)

    Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P

    2004-01-01

    Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values

  19. Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2004-09-01

    Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

  20. Assessment of full power turbine trip start-up test for C. Trillo 1 with RELAP5/MOD2

    International Nuclear Information System (INIS)

    Lozano, M.F.; Moreno, P.; de la Cal, C.; Larrea, E.; Lopez, A.; Santamaria, J.G.; Lopez, E.; Novo, M.

    1993-07-01

    C. Trillo I has developed a model of the plant with RELAP5/MOD2/36.04. This model will be validated against a selected set of start-up tests. One of the transients selected to that aim is the turbine trip, which presents very specific characteristics that make it significantly different from the same transient in other PWRs of different design, the main difference being that the reactor is not tripped: a reduction in primary power is carried out instead. Pre-test calculations were done of the Turbine Trip Test and compared against the actual test. Minor problems in the first model, specially in the Control and Limitation Systems, were identified and post-test calculations had been carried out. The results show a good agreement with data for all the compared variables

  1. Simulation of tokamak runaway-electron events

    International Nuclear Information System (INIS)

    Bolt, H.; Miyahara, A.; Miyake, M.; Yamamoto, T.

    1987-08-01

    High energy runaway-electron events which can occur in tokamaks when the plasma hits the first wall are a critical issue for the materials selection of future devices. Runaway-electron events are simulated with an electron linear accelerator to better understand the observed runaway-electron damage to tokamak first wall materials and to consider the runaway-electron issue in further materials development and selection. The electron linear accelerator produces beam energies of 20 to 30 MeV at an integrated power input of up to 1.3 kW. Graphite, SiC + 2 % AlN, stainless steel, molybdenum and tungsten have been tested as bulk materials. To test the reliability of actively cooled systems under runaway-electron impact layer systems of graphite fixed to metal substrates have been tested. The irradiation resulted in damage to the metal compounds but left graphite and SiC + 2 % AlN without damage. Metal substrates of graphite - metal systems for actively cooled structures suffer severe damage unless thick graphite shielding is provided. (author)

  2. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

  3. D-D tokamak reactor assessment

    International Nuclear Information System (INIS)

    Baxter, D.C.; Dabiri, A.E.

    1983-01-01

    A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

  4. Microscopic observation drug susceptibility assay (MODS for early diagnosis of tuberculosis in children.

    Directory of Open Access Journals (Sweden)

    Dang Thi Minh Ha

    2009-12-01

    Full Text Available MODS is a novel liquid culture based technique that has been shown to be effective and rapid for early diagnosis of tuberculosis (TB. We evaluated the MODS assay for diagnosis of TB in children in Viet Nam. 217 consecutive samples including sputum (n = 132, gastric fluid (n = 50, CSF (n = 32 and pleural fluid (n = 3 collected from 96 children with suspected TB, were tested by smear, MODS and MGIT. When test results were aggregated by patient, the sensitivity and specificity of smear, MGIT and MODS against "clinical diagnosis" (confirmed and probable groups as the gold standard were 28.2% and 100%, 42.3% and 100%, 39.7% and 94.4%, respectively. The sensitivity of MGIT and MODS was not significantly different in this analysis (P = 0.5, but MGIT was more sensitive than MODS when analysed on the sample level using a marginal model (P = 0.03. The median time to detection of MODS and MGIT were 8 days and 13 days, respectively, and the time to detection was significantly shorter for MODS in samples where both tests were positive (P<0.001. An analysis of time-dependent sensitivity showed that the detection rates were significantly higher for MODS than for MGIT by day 7 or day 14 (P<0.001 and P = 0.04, respectively. MODS is a rapid and sensitive alternative method for the isolation of M.tuberculosis from children.

  5. Ignition experiment - alternatives

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1979-10-01

    This report comprises three short papers on cost estimates, integral burn time and alternative versions of Tokamak ignition experiments. These papers were discussed at the ZEPHYR workshop with participants from IPP Garching, MIT Cambridge and PPPL Princeton (Garching July 30 - August 2 1979) (Chapters A, B, C). It is shown, that starting from a practical parameter independent minimum integral burn time of Tokamak ignition experiments (some 10 3 s) by adding a shield for protection of the magnet insulation (permitted neutron dose 10 9 rad) an integral burn time of some 10 4 s can be achieved for only about 30% more outlay. For a substantially longer integral burn time the outlay approaches rather quickly that for a Tokamak reactor. Some examples for alternatives to ZEPHYR are being given, including some with low or no compression. In a further chapter D some early results of evaluating an ignition experiment on the basis of the energy confinement scaling put forward by Coppi and Mazzucato are presented. As opposed to the case of the Alcator scaling used in chapters A through C the minimum integral burn time of Tokamak ignition experiments here depends on the plasma current. Provided neutral injectors up to about 160 keV are available compression boosting is not required with this scaling. The results presented have been obtained neglecting the effects of the toroidal field ripple. (orig.) 891 HT/orig. 892 RKD [de

  6. Controlled thermonuclear fusion in TOKAMAK type reactors, the European example: Joint European Torus (JET)

    International Nuclear Information System (INIS)

    Paris, P.J.; Yassen, F.; Assis, A.S. de; Raposo, C.

    1988-07-01

    The development of controlled thermonuclear reaction in TOKAMAK type reactors, and the main projects in the world are presented. The main characteristics of the JET (Joint European Torus) program, the perspectives for energy production, and the international cooperation for viable use of the TOKAMAK are analysed. (M.C.K.) [pt

  7. Hot particle dose calculations using the computer code VARSKIN Mod 2

    International Nuclear Information System (INIS)

    Durham, J.S.

    1991-01-01

    The only calculational model recognised by the Nuclear Regulatory Commission (NRC) for hot particle dosimetry is VARSKIN Mod 1. Because the code was designed to calculate skin dose from distributed skin contamination and not hot particles, it is assumed that the particle has no thickness and, therefore, that no self-absorption occurs within the source material. For low energy beta particles such as those emitted from 60 Co, a significant amount of self-shielding occurs in hot particles and VARSKIN Mod 1 overestimates the skin dose. In addition, the presence of protective clothing, which will reduce the calculated skin dose for both high and low energy beta emitters, is not modelled in VARSKIN Mod 1. Finally, there is no provision in VARSKIN Mod 1 to calculate the gamma contribution to skin dose from radionuclides that emit both beta and gamma radiation. The computer code VARSKIN Mod 1 has been modified to model three-dimensional sources, insertion of layers of protective clothing between the source and skin, and gamma dose from appropriate radionuclides. The new code, VARSKIN Mod 2, is described and the sensitivity of the calculated dose to source geometry, diameter, thickness, density, and protective clothing thickness are discussed. Finally, doses calculated using VARSKIN Mod 2 are compared to doses measured from hot particles found in nuclear power plants. (author)

  8. Teaching Mods with Class

    DEFF Research Database (Denmark)

    Champion, Erik

    2012-01-01

    from around the world, representing fields as diverse as architecture, ethnography, puppetry, cultural studies, music education, interaction design and industrial design. How can we design, play with and reflect on the contribution of game mods, related tools and techniques, to both game studies...

  9. C-Vitamin mod åreforkalkning

    DEFF Research Database (Denmark)

    Frikke-Schmidt, Henriette Rønne; Lykkesfeldt, Jens

    2009-01-01

    Overraskende mange mennesker i den vestlige verden får ikke C-vitamin nok. Muligvis vil tilskud med C-vitamin kunne forebygge hjertekarsygdomme.......Overraskende mange mennesker i den vestlige verden får ikke C-vitamin nok. Muligvis vil tilskud med C-vitamin kunne forebygge hjertekarsygdomme....

  10. Evaluation of the applicability of cladding deformation model in RELAP5/MOD3.2 code for VVER-1000 fuel

    International Nuclear Information System (INIS)

    Vorob'ev, Yu.; Zhabin, O.

    2015-01-01

    Applicability of cladding deformation model in RELAP5/MOD3.2 code is analyzed for VVER-1000 fuel cladding from Zr+1%Nb alloy. Experimental data and calculation model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the cladding temperature range from 600 to 1200 deg C and pressure range from 1 to 12 MPa. Evaluation results demonstrate limited applicability of built-in RELAP5/MOD3.2 cladding deformation model to the estimation of Zr+1%Nb cladding rupture conditions. The limitations found shall be considered in application of RELAP5/MOD3.2 cladding deformation model in the design-basis accident analysis of VVER reactors

  11. Méditerranée asiatique : un modèle urbain polycentrique

    OpenAIRE

    Gipouloux , François

    2012-01-01

    Publication de la revue "Diplomatie. Affaires stratégiques et relations internationales" http://www.diplomatie-presse.com/; International audience; En Asie orientale, un corridor maritime a pris forme autour de plates-formes logistiques et portuaires, fidèles au modèle de Hong Kong. C'est là que bat le coeur de l'économie régionale.

  12. NAUA Mod 4

    International Nuclear Information System (INIS)

    Bunz, H.; Koyro, M.; Schoeck, W.

    1983-08-01

    This report describes the computer program NAUA Mod4. Its purpose is to calculate the behaviour of a polydisperse aerosol system in a closed vessel containing a condensing atmosphere as a function of the time. The main object is to explain the physical background and to describe the structure of the code and the input and output in detail. (orig.) [de

  13. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1994-01-01

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  14. Developmental assessment of RELAP5/MOD3 using the semiscale natural circulation tests

    International Nuclear Information System (INIS)

    Carlson, K.E.

    1990-01-01

    A code development effort creating RELAP5/MOD3 from RELAP5/MOD2 has been completed. Upon completion, a developmental assessment task was performed. One of the problems used for the developmental assessment was the Semiscale Natural Circulation Test. Calculated results from RELAP5/MOD3 are compared to measured data and previously calculated results from RELAP5/MOD2. 10 refs., 6 figs., 1 tab

  15. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

    The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)

  16. Liquid tin limiter for FTU tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Vertkov, A., E-mail: avertkov@yandex.ru [JSC “Red Star”, Moscow (Russian Federation); Lyublinski, I. [JSC “Red Star”, Moscow (Russian Federation); NRNU MEPhI, Moscow (Russian Federation); Zharkov, M. [JSC “Red Star”, Moscow (Russian Federation); Mazzitelli, G.; Apicella, M.L.; Iafrati, M. [Associazione EURATOM-ENEA sulla Fusione, C. R. Frascati, Frascati, Rome, Italy, (Italy)

    2017-04-15

    Highlights: • First steady state operating liquid tin limiter TLL is under study on FTU tokamak. • The cooling system with water spray coolant for TLL has been developed and tested. • High corrosion resistance of W and Mo in molten Sn confirmed up to 1000 °C. • Wetting process with Sn has been developed for Mo and W. - Abstract: The liquid Sn in a matrix of Capillary Porous System (CPS) has a high potential as plasma facing material in steady state operating fusion reactor owing to its physicochemical properties. However, up to now it has no experimental confirmation in tokamak conditions. First steady state operating limiter based on the CPS with liquid Sn installed on FTU tokamak and its experimental study is in progress. Several aspects of the design, structural materials and operation parameters of limiter based on tungsten CPS with liquid Sn are considered. Results of investigation of corrosion resistance of Mo and W in Sn and their wetting process are presented. The heat removal for limiter steady state operation is provided by evaporation of flowing gaswater spray. The effectiveness of such heat removal system is confirmed in modelling tests with power flux up to 5 MW/m2.

  17. Multilayer mirror based monitors for impurity controls in large fusion reactor type devices

    International Nuclear Information System (INIS)

    Regan, S.P.; May, M.J.; Soukhanovskii, V.; Finkenthal, M.; Moos, H.W.

    1995-01-01

    Multilayer Mirror (MLM) based monitors are compact, high throughput diagnostics capable of extracting XUV emissions (the wavelength range including the soft-x-ray and the extreme ultraviolet, 10 angstrom to 304 angstrom) of impurities from the harsh environment of large fusion reactor type devices. For several years the Plasma Spectroscopy Group at Johns Hopkins University has investigated the application of MLM based XUV spectroscopic diagnostics for magnetically confined fusion plasmas. MLM based monitors have been constructed for and extensively used on DIII-D, Alcator C-mod, TEXT, Phaedrus-T, and CDX-U tokamaks to study the impurity behavior of elements ranging from He to Mo. On ITER MLM based devices would be used to monitor the spectral line emissions from Li I-like to F I-like charge states of Fe, Cr, and Ni, as well as extractors for the bands of emissions from high Z elements such as Mo or W for impurity controls of the fusion plasma. In addition to monitoring the impurity emissions from the main plasma, MLM based devices can also be adapted for radiation measurements of low Z elements in the divertor. The concepts and designs of these MLM based monitors for impurity controls in ITER will be presented. The results of neutron irradiation experiments of the MLMs performed in the Los Alamos Spallation Radiation Effects Facility (LASREF) at the Los Alamos National Laboratory will also be discussed. These preliminary neutron exposure studies show that the dispersive and reflective qualities of the MLMs were not affected in a significant manner

  18. Edge transport and mode structure of a QCM-like fluctuation driven by the Shoelace antenna

    Science.gov (United States)

    Golfinopoulos, T.; LaBombard, B.; Brunner, D.; Terry, J. L.; Baek, S. G.; Ennever, P.; Edlund, E.; Han, W.; Burke, W. M.; Wolfe, S. M.; Irby, J. H.; Hughes, J. W.; Fitzgerald, E. W.; Granetz, R. S.; Greenwald, M. J.; Leccacorvi, R.; Marmar, E. S.; Pierson, S. Z.; Porkolab, M.; Vieira, R. F.; Wukitch, S. J.; The Alcator C-Mod Team

    2018-05-01

    The Shoelace antenna was built to drive edge fluctuations in the Alcator C-Mod tokamak, matching the wavenumber (k\\perp≈1.5 cm‑1) and frequency (30≲ f ≲ 200 kHz) of the quasi-coherent mode (QCM), which is responsible for regulating transport across the plasma boundary in the steady-state, ELM-free Enhanced D α (EDA) H-mode. Initial experiments in 2012 demonstrated that the antenna drove a resonant response in the edge plasma in steady-state EDA and transient, non-ELMy H-modes, but transport measurements were unavailable. In 2016, the Shoelace antenna was relocated to enable direct measurements of driven transport by a reciprocating Mirror Langmuir Probe, while also making available gas puff imaging and reflectometer data to provide additional radial localization of the driven fluctuation. This new data suggests a  ∼4 mm-wide mode layer centered on or just outside the separatrix. Fluctuations coherent with the antenna produced a radial electron flux with {Γ_e}/{n_e}∼4 m s‑1 in EDA H-mode, smaller than but comparable to the QCM level. But in transient ELM-free H-mode, {Γ_e}/{n_e} was an order of magnitude smaller, and driven fluctuations reduced by a factor of ≳ 3. The driven mode is quantitatively similar to the intrinsic QCM across measured spectral quantities, except that it is more coherent and weaker. This work informs the prospect of achieving control of edge transport by direct coupling to edge modes, as well as the use of such active coupling for diagnostic purposes.

  19. Københavns Kommunes indsats mod social dumping - målopfyldelsesevaluering

    DEFF Research Database (Denmark)

    Baadsgaard, Kelvin; Jørgensen, Henning

    2016-01-01

    Evaluering af, om de politiske intentioner med indsats mod social dumping i Københavns Kommune er blevet indfriet......Evaluering af, om de politiske intentioner med indsats mod social dumping i Københavns Kommune er blevet indfriet...

  20. Modifications of plasma edge electric field and confinement properties by limiter biasing on the KT-5C tokamak

    International Nuclear Information System (INIS)

    Hui Gao; Kan Zhai; Yizhi Wen; Shude Wan; Guiding Wang; Changxun Yu

    1995-01-01

    Experiments using a biased multiblock limiter in the KT-5C tokamak show that positive biasing is more effective than negative biasing in modifying the edge electric field, suppressing fluctuations and improving plasma confinement. The biasing effect varies with the limiter area, the toroidal magnetic field and the biasing voltage. By positive biasing, the edge profiles of the plasma potential, the electron temperature and the density become steeper, resulting in a reduced edge particle flux, an increased global particle confinement time and lower fluctuation levels of the edge plasma. (author)

  1. RELAP5/MOD3 AP600 problems

    International Nuclear Information System (INIS)

    Riemke, R.A.

    1993-01-01

    RELAP5/MOD3 is a reactor systems analysis code that has been developed jointly by the US Nuclear Regulatory Commission (USNRC) and a consortium consisting of several of the countries and domestic organizations that were members of the International Code Assessment and Applications Program (ICAP). The code is currently being used to simulate transients for the next generation of advanced light water reactors (ALWR's). One particular reactor design is the Westinghouse AP600 pressurized water reactor (PWR), which consists of two hot legs and four cold legs as well as passive emergency core cooling (ECC) systems. Initial calculations with RELAP5/MOD3 indicated that the code was not as robust as RELAP5/MOD2.5 with regard to AP600 calculations. Recent modifications in the areas of condensation wall heat transfer, interfacial heat transfer in the presence of noncondensibles, bubbly flow interfacial heat transfer, and time smoothing of both interfacial drag and interfacial heat transfer have improved the robustness, although more reliability is needed

  2. Extension of the lod score: the mod score.

    Science.gov (United States)

    Clerget-Darpoux, F

    2001-01-01

    In 1955 Morton proposed the lod score method both for testing linkage between loci and for estimating the recombination fraction between them. If a disease is controlled by a gene at one of these loci, the lod score computation requires the prior specification of an underlying model that assigns the probabilities of genotypes from the observed phenotypes. To address the case of linkage studies for diseases with unknown mode of inheritance, we suggested (Clerget-Darpoux et al., 1986) extending the lod score function to a so-called mod score function. In this function, the variables are both the recombination fraction and the disease model parameters. Maximizing the mod score function over all these parameters amounts to maximizing the probability of marker data conditional on the disease status. Under the absence of linkage, the mod score conforms to a chi-square distribution, with extra degrees of freedom in comparison to the lod score function (MacLean et al., 1993). The mod score is asymptotically maximum for the true disease model (Clerget-Darpoux and Bonaïti-Pellié, 1992; Hodge and Elston, 1994). Consequently, the power to detect linkage through mod score will be highest when the space of models where the maximization is performed includes the true model. On the other hand, one must avoid overparametrization of the model space. For example, when the approach is applied to affected sibpairs, only two constrained disease model parameters should be used (Knapp et al., 1994) for the mod score maximization. It is also important to emphasize the existence of a strong correlation between the disease gene location and the disease model. Consequently, there is poor resolution of the location of the susceptibility locus when the disease model at this locus is unknown. Of course, this is true regardless of the statistics used. The mod score may also be applied in a candidate gene strategy to model the potential effect of this gene in the disease. Since, however, it

  3. Disruption Studies in JT-60U

    International Nuclear Information System (INIS)

    Kawano, Y.; Yoshino, R.; Neyatani, Y.; Nakamura, Y.; Tokuda, S.; Tamai, H.

    2002-01-01

    Intensive studies on the physics of disruptions and developments of avoidance/mitigation methods of disruption-related phenomena have being carried out in JT-60U. The characteristics of the disruption sequence were well understood from the observation of the relationship between the heat pulse onto divertor plates during thermal quench and the impurity influx into the plasma, which determined the speed of the following current quench. A fast shutdown was first demonstrated by injecting impurity ice pellets to the plasma and intensively reducing the heat flux on first wall. The halo current and its toroidal asymmetry were precisely measured, and the halo current database was made for ITER in a wide parameter range. It was found that TPF x I h /I p0 was 0.52 at the maximum in a large tokamak like the JT-60U, whereas the higher factor of 0.75 had been observed in medium-sized tokamaks such as Alcator C-Mod and ASDEX-Upgrade. The vertical displacement event (VDE) at the start of the current quench was carefully investigated, and the neutral point where the VDE hardly occurs was discovered. MHD simulations clarified the onset mechanisms of the VDE, in which the eddy current effect of the up-down asymmetric resistive shell was essential. The real-time Z j measurement was improved for avoiding VDEs during slow current quench, and plasma-wall interaction was avoided by a well-optimized plasma equilibrium control. Magnetic fluctuations that were spontaneously generated at the disruption and/or enhanced by the externally applied helical field have been shown to avoid the generation of runaway electrons. Numerical analysis clarified an adequate rate of collisionless loss of runaway electrons in turbulent magnetic fields, which was consistent with the avoidance of runaway electron generation by magnetic fluctuations observed in JT-60U. Once generated, runaway electrons were suppressed when the safety factor at the plasma surface was reduced to 3 or 2

  4. Development of long-pulse heating & current drive actuators & operational techniques compatible with a high-Z divertor & first wall

    Energy Technology Data Exchange (ETDEWEB)

    Tynan, George [Univ. of California, San Diego, CA (United States)

    2018-01-09

    This was a collaboration between UCSD and MIT to study the effective application of ion-cyclotron heating (ICRH) on the EAST tokamak, located in China. The original goal was for UCSD to develop a diagnostic that would allow measurement of the steady state, or DC, convection pattern that develops on magnetic field lines that attach or connect to the ICRH antenna. This diagnostic would then be used to develop techniques and approaches that minimize or even eliminate such DC convection during application of strong ICRH heating. This was thought to then indicate reduction or elimination of parasitic losses of heating power, and thus be an indicator of effective RF heating. The original plan to use high speed digital gas-puff imaging (GPI) of the antenna-edge plasma region in EAST was ultimately unsuccessful due to limitations in machine and camera operations. We then decided to attempt the same experiment on the ALCATOR C-MOD tokamak at MIT which had a similar instrument already installed. This effort was ultimately successful, and demonstrated that the underlying idea of using GPI as a diagnostic for ICRH antenna physics would, in fact, work. The two-dimensional velocity fields of the turbulent structures, which are advected by RF-induced E x B flows, are obtained via the time-delay estimation (TDE) techniques. Both the magnitude and radial extension of the radial electric field E-r were observed to increase with the toroidal magnetic field strength B and the ICRF power. The TDE estimations of RF-induced plasma potentials are consistent with previous results based on the probe measurements of poloidal phase velocity. The results suggest that effective ICRH heating with reduced impurity production is possible when the antenna/box system is designed so as to reduce the RF-induced image currents that flow in the grounded conducting antenna frame elements that surround the RF antenna current straps.

  5. Annual status report 31 December 1975

    International Nuclear Information System (INIS)

    1976-06-01

    The desired current distribution in the high-β experiment SPICA can be maintained on a 100 μs time-scale. In the turbulent heating experiment, a plasma with parameters not unlike those in current tokamak experiments was produced in a much shorter time-scale. Gas-blanket research was carried out both in the Institute's Ringboog facility and in Alcator at MIT. A comparison of the achievements of these experiments underlines the advantage of the Alcator scheme. Work with Ringboog now aims at understanding the mechanisms responsible for the energy loss (both radiation and instability being significant) and at controlling the current-density profile by means of turbulent heating of a skin. The emphasis of the work in the theoretical group has been in the fields of magnetohydrodynamics and transport theory. In particular, the problem of the β-limits for stable equilibria in tokamak and screw-pinch configurations was investigated. It was confirmed that, from this point of view, a screw pinch has more favourable properties than a tokamak. Studies into the effect of non-circular cross-sections were started. As regards transport phenomena due to binary collisions, work was initiated to complete the picture for different possible regimes. The studies on transport due to impurity-driven dissipative drift modes were continued. The investigation of the wave dynamics in an homogeneous plasma was extended to include quasi-linear effects. Research was also started to evaluate the potentialities of using advanced fuels in fusion reactors

  6. Magnet power system for the Microwave Tokamak Experiment (MTX)

    International Nuclear Information System (INIS)

    Jackson, M.C.; Musslewhite, R.C.

    1987-01-01

    The system configuration, layout, and general philosophy for the MTX magnet power system is described. The vast majority of the magnet power equipment was quite successfully used on the ALCATOR-C experiment at the Massachusetts Institute of Technology. The AC power for the magnet system at MIT was obtained from a 225MVA alternator. The power for the system at LLNL is obtained directly from the local utility's 230 kV line. This installation, therefore, necessitates the addition of a great deal of equipment in ranges from new switchgear in the substation to using existing switchgear obtained from MIT as contractors for intershop electrical isolation as well as safety isolation for personnel entry into the experimental area. Additionally, some discussion is made of the unique layout of this facility and the tradeoffs made to accommodate them. 2 refs., 6 figs

  7. Progress of the ECH·ECCD experiments. Research progress of the ECH·ECCD experiments in tokamaks and spherical tokamaks

    International Nuclear Information System (INIS)

    Isayama, Akihiko; Tanaka, Hitoshi

    2009-01-01

    Recent progress in the ECH·ECCD study in tokamak and spherical tokamak devices is described. As for the tokamak study, results on the control of neoclassical tearing modes and sawtooth oscillations, the current profile, the internal transport barrier, the plasma start-up and the discharge cleaning are given. As for the spherical tokamak study, the plasma start-up by ECH·ECCD and the electron-Bernstein-wave heating and the current drive are described. (T.I.)

  8. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  9. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  10. Vaccine mod halthed testes i besætning

    DEFF Research Database (Denmark)

    Lauritsen, Klara Tølbøll

    2012-01-01

    Ny vaccine mod ledbetændelse forårsaget af Mycoplasma hyosynoviae testes nu hos 200 svin i en problembesætning. Håbet er færre halte svin og en nedbringelse af antibiotikaforbruget.......Ny vaccine mod ledbetændelse forårsaget af Mycoplasma hyosynoviae testes nu hos 200 svin i en problembesætning. Håbet er færre halte svin og en nedbringelse af antibiotikaforbruget....

  11. Analysis on Θ pumping for tokamak current drive

    International Nuclear Information System (INIS)

    Miyamoto, Kenro; Naito, Osamu

    1986-01-01

    Analytical results of Θ pumping for the tokamak current drive are presented. Diffusion of externally applied oscillating electric field into the tokamak plasma is examined when the plasma is normal. When the oscillating electric field is parallel to the stationary toroidal plasma current and the induced current density by the applied electric field becomes larger than the average density of the toroidal plasma current over the plasma cross section, the radial profile of the safety factor has the extremum near the plasma boundary region and MHD instabilities are excited. It is assumed that anomalous diffusion of the induced current localized in the plasma boundary region takes place, so that the extreme value in the radial profile of the safety factor disappears. The anomalously diffused electric field due to this relaxation process has net d. c component and its non-zero value of the time average is estimated. Then the condition of the tokamak current drive by Θ pumping is derived. Some numerical results are presented for an example of a fusion grade plasma. (author)

  12. Experimental measurements of the ion cyclotron antennas' coupling and rf characteristics

    International Nuclear Information System (INIS)

    Hoffman, D.J.; Baity, F.W.; Becraft, W.R.; Caughman, J.B.O.; Owens, T.L.

    1985-01-01

    The rf coupling capabilities and characteristics of various antennas have been measured. The tested antenna configurations include the simple loop antenna operated at resonant lengths as used on Alcator-C, the cavity antenna proposed for Doublet III-D and the resonant double loop, asymmetric resonant double loop, and U-slot antennas. Models of the voltage, magnetic fields outside the structure, and current have been correlated with the measurements made on these antennas. From these measurements and from typical observations of ICRH coupling in tokamaks, we are studying power and frequency limitations on each antenna and the causes of the limitations. A comparison of the technology, performance, and power limitations of each type of antenna is presented

  13. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  14. Functional overview of the Production Planning Model (ProdMod)

    International Nuclear Information System (INIS)

    Gregory, M.V.; Paul, P.K.

    1995-09-01

    The Production Planning Model (ProdMod) has been developed by SRTC for use by High Level Waste Program Management and High Level Waste Engineering as a fast running, integrated, comprehensive model of the entire SRS high level waste (HLW) complex. ProdMod can simulate the response of the HLW complex from its current state to the end of tank clean-up or to any intermediate point. The present document describes the initial release of ProdMod at the end of FY95: a model version that contains all the significant elements from the High-level Waste System Plan Revision 5 and is capable of running the simulation all the way to the postulated completion of waste removal. For the scenario represented by this release, that simulates approximately 70 years of operation of the HLW complex (out to FY2065). This initial release of ProdMod will serve as the immediate starting point for the modeling of the High-Level Waste System Plan Revision 6. Thus ProdMod is expected to be in a state of continuous change and improvement.the initial goal has been to generate a simulation of the processes of interest, with the emphasis on mass and volume balances tracked throughout the HLW complex. That has been accomplished. Future development will add a set of cost equations to the process equations and extend the model for use as a linear programming (optimization) application. The goal of this later phase will be to free the ProdMod user to some extent from the need to set up detailed simulation scenarios: the model will automatically make operational choices which minimize or maximize a given objective function. Appendix A contains the source code

  15. Comparision of calculations for the ROSA-IV LSTF with RELAP5/MOD0 and RELAP5/MOD1 (cycle 1)

    International Nuclear Information System (INIS)

    Fineman, C.P.; Tanaka, Mitsugu; Tasaka, Kanji

    1982-03-01

    10% and 2.5% cold leg break analyses have been completed for the ROSA-IV Large Scale Test Facility (LSTF) with the RELAP5/MOD0 and RELAP5/MOD1, cycle 1, computer codes. Comparisons between the calculations were made to determine any differences in the results obtained from the two versions of RELAP5. Differences in the two calculations were found which can be attributed to changes in the flow regime maps and critical flow model. (author)

  16. Modeling of Nonlinear Beat Signals of TAE's

    Science.gov (United States)

    Zhang, Bo; Berk, Herbert; Breizman, Boris; Zheng, Linjin

    2012-03-01

    Experiments on Alcator C-Mod reveal Toroidal Alfven Eigenmodes (TAE) together with signals at various beat frequencies, including those at twice the mode frequency. The beat frequencies are sidebands driven by quadratic nonlinear terms in the MHD equations. These nonlinear sidebands have not yet been quantified by any existing codes. We extend the AEGIS code to capture nonlinear effects by treating the nonlinear terms as a driving source in the linear MHD solver. Our goal is to compute the spatial structure of the sidebands for realistic geometry and q-profile, which can be directly compared with experiment in order to interpret the phase contrast imaging diagnostic measurements and to enable the quantitative determination of the Alfven wave amplitude in the plasma core

  17. Observations of a quasi-coherent fluctuation mode in the KT-5C tokamak during -90 deg. phase shift feedback

    International Nuclear Information System (INIS)

    Zhai Kan; Wen Yizhi; Yu Changxuan; Liu Wandong; Wan Shude; Zhuang Ge; Yu Wen; Xu Zhizhan

    1997-01-01

    A new fluctuation phenomenon is observed through Langmuir probe measurements at the edge plasma in the KT-5C tokamak by applying a -90 deg. phase shift feedback. Using a two point correlation technique, it is found that this fluctuation mode has a longer poloidal wavelength and a definite frequency when compared with the usual edge turbulence. It is also found through bispectral analysis that this mode is a spontaneously excited quasi-coherent mode, which has almost no contribution to the cross-field particle flux. (author)

  18. Modding a free and open source software video game: "Play testing is hard work"

    Directory of Open Access Journals (Sweden)

    Giacomo Poderi

    2014-03-01

    Full Text Available Video game modding is a form of fan productivity in contemporary participatory culture. We see modding as an important way in which modders experience and conceptualize their work. By focusing on modding in a free and open source software video game, we analyze the practice of modding and the way it changes modders' relationship with their object of interest. The modders' involvement is not always associated with fun and creativity. Indeed, activities such as play testing often undermine these dimensions of modding. We present a case study of modding that is based on ethnographic research done for The Battle for Wesnoth, a free and open source software strategy video game entirely developed by a community of volunteers.

  19. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

  20. Ny vaccine mod ledbetændelse er ikke effektiv

    DEFF Research Database (Denmark)

    Nielsen, Elisabeth Okholm; Lauritsen, Klara Tølbøll

    2013-01-01

    En ny mulighed for at vaccinere mod mykoplasma-ledbetændelse er undersøgt hos en slagtesvineproducent. Vaccinen kunne desværre ikke forebygge halthed eff ektivt.......En ny mulighed for at vaccinere mod mykoplasma-ledbetændelse er undersøgt hos en slagtesvineproducent. Vaccinen kunne desværre ikke forebygge halthed eff ektivt....

  1. Development of wide area reaction system for Reel-to-Reel TFA-MOD process

    International Nuclear Information System (INIS)

    Nomoto, Sukeharu; Aoki, Yuji; Teranishi, Ryo; Sato, Akihiro; Izumi, Teruo; Shiohara, Yuh

    2006-01-01

    The previously developed numerical simulation method for the TFA-MOD process, which calculated the YBCO growth kinetics, gas element diffusion and gas flow, was applied to study the suitable gas flow mode for a multi-turning Reel-to-Reel tape conveyance system of a long YBCO coated conductors. The high YBCO production rate with uniform J c distribution among tape lines is desired in the system. It was found by the numerical simulation for the vertical gas flow onto the tape surface to realize the above demands even in a wider reaction area. We developed a new wide area reaction tube for the Reel-to-Reel TFA-MOD process according to the numerically designed gas flow configuration. The demand for the new tube was confirmed to be satisfied by experiments

  2. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  3. GeoMod 2014 - Modelling in geoscience

    Science.gov (United States)

    Leever, Karen; Oncken, Onno

    2016-08-01

    GeoMod is a biennial conference to review and discuss latest developments in analogue and numerical modelling of lithospheric and mantle deformation. GeoMod2014 took place at the GFZ German Research Centre for Geosciences in Potsdam, Germany. Its focus was on rheology and deformation at a wide range of temporal and spatial scales: from earthquakes to long-term deformation, from micro-structures to orogens and subduction systems. It also addressed volcanotectonics and the interaction between tectonics and surface processes (Elger et al., 2014). The conference was followed by a 2-day short course on "Constitutive Laws: from Observation to Implementation in Models" and a 1-day hands-on tutorial on the ASPECT numerical modelling software.

  4. Topology of tokamak orbits

    International Nuclear Information System (INIS)

    Rome, J.A.; Peng, Y.K.M.

    1978-09-01

    Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well

  5. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

    This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

  6. Assessment and improvement of condensation models in RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ki Yong; Park, Hyun Sik; Kim, Sang Jae; No, Hee Chen [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    The condensation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condensation phenomena. The default model of the laminar film condensation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data. Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data. The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP6/MOD3.2. The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to consider the effects of the gas velocity and the film thickness. 2 refs., 5 figs., 1 tab. (Author)

  7. Assessment and improvement of condensation models in RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ki Yong; Park, Hyun Sik; Kim, Sang Jae; No, Hee Chen [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    The condensation models in the standard RELAP5/MOD3.2 code are assessed and improved based on the database, which is constructed from the previous experimental data on various condensation phenomena. The default model of the laminar film condensation in RELAP5/MOD3.2 does not give any reliable predictions, and its alternative model always predicts higher values than the experimental data. Therefore, it is needed to develop a new correlation based on the experimental data of various operating ranges in the constructed database. The Shah correlation, which is used to calculate the turbulent film condensation heat transfer coefficients in the standard RELAP5/MOD3.2, well predicts the experimental data in the database. The horizontally stratified condensation model of RELAP5/MOD3.2 overpredicts both cocurrent and countercurrent experimental data. The correlation proposed by H.J.Kim predicts the database relatively well compared with that of RELAP6/MOD3.2. The RELAP5/MOD3.2 model should use the liquid velocity for the calculation of the liquid Reynolds number and be modified to consider the effects of the gas velocity and the film thickness. 2 refs., 5 figs., 1 tab. (Author)

  8. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

  9. START: the creation of a spherical tokamak

    International Nuclear Information System (INIS)

    Sykes, Alan

    1992-01-01

    The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)

  10. Tokamaks (Second Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Stott, Peter [JET, UK (United Kingdom)

    1998-10-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  11. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  12. Using computer program RELAP5/MOD2 on microcomputers

    International Nuclear Information System (INIS)

    Grgic, D.; Bajs, T; Cavlina, N.; Debrecin, N.

    1990-01-01

    Our work on installation of RELAP5/MOD2 code on IBM4341, mVAX 11, MGT-386 and COMPAQ-386/20e computers is described. Main characteristics of RELAP5/MOD2 structure programming style and differences between FORTRAN VS, VAX-11 FORTRAN and NDP FORTRAN 386 are presented. We discussed basic philosophy used in modification and testing and test results. (author)

  13. Characterization of the Hamamatsu 8" R5912-MOD Photomultiplier tube

    Science.gov (United States)

    Kaptanoglu, Tanner

    2018-05-01

    Current and future neutrino and direct detection dark matter experiments hope to take advantage of improving technologies in photon detection. Many of these detectors are large, monolithic optical detectors that use relatively low-cost, large-area, and efficient photomultiplier tubes (PMTs). A candidate PMT for future experiments is a newly developed prototype Hamamatsu PMT, the R5912-MOD. In this paper we describe measurements made of the single photoelectron time and charge response of the R5912-MOD, as well as detail some direct comparisons to similar PMTs. Most of these measurements were performed on three R5912-MOD PMTs operating at gains close to 1 × 107. The transit time spread (σ) and the charge peak-to-valley were measured to be on average 680ps and 4.2 respectively. The results of this paper show the R5912-MOD is an excellent candidate for future experiments in several regards, particularly due to its narrow spread in timing.

  14. Brazilian Irradiation Project: CAFE-MOD1 validation experimental program

    International Nuclear Information System (INIS)

    Mattos, Joao Roberto Loureiro de; Costa, Antonio Carlos L. da; Esteves, Fernando Avelar; Dias, Marcio Soares

    1999-01-01

    The Brazilian Irradiation Project whose purpose is to provide Brazil with a minimal structure to qualify the design, fabrication and quality procedures of nuclear fuels, consists of three main facilities: IEA-R1 reactor of IPEN-CNEN/SP, CAFE-MOD1 irradiation device and a unit of hot cells. The CAFE-MOD1 is based on concepts successfully used for more than 20 years in the main nuclear institutes around the world. Despite these concepts are already proved it should be adapted to each reactor condition. For this purpose, there is an ongoing experimental program aiming at the certification of the criteria and operational limits of the CAFE-MOD1 in order to get the allowance for its installation at the IEA-R1 reactor. (author)

  15. 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

  16. RELAP5/MOD2: for PWR transient analysis

    International Nuclear Information System (INIS)

    Ransom, V.H.

    1983-01-01

    RELAP5 is a light water reactor system transient simulation code for use in nuclear plant safety analysis. Development of a new version, RELAP5/MOD2, has been completed and will be released to the United States Nuclear Regulatory Commission during September of 1983. The new and improved modeling capability of RELAP5/MOD2 is described and some developmental assessment results are presented. The future plans for extension to severe accident modeling are briefly discussed

  17. Modelling dust transport in tokamaks

    International Nuclear Information System (INIS)

    Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

    2008-01-01

    The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

  18. Compact tokamak reactors

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1997-01-01

    The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

  19. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1994-01-01

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  20. L'utilisation d'un modèle spécifique: l'exemple de la Bolivie

    Directory of Open Access Journals (Sweden)

    1996-01-01

    Full Text Available La modélisation graphique connaît aujourd’hui un essor certain depuis la publication de la Géographie Universelle sous la direction de R. Brunet en France. Les utilisations en sont diverses, qu’elles aient des fins pédagogiques, expérimentales ou prospectives, mais il n’en demeure pas moins qu’à la différence du langage mathématique, la pratique du langage graphique constitue un handicap lorsqu’il s’agit de valider ce type de modèles. Pourtant cette démarche est non seulement possible mais souhaitable, un modèle non validable n’étant plus un modèle. Cependant les tests de validation/vérification et de validation/exploration utilisés pour analyser le modèle de l’espace bolivien sont globalement satisfaisants. LA UTILIZACIÓN DE UN MODELO ESPECÍFICO: EL EJEMPLO DE BOLIVIA. La modelización gráfica tiene en Francia un verdadero desarrollo con la publicación de la Geografía Universal dirigida por R. Brunet. Las utilizaciones son variadas: pedagógicas, experimentales o de investigación pero la práctica del código gráfico es mucho más problemático que la práctica del código matemático cuando se necesita validar un modelo. Sin embargo las pruebas de validación/comprobación y de validación/investigación utilizadas para analizar el modelo del espacio boliviano son bastante satisfactorias. USING SPECIFIC MODELIZATION IN THE CASE OF BOLIVIA. The use of grafic modelization has grown to be acknowledged in France since the publishing of Géographie Universelle edited by R. Brunet. However, in spite of its various applications, whether pedagogical, experimental or prospective, evidence is there to show that graphic models do not as yet compare with mathematical models as regards the validating process. Nonetheless, the two-step testing based on validation/verification and validation/ exploration carried out here on a Bolivian space model does give quite satisfying results.