Science.gov

Sample records for alcator c-mod diii-d

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

    SciTech Connect

    Pankin, A. Y.; Rafiq, T.; Kritz, A. H.; Park, G. Y.; Chang, C. S.; Ku, S.; Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L.; Groebner, R. J.

    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.

  2. Non-axisymmetric Field Effects on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Wolfe, S.; Hutchinson, I.; Granetz, R.; Rice, J.; Hubbard, A.; Irby, J.; Vieira, R.; Cochran, W.; Gwinn, D.; Rosati, J.; Lynn, A.

    2003-10-01

    A set of coils capable of producing non-axisymmetric, predominantly n=1, fields with different toroidal phase and a range of poloidal mode (m) spectra has been installed on Alcator C-Mod. This coilset has been used to suppress locked modes during low density or high current operation and also to induce locked modes in normally stable configurations in order to study error field effects. Locked modes are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement. The inferred value of the threshold perturbation for producing a locked mode is of order B_21/B_T ˜ 10-4, where B_21 is the helically resonant m/n=2/1 field evaluated at the q=2 surface. This value is comparable to extrapolations based on experiments on JET and DIII-D, but is inconsistent with stronger BT and size scaling inferred from Compass-D results(R. J. Buttery, et al., 17th Fusion Energy Conference, Oct. 1998, Yokohama (IAEA-CN-69) EX8/5). The C-Mod result therefore has favorable implications for the locked mode threshold in ITER.

  3. Axisymmetric Control in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Tinios, Gerasimos

    1995-01-01

    This thesis investigates the degree to which linear axisymmetric modeling of the response of a tokamak plasma can reproduce observed experimental behavior. The emphasis is on the vertical instability. The motivation for this work lies in the fact that, once dependable models have been developed, modern control theory methods can be used to design feedback laws for more effective and efficient tokamak control. The models are tested against experimental data from the Alcator C-Mod tokamak. A linear model for each subsystem of the closed-loop system constituting an Alcator C-Mod discharge under feedback control has been constructed. A non-rigid, approximately flux-conserving, perturbed equilibrium plasma response model is used in the comparison to experiment. A detailed toroidally symmetric model of the vacuum vessel and the supporting superstructure is used. Modeling of the power supplies feeding the active coils has been included. Experiments have been conducted with vertically unstable plasmas where the feedback was turned off and the plasma response was observed in an open -loop configuration. The closed-loop behavior has been examined by injecting step perturbations into the desired vertical position of the plasma. The agreement between theory and experiment in the open-loop configuration was very satisfactory, proving that the perturbed equilibrium plasma response model and a toroidally symmetric electromagnetic model of the vacuum vessel and the structure can be trusted for the purpose of calculations for control law design. When the power supplies and the feedback computer hardware are added to the system, however, as they are in the closed-loop configuration, they introduce nonlinearities that make it difficult to explain observed behavior with linear theory. Nonlinear simulation of the time evolution of the closed-loop experiments was able to account for the discrepancies between linear theory and experiment. (Copies available exclusively from MIT Libraries

  4. SOL Reflectometer for Alcator C-Mod

    SciTech Connect

    Hanson, Gregory R; Wilgen, John B; Wukitch, Dr. Steve; Lin, Dr. Yijun; Lau, Cornwall H; Wallace, Gregory M

    2008-01-01

    A two-frequency x-mode reflectometer operating from 100 146 GHz is being deployed on Alcator C-Mod to measure the density profile and fluctuations in the scrape-off layer (SOL) immediately in-front of the new J-port ICRF antenna and the new B-port Lower Hybrid launcher. The reflectometer will cover densities from 1016 to 1020 m-3 at 5 5.4 T. To provide the greatest flexibility and capability to deal with density fluctuations approaching 100% peak to peak in the SOL, both full-phase and differential-phase measurement capabilities with sweep speeds of ~10 s to >1 ms are being implemented. The differential-phase measurement will use a difference-frequency of 500 MHz, corresponding to cutoff layer separations ranging from about 0.1 mm to 1 mm. The reflectometer will have 6 sets of launchers: 3 on the J-port ICRF antenna and 3 on the B-port LHRF launcher. The ICRF and LHRF antennas will incorporate reflectometer antennas at their top, bottom and mid-plane locations.

  5. SOL Reflectometer for Alcator C-MOD

    SciTech Connect

    Lau, Cornwall H; Hanson, Gregory R; Wilgen, John B; Lin, Dr. Yijun; Wukitch, Dr. 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 {micro}s to 1 ms, and will cover a density range of approximately 10{sup 16}-10{sup 20} m{sup -3} at B{sub 0} = 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 (TE{sub 01}) 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.

  6. Disruptions and halo currents in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Granetz, R. S.; Hutchinson, I. H.; Sorci, J.; Irby, J. H.; La Bombard, B.; Gwinn, D.

    1996-05-01

    Disruptions in Alcator C-Mod can generate large eddy currents in the highly conducting vacuum vessel and internal structures, including a significant poloidal component due to halo currents. In order to understand better the stresses arising from the resulting J*B forces, Alcator C-Mod has been fitted with a comprehensive set of sensors to measure the spatial distribution and temporal behaviour of the halo currents. It is found that they are toroidally asymmetric, with a typical peaking factor of 2. The asymmetric pattern usually rotates toroidally at a few kilohertz, thus ruling out first wall non-uniformities as the cause of the asymmetry. Analysis of the information compiled in the C-Mod disruption database indicates that the maximum halo current during a disruption scales roughly as either Ip2/Bphi or Ip/q95, but that there is a large amount of variation that is not yet understood

  7. Divertor bypass in the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Pitcher, C. S.; LaBombard, B.; Danforth, R.; Pina, W.; Silveira, M.; Parkin, B.

    2001-01-01

    The Alcator C-Mod divertor bypass has for the first time allowed in situ variations to the mechanical baffle design in a tokamak. The design utilizes small coils which interact with the ambient magnetic field inside the vessel to provide the torque required to control small flaps of a Venetian blind geometry. Plasma physics experiments with the bypass have revealed the importance of the divertor baffling to maintain high divertor gas pressures. These experiments have also indicated that the divertor baffling has only a limited effect on the main chamber pressure in C-Mod.

  8. Analysis tools for turbulence studies at Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Burns, C.; Shehata, S.; White, A. E.; Cziegler, I.; Dominguez, A.; Terry, J. L.; Pace, D. C.

    2010-11-01

    A new suite of analysis tools written in IDL is being developed to support experimental investigation of turbulence at Alcator C-Mod. The tools include GUIs for spectral analysis (coherence, cross-phase and bicoherence) and characteristic frequency calculations. A user-friendly interface for the GENRAY code, to facilitate in-between shot ray-tracing analysis, is also being developed. The spectral analysis tool is being used to analyze data from existing edge turbulence diagnostics, such as the O-mode correlation reflectometer and Gas Puff Imaging, during I-mode, ITB and EDA H-mode plasmas. GENRAY and the characteristic frequency tool are being used to study diagnostic accessibility limits set by wave propagation and refraction for X-mode Doppler Backscattering and Correlation Electron Cyclotron Emission (CECE) systems that are being planned for core turbulence studies at Alcator C-Mod.

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

    SciTech Connect

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

    2004-08-10

    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

  10. Overview of Alcator C-Mod Research

    NASA Astrophysics Data System (ADS)

    White, Anne

    2015-11-01

    Research on C-Mod supports next-step-devices: RF heating, current and flow drive, divertor/PMI physics, non-ELMing regimes with enhanced confinement, and disruption mitigation/runaway dynamics. Disruption mitigation experiments in MHD-unstable plasmas show MGI works equally well with and without locked modes. The L-I-mode threshold is found to be independent of magnetic field, opening an expanded operating range at high field. The toroidal and radial structure of power deposition of RF waves into the edge plasma has been systematically quantified, through the use of a unique set of fast time resolution edge diagnostics. Progress in understanding multi-channel core transport has been significant. Full-physics, ITG/TEM/ETG gyrokinetic simulations show that nonlinear cross-scale coupling enhances both ion and electron heat flux to match experiments, explaining the origin of electron heat flux and stiffness. Dynamic, passive measurements of the core rotation velocity profiles with X-ray imaging crystal spectroscopy show the direction of intrinsic rotation reversals depends on central safety factor, not on the magnetic shear. Design studies for ADX and SPARC are establishing the engineering, economics and physics for a fusion energy development path leveraging new superconducting magnet technologies. This work is supported by the US DOE under DE- FC02-99ER54512-CMOD.

  11. Turbulent impurity transport modeling for Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Fu, X. R.; Horton, W.; Bespamyatnov, I. O.; Rowan, W. L.; Benkadda, S.; Fiore, C. L.; Futatani, S.; Liao, K. T.; Liao

    2013-10-01

    Turbulent particle transport is investigated with a quasilinear theory that is motivated by the boron impurity transport experiments in the Alcator C-Mod. Eigenvalue problems for sets of reduced fluid equations for multi-component plasmas are solved for the self-consistent fluctuating field vectors composed of the electric potential φ, the main ion density δni , the impurity density δnz and the ion temperature fluctuation δTi . For Alcator C-Mod parameters, we investigate two drift wave models: (1) the density-gradient-driven impurity drift wave and (2) the ion-temperature-gradient-driven ion temperature gradient (ITG) mode. Analytic and numerical results for particle transport coefficients are derived and compared with the transport data and the neoclassical theory. We explore the ability of the model to explain impurity density profiles in three confinement regimes: H-mode, I-mode and internal transport barrier (ITB) regime in C-Mod. Related experiments reported on the Large Helical Device are briefly discussed.

  12. Divertor IR thermography on Alcator C-Mod.

    PubMed

    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. PMID:21034041

  13. Divertor IR thermography on Alcator C-Mod

    SciTech Connect

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

    2010-10-15

    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 deg. toroidal sector has been given a 2 deg. 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.

  14. Benchmarking Nonlinear Turbulence Simulations on Alcator C-Mod

    SciTech Connect

    M.H. Redi; C.L. Fiore; W. Dorland; M.J. Greenwald; G.W. Hammett; K. Hill; D. McCune; D.R. Mikkelsen; G. Rewoldt; J.E. Rice

    2004-06-22

    Linear simulations of plasma microturbulence are used with recent radial profiles of toroidal velocity from similar plasmas to consider nonlinear microturbulence simulations and observed transport analysis on Alcator C-Mod. We focus on internal transport barrier (ITB) formation in fully equilibrated H-mode plasmas with nearly flat velocity profiles. Velocity profile data, transport analysis and linear growth rates are combined to integrate data and simulation, and explore the effects of toroidal velocity on benchmarking simulations. Areas of interest for future nonlinear simulations are identified. A good gyrokinetic benchmark is found in the plasma core, without extensive nonlinear simulations. RF-heated C-Mod H-mode experiments, which exhibit an ITB, have been studied with the massively parallel code GS2 towards validation of gyrokinetic microturbulence models. New, linear, gyrokinetic calculations are reported and discussed in connection with transport analysis near the ITB trigger time of shot No.1001220016.

  15. Extension of Alcator C-Mod's ICRF experimental capability

    SciTech Connect

    Schilling, G.; Hosea, J. C.; Wilson, J. R.; Bonoli, P. T.; Lee, W. D.; Nelson-Melby, E.; Porkolab, M.; Wukitch, S. J.

    1999-09-20

    A new 4-strap single-ended ICRF antenna has been added to the Alcator C-Mod tokamak. PPPL designed, fabricated, and tested the antenna up to 45 kV on an rf test stand. It is capable of symmetric phasing for ICRF heating studies, and asymmetric phasing with an improved directed wave spectrum for current drive. Two new 2 MW transmitters, tunable from 40-80 MHz, allow operation in plasma at 43, 60, and 78 MHz to match a variety of toroidal fields and plasma conditions. This addition increases the total available ICRF power to 4 MW at 80 MHz plus 4 MW at 40-80 MHz. Plasma heating and current drive experiments at the extended power levels and new frequencies are planned, and initial system performance will be discussed. (c) 1999 American Institute of Physics.

  16. Extension of Alcator C-mod's ICRF Experimental Capability

    SciTech Connect

    Schilling, G.; Hosea, J.C.; Wilson, J.R.; Bonoli, P.T.; Lee, W.D.

    1999-06-01

    A new 4-strap single-ended ICRF antenna has been added to the Alcator C-Mod tokamak. PPPL designed, fabricated, and tested the antenna up to 45 kV on an rf test stand. It is capable of symmetric phasing for ICRF heating studies, and asymmetric phasing with an improved directed wave spectrum for current drive. Two new 2 MW transmitters, tunable from 40-80 MHz, allow operation in plasma at 43, 60, and 78 MHz to match a variety of toroidal fields and plasma conditions. This addition increases the total available ICRF power to 4 MW at 80 MHz plus 4 MW at 40-80 MHz. Plasma heating and current drive experiments at the extended power levels and new frequencies are planned, and initial system performance will be discussed.

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

    SciTech Connect

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

    2012-12-15

    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.

  18. Edge Minority Heating Experiment in Alcator C-Mod

    SciTech Connect

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

    2005-03-25

    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 radio-frequency power is less than or equal to 5.5 MW. A preliminary analysis of these results is presented and some ideas for improvement are discussed.

  19. Lower Hybrid Coupling Experiments on Alcator C-Mod

    SciTech Connect

    Wallace, G.; Bonoli, P.; Parisot, A.; Parker, R.; Schmidt, A.; Wilson, J. R.

    2007-09-28

    The Alcator C-Mod Lower Hybrid Current Drive experiment launches RF waves at 4.6 GHz via 4 rows of 22 phased waveguides. Forward and reflected power is measured with 156 directional couplers in the launcher structure. Langmuir probes mounted to the front of the antenna monitor density at the plasma edge and act as RF probes for the observation of parametric decay instability. Measurements of the coupling of lower hybrid waves have been performed at power levels approaching 1 MW. Edge density, launched n{sub parallel} spectrum, and plasma shape have been adjusted to optimize coupling in L-mode plasmas. Experimentally observed coupling results will be compared to simulations from the Brambilla Grill code [1].

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

    PubMed

    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 P(rad) of nearly 50% by the diodes compared to P(rad) determined using resistive bolometers. PMID:19044619

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

    SciTech Connect

    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.

  2. Overview of Recent Alcator C-Mod Results

    NASA Astrophysics Data System (ADS)

    Marmar, Earl; Alcator C-Mod Team

    2014-10-01

    Alcator C-Mod research currently emphasizes RF heating, current and flow drive, divertor/PMI issues, non-ELMing pedestal regimes with enhanced confinement, and disruption mitigation/runaway dynamics. Stability analysis of I-mode pedestals shows pressures well below the peeling-ballooning limit, as well as expected kinetic ballooning mode thresholds, consistent with the lack of ELMs. Results with the magnetic field aligned ICRF antenna show reductions in high-Z metallic impurities. Implementation of novel ``mirror-probe'' electronics has enabled simultaneous measurements of Te, ne and φ with 1 μs time response using a single probe tip, revealing important properties of the Quasi-Coherent-Mode (QCM) which regulates edge particle transport in EDA H-mode. An Accelerator-based In-situ Material Surveillance diagnostic has been deployed, providing the first between-shot measurements of surface evolution of the all-metal wall. We have observed suppression of boundary turbulence and τE improvement using LHRF into high-density H-modes, with H-factor increases up to 30%. Upgrades which are ready for construction and near term installation on C-Mod include: an off-midplane LH launcher to test theories of improved current drive at high density and an actively heated (900 K) tungsten DEMO-like outer divertor. We are proposing a new facility, ADX, based on Alcator technology, to access advanced magnetic topologies to solve the divertor PMI problem, combined with high-field launch LHCD and ICRF to extend the tokamak to steady-state with reactor relevant tools. Supported by USDOE.

  3. Correlation ECE diagnostic in Alcator C-Mod

    DOE PAGESBeta

    Sung, C.; White, A. E.; Howard, N. T.; Mikkelsen, D.; Irby, J.; Leccacorvi, R.; Vieira, R.; Oi, C.; Rice, J.; Reinke, M.; et al

    2015-03-12

    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-modemore » 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.« less

  4. Implementation of LHCD Experiments on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Parker, R.; Basse, N.; Beck, W.; Bernabei, S.; Childs, R.; Ellis, R.; Fredd, E.; Greenough, N.; Grimes, M.; Gwinn, D.; Hosea, J.; Irby, J.; Koert, P.; Kung, C. C.; Labombard, B.; Liptac, J.; Loesser, G. D.; Marmar, E.; Schilling, G.; Terry, D.; Terry, J.; Vieira, R.; Wallace, G.; Wilson, J. R.; Zaks, J.

    2005-09-01

    An antenna-transmitter system for driving current in the LHRF has been installed in Alcator C-Mod. The antenna is a grill consisting of 4 poloidal rows of waveguides, each with 24 guides in the toroidal direction. Power is supplied by 12 klystrons capable of 250 kW operation at a frequency of 4.6 GHz. Thus the total source power is 3 MW, with about 1.5 MW available to be coupled to the plasma. Power supply and heat throughput limits in C-Mod limit the pulse length to 5 s, which however represents several current redistribution times. With 90° phasing, the n∥ spectrum is sharply peaked at 2.3 and the range 1.5 < n∥ < 3.5 can be accessed dynamically by varying the phase of the klystrons. The system is in the commissioning phase with klystron power limited to ˜20 kW and pulse length to 10 ms. Early results from plasma operation are discussed.

  5. Implementation of LHCD Experiments on Alcator C-Mod

    SciTech Connect

    Parker, R.; Basse, N.; Beck, W.; Childs, R.; Grimes, M.; Gwinn, D.; Irby, J.; Koert, P.; Labombard, B.; Liptac, J.; Marmar, E.; Terry, D.; Terry, J.; Vieira, R.; Wallace, G.; Wilson, J.R.; Zaks, J.; Bernabei, S.; Ellis, R.; Fredd, E.

    2005-09-26

    An antenna-transmitter system for driving current in the LHRF has been installed in Alcator C-Mod. The antenna is a grill consisting of 4 poloidal rows of waveguides, each with 24 guides in the toroidal direction. Power is supplied by 12 klystrons capable of 250 kW operation at a frequency of 4.6 GHz. Thus the total source power is 3 MW, with about 1.5 MW available to be coupled to the plasma. Power supply and heat throughput limits in C-Mod limit the pulse length to 5 s, which however represents several current redistribution times. With 90 deg. phasing, the n parallel spectrum is sharply peaked at 2.3 and the range 1.5 < n parallel < 3.5 can be accessed dynamically by varying the phase of the klystrons. The system is in the commissioning phase with klystron power limited to {approx}20 kW and pulse length to 10 ms. Early results from plasma operation are discussed.

  6. Correlation ECE diagnostic in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Sung, C.; White, A. E.; Howard, N. T.; Mikkelsen, D.; 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.

    2015-03-01

    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.

  7. Correlation ECE diagnostic in Alcator C-Mod

    SciTech Connect

    Sung, C.; White, A. E.; Howard, N. T.; Mikkelsen, D.; 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.

    2015-03-12

    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.

  8. Investigation of the transport shortfall in Alcator C-Mod L-mode plasmas

    SciTech Connect

    Howard, N. T.; White, A. E.; Greenwald, M.; Reinke, M. L.; Walk, J.; Holland, C.; Candy, J.; Goerler, T.

    2013-03-15

    A so-called 'transport shortfall,' where ion and electron heat fluxes and turbulence are underpredicted by gyrokinetic codes, has been robustly identified in DIII-D L-mode plasmas for {rho}>0.55[T. L. Rhodes et al., Nucl. Fusion 51(6), 063022 (2011); and C. Holland et al., Phys. Plasmas 16(5), 052301 (2009)]. To probe the existence of a transport shortfall across different tokamaks, a dedicated scan of auxiliary heated L-mode discharges in Alcator C-Mod are studied in detail with nonlinear gyrokinetic simulations for the first time. Two discharges, only differing by the amount of auxiliary heating are investigated using both linear and nonlinear simulation of the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)]. Nonlinear gyrokinetic simulation of the low and high input power discharges reveals a discrepancy between simulation and experiment in only the electron heat flux channel of the low input power discharge. However, both discharges demonstrate excellent agreement in the ion heat flux channel, and the high input power discharge demonstrates simultaneous agreement with experiment in both the electron and ion heat flux channels. A summary of linear and nonlinear gyrokinetic results and a discussion of possible explanations for the agreement/disagreement in each heat flux channel is presented.

  9. Investigation of the transport shortfall in Alcator C-Mod L-mode plasmas

    NASA Astrophysics Data System (ADS)

    Howard, N. T.; White, A. E.; Greenwald, M.; Reinke, M. L.; Walk, J.; Holland, C.; Candy, J.; Görler, T.

    2013-03-01

    A so-called "transport shortfall," where ion and electron heat fluxes and turbulence are underpredicted by gyrokinetic codes, has been robustly identified in DIII-D L-mode plasmas for ρ >0.55 [T. L. Rhodes et al., Nucl. Fusion 51(6), 063022 (2011); and C. Holland et al., Phys. Plasmas 16(5), 052301 (2009)]. To probe the existence of a transport shortfall across different tokamaks, a dedicated scan of auxiliary heated L-mode discharges in Alcator C-Mod are studied in detail with nonlinear gyrokinetic simulations for the first time. Two discharges, only differing by the amount of auxiliary heating are investigated using both linear and nonlinear simulation of the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)]. Nonlinear gyrokinetic simulation of the low and high input power discharges reveals a discrepancy between simulation and experiment in only the electron heat flux channel of the low input power discharge. However, both discharges demonstrate excellent agreement in the ion heat flux channel, and the high input power discharge demonstrates simultaneous agreement with experiment in both the electron and ion heat flux channels. A summary of linear and nonlinear gyrokinetic results and a discussion of possible explanations for the agreement/disagreement in each heat flux channel is presented.

  10. Alcator C-Mod: A high-field divertor tokamak

    NASA Astrophysics Data System (ADS)

    Lipschultz, B.; Becker, H.; Bonoli, P.; Coleman, J.; Fiore, C.; Golovato, S.; Granetz, R.; Greenwald, M.; Gwinn, D.; Humphries, D.; Hutchinson, I.; Irby, J.; Marmar, E.; Montgomery, D. B.; Najmabadi, F.; Parker, R.; Porkolab, M.; Rice, J.; Sevillano, E.; Takase, Y.; Terry, J.; Watterson, R.; Wolfe, S.

    1989-04-01

    The Alcator C-Mod tokamak is a new device presently under construction at Massachusetts Institute of Technology (M.I.T.) which is scheduled to begin operation in mid-1990. The projected operating parameters are as follows: Toroidal field of 9 T; Ip ≤ 3 MA, R = 66.5 cm, a = 21 cm, κ ≤ 2.0, δ ≤ 0.5, ne ≤ 10 21m-3, PICRF ≤ 6 MW. The divertor configuration includes mechanical baffling as opposed to an 'open' geometry. Under strictly ohmic heating conditions, central Ti and Te are predicted to be in the range 2.5-3.5 keV over the density range (4-8) × 10 20m-3. With the addition of 6 MW of ICRF heating, Ti should vary from 4-8 keV over the same density range (assuming either Kaye-Goldston or Neo-Alcator scalings for electron confinement). Based on edge plasma characterizations from Alcator-C and divertor tokamaks, the scrape-off layer (SOL) properties are predicted to be: λn ≈ 10mm, density at the divertor plate < 2 × 10 21m-3, H 0 ionization mean free path between 1 and 10 mm. Maximum heat loads on various internal components are predicted to be in the range 5-10 MW/m 2. The flexibility of the poloidal field system in forming a number of flux surface geometries will provide further comparisons of the relative impurity control capabilities of double-null, single-null and limiter plasmas.

  11. Charge exchange recombination spectroscopy on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Eisner, E. C.; Rowan, William L.

    2001-01-01

    The optical systems for charge-exchange recombination spectroscopy (CXRS) on Alcator C-Mod have been developed and installed. They will provide measurements of Ti, νθ, and νφ. With the addition of the motional Stark effect diagnostic to determine Bp, Er can be inferred from the ion pressure balance equation, Er=(Zenz)-1∇rpz+νφzBθ-νθzBφ. The optical systems are simple and have high throughput. In particular, the toroidally viewing systems must be designed for invessel installation close to the plasma where they are subject to large forces and are inaccessible between vacuum vents. Two optical systems, located invessel, provide 20 channels of Ti and νφ data from 67.3 cm

  12. Scrape-off layer reflectometer for Alcator C-Mod.

    PubMed

    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 10(16)-10(20) m(-3) at B(0)=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 (TE(01)) 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. PMID:21033950

  13. Scrape-off layer reflectometer for Alcator C-Mod

    SciTech Connect

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

    2010-10-15

    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 {mu}s to 1 ms, and will cover a density range of approximately 10{sup 16}-10{sup 20} m{sup -3} at B{sub 0}=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 (TE{sub 01}) 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.

  14. Scrape-off layer reflectometer for Alcator C-Mod.

    PubMed

    Hanson, G R; Wilgen, J B; Lau, C; Lin, Y; Wallace, G M; Wukitch, S J

    2008-10-01

    A two-frequency x-mode reflectometer operating from 100 to 146 GHz is deployed on Alcator C-Mod to measure the density profile and fluctuations in the scrape-off layer (SOL) immediately in front of the new J-port ICRF antenna and the new C-port lower hybrid launcher. The reflectometer covers densities from 10(16) to 10(20) m(-3) at 5-5.4 T. To provide the greatest flexibility and capability to deal with density fluctuations approaching 100% peak-to-peak in the SOL, both full-phase and differential-phase measurement capabilities with sweep speeds of approximately 10 micros to >1 ms are implemented. The differential-phase measurement uses a difference frequency of 500 MHz, corresponding to cutoff layer separations ranging from about 0.1 to 1 mm. The reflectometer has six sets of launchers: three on the ICRF antenna and three on the lower hybrid launcher. Both the ICRF antenna and the lower hybrid launcher incorporate reflectometer antennas at their top, bottom, and midplane locations. PMID:19044598

  15. Scrape-off layer reflectometer for Alcator C-Mod

    SciTech Connect

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

    2008-10-15

    A two-frequency x-mode reflectometer operating from 100 to 146 GHz is deployed on Alcator C-Mod to measure the density profile and fluctuations in the scrape-off layer (SOL) immediately in front of the new J-port ICRF antenna and the new C-port lower hybrid launcher. The reflectometer covers densities from 10{sup 16} to 10{sup 20} m{sup -3} at 5-5.4 T. To provide the greatest flexibility and capability to deal with density fluctuations approaching 100% peak-to-peak in the SOL, both full-phase and differential-phase measurement capabilities with sweep speeds of {approx}10 {mu}s to >1 ms are implemented. The differential-phase measurement uses a difference frequency of 500 MHz, corresponding to cutoff layer separations ranging from about 0.1 to 1 mm. The reflectometer has six sets of launchers: three on the ICRF antenna and three on the lower hybrid launcher. Both the ICRF antenna and the lower hybrid launcher incorporate reflectometer antennas at their top, bottom, and midplane locations.

  16. Lower Hybrid Wave Induced Rotation on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Parker, Ron; Podpaly, Yuri; Rice, John; Schmidt, Andrea

    2009-11-01

    Injection of RF power in the vicinity of the lower hybrid frequency has been observed to cause strong counter current rotation in Alcator C-Mod plasmas [1,2]. The spin-up rate is consistent with the rate at which momentum is injected by the LH waves, and also the rate at which fast electron momentum is transferred to the ions. A momentum diffusivity of ˜ 0.1 m^2/s is sufficient to account for the observed steady-state rotation. This value is also comparable with that derived from an analysis of rotation induced by RF mode conversion [3]. Radial force balance requires a radial electric field, suggesting a buildup of negative charge in the plasma core. This may be the result of an inward pinch of the LH produced fast electrons, as would be expected for resonant trapped particles. Analysis of the fast-electron-produced bremsstrahlung during LH power modulation experiments yields an inward pinch velocity of ˜ 1 m/s, consistent with the estimated trapped particle pinch velocity. [4pt] [1] A. Ince-Cushman, et.al., Phys. Rev. Lett., 102, 035002 (2009)[0pt] [2] J. E. Rice, et. al., Nucl. Fusion 49, 025004 (2009)[0pt] [3] Y. Lin, et.al., this meeting

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

    SciTech Connect

    J.R. Wilson, S. Bernabei, P. Bonoli, A. Hubbard, R. Parker, A. Schmidt, G. Wallace, J. Wright, and the Alcator C-Mod Team

    2007-10-09

    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|| ~ 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 n20IR/P ~ 0.3. Modeling using the CQL3D code supports these efficiencies. Sawtooth oscillations vanish, accompanied with peaking of the electron temperature (Te0 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).

  18. Edge Turbulence Imaging in the Alcator C-Mod Tokamak

    SciTech Connect

    S.J. Zweben; D.P. Stotler; J.L. Terry; B. LaBombard; M. Greenwald; M. Muterspaugh; C.S. Pitcher; the Alcator C-Mod Group; K. Hallatschek; R.J. Maqueda; B. Rogers; J.L. Lowrance; V.J. Mastrocola; G.F. Renda

    2001-11-26

    The 2-D radial vs. poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I.H. Hutchinson, R. Boivin, P.T. Bonoli et al., Nuclear Fusion 41(2001) 1391] was measured using fast cameras and compared with 3-D numerical simulations of edge plasma turbulence. The main diagnostic is Gas Puff Imaging (GPI), in which the visible D(subscript alpha) emission from a localized D(subscript 2) gas puff is viewed along a local magnetic field line. The observed D(subscript alpha) fluctuations have a typical radial and poloidal scale of approximately 1 cm, and often have strong local maxima (''blobs'') in the scrape-off layer. The motion of this 2-D structure motion has also been measured using an ultra-fast framing camera with 12 frames taken at 250,000 frames/sec. 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.

  19. Overview of Recent Alcator C-Mod Highlights

    NASA Astrophysics Data System (ADS)

    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.

  20. Overview of recent Alcator C-Mod research

    NASA Astrophysics Data System (ADS)

    Marmar, E. S.; Bai, B.; Boivin, R. L.; Bonoli, P. T.; Boswell, C.; Bravenec, R.; Carreras, B.; Ernst, D.; Fiore, C.; Gangadhara, S.; Gentle, K.; Goetz, J.; Granetz, R.; Greenwald, M.; Hallatschek, K.; Hastie, J.; Hosea, J.; Hubbard, A.; Hughes, J. W.; Hutchinson, I.; In, Y.; Irby, J.; Jennings, T.; Kopon, D.; Kramer, G.; La Bombard, B.; Lee, W. D.; Lin, Y.; Lipschultz, B.; Liptac, J.; Lynn, A.; Marr, K.; Maqueda, R.; Melby, E.; Mikkelsen, D.; Mossessian, D.; Nazikian, R.; Nevins, W. M.; Parker, R.; Pedersen, T. S.; Phillips, C. K.; Phillips, P.; Pitcher, C. S.; Porkolab, M.; Ramos, J.; Redi, M.; Rice, J.; Rogers, B. N.; Rowan, W. L.; Sampsell, M.; Schilling, G.; Scott, S.; Snipes, J.; Snyder, P.; Stotler, D.; Taylor, G.; Terry, J. L.; Wilson, H.; Wilson, J. R.; Wolfe, S. M.; Wukitch, S.; Xu, X. Q.; Youngblood, B.; Yuh, H.; Zhurovich, K.; Zweben, S.

    2003-12-01

    Research on the Alcator C-Mod tokamak [1] 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-H transition, with timescale comparable to that for energy transport. In discharges which develop internal transport barriers, 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 enhanced D-Alpha (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 modelling indicates that these are due to intermediate n peeling-ballooning modes. Phase contrast imaging 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 in 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 modelling indicates that regimes with high bootstrap fraction can be produced.

  1. Overview of the Alcator C-Mod Research Program

    NASA Astrophysics Data System (ADS)

    Marmar, E.; Bader, A.; Bakhtiari, M.; Barnard, H.; Beck, W.; Bespamyatnov, I.; Binus, A.; Bonoli, P.; Bose, B.; Bitter, M.; Cziegler, I.; Dekow, G.; Dominguez, A.; Duval, B.; Edlund, E.; Ernst, D.; Ferrara, M.; Fiore, C.; Fredian, T.; Graf, A.; Granetz, R.; Greenwald, M.; Grulke, O.; Gwinn, D.; Harrison, S.; Harvey, R.; Hender, T. C.; Hosea, J.; Hill, K.; Howard, N.; Howell, D. F.; Hubbard, A.; Hughes, J. W.; Hutchinson, I.; Ince-Cushman, A.; Irby, J.; Izzo, V.; Kanojia, A.; Kessel, C.; Ko, J. S.; Koert, P.; La Bombard, B.; Lau, C.; Lin, L.; Lin, Y.; Lipschultz, B.; Liptac, J.; Ma, Y.; Marr, K.; May, M.; McDermott, R.; Meneghini, O.; Mikkelsen, D.; Ochoukov, R.; Parker, R.; Phillips, C. K.; Phillips, P.; Podpaly, Y.; Porkolab, M.; Reinke, M.; Rice, J.; Rowan, W.; Scott, S.; Schmidt, A.; Sears, J.; Shiraiwa, S.; Sips, A.; Smick, N.; Snipes, J.; Stillerman, J.; Takase, Y.; Terry, D.; Terry, J.; Tsujii, N.; Valeo, E.; Vieira, R.; Wallace, G.; Whyte, D.; Wilson, J. R.; Wolfe, S.; Wright, G.; Wright, J.; Wukitch, S.; Wurden, G.; Xu, P.; Zhurovich, K.; Zaks, J.; Zweben, S.

    2009-10-01

    This paper summarizes highlights of research results from the Alcator C-Mod tokamak covering the period 2006-2008. Active flow drive, using mode converted ion cyclotron waves, has been observed for the first time in a tokamak plasma, using a mix of D and 3He ion species; toroidal and poloidal flows are driven near the location of the mode conversion layer. ICRF induced edge sheaths are implicated in both the erosion of thin boron coatings and the generation of metallic impurities. Lower hybrid range of frequencies (LHRF) microwaves have been used for efficient current drive, current profile modification and toroidal flow drive. In addition, LHRF has been used to modify the H-mode pedestal, increasing temperature, decreasing density and lowering the pedestal collisionality. Studies of hydrogen isotope retention in solid metallic plasma facing components reveal significantly higher retention than expected from ex situ laboratory studies; a model to explain the results, based on plasma/neutral induced lattice damage, has been developed and tested. During gas-puff mitigation of disruptions, induced MHD instabilities cause the magnetic field to become stochastic, resulting in reduction of halo currents, spreading of plasma power loading and loss of runaway electrons before they cause damage. Detailed pedestal rotation profile measurements have been used to infer Er profiles, and correlation with global H-mode confinement. An improved L-mode regime, obtained at q95 <= 3 with ion drift away from the active X-point, shows very good energy confinement with a strong temperature pedestal, a weak density pedestal, and no evidence of particle or impurity accumulation, without the need for ELMs or any additional edge density regulation mechanism.

  2. Overview of the Alcator C-MOD Research Program

    SciTech Connect

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

    2007-11-13

    Recent research on the high-field, high-density diverted Alcator C-MOD tokamak has focussed 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 < 900 kW) have demonstrated fully noninductive current drive at Ιρ ~ 1.0 MA with good efficiency, Ιdrive = 0.4PLH/neoR (MA,MW,1020m-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 Ι 2(over)ρ, 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.

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

    SciTech Connect

    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.

  4. Overview of the Alcator C-MOD research programme

    NASA Astrophysics Data System (ADS)

    Scott, S.; Bader, A.; Bakhtiari, M.; Basse, N.; Beck, W.; Biewer, T.; Bernabei, S.; Bonoli, P.; Bose, B.; Bravenec, R.; Bespamyatnov, I.; Childs, R.; Cziegler, I.; Doerner, R.; Edlund, E.; Ernst, D.; Fasoli, A.; Ferrara, M.; Fiore, C.; Fredian, T.; Graf, A.; Graves, T.; Granetz, R.; Greenough, N.; Greenwald, M.; Grimes, M.; Grulke, O.; Gwinn, D.; Harvey, R.; Harrison, S.; Hender, T. C.; Hosea, J.; Howell, D. F.; Hubbard, A. E.; Hughes, J. W.; Hutchinson, I.; Ince-Cushman, A.; Irby, J.; Jernigan, T.; Johnson, D.; Ko, J.; Koert, P.; La Bombard, B.; Kanojia, A.; Lin, L.; Lin, Y.; Lipschultz, B.; Liptac, J.; Lynn, A.; MacGibbon, P.; Marmar, E.; Marr, K.; May, M.; Mikkelsen, D. R.; McDermott, R.; Parisot, A.; Parker, R.; Phillips, C. K.; Phillips, P.; Porkolab, M.; Reinke, M.; Rice, J.; Rowan, W.; Sampsell, M.; Schilling, G.; Schmidt, A.; Smick, N.; Smirnov, A.; Snipes, J.; Stotler, D.; Stillerman, J.; Tang, V.; Terry, D.; Terry, J.; Ulrickson, M.; Vieira, R.; Wallace, G.; Whyte, D.; Wilson, J. R.; Wright, G.; Wright, J.; Wolfe, S.; Wukitch, S.; Wurden, G.; Yuh, H.; Zhurovich, K.; Zaks, J.; Zweben, S.

    2007-10-01

    Alcator C-MOD has compared plasma performance with plasma-facing components (PFCs) coated with boron to all-metal PFCs to assess projections of energy confinement from current experiments to next-generation burning tokamak plasmas. Low-Z coatings reduce metallic impurity influx and diminish radiative losses leading to higher H-mode pedestal pressure that improves global energy confinement through profile stiffness. 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 < 900 kW) in preparation for future advanced-tokamak studies have demonstrated fully non-inductive current drive at Ip ~ 1.0 MA with good efficiency, Idrive = 0.4 PLH/neoR (MA, MW, 1020 m-3,m). The potential to mitigate disruptions in ITER through massive gas-jet impurity puffing has been extended to significantly higher plasma pressures and shorter disruption times. The fraction of total plasma energy radiated increases with the Z of the impurity gas, reaching 90% for krypton. A positive major-radius scaling of the error field threshold for locked modes (Bth/B ~ R0.68±0.19) is inferred from its measured variation with BT that implies a favourable threshold value for ITER. A phase contrast imaging diagnostic has been used to study the structure of Alfvén cascades and turbulent density fluctuations in plasmas with an internal transport barrier. Understanding the mechanisms responsible for regulating the H-mode pedestal height is also crucial for projecting performance in ITER. Modelling of H-mode edge fuelling 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, including a weak variation of pedestal height and constant width. Pressure gradients in the near SOL of Ohmic L-mode plasmas are observed to scale consistently as I_p^2

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

    SciTech Connect

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

    2005-06-27

    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.

  6. Observation of Co and Counter Rotation Produced by Lower Hybrid Waves in Alcator C-Mod

    SciTech Connect

    Parker, R. R.; Podpaly, Y.; Lee, J.; Reinke, M. L.; Rice, J. E.; Bonoli, P. T.; Meneghini, O.; Shiraiwa, S.; Wallace, G. M.; Wilson, J. R.

    2011-12-23

    Lower hybrid waves launched uni-directionally into tokamak plasmas impart momentum to the electrons. This momentum can be transferred to the ions, leading to substantial counter current rotation. Observations of LH-induced counter rotation have been previously reported [1], and the initial rate of increase has been found to be consistent with the calculated rate of wave momentum injection [2]. However, in recent experiments in Alcator C-Mod it has been found that application of LH waves to relatively low current (I{sub p}{approx}0.4-0.6 MA) plasmas can result in a co-current change of rotation, which implies a different mechanism than that described above. This appears to be linked to the so-called intrinsic rotation commonly observed in Alcator C-Mod and other tokamaks [3]. In addition to the change in direction at low current, some dependence on the magnetic configuration (USL vs. LSN) has been observed.

  7. Extended pulse-length operation of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Wolfe, S.; Irby, J.; Terry, J.; Labombard, B.; Wukitch, S.; Marmar, E.; Cochran, W.; Dekow, G.; Gwinn, D.; Maqueda, R.

    2001-10-01

    The C-Mod Advanced Tokamak program depends on the unique capability of the C-Mod facility to operate with plasma pulse lengths corresponding to multiple skin times with high performance parameters. Specifically, pulse lengths with current and toroidal field flattops of order 5 seconds, with toroidal field of 4 tesla, are proposed. In the case of the AT program, these plasmas would have current sustained non-inductively, i.e. by a combination of RF (lower hybrid) current drive and pressure-driven current. Experiments during the 2001 experimental campaign will extend the plasma pulse length to the maximum possible with only inductive current drive. The purpose of these experiments is to test and demonstrate the long-pulse capability of the coils, power system, control system, etc., and to test power and particle handling performance under long pulse conditions. In support of the latter goal, we will benchmark divertor surface heating during medium-power operation and assess the effectiveness of X-point sweeping and N2 puffing for dissipating the divertor heat loads. Results of these experiments will be presented.

  8. Performance evaluation of MSE on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Scott, S.; Mumgaard, R.; Granetz, R.; Shugayev, R.; Chung, M.

    2011-10-01

    Performance of the Motional Stark Effect (MSE) diagnostic has been compared to first-principle numerical simulations by moving the Stark-split beam spectrum past MSE's narrow bandpass filters. The relative position of the Stark spectrum to the filters was varied by ramping the toroidal field at fixed q; changing the filter temperature; and swapping one filter for another. We compare the signal strength and polarization fraction of beam emission to simulations that model the MSE emission using a modified version of the PERF code (NC Hawkes et al., Rev. Sci. Instrum. 70, 894 (1999)). The simulation includes models for beam attenuation, beam excitation, Stark emission, aperture broadening, the measured filter response function, and background visible Bremsstrahlung emission. This work confirms analytic comparisons of the expected MSE signal strength between C-Mod and other tokamaks which suggest that the polarized signal intensity on C-Mod is unexpectedly low by a factor of several. Measurements of the beam performance using calorimeters, beam-target neutron production during deuterium beam injection, and IR measurements of a stainless steel target heated by the beam will be discussed. Supported by US DoE awards DE-FC02-99-ER54512 and DE-AC02-09CH11446.

  9. Initial Active MHD Spectroscopy Experiments on Alcator C-MOD

    NASA Astrophysics Data System (ADS)

    Schmittdiel, D. A.; Snipes, J. A.; Granetz, R. S.; Parker, R. R.; Wolfe, S. M.; Fasoli, A.

    2002-11-01

    The Active MHD Spectroscopy system is a new diagnostic on C-MOD that will be used to study low frequency MHD modes and TAE's present at high B_tor, n_e, and Te ˜= T_i. The present system consists of two antennas, power amplifiers, and an impedance matching network. Each antenna is 15 × 25 cm with five turns, an inductance of ˜10 μH, and is covered by boron nitride tiles. The two antennas are placed at the same toroidal location, symmetrically above and below the midplane. Each antenna is driven by a ˜1 kW power amplifier in the range of 1 kHz - 1 MHz with an expected antenna current ˜10 A, which will produce a vacuum field of ˜0.5 G at the q = 1.5 surface. This diagnostic is designed to excite high n ( ˜20) stable TAE's and initial results regarding their frequency, mode structure, and damping rate will be presented. Evolution of these modes could also provide information on the q profile to compare with MSE measurements, which will be important for planned lower hybrid current drive operation in 2003.

  10. Integrated modeling of LHCD experiment on Alcator C-Mod

    SciTech Connect

    Shiraiwa, S.; Bonoli, P.; Parker, R.; Wallace, G.

    2014-02-12

    Recent progress in integrating the latest LHCD model based on ray-tracing into the Integrated Plasma Simulator (IPS) is reported. IPS, a python based framework for time dependent tokamak simulation, was expanded recently to incorporate LHCD simulation using GENRAY/CQL3D (ray-tracing/3D Fokker-Planck package). Using GENRAY/CQL3D in the IPS framework, it becomes possible to include parasitic LHCD power loss near the plasma edge, which was found to be important in experiments particularly at high density as expected on reactors. Moreover, it allows for evolving the velocity distribution function in 4 D (ν{sub ∥}, ν⊥, r/a, t) space self-consistently. In order to validate the code, IPS is applied to LHCD experiments on Alctor C-Mod. In this paper, a LHCD experiment performed at the density of n{sub e}∼0.5×10{sup 20}m{sup −3} where good LHCD efficiency and the development of internal transport barrier (ITB) was reported, is modelled in a predictive mode and the result is compared with experiment.

  11. Electron Transport Dominated Regimes in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Porkolab, M.; Dorris, J.; Bonoli, P. T.; Ennever, P.; Fiore, C.; Greenwald, M.; Hubbard, A.; Ma, Y.; Reinke, M. L.; Rice, J.; Rost, J.; Tsujii, N.; Lin, L.; Candy, J.; Waltz, R.; Diamond, P.; Lee, C. J.

    2010-11-01

    In ohmically heated low density plasmas where τE ne, the so-called neo-Alcator regime, TRANSP results indicate that χi<< χe, while nonlinear gyrokinetic analysis for the measured profiles predicts the opposite inequality [1]. This regime is of great interest for transport studies since Ti < Te, and the electron and ion transport channels can be separated and studied separately. At the same time, measurements of turbulent fluctuations with Phase Contrast Imaging diagnostic (PCI) indicated reasonable agreement with GYRO predictions at frequencies 80-250 kHz, corresponding to core ITG turbulence. The turbulent spectrum at lower frequencies could not be identified since the PCI technique does not allow separation of the core plasma fluctuations from those at the edge. Here we present measurements and analysis from a more extensive set of plasma regimes than previously. Of particular current interest is the role of electron drift wave turbulence driven by ohmic electron drift, U [2], since in these low density regimes U/Cs <= 6, and experimentally we find that the global confinement τE Cs/U where Cs = (Te/mi)^1/2. [1] L. Lin, Invited talk, APS-DPP, 11, 2009, Atlanta, GA. [2] C.J. Lee, P. Diamond, M. Porkolab, presented at TTF workshop, 2010.

  12. 20 years of research on the Alcator C-Mod tokamaka)

    NASA Astrophysics Data System (ADS)

    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

  13. 20 years of research on the Alcator C-Mod tokamak

    SciTech Connect

    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.; and others

    2014-11-15

    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

  14. Design of a Compact Lower Hybrid Coupler for Alcator C-Mod

    SciTech Connect

    Bernabei, Stefano; Hosea, Joel C.; Kung, Chun Chieh; Loesser, George D.; Rushinski, Joseph; Wilson, James R.; Parker, Ronald R.; Porkolab, Miklos

    2003-03-15

    Princeton Plasma Physics Laboratory and the Massachusetts Institute of Technology are preparing an experiment of current profile control using lower hybrid waves to produce and sustain advanced tokamak regimes in steady-state conditions in Alcator C-Mod. Unlike the Joint European Torus, ToreSupra, and JT60 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.

  15. ICRF Heating with {omega}<{omega}{sub ci} in Alcator C-Mod

    SciTech Connect

    Phillips, C. K.; Hosea, J. C.; Valeo, E. J.; Wilson, J. R.; Bonoli, P. T.; Lin, Y.; Porkolab, M.; Wright, J. C.; Wukitch, S. J.

    2007-09-28

    The TORIC 2D full wave simulation code has been used to study the dynamics of waves with {omega}<{omega}{sub ci} everywhere for all ions in the Alcator C-Mod tokamak. This potential heating regime can be accessed uniquely on C-Mod, because of its high magnetic field capability, B{sub T}{<=}8 T, and variable ICRF source frequency, 40-80 MHz. The simulations indicate that the launched fast waves can mode convert to a short wavelength slow wave on the high field side of the discharge that damps primarily on electrons. The degree to which the mode converted wave penetrates into the core of the plasma is found to depend on the equilibrium density profile.

  16. ECE Temperature Fluctuations associated with EDA H-Mode discharges in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Phillips, P. E.; Lynn, A. G.

    2006-10-01

    Alcator C-Mod exhibits an ELM-free H-mode with ``enhanced,,lpha'' emission accompanied by a quasi-coherent mode (QCM) edge relaxation mechanism. This steady state H-mode lowers the peak heat load to the diverters which is advantageous for reactor operations. A high-resolution heterodyne electron-cyclotron-emission (ECE) radiometer with 32 channels (δR˜7mm) and a bandwidth up to 1MHz covering the full radius of C-Mod has observed spatial resolved temperature fluctuations that are highly correlated with the edge QCM mode. The QCM mode is also directly observed by the edge ECE channels though the changes in optical depth due to the large density fluctuations in the QCM (˜30%). Details of these measurements will be presented in this poster.

  17. Alcator C-Mod: research in support of ITER and steps beyond

    NASA Astrophysics Data System (ADS)

    Marmar, E. S.; Baek, S. G.; Barnard, H.; Bonoli, P.; Brunner, D.; Candy, J.; Canik, J.; Churchill, R. M.; Cziegler, I.; Dekow, G.; Delgado-Aparicio, L.; Diallo, A.; Edlund, E.; Ennever, P.; Faust, I.; Fiore, C.; Gao, Chi; Golfinopoulos, T.; Greenwald, M.; Hartwig, Z. S.; Holland, C.; Hubbard, A. E.; Hughes, J. W.; Hutchinson, I. H.; Irby, J.; LaBombard, B.; Lin, Yijun; Lipschultz, B.; Loarte, A.; Mumgaard, R.; Parker, R. R.; Porkolab, M.; Reinke, M. L.; Rice, J. E.; Scott, S.; Shiraiwa, S.; Snyder, P.; Sorbom, B.; Terry, D.; Terry, J. L.; Theiler, C.; Vieira, R.; Walk, J. R.; Wallace, G. M.; White, A.; Whyte, D.; Wolfe, S. M.; Wright, G. M.; Wright, J.; Wukitch, S. J.; Xu, P.

    2015-10-01

    This paper presents an overview of recent highlights from research on Alcator C-Mod. Significant progress has been made across all research areas over the last two years, with particular emphasis on divertor physics and power handling, plasma-material interaction studies, edge localized mode-suppressed pedestal dynamics, core transport and turbulence, and RF heating and current drive utilizing ion cyclotron and lower hybrid tools. Specific results of particular relevance to ITER include: inner wall SOL transport studies that have led, together with results from other experiments, to the change of the detailed shape of the inner wall in ITER; runaway electron studies showing that the critical electric field required for runaway generation is much higher than predicted from collisional theory; core tungsten impurity transport studies reveal that tungsten accumulation is naturally avoided in typical C-Mod conditions.

  18. Correlation ECE and Doppler Backscattering Diagnostics for Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    White, A. E.; Hubbard, A.; Phillips, P.; Irby, J.; Greenwald, M.; Mikkelsen, D. R.; Howard, N. T.

    2010-11-01

    A Correlation Electron Cyclotron Emission (CECE) diagnostic and an X-mode reflectometer system are being developed to measure long wavelength (kθρs) core electron temperature fluctuations and density fluctuations, respectively, at Alcator C-Mod. These new diagnostics will allow for detailed two-field core turbulence measurements and validation studies. Adjustable optics will allow the reflectometer to be configured as a Doppler backscattering system. Global, nonlinear gyrokinetic turbulence simulations (GYRO) and synthetic diagnostics are used to model the diagnostics' expected responses to turbulence in a variety of operating regimes. The challenges associated with the high frequency systems required for core turbulence studies at C-Mod and the feasibility of combining these complementary diagnostics into a single transmission system will be assessed. Accessibility limits, expected wavenumber sensitivity and waveguide/antennae configurations are discussed.

  19. New correlation electron cyclotron emission temperature fluctuation diagnositc for Alcator C-mod

    NASA Astrophysics Data System (ADS)

    Sung, C.; White, A. E.; Irby, J.; Peebles, W. A.; Nguyen, X.

    2011-10-01

    A new Correlation Electron Cyclotron Emission (CECE) system for the measurement of electron temperature fluctuation is planned for Alcator C-mod. The multi-channel CECE radiometer will use the spectral decorrelation technique to measure turbulent fluctuations that are below thermal noise levels. The design of the optics and Intermediate Frequency (IF) section was constrained using predictions from nonlinear gyrokinetic turbulence simulations using the GYRO code. A Gaussian optical system will provide high poloidal spatial resolution (ω0 < 0 . 5 cm) needed to measure long-wavelength core turbulence at C-Mod, kθρs < 0 . 5 . The IF section will employ tunable band-pass filters to optimize turbulence measurements. We will present details of the new CECE system design and laboratory tests of the optics and IF section.

  20. Simulated plasma facing component measurements for an in situ surface diagnostic on Alcator C-Mod

    SciTech Connect

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

    2010-10-15

    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 ({approx}10 min) time scale with {approx}1 {mu}m depth and {approx}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.

  1. Preliminary results from the soft x-ray crystal spectrometer on Alcator C-Mod

    SciTech Connect

    Ince-Cushman, A.; Rice, J. E.; Lee, S. G.; Bitter, M.; Reinke, M.; Podpaly, Y.

    2006-10-15

    A high resolution ({lambda}/{delta}{lambda}{approx}2000) soft x-ray spectrometer has been installed on the Alcator C-Mod tokamak. The system was designed to measure emission line spectra from heliumlike neon (1s{sup 1}2p{sup 1}{yields}1s{sup 2}) in a narrow spectral band centered on {lambda}=13.5 A (920 eV). The instrument is mounted with a poloidal view 20 cm below the midplane (r/a{approx}0.85). In addition to the neon emission lines, strong fluorine, xenon, iron, and manganese lines have been observed.

  2. Plasma wave simulation based on versatile FEM solver on Alcator C-mod

    SciTech Connect

    Shiraiwa, S.; Meneghini, O.; Parker, R.; Wallace, G.; Wilson, J.

    2009-11-26

    The finite element method (FEM) has the potential of simulating plasma waves seamlessly from the core to the vacuum and antenna regions. We explored the possibility of using a versatile FEM solver package, COMSOL, for lower hybrid (LH) wave simulation. Special care was paid to boundary conditions to satisfy toroidal symmetry. The non-trivial issue of introducing hot plasma effects was addressed by an iterative algorithm. These techniques are verified both analytically and numerically. In the lower hybrid (LH) grill antenna coupling problem, the FEM solver successfully reproduced the solution that was obtained analytically. Propagation of LH waves on the Alcator C and Alcator C-MOD plasmas was compared with a ray-tracing code, showing good consistency. The approach based on the FEM is computationally less intensive compared to spectral domain solvers, and more suitable for the simulation of larger device such as ITER.

  3. Design of a CO{sub 2}-laser Thomson scattering ion-tail diagnostic for Alcator C-Mod

    SciTech Connect

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

    1994-09-01

    A CO{sub 2}-laser Thomson scattering diagnostic has been designed for the measurement of the ICRH-produced ion tail on Alcator C-Mod. The plasma parameters and port access require that the detection of scattered radiation be made at small angles, typically one degree or less. The receiver system consists of five heterodyne detectors and the source laser produces an energy of 10 Joules per pulse with a 1 to 5 microsecond pulse length. The scattering system is currently being installed on the Alcator C-Mod experiment. Details of the diagnostic, calculations of the expected measurements, and application of the diagnostic for ITER are presented.

  4. Design of a new optical system for Alcator C-Mod motional Stark effect diagnostic

    SciTech Connect

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

    2008-10-15

    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.

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

    SciTech Connect

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

    2009-11-12

    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.

  6. Fuel retention measurements in Alcator C-Mod using accelerator-based in situ materials surveillance

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

    This paper presents the first in situ time- and space-resolved measurements of deuterium (D) fuel retention in plasma-facing component (PFC) surfaces using Accelerator-based In-situ Materials Surveillance (AIMS) on the Alcator C-Mod tokamak. AIMS is a novel in situ materials diagnostic technique based on the spectroscopic analysis of nuclear reaction products induced in PFC surfaces using an ∼MeV beam of deuterons from a compact linear accelerator in between plasma shots. AIMS measurements of D retention on inner wall PFCs were acquired during diverted and limited plasma operations and during wall conditioning experiments. Intershot measurements demonstrate the local erosion and codeposition of boron films on PFC surfaces with a constant D / B ratio. This is consistent with previous results suggesting that D codeposition with boron is insufficient to account for the net retention observed in Alcator C-Mod. Changes in deuterium concentration during boronization, electron cyclotron and glow cleanings were also measured.

  7. Second Harmonics of Reversed Shear TAE in Alcator C-Mod Geometry

    NASA Astrophysics Data System (ADS)

    Chen, Eugene; Berk, Herbert; Breizman, Boris; Zheng, Linjin

    2009-11-01

    Experiments on Alcator C-Mod, operating with reversed magnetic shear, reveal Toroidal Alfven Eigenmodes (TAE) together with signals at twice the mode frequency. The double frequency signals can be viewed as second harmonic sidebands driven by quadratic non-linear terms in the MHD equations, in analogy with a corresponding theory for Alfven Cascades [1]. However, these nonlinear sidebands have not yet been quantified by any of the existing codes. In this work, we extend AEGIS code [2] to capture nonlinear effects iteratively by treating the nonlinear terms as a driving source in the linear MHD solver. We first compute the TAE mode structure for realistic geometry and q-profile and then use it to find the spatial structure of the second harmonic density perturbation, which can be directly compared with PCI measurements at Alcator C-Mod. [1] H. Smith, B. N. Breizman, M. Lisak and D. Anderson, Physics of Plasmas 13 042504 (2006) [2] L. J. Zheng and M. Kotschenreuther, Journal of Computational Physics 211 (2006) 748-766

  8. Applications of visible CCD cameras on the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Boswell, C. J.; Terry, J. L.; Lipschultz, B.; Stillerman, J.

    2001-01-01

    Five 7 mm diameter remote-head visible charge-coupled device (CCD) cameras are being used on Alcator C-Mod for several different diagnostic purposes. All of the cameras' detectors and optics are placed inside a magnetic field of up to 4 T. Images of the cameras are recorded simultaneously using two three-channel color framegrabber cards. Two CCD cameras are used typically to generate two-dimensional emissivity profiles of deuterium line radiation from the divertor. Interference filters are used to select the spectral line to be measured. The local emissivity is obtained by inverting the measured brightnesses assuming toroidal symmetry of the emission. Another use of the cameras is the identification and localization of impurity sources generated by the ion cyclotron radio frequency (ICRF) antennas, which supply the auxiliary heating on Alcator C-Mod. The impurities generated by the antennas are identified by correlating in time the injections seen at the cameras with measurements made with core diagnostics. Fibers whose views aligned with the camera views and whose outputs are coupled to a visible spectrometer are also used to identify the species of the impurities injected.

  9. ICRF antenna matching system with ferrite tuners for the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Lin, Y.; Binus, A.; Wukitch, S. J.; Koert, P.; Murray, R.; Pfeiffer, A.

    2015-12-01

    Real-time fast ferrite tuning (FFT) has been successfully implemented on the ICRF antennas on Alcator C-Mod. The former prototypical FFT system on the E-port 2-strap antenna has been upgraded using new ferrite tuners that have been designed specifically for the operational parameters of the Alcator C-Mod ICRF system (˜ 80 MHz). Another similar FFT system, with two ferrite tuners and one fixed-length stub, has been installed on the transmission line of the D-port 2-strap antenna. These two systems share a Linux-server-based real-time controller. These FFT systems are able to achieve and maintain the reflected power to the transmitters to less than 1% in real time during the plasma discharges under almost all plasma conditions, and help ensure reliable high power operation of the antennas. The innovative field-aligned (FA) 4-strap antenna on J-port has been found to have an interesting feature of loading insensitivity vs. plasma conditions. This feature allows us to significantly improve the matching for the FA J-port antenna by installing carefully designed stubs on the two transmission lines. The reduction of the RF voltages in the transmission lines has enabled the FA J-port antenna to deliver 3.7 MW RF power to plasmas out of the 4 MW source power in high performance I-mode plasmas.

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

    SciTech Connect

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

    1999-11-03

    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.

  11. BOUT++ Simulations of Edge Turbulence in Alcator C-Mod's EDA H-Mode

    NASA Astrophysics Data System (ADS)

    Davis, Evan; Porkolab, Miklos; Hughes, Jerry; Tsujii, Naoto; Ennever, Paul; Golfinopoulos, Ted; Gyou Baek, Seung; Terry, Jim; Xu, Xueqiao

    2012-10-01

    Energy confinement in tokamaks is believed to be strongly controlled by plasma transport in the edge region, just inside the last closed magnetic flux surface. The Boundary-plasma Turbulence (BOUT++) code is capable of simulating nonlinear fluid turbulence in this region and is well-suited to Alcator C-Mod's Enhanced D-Alpha (EDA) H-mode (&*circ;>1). The EDA H-mode is always accompanied by the quasi-coherent mode (QCM), an edge fluctuation believed to reduce impurity confinement and allow steady-state H-mode operation. Using experimentally measured profiles as input, BOUT++ calculations show that typical C-Mod EDA H-modes are ideal MHD stable but become linearly unstable when the pedestal resistivity is included (η>10-7 φ-m). The computed growth rate in these resistive ballooning modes is found to be consistent with theory, while incorporation of experimentally measured flow profiles has allowed the self-consistent temporal evolution of the edge radial electric field. Nonlinear simulations have reached turbulent steady state, and the computed turbulence spectrum will be compared with measurements from relevant C-Mod diagnostics, such as phase contrast imaging (PCI), reflectometry, gas puff imaging (GPI), and magnetic probes.

  12. Multi-channel transport experiments at Alcator C-Mod and comparison with gyrokinetic simulationsa)

    NASA Astrophysics Data System (ADS)

    White, A. E.; Howard, N. T.; Greenwald, M.; Reinke, M. L.; Sung, C.; Baek, S.; Barnes, M.; Candy, J.; Dominguez, A.; Ernst, D.; Gao, C.; Hubbard, A. E.; Hughes, J. W.; Lin, Y.; Mikkelsen, D.; Parra, F.; Porkolab, M.; Rice, J. E.; Walk, J.; Wukitch, S. J.; Team, Alcator C-Mod

    2013-05-01

    Multi-channel transport experiments have been conducted in auxiliary heated (Ion Cyclotron Range of Frequencies) L-mode plasmas at Alcator C-Mod [Marmar and Alcator C-Mod Group, Fusion Sci. Technol. 51(3), 3261 (2007)]. These plasmas provide good diagnostic coverage for measurements of kinetic profiles, impurity transport, and turbulence (electron temperature and density fluctuations). In the experiments, a steady sawtoothing L-mode plasma with 1.2 MW of on-axis RF heating is established and density is scanned by 20%. Measured rotation profiles change from peaked to hollow in shape as density is increased, but electron density and impurity profiles remain peaked. Ion or electron heat fluxes from the two plasmas are the same. The experimental results are compared directly to nonlinear gyrokinetic theory using synthetic diagnostics and the code GYRO [Candy and Waltz, J. Comput. Phys. 186, 545 (2003)]. We find good agreement with experimental ion heat flux, impurity particle transport, and trends in the fluctuation level ratio (T˜e/Te)/(n ˜e/ne), but underprediction of electron heat flux. We find that changes in momentum transport (rotation profiles changing from peaked to hollow) do not correlate with changes in particle transport, and also do not correlate with changes in linear mode dominance, e.g., Ion Temperature Gradient versus Trapped Electron Mode. The new C-Mod results suggest that the drives for momentum transport differ from drives for heat and particle transport. The experimental results are inconsistent with present quasilinear models, and the strong sensitivity of core rotation to density remains unexplained.

  13. Perturbative thermal diffusivity from partial sawtooth crashes in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Creely, A. J.; White, A. E.; Edlund, E. M.; Howard, N. T.; Hubbard, A. E.

    2016-03-01

    Perturbative thermal diffusivity has been measured on Alcator C-Mod via the use of the extended-time-to-peak method on heat pulses generated by partial sawtooth crashes. Perturbative thermal diffusivity governs the propagation of heat pulses through a plasma. It differs from power balance thermal diffusivity, which governs steady state thermal transport. Heat pulses generated by sawtooth crashes have been used extensively in the past to study heat pulse thermal diffusivity (Lopes Cardozo 1995 Plasma Phys. Control. Fusion 37 799), but the details of the sawtooth event typically lead to non-diffusive ‘ballistic’ transport, making them an unreliable measure of perturbative diffusivity on many tokamaks (Fredrickson et al 2000 Phys. Plasmas 7 5051). Partial sawteeth are common on numerous tokamaks, and generate a heat pulse without the ‘ballistic’ transport that often accompanies full sawteeth (Fredrickson et al 2000 Phys. Plasmas 7 5051). This is the first application of the extended-time-to-peak method of diffusivity calculation (Tubbing et al 1987 Nucl. Fusion 27 1843) to partial sawtooth crashes. This analysis was applied to over 50 C-Mod shots containing both L- and I-Mode. Results indicate correlations between perturbative diffusivity and confinement regime (L- versus I-mode), as well as correlations with local temperature, density, the associated gradients, and gradient scale lengths (a/L Te and a/L n ). In addition, diffusivities calculated from partial sawteeth are compared to perturbative diffusivities calculated with the nonlinear gyrokinetic code GYRO. We find that standard ion-scale simulations (ITG/TEM turbulence) under-predict the perturbative thermal diffusivity, but new multi-scale (ITG/TEM coupled with ETG) simulations can match the experimental perturbative diffusivity within error bars for an Alcator C-Mod L-mode plasma. Perturbative diffusivities extracted from heat pulses due to partial sawteeth provide a new constraint that can be used to

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

    SciTech Connect

    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.; Mikkelsen, D.; Rowan, W. L.; Bespamyatnov, I.

    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.

  15. Multi-channel transport experiments at Alcator C-Mod and comparison with gyrokinetic simulations

    SciTech Connect

    White, A. E.; Howard, N. T.; Greenwald, M.; Reinke, M. L.; Sung, C.; Baek, S.; Barnes, M.; Dominguez, A.; Ernst, D.; Gao, C.; Hubbard, A. E.; Hughes, J. W.; Lin, Y.; Parra, F.; Porkolab, M.; Rice, J. E.; Walk, J.; Wukitch, S. J.; Team, Alcator C-Mod; Candy, J.; and others

    2013-05-15

    Multi-channel transport experiments have been conducted in auxiliary heated (Ion Cyclotron Range of Frequencies) L-mode plasmas at Alcator C-Mod [Marmar and Alcator C-Mod Group, Fusion Sci. Technol. 51(3), 3261 (2007)]. These plasmas provide good diagnostic coverage for measurements of kinetic profiles, impurity transport, and turbulence (electron temperature and density fluctuations). In the experiments, a steady sawtoothing L-mode plasma with 1.2 MW of on-axis RF heating is established and density is scanned by 20%. Measured rotation profiles change from peaked to hollow in shape as density is increased, but electron density and impurity profiles remain peaked. Ion or electron heat fluxes from the two plasmas are the same. The experimental results are compared directly to nonlinear gyrokinetic theory using synthetic diagnostics and the code GYRO [Candy and Waltz, J. Comput. Phys. 186, 545 (2003)]. We find good agreement with experimental ion heat flux, impurity particle transport, and trends in the fluctuation level ratio (T(tilde sign){sub e}/T{sub e})/(ñ{sub e}/n{sub e}), but underprediction of electron heat flux. We find that changes in momentum transport (rotation profiles changing from peaked to hollow) do not correlate with changes in particle transport, and also do not correlate with changes in linear mode dominance, e.g., Ion Temperature Gradient versus Trapped Electron Mode. The new C-Mod results suggest that the drives for momentum transport differ from drives for heat and particle transport. The experimental results are inconsistent with present quasilinear models, and the strong sensitivity of core rotation to density remains unexplained.

  16. High density LHRF experiments in Alcator C-Mod and implications for reactor scale devices

    NASA Astrophysics Data System (ADS)

    Baek, S. G.; Parker, R. R.; Bonoli, P. T.; Shiraiwa, S.; Wallace, G. M.; LaBombard, B.; Faust, I. C.; Porkolab, M.; Whyte, D. G.

    2015-04-01

    Parametric decay instabilities (PDI) appear to be an ubiquitous feature of lower hybrid current drive (LHCD) experiments at high density. In density ramp experiments in Alcator C-Mod and other machines the onset of PDI activity has been well correlated with a decrease in current drive efficiency and production of fast electron bremsstrahlung. However whether PDI is the primary cause of the ‘density limit’, and if so by exactly what mechanism (beyond the obvious one of pump depletion) has not been clearly established. In order to further understand the connection, the frequency spectrum of PDI activity occurring during Alcator C-Mod LHCD experiments has been explored in detail by means of a number of RF probes distributed around the periphery of the C-Mod tokamak including a probe imbedded in the inner wall. The results show that (i) the excited spectra consists mainly of a few discrete ion cyclotron (IC) quasi-modes, which have higher growth than the ion sound branch; (ii) PDI activity can begin either at the inner or outer wall, depending on magnetic configuration; (iii) the frequencies of the IC quasi-modes correspond to the magnetic field strength close to the low-field side (LFS) or high-field side separatrix; and (iv) although PDI activity may initiate near the inner separatrix, the loss in fast electron bremsstrahlung is best correlated with the appearance of IC quasi-modes characteristic of the magnetic field strength near the LFS separatrix. These data, supported by growth rate calculations, point to the importance of the LFS scrape-off layer (SOL) density in determining PDI onset and degradation in current drive efficiency. By minimizing the SOL density it is possible to extend the core density regime over which PDI can be avoided, thus potentially maximizing the effectiveness of LHCD at high density. Increased current drive efficiency at high density has been achieved in FTU and EAST through lithium coating and special fuelling methods, and in recent

  17. Effects of the q Profile on Toroidal Rotation in Alcator C-Mod LHCD Plasmas

    NASA Astrophysics Data System (ADS)

    Rice, John

    2015-11-01

    Changes in the core toroidal rotation velocity profiles following injection of lower hybrid (LH) waves have been documented in Alcator C-Mod plasmas. Shot by shot scans of LH input power have been performed at fixed magnetic field and electron density for several plasma currents. If the input power is low enough that there are still sawtooth oscillations, the change in the core rotation is in the counter-current direction, consistent in sign and magnitude with direct momentum input from the LH waves. If the power level is high enough that there are significant changes to the q profile, the change in the toroidal rotation is in the co-current direction, consistent with changes in the residual stress through its dependence on the current density profile. The direction of the rotation changes depends on the whether q0 is below or above unity, and seemingly not on the magnetic shear.

  18. RF current profile control studies in the alcator C-mod tokamak

    NASA Astrophysics Data System (ADS)

    Bonoli, P. T.; Porkolab, M.; Wukitch, S. J.; Bernabei, S.; Kaita, R.; Mikkelsen, D.; Phillips, C. K.; Schilling, G.

    1999-09-01

    Time dependent calculations of lower hybrid (LH) current profile control in Alcator C-Mod have been done using the TRANSP [1], FPPRF [2], and LSC [3] codes. Up to 3 MW of LH current drive power was applied in plasmas with high power ICRF minority heating (PICH=1.8-3 MW) and fast current ramp up. Using the experimentally measured temperature profiles, off-axis current generation resulted in nonmonotonic q-profiles with qmin~=1.6. Self-consistent effects of off-axis electron heating by the LH power were also included in the analysis and significant broadening of the electron temperature profile was found with qmin>~2 and a larger shear reversal radius.

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

    SciTech Connect

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

    2001-05-04

    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.

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

    SciTech Connect

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

    2011-07-25

    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 {approx}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).

  1. Divertor heat flux footprints in EDA H-mode discharges on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Labombard, B.; Terry, J. L.; Hughes, J. W.; Brunner, D.; Payne, J.; Reinke, M. L.; Lin, Y.; Wukitch, S.

    2011-08-01

    The physics that sets the width of the power exhaust channel in a tokamak scrape-off layer and its scaling with engineering parameters is of fundamental importance for reactor design, yet it remains to be understood. An extensive array of divertor heat flux diagnostics was recently commissioned in Alcator C-Mod with the aim of improving our understanding. Initial results are reported from EDA H-mode discharges in which plasma current, input power, toroidal field and magnetic topology were varied. The integral width of the outer divertor heat flux footprint is found to lie in the range of 3-5 mm mapped to the mid-plane. Widths are insensitive to single versus double-null topology and the magnitude of toroidal field. Pedestal physics appears to largely determine these widths; a dependence of width on plasma thermal energy is noted, yielding a reduction in width as plasma current is increased for the best EDA H-modes.

  2. First results of the SOL reflectometer on Alcator C-Mod.

    PubMed

    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 ∼10(16)-10(20) m(-3) at B(0) = 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. PMID:23126969

  3. Control of Internal Profiles via LHCD on Alcator C-Mod

    SciTech Connect

    Wilson, J. R.; Kessel, C.; Scott, S.; Wallace, G. M.; Parker, R. R.; Bonoli, P. T.; Hubbard, A. E.; Hughes, J. W.; Ince-Cushman, A.; Ko, J. S.; Meneghini, O.; Porkolab, M.; Reinke, M.; Rice, J. E.; Schmidt, A. E.; Shiraiwa, S.; Wright, J. C.

    2009-11-26

    LHCD on Alcator C-Mod is being used in plasmas with parameters similar to those expected on ITER for the purpose of tailoring the plasma current profile. LHCD experiments have also produced intriguing results related to momentum transport and edge pedestal physics that affect the toroidal rotation profile and the temperature and density profiles. Quantitative comparisons between local measurements and theory/simulation have been performed, confirming the off-axis localization of the current drive, as well as its magnitude and location dependence on the launched n{sub parallel} spectrum and electron temperature. Applying LHCD during the current ramp saves volt-seconds and delays the peaking of the current profile. Counter current toroidal rotation during LHCD has been observed in both L and H-mode plasmas. In H-mode plasmas the edge pedestal collisionality is reduced while the overall pressure in the pedestal increases slightly.

  4. Design of a correlation electron cyclotron emission diagnostic for Alcator C-Mod

    SciTech Connect

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

    2012-10-15

    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 k{sub {theta}}{<=} 4.8 cm{sup -1} (k{sub {theta}}{rho}{sub s} < 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.

  5. Design of a correlation electron cyclotron emission diagnostic for Alcator C-Mod.

    PubMed

    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 k(θ) ≤ 4.8 cm(-1) (k(θ)ρ(s) < 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. PMID:23126971

  6. First Results From the Alcator C-Mod Lower Hybrid Experiment

    NASA Astrophysics Data System (ADS)

    Parker, Ronald; Bernabei, Stefano; Grimes, Montgomery; Hosea, Joel; Johnson, David; Wilson, Randy

    2005-10-01

    A lower hybrid system operating at 4.6 GHz and capable of 3 MW source power has been installed on Alcator C-Mod. The grill facing the plasma consists of 4 rows of 24 waveguides. Electronic control of the amplitude and phase of the 12 klystrons supplying the RF power enables the launcher's n|| spectrum to be dynamically controlled over a wide range with a time response of 1 ms. Since the deposition of current depends on n|| as well as the temperature profile, the spatial distribution of the driven current can be varied with the same time response. Detection of fast electron Bremsstrahlung is the primary means of monitoring the driven current profile. Initial measurements at the 100 kW power level show that reflection coefficients as low as 7% are obtained at optimal phasing and density at the grill mouth. Comparison of these results with modeling predictions will be presented in a companion paper.

  7. Multi-scale gyrokinetic simulation of Alcator C-Mod tokamak discharges

    NASA Astrophysics Data System (ADS)

    Howard, N. T.; White, A. E.; Greenwald, M.; Holland, C.; Candy, J.

    2014-03-01

    Alcator C-Mod tokamak discharges have been studied with nonlinear gyrokinetic simulation simultaneously spanning both ion and electron spatiotemporal scales. These multi-scale simulations utilized the gyrokinetic model implemented by GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and the approximation of reduced electron mass (μ = (mD/me).5 = 20.0) to qualitatively study a pair of Alcator C-Mod discharges: a low-power discharge, previously demonstrated (using realistic mass, ion-scale simulation) to display an under-prediction of the electron heat flux and a high-power discharge displaying agreement with both ion and electron heat flux channels [N. T. Howard et al., Nucl. Fusion 53, 123011 (2013)]. These multi-scale simulations demonstrate the importance of electron-scale turbulence in the core of conventional tokamak discharges and suggest it is a viable candidate for explaining the observed under-prediction of electron heat flux. In this paper, we investigate the coupling of turbulence at the ion (kθρs˜O(1.0)) and electron (kθρe˜O(1.0)) scales for experimental plasma conditions both exhibiting strong (high-power) and marginally stable (low-power) low-k (kθρs < 1.0) turbulence. It is found that reduced mass simulation of the plasma exhibiting marginally stable low-k turbulence fails to provide even qualitative insight into the turbulence present in the realistic plasma conditions. In contrast, multi-scale simulation of the plasma condition exhibiting strong turbulence provides valuable insight into the coupling of the ion and electron scales.

  8. Multi-scale gyrokinetic simulation of Alcator C-Mod tokamak discharges

    SciTech Connect

    Howard, N. T. White, A. E.; Greenwald, M.; Holland, C.; Candy, J.

    2014-03-15

    Alcator C-Mod tokamak discharges have been studied with nonlinear gyrokinetic simulation simultaneously spanning both ion and electron spatiotemporal scales. These multi-scale simulations utilized the gyrokinetic model implemented by GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and the approximation of reduced electron mass (μ = (m{sub D}/m{sub e}){sup .5} = 20.0) to qualitatively study a pair of Alcator C-Mod discharges: a low-power discharge, previously demonstrated (using realistic mass, ion-scale simulation) to display an under-prediction of the electron heat flux and a high-power discharge displaying agreement with both ion and electron heat flux channels [N. T. Howard et al., Nucl. Fusion 53, 123011 (2013)]. These multi-scale simulations demonstrate the importance of electron-scale turbulence in the core of conventional tokamak discharges and suggest it is a viable candidate for explaining the observed under-prediction of electron heat flux. In this paper, we investigate the coupling of turbulence at the ion (k{sub θ}ρ{sub s}∼O(1.0)) and electron (k{sub θ}ρ{sub e}∼O(1.0)) scales for experimental plasma conditions both exhibiting strong (high-power) and marginally stable (low-power) low-k (k{sub θ}ρ{sub s} < 1.0) turbulence. It is found that reduced mass simulation of the plasma exhibiting marginally stable low-k turbulence fails to provide even qualitative insight into the turbulence present in the realistic plasma conditions. In contrast, multi-scale simulation of the plasma condition exhibiting strong turbulence provides valuable insight into the coupling of the ion and electron scales.

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

    NASA Astrophysics Data System (ADS)

    Mikkelsen, D. R.; Bitter, M.; Delgado-Aparicio, L.; Hill, K. W.; Greenwald, M.; Howard, N. T.; Hughes, J. W.; Rice, J. E.; Reinke, M. L.; Podpaly, Y.; Ma, Y.; Candy, J.; Waltz, R. E.

    2015-06-01

    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.

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

    SciTech Connect

    Mikkelsen, D. R. Bitter, M.; Delgado-Aparicio, L.; Hill, K. W.; Greenwald, M.; Howard, N. T.; Hughes, J. W.; Rice, J. E.; Reinke, M. L.; Podpaly, Y.; Ma, Y.; Candy, J.; Waltz, R. E.

    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.

  11. Vertical localization of phase contrast imaging diagnostic in Alcator C-Mod

    SciTech Connect

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

    2006-10-15

    Phase contrast imaging (PCI) diagnostic has been used to study mode conversion physics of ion cyclotron range of frequency waves [E. Nelson-Melby et al., Phys. Rev. Lett. 90, 155004 (2003)], plasma turbulence [A. Mazurenko et al., Phys. Rev. Lett. 89, 225004 (2002); N. Basse et al., Phys. Plasmas 12, 052512 (2005)], and Alfven Cascades [M. Porkolab et al., IEEE Trans. Plasma Sci. 34, 229 (2006)] in Alcator C-Mod. The C-Mod PCI system measures line integrated electron density fluctuations along 32 vertical chords with a sampling frequency of 10 MHz and wavenumber resolution up to 30 cm{sup -1}. Although PCI normally lacks localization along the chords, the vertical variation of the magnetic field pitch angle allows for localized measurements for large k{sub perpendicular} fluctuations. A system consisting of a partially masked phase plate on a rotatable stage has been installed and quasicoherent modes with wave number {approx}5 cm{sup -1} associated with the enhanced D{alpha}H mode at the top and bottom of the plasma have been differentiated. In future experiments, for k{approx}30 cm{sup -1}, a spatial resolution of r/a{approx}0.3 can be achieved under ideal conditions.

  12. Spectroscopic measurement of impurity transport coefficients and penetration efficiencies in Alcator C-Mod plasmas

    NASA Astrophysics Data System (ADS)

    Graf, M. A.; Rice, J. E.; Terry, J. L.; Marmar, E. S.; Goetz, J. A.; McCracken, G. M.; Bombarda, F.; May, M. J.

    1995-01-01

    Impurity transport coefficients and the penetration efficiencies of intrinsic and injected impurities through the separatrix of diverted Alcator C-Mod discharges have been measured using x-ray and vacuum ultraviolet (VUV) spectroscopic diagnostics. The dominant low Z intrinsic impurity in C-Mod is carbon which is found to be present in concentrations of less than 0.5%. Molybdenum, from the plasma facing components, is the dominant high Z impurity and is typically found in concentrations of about 0.02%. Trace amounts of medium and high Z nonrecycling impurities can be injected at the midplane using the laser blow-off technique and calibrated amounts of recycling, gaseous impurities can be introduced through fast valves either at the midplane or at various locations in the divertor chamber. A five chord crystal x-ray spectrometer array with high spectral resolution is used to provide spatial profiles of high charge state impurities. An absolutely calibrated, grazing incidence VUV spectrograph with high time resolution and a broad spectral range allows for the simultaneous measurement of many impurity lines. Various filtered soft x-ray diode arrays allow for spatial reconstructions of plasma emissivity. The observed brightnesses and emissivities from a number of impurity lines are used together with the mist transport code and a collisional-radiative atomic physics model to determine charge state density profiles and impurity transport coefficients. Comparisons of the deduced impurity content with the measured Zeff and total radiated power of the plasma are made.

  13. Wall scanning probe for high-field side plasma measurements on Alcator C-Mod.

    PubMed

    Smick, Noah; LaBombard, Brian

    2009-02-01

    A new, high-field side scanning probe has been added to Alcator C-Mod's complement of edge diagnostics. The wall scanning probe is designed to provide all the benefits of a linear plunge, multielectrode scanning probe while working from the confined space of the inner tokamak wall. The drive mechanism is an embedded coil which produces a torque with the ambient toroidal magnetic field when energized, thus allowing the probe to plunge to different preprogramed depths at different times during a plasma discharge. The probe tip is designed for easy replacement and is presently configured to operate as a modified, high heat-flux "Gundestrup-type" probe with four tungsten electrodes. The probe has demonstrated the ability to obtain cross-field profiles for electron temperature, density, floating potential, and plasma flow information (parallel and perpendicular to B) up to a depth of a few millimiters inside the last-closed flux surface in standard C-Mod discharges. The tungsten-tipped probe has proved very robust and shows little or no damage though it routinely handles surface heat fluxes on the order of 100 MW/m(2) at peak insertion. PMID:19256643

  14. Perturbative Thermal Transport Studies on Alcator C-Mod and ASDEX Upgrade

    NASA Astrophysics Data System (ADS)

    Creely, A. J.; White, A. E.; Edlund, E. M.; Howard, N. T.; Hubbard, A. E.; Ryter, F.

    2015-11-01

    Perturbative thermal diffusivity has been measured on Alcator C-Mod and ASDEX Upgrade via the extended-time-to-peak method with heat pulses generated by partial sawtooth crashes. Heat pulses generated by sawtooth crashes have been used extensively in the past to study perturbative diffusivity, but the details of the sawtooth event lead to non-diffusive ``ballistic'' transport, invalidating their use for measuring perturbative diffusivity. Partial sawteeth generate a heat pulse without the ballistic transport of full sawteeth [Fredrickson 2000]. Partial sawtooth analysis was applied to over 50 C-Mod shots containing both L- and I-Mode, as well as ASDEX Upgrade plasmas, though partial sawteeth were less common on AUG. Results indicate correlations between perturbative diffusivity and confinement regime (L- vs. I-mode), as well as correlations with local temperature, density, the associated gradients, and gradient scale lengths (a/LTe and a/Ln). Finally, diffusivities calculated from partial sawteeth were compared to perturbative diffusivities calculated with the GYRO gyrokinetic code, leading to quantitative agreement with multi-scale GYRO simulations. This work is supported by the US DOE under grants DE-SC0006419 and DE-FC02-99ER54512-CMOD.

  15. Advanced Tokamak Regimes in Alcator C-Mod with Lower Hybrid Current Drive

    NASA Astrophysics Data System (ADS)

    Parker, R.; Bonoli, P.; Gwinn, D.; Hutchinson, I.; Porkolab, M.; Ramos, J.; Bernabei, S.; Hosea, J.; Wilson, R.

    1999-11-01

    Alcator C-Mod has been proposed as a test-bed for developing advanced tokamak scenarios owing to its strong shaping, relatively long pulse length capability at moderate field, e.g. t ~ L/R at B = 5T and T_eo ~ 7keV, and the availability of strong ICRF heating. We plan to exploit this capability by installing up to 4 MW RF power at 4.6 GHz for efficient off-axis current drive by lower hybrid waves. By launching LH waves with a grill whose n_xx spectrum can be dynamically controlled over the range 2 < n_xx < 3.5, the driven current profile can be modified so that, when combined with bootstrap current in high ɛβ_pol regimes, q_min > 2. Such reversed or nearly zero shear regimes have already been proposed as the basis of an advanced tokamak burning-plasma experiment-ATBX (M. Porkolab et al, IAEA-CN-69/FTP/13, IAEA,Yokohama 1998.), and could provide the basis for a demonstration power reactor. Theoretical and experimental basis for this advanced tokamak research program on C-Mod, including design of the lower hybrid coupler, its spectrum and current drive capabilities will be presented.

  16. First measurements of core electron temperature fluctuations in Alcator C-Mod via Correlation ECE

    NASA Astrophysics Data System (ADS)

    Sung, C.; White, A.; Howard, N.; Oi, C.; Rice, J.; Gao, C.; Ennever, P.; Porkolab, M.; Hubbard, A.; Greenwald, M.

    2012-10-01

    A new Correlation Electron Cyclotron Emission (CECE) diagnostic has been installed at Alcator C-Mod to measure broadband, turbulent electron temperature fluctuations in the core region (ρ<0.9). This diagnostic has high poloidal resolution (φ0< 0.5cm ), and can measure long wavelength turbulence (kθρs< 0.4). We obtained the first electron temperature fluctuation data in Ohmic and ICRH plasmas. It was observed that the core (ρ˜0.8) electron temperature fluctuation level decreases as the Ohmic confinement regime transitions from Linear Ohmic Confinement (LOC) to Saturated Ohmic Confinement (SOC), while edge fluctuation levels (ρ> 0.9) stay similar in amplitude. These first quantitative, local electron temperature fluctuation measurements in C-Mod demonstrate that core turbulence characteristics change between LOC and SOC. In order to interpret the data, linear stability analysis is being performed, and the relation between fluctuation level and other plasma parameters is being investigated. This talk will describe the new CECE diagnostic, the SOC/LOC measurements, and associated gyrokinetic analyses.

  17. Modification of Current Profile, Toroidal Rotation and Pedestal by Lower Hybrid Waves in Alcator C-Mod

    SciTech Connect

    Parker, R.; Bonoli, P. T.; Meneghini, O.; Porkolab, M.; Schmidt, A. E.; Shiraiwa, S.; Wallace, G.; Hubbard, A. E.; Hughes, J. W.; Ko, J.-S.; McDermott, R. M.; Reinke, M. L.; Rice, J. E.; Wilson, J. R.; Scott, S.

    2009-11-26

    Recent results from the lower hybrid current drive experiments on Alcator C-Mod are presented. These include i) MSE measurements of broadened LHCD current profiles; ii) development of counter rotation comparable to the rate of injected wave momentum; iii) modification of pedestals and rotation in H-mode; and iv) development of a new FEM-based code that models LH wave propagation from the RF source to absorption in the plasma. An improved antenna concept that will be used in the upcoming C-Mod campaigns is also briefly described.

  18. The Submillimeter Wave Electron Cyclotron Emission Diagnostic for the Alcator C-Mod Tokamak.

    NASA Astrophysics Data System (ADS)

    Hsu, Thomas C.

    This thesis describes the engineering design, construction, and operation of a high spatial resolution submillimeter wave diagnostic for electron temperature measurements on Alcator C-Mod. Alcator C-Mod is a high performance compact tokamak capable of producing diverted, shaped plasmas with a major radius of 0.67 meters, minor radius of 0.21 centimeters, plasma current of 3 MA. The maximum toroidal field is 9 Tesla on the magnetic axis. The ECE diagnostic includes three primary components: a 10.8 meter quasioptical transmission line, a rapid scanning Michelson interferometer, and a vacuum compatible calibration source. Due to the compact size and high field of the tokamak the ECE system was designed to have a spectral range from 100 to 1000 GHz with frequency resolution of 5 GHz and spatial resolution of one centimeter. The beamline uses all reflecting optical elements including two off-axis parabolic mirrors with diameters of 20 cm. and focal lengths of 2.7 meters. Techniques are presented for grinding and finishing the mirrors to sufficient surface quality to permit optical alignment of the system. Measurements of the surface figure confirm the design goal of 1/4 wavelength accuracy at 1000 GHz. Extensive broadband tests of the spatial resolution of the ECE system are compared to a fundamental mode Gaussian beam model, a three dimensional vector diffraction model, and a geometric optics model. The Michelson interferometer is a rapid scanning polarization instrument which has an apodized frequency resolution of 5 GHz and a minimum scan period of 7.5 milliseconds. The novel features of this instrument include the use of precision linear bearings to stabilize the moving mirror and active counterbalancing to reduce vibration. Beam collimation within the instrument is done with off-axis parabolic mirrors. The Michelson also includes a 2-50 mm variable aperture and two signal attenuators constructed from crossed wire grid polarizers. To make full use of the advantages

  19. Lower Hybrid Heating and Current Drive on the Alcator C-Mod Tokamak

    SciTech Connect

    R. Wilson, R. Parker, M. Bitter, P.T. Bonoli, C. Fiore, R.W. Harvey, K. Hill, A.E. Hubbard, J.W. Hughes, A. Ince-Cushman, C. Kessel, J.S. Ko, O. Meneghini, C.K. Phillips, M. Porkolab, J. Rice, A.E. Schmidt, S. Scott,S. Shiraiwa, E. Valeo, G.Wallace, J.C. Wright and the Alcator C-Mod Team

    2009-11-20

    On the Alcator C-Mod tokamak, lower hybrid current drive (LHCD) is being used to modify the current profile with the aim of obtaining advanced tokamak (AT) performance in plasmas with parameters similar to those that would be required on ITER. To date, power levels in excess of 1 MW at a frequency of 4.6 GHz have been coupled into a variety of plasmas. Experiments have established that LHCD on C-Mod behaves globally as predicted by theory. Bulk current drive efficiencies, n20IlhR/Plh ~ 0.25, inferred from magnetics and MSE are in line with theory. Quantitative comparisons between local measurements, MSE, ECE and hard x-ray bremsstrahlung, and theory/simulation using the GENRAY, TORIC-LH CQL3D and TSC-LSC codes have been performed. These comparisons have demonstrated the off-axis localization of the current drive, its magnitude and location dependence on the launched n|| spectrum, and the use of LHCD during the current ramp to save volt-seconds and delay the peaking of the current profile. Broadening of the x-ray emission profile during ICRF heating indicates that the current drive location can be controlled by the electron temperature, as expected. In addition, an alteration in the plasma toroidal rotation profile during LHCD has been observed with a significant rotation in the counter current direction. Notably, the rotation is accompanied by peaking of the density and temperature profiles on a current diffusion time scale inside of the half radius where the LH absorption is taking place.

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

    SciTech Connect

    Ochoukov, R.; Whyte, D. G.; Brunner, D.; LaBombard, B.; Lipschultz, B.; Terry, J. L.; Wukitch, S. J.; D'Ippolito, D. A.; Myra, J. R.

    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.

  1. The development of an Omegratron plasma ion mass spectrometer for Alcator C-Mod

    SciTech Connect

    Thomas, E.E. Jr.

    1993-05-01

    A new diagnostic device, the Omegatron Probe, has been developed to investigate relative impurity levels and impurity charge state distribution in the Alcator C-Mod Tokamak edge plasma. The Omegatron probe consists of two principal components, a ``front-end`` of independently biased grids, arranged in a gridded energy analyzer fashion and a large collection cavity. Particles enter the probe in a thin ``ribbon`` through a knife-edge slit. The grids provide a means to measure and control the parallel energy distribution of the ions. In the collection cavity, an oscillating electric field is applied perpendicularly to the ambient magnetic field. Ions whose cyclotron frequencies are resonant with this electric field oscillation will gain perpendicular energy and be collected. In this way, the probe can be operated in two modes: first, by fixing the potentials on the grids and sweeping frequencies to obtain a `` Z/m spectrum`` of ion species and second, by fixing the frequency and sweeping the grid potentials to obtain the distribution function of an individual impurity species. The Omegatron probe performed successfully in tests on a Hollow Cathode Discharge (HCD) linear plasma column. It obtained measurements of T{sub e} {approx} 5 eV, T{sub i} (H{sup +}) {approx} 2.0 {plus_minus} 0.2 eV, n{sub 0} {approx} 9 {times} 10{sup 15} m{sup {minus}3}, RMS potential fluctuation levels of {approximately} 0.5 {plus_minus} 0.05 {plus_minus} T{sub e}, and obtained ``Z/m`` spectra for the plasma ions (H{sup +}, H{sub 2}{sup +}, He{sup +}). Additional experiments confirmed the theoretical scalings of the f/{delta}f resolution with the applied electric field and magnetic field strengths. The instrument yielded an absolute level of resolution, f/{delta}f, of approximately 2.5 to 3 times the theoretical values. Finally, the results from the HCD are used to project operation on Alcator C-Mod.

  2. Lower Hybrid wave edge power loss quantification on the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Faust, I. C.

    2015-11-01

    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 <τE) response of the scrape-off-layer (SOL) plasma to 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 key for the LHRF edge power deposition physics. These observations support the existence a loss mechanism near the edge for LHRF at high density (ne > 1 . 0 .1020 [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 motivates the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch. This work was performed on the Alcator C-Mod tokamak, a DoE Office of Science user facility, and is supported by USDoE award DE-FC02-99ER54512.

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

    SciTech Connect

    Delgado-Aparicio, L.; Bitter, M.; Gates, D.; Hill, K.; Pablant, N.; Granetz, R.; Reinke, M.; Podpaly, Y.; Rice, J.; Beiersdorfer, P.; Sugiyama, L.

    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.

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

    SciTech Connect

    Loarte, A.; Polevoi, A. R.; Hosokawa, M.; Reinke, M. L.; Chilenski, M.; Howard, N.; Hubbard, A.; Hughes, J. W.; Rice, J. E.; Walk, J.; Köchl, F.; Pütterich, T.; Dux, R.; Zhogolev, V. E.

    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.

  5. Ion Temperature and Plasma Rotation in EDA H-Mode and ITB Discharges in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Rowan, William L.; Bravenec, R. V.; Phillips, P. E.; Sampsell, M. B.; Granetz, R. S.; Lipschultz, B.; McDermott, R. M.

    2004-11-01

    In the course of the 2004 campaign on Alcator C-Mod, extensive ion temperature and plasma rotation measurements were made in the outer half of the plasma via CXRS and spectroscopy of ambient species. The most significant of these results are for the unique C-Mod modes: the EDA H-Mode and the RF induced internal transport barrier. The CXRS data was taken with a 50 ms beam pulse which occurs just once during the typical 1.5 s C-Mod discharge. Ion temperature and rotation are inferred from measurements of ambient spectra to fill out the time series for a discharge. Due regard is given to spatial averaging in using this data. Consistency of the data is checked where possible via the momentum balance equation. Thermal transport analysis (TRANSP) is included as well. A long pulse beam is planned for installation in fall 2004. Expected improvements in CXRS will be discussed.

  6. Compact multichannel neutral particle analyzer for measurement of energetic charge-exchanged neutrals in Alcator C-Mod

    SciTech Connect

    Tang, V.; Liptac, J.; Parker, R. R.; Bonoli, P. T.; Fiore, C. L.; Granetz, R. S.; Irby, J. H.; Lin, Y.; Wukitch, S. J.; Frenje, J. A.; Leiter, R.; Mcduffee, S.; Petrasso, R. D.

    2006-08-15

    A four-channel compact neutral particle analyzer (CNPA) based on operating small Si diode detectors in pulse-height analysis (PHA) mode is used to measure energetic hydrogen minority ions with energies between {approx}50 and 350 keV stemming from ion-cyclotron range-of-frequency heated D(H) Alcator C-Mod plasmas with both active and passive charge exchange (CX). First core minority ion distribution results from Alcator C-Mod discharges and a detailed description of the diagnostic are presented. The diagnostic employs integrated electronics and fast digitization of the shaping amplifier voltage. The digitized data are stored for postshot PHA, which removes the constraints of real-time PHA and allows for improved performance via elimination of base line shift effects and potentially relieving pileup through Gaussian fitting routines. The CNPA is insensitive to the large gamma and neutron background in Alcator C-Mod discharges but is susceptible to the plasma's soft x-ray flux. The soft x-ray flux limits the CNPA energy resolution to {approx}15-20 keV. A simple model is used to interpret the active CNPA data which permits rapid estimates of the core hydrogen minority temperatures and anisotropy with a time resolution of {approx}100 ms. Hydrogenlike boron is identified as an important electron donor for the CX signal.

  7. Studies of Electron Temperature Fluctuations in the Core of Alcator C-Mod Plasmas via Correlation ECE

    NASA Astrophysics Data System (ADS)

    Oi, C.; Sung, C.; Howard, N.; White, A. E.; Irby, J.; Leccacorvi, R.; Vieira, R.; Rice, J.; Gao, C.

    2012-10-01

    A correlation electron cyclotron emission (CECE) diagnostic has been designed and installed at the Alcator C-Mod tokamak to measure long wavelength fluctuations in electron temperature [C. Sung, this conference, C-Mod oral session]. It is important to characterize turbulent fluctuations in the plasma in order to better understand and predict transport, since the cross-magnetic field transport of particles and energy in fusion plasmas exceeds the values predicted by neoclassical theory. The first electron temperature fluctuation data collected from Alcator C-Mod ohmic plasmas has shown that as the ohmic confinement regime transitions from Linear Ohmic Confinement (LOC) to Saturated Ohmic Confinement (SOC) there is a decrease in the core electron temperature fluctuation level. The edge electron temperature fluctuations are similar between the two modes, suggesting that mainly the core turbulence characteristics change when there is a shift between LOC and SOC regimes. In both ohmic and ICRH plasmas, the measured spectrum is seen to broaden in response to increases in plasma rotation; and fluctuation level increases with radius. A description of the C-Mod CECE diagnostic will be presented, along with the initial measurements of turbulent electron temperature fluctuations.

  8. Imaging of X-point turbulence in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Ballinger, Sean; Terry, James; White, Anne; Zweben, Stewart

    2015-11-01

    A nearly tangential view of the lower X-point region of Alcator C-Mod has been coupled to a high-speed camera filtered for D-alpha line emission. Recording at ~400,000 frames per second, the system detects filaments propagating in the private flux region that are approximately aligned with the local magnetic field. This behavior appears similar to what has recently been observed in the MAST tokamak. Turbulence and transport into the private flux region is potentially important. It may be a mechanism to spread heat across field lines and reduce peak heat fluxes on divertor targets. It may also explain how transport-driven flows seen in the high-field side scrape-off layer are accommodated, being otherwise too large compared to the particle flux arriving at the inner divertor target plates. The dynamics of these filaments are analyzed, as is the rate at which they are generated. Correlation analysis is used to determine the speed and trajectories of the filaments. Radial speeds of ~1 km/s are found. Clear changes are observed in the X-point-region fluctuations at the L-to-H-mode transition.

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

    NASA Astrophysics Data System (ADS)

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

    2009-11-01

    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.

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

    SciTech Connect

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

    2009-11-26

    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.

  11. Collisionality dependence of impurity transport in Alcator C-Mod H-modes

    NASA Astrophysics Data System (ADS)

    Chilenski, M. A.; Greenwald, M.; Howard, N. T.; Reinke, M. L.; Rice, J.; White, A. E.; Marzouk, Y.

    2014-10-01

    Understanding and actuating impurity transport is of particular interest for future machines because of the concern that core accumulation of heavy impurities will lead to radiative collapse and higher disruptivity. This problem is expected to be especially pronounced at low collisionality, where a strong peaking of the electron density profile has previously been observed (Greenwald et al., Nucl. Fusion 47, L26 (2007)). To investigate this issue several experiments have been performed in Alcator C-Mod to measure the behavior of mid-Z (Ar, Ca) and high-Z (Mo, W) impurities in H-mode plasmas of varying collisionality (2 <νeff < 40). These plasmas are of particular interest to this problem because they are entirely RF heated and lack core particle sources. Impurities are injected using laser blow-off or gas injection and the evolution of the impurity density profile is constrained with an X-ray imaging crystal spectrometer. These diagnostics combined with analysis using STRAHL allows detailed study of the transport. Furthermore, analysis of the background ne, Te profiles is conducted using advanced techniques including Gaussian process regression. An outline of this analysis scheme will be presented and recent results obtained from its application will be shown. Supported by USDOE Award DE-FC02-99ER54512.

  12. Effects of the q profile on toroidal rotation in Alcator C-Mod LHCD plasmas

    NASA Astrophysics Data System (ADS)

    Rice, J. E.; Gao, C.; Mumgaard, R.; Parker, R. R.; Scott, S. D.; Shiraiwa, S.; Wallace, G. M.; Bonoli, P. T.; Delgado-Aparicio, L.; Fenzi, C.; Granetz, R. S.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Lee, J. P.; Marmar, E. S.; Reinke, M. L.; Wolfe, S. M.

    2016-03-01

    Changes in the core toroidal rotation profiles following injection of lower hybrid (LH) waves have been documented in Alcator C-Mod plasmas. Shot by shot scans of LH input power have been performed at fixed magnetic field and electron density for several plasma currents. For sawtoothing target plasmas, if the input power is low enough that the central safety factor q 0 remains below 1, the change in the core rotation is in the counter-current direction, consistent in sign, magnitude and LH power scaling with direct momentum input from the LH waves. If the power level is high enough that there are significant changes to the q profile, including the termination of sawtooth oscillations, the change in the toroidal rotation is in the co-current direction, consistent with changes in the momentum flux through its dependence on the current density profile. The direction of the rotation changes depends on whether q 0 is below or above unity, and seemingly not on the magnetic shear, nor the Ohmic confinement regime of the target plasma.

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

    SciTech Connect

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

    2002-05-07

    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.

  14. ICRF heating in Alcator C-Mod: Present status and future prospects

    SciTech Connect

    Porkolab, M.; Fiore, C.; Greenwald, M.; Hosea, J. C.; Hubbard, A.; Hutchinson, I.; Irby, J.; Nelson-Melby, E.; Marmar, E.; Phillips, C. K.

    1999-09-20

    Alcator C-Mod, the high field, high density, diverted, compact tokamak in the world's portfolio of high performance plasma fusion devices, is heated exclusively with ICRF auxiliary power. In this paper an overview of recent results is summarized, with particular attention given to the importance of RF operation and the flexibility afforded by different heating scenarios. Besides the routine minority heating operation, results in the mode conversion heating regime are also presented (mainly direct electron heating through mode converted ion Bernstein waves). Recent attempts at improving plasma performance by establishing internal transport barriers (ITBs) by various transient profile control techniques (the so-called Advanced Tokamak mode of operation) are also presented. Future improvements in performance afforded by the recent addition of a new 4-strap antenna and 4 MW of tunable (40-80 MHz) ICRF power are also discussed. Mode-conversion current drive (MCCD) and fast wave current drive (FWCD) will be among the many new options that will be tested with the goal of improving plasma performance. (c) 1999 American Institute of Physics.

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

    SciTech Connect

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

    2001-06-12

    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.

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

    SciTech Connect

    G. Schilling; S.J. Wukitch; R.L. Boivin; J.A. Goetz; J.C. Hosea; J.H. Irby; Y. Lin; A. Parisot; M. Porkolab; J.R. Wilson; the Alcator C-Mod Team

    2003-07-31

    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 {approx}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.

  17. Status of the Alcator C-MOD scanning two-dimensional Thomson scattering diagnostic

    SciTech Connect

    Watterson, R. ); Chen, K. )

    1990-10-01

    The detailed study of noncircular tokamak discharges will require measurements of local {ital T}{sub {ital e}} and {ital n}{sub {ital e}} over a two-dimensional region. A novel scanning Thomson scattering system is under construction for use on the Alcator C-Mod device. A multipulse (50 Hz pulse rate) Nd:YAG laser (1 J, 15 nS) will be used to scan the plasma major radius within a 14 cm range during a 160 mS period with a spatial step size of 2 cm. The laser beam sweep, which is limited by the size of the diagnostic ports, will cover {approximately}50% of the plasma discharge cross section. The scattered laser light spectra at various positions along the vertical laser beam path will be analyzed by a set of up to 11 (initial operation is planned for six) Wadsworth spectrometers. A prototype spectrometer with associated light collection optics has been tested and the total Thomson system optical transmission has been measured to be {approximately}12--15%. System details and performance measurements will be presented.

  18. Design of off-midplane launcher (LH3) for Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Shiraiwa, S.; Bonoli, P. T.; Hillairet, J.; Meneghini, O.; Parker, R. R.; Wallace, G. M.; Wilson, J. R.; Harvey, R. W.; Smirnov, A. P.

    2012-10-01

    Improving LHCD efficiency at density above 10^20m-3 is crucial for approaching AT regimes on Alcator C-Mod. In this density regime, an unexpected loss of LHCD efficency has been observed. Modeling and experiments suggest that such loss is intrinsic to the strong multipass absorption regime and can be overcome by improving the wave single pass absorption. For this purpose, an additional launcher (LH3) was designed. Besides doubling total injected LH power, LH3 will be located off-midplane and will enhance the waves single pass absorption by velocity space synergy with the existing launcher (LH2). The poloidal location and launched N// were selected based on a large parameter scan using a ray-tracing code. About 300kA of LH driven current was predicted at 1.4e20m-3 using 1.3MW of forward power. LH3 has a grill of 4x16 active waveguides and is based on a eight-way splitter design, which splits the microwave power into four ways in the poloidal direction and two ways in the toroidal direction. Coupling studies using commercial FEM software and the ALOHA code predict good coupling in a wide range of N// (from 2.2 to 2.7). Details of RF design and progress of launcher fabrication will be reported.

  19. Rapid shutdown experiments with one and two gas jets on Alcator C-Mod

    SciTech Connect

    OlynykPlasma Scien, G.; Granetz, R. S.; Reinke, M. L.; Whyte, D. G.; Golfinopoulos, T.; Hughes, J.; Walk, J.; Izzo, V. A.; Combs, Stephen Kirk; Milora, Stanley L; Brookman, M. W.

    2013-01-01

    Massive gas injection rapid shutdown experiments have been conducted on the Alcator C-Mod tokamak using two toroidally separated gas injectors, in order to investigate the effect of multiple gas injection locations on the toroidal asymmetry in the radiated power. Toroidal radiation asymmetry is diagnosed by an array of six single-channel photodiodes mounted on the vessel wall. The presence of magnetohydrodynamic (MHD) activity is diagnosed using an array of magnetic pickup (Mirnov) coils, mounted on stalks on the vessel wall. Scans were conducted of the relative timing between the two jets, of the 95th percentile safety factor, and of the plasma elongation. It is observed that firing the two gas jets so that the injected impurities arrive at the plasma at nearly the same time produced an increase in the toroidal radiation asymmetry. In addition, the radiation asymmetry in the thermal quench phase correlates with the growth rate of low toroidal mode number MHD modes, indicating that these mode(s) are playing a role in setting the radiation asymmetry.

  20. Excitation of unstable TAEs and stable n=0 modes in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Sears, J.; Bader, A.; Parker, R. R.; Kramer, G. J.

    2009-11-01

    Toroidicity-induced Alfv'en Eigenmodes (TAEs) are weakly damped MHD modes in tokamak plasmas. The modes occur at discrete frequencies near φTAE=vA/2qR, ( vA=B/√μ0ρ ) in a gap of the continuous spectrum of Alfv'en waves. Unstable TAEs are excited by ICRF heating of at least 3.5 MW in Alcator C-Mod L-mode plasmas. These unstable modes have toroidal mode numbers in the range of n=-6 to n=6. In contrast, stable resonant modes that are observed in these plasmas at similar and lower ICRF powers by the Active MHD diagnostic in the TAE frequency range commonly have toroidal mode numbers of n=0, which precludes a TAE or EAE identity. The origin of these modes is explored with the NOVA-K code, and the destabilizing role of the energetic hydrogen tail as measured by the Neutral Particle Analyzer is presented.

  1. Tomography of Light Emission from the Plasma Edge of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Kurz, Christian

    1995-01-01

    This thesis focuses on the development of a tomographic technique used on the Alcator C-Mod tokamak to reconstruct local emissivity profiles from line-of-sight integrated brightness measurements of H_{alpha } and low-Z plasma impurities. The knowledge of H_{alpha} emissivity profiles in particular opens up the possibility to calculate important plasma parameters such as the particle confinement time, parallel plasma flow velocities in the edge, and local neutral densities. These calculations depend on the knowledge of plasma temperature and density in the respective regions. A simple plasma model for the edge region is developed to interpolate between measurement data of Langmuir probes at the divertor target and upstream. The neutral density profile along the outer divertor target plate is examined during attached and detached divertor plasma conditions. The neutral densities found in the divertor are consistent with neutral pressure data from pressure gauges at the midplane and at the bottom of the divertor. Application of the inversion algorithm to CII and CIII impurity radiation is successfully demonstrated. Total radiation profiles measured by bolometers are examined during methane impurity gas puffing and plasma detachment, and found to agree well with reconstructed CIII profiles. (Copies available exclusively from MIT Libraries, Rm. 14-0551, Cambridge, MA 02139-4307. Ph. 617-253-5668; Fax 617-253 -1690.).

  2. BOUT++ simulations of edge turbulence in Alcator C-Mod's EDA H-mode

    NASA Astrophysics Data System (ADS)

    Davis, E. M.; Porkolab, M.; Hughes, J. W.; Labombard, B.; Snyder, P. B.; Xu, X. Q.; MIT PSFC Team; Atomics Team, General; LLNL Team

    2013-10-01

    Energy confinement in tokamaks is believed to be strongly controlled by plasma transport in the pedestal. The pedestal of Alcator C-Mod's Enhanced Dα (EDA) H-mode (ν* > 1) is regulated by a quasi-coherent mode (QCM), an edge fluctuation believed to reduce particle confinement and allow steady-state H-mode operation. ELITE calculations indicate that EDA H-modes sit well below the ideal peeling-ballooning instability threshold, in contrast with ELMy H-modes. Here, we use a 3-field reduced MHD model in BOUT++ to study the effects of nonideal and nonlinear physics on EDA H-modes. In particular, incorporation of realistic pedestal resistivity is found to drive resistive ballooning modes (RBMs) and increase linear growth rates above the corresponding ideal rates. These RBMs may ultimately be responsible for constraining the EDA pedestal gradient. However, recent high-fidelity mirror Langmuir probe measurements indicate that the QCM is an electron drift-Alfvén wave - not a RBM. Inclusion of the parallel pressure gradient term in the 3-field reduced MHD Ohm's law and various higher field fluid models are implemented in an effort to capture this drift wave-like response. This work was performed under the auspices of the USDoE under awards DE-FG02-94-ER54235, DE-AC52-07NA27344, DE-AC52-07NA27344, and NNSA SSGF.

  3. BOUT++ Simulations of Edge Turbulence in Alcator C-Mod's EDA H-Mode

    NASA Astrophysics Data System (ADS)

    Davis, E. M.; Porkolab, M.; Hughes, J. W.; Labombard, B.; Snyder, P. B.; Xu, X. Q.

    2013-10-01

    Energy confinement in tokamaks is believed to be strongly controlled by plasma transport in the pedestal. The pedestal of Alcator C-Mod's Enhanced Dα (EDA) H-mode (ν* > 1) is regulated by a quasi-coherent mode (QCM), an edge fluctuation believed to reduce particle confinement and allow steady-state H-mode operation. ELITE calculations indicate that EDA H-modes sit well below the ideal peeling-ballooning instability threshold, in contrast with ELMy H-modes. Here, we use a 3-field reduced MHD model in BOUT++ to study the effects of nonideal and nonlinear physics on EDA H-modes. In particular, incorporation of realistic pedestal resistivity is found to drive resistive ballooning modes (RBMs) and increase linear growth rates above the corresponding ideal rates. These RBMs may ultimately be responsible for constraining the EDA pedestal gradient. However, recent high-fidelity mirror Langmuir probe measurements indicate that the QCM is an electron drift-Alfvén wave - not a RBM. Inclusion of the parallel pressure gradient term in the 3-field reduced MHD Ohm's law and various higher field fluid models are implemented in an effort to capture this drift wave-like response. This work was performed under the auspices of the USDoE under awards DE-FG02-94-ER54235, DE-AC52-07NA27344, DE-AC52-07NA27344, and NNSA SSGF.

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

    SciTech Connect

    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.; Edlund, E.

    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.

  5. Modeling of particle and energy transport in the edge plasma of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Umansky, M. V.; Krasheninnikov, S. I.; LaBombard, B.; Lipschultz, B.; Terry, J. L.

    1999-07-01

    In the present study recycling and transport in the edge plasma of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] is modeled and analyzed with the multi-fluid code UEDGE [T. D. Rognlien et al., J. Nucl. Mater. 196-198, 347 (1992)]. Matching the experimental plasma density profiles in the scrape-off layer (SOL) requires a spatially dependent effective anomalous diffusion coefficient D⊥ growing rapidly towards the wall. The midplane pressure of neutral gas, Pmid, is a key parameter that reflects the magnitude of anomalous transport of plasma from the core. Recycling of plasma on the main chamber wall appears to be quite significant, especially in the case of high Pmid˜0.3 mTorr when the main wall provides ˜70% of recycling neutrals in the main chamber. In the upper SOL (well above the x point) draining of particles by the poloidal flow is weak and thus the particle balance is predominantly radial. For the radial heat transport it is found that energy flux carried by radial plasma convection and by charge-exchange (CX) neutrals is quite significant in SOL. In the high Pmid case, heat conduction by CX neutrals along with radial heat convection by plasma carries most of the power flux (˜75%) across the last closed flux surface. Even in the low Pmid case, heat conduction by CX neutrals dominates the radial heat flux far out in the SOL.

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

    NASA Astrophysics Data System (ADS)

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

    2001-10-01

    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 2/2000, 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 7/2000, with the installation of BN plasma-facing tiles and radiofrequency 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 2/2001 and will be completed by this meeting. All the above upgrades and their effect on antenna performance will be presented.

  7. First ion temperature profiles in the boundary of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Brunner, Dan; Labombard, Brian; Churchill, Michael; Lipschultz, Bruce; Ochoukov, Roman; Theiler, Christian; Whyte, Dennis

    2012-10-01

    The ion temperature is an important parameter in the boundary of magnetic fusion devices, playing a role in the sheath heat flux, total plasma pressure, plasma potential, and sound speed. High spatial resolution measurements of Ti and Te profiles are needed to evaluate these quantities and to unfold the rates of cross-field heat transport in each species. To this end, we have developed two specialized scanning probes: an Ion Sensitive Probe (ISP) and a Retarding Field Analyzer (RFA). Alcator C-Mod is a challenging environment for probes, with parallel heat fluxes comparable to that expected in reactors, 100's MW/m^2, necessitating the use of refractory materials in compact, optimized geometries. The two probes, along with Charge-eXchange Recombination Spectroscopy (CXRS) measurements of B^5+ ions, allow important cross-checks to be performed. Preliminary results indicate good agreement between the ISP and RFA. Comparison with B^5+ CXRS is also favorable. In ohmic-heated, sheath-limited plasmas the ratio of Ti/Te is ˜3 (150 and 50 eV) at the seperatrix and increases to ˜5 (50 and 10 eV) at the limiter boundary. Experiments with conduction-limited plasma conditions are planned and will be reported. Implications for boundary electron and ion heat transport will also be discussed.

  8. Effect on plasma rotation of lower hybrid (LH) waves in Alcator C-Mod

    SciTech Connect

    Lee, J. P.; Barnes, M.; Parker, R. R.; Rice, J. E.; Parra, F. I.; Bonoli, P. T.; Reinke, M. L.

    2014-02-12

    The injection of LH waves for current drive into a tokamak changes the ion toroidal rotation. In Alcator C-Mod, the direction of the steady state rotation change due to LH waves depends on the plasma current and the density. The change in rotation can be estimated by balancing the external torque of lower hybrid waves with the turbulent radial transport of the momentum. For high plasma current, the turbulent pinch and diffusion of the injected counter-current momentum are sufficient to explain the rotation change. However, for low plasma current, the change in the the intrinsic momentum transport (residual stress) for a non-rotating state is required to explain the co-current rotation change. Accordingly, we investigate the intrinsic momentum transport for the non-rotating state when diamagnetic flow and ExB flow cancel each other. The change in the intrinsic momentum transport due to lower hybrid waves is significant when the plasma current is low, which may explain the rotation reversal for low plasma current. The effect of changed q (safety factor) profile by lower hybrid on the intrinsic momentum transport is estimated by gyrokinetics.

  9. First Measurements of Edge Transport Driven by the Shoelace Antenna on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Golfinopoulos, T.; Labombard, B.; Parker, R. R.; Burke, W. M.; Hughes, J. W.; Brunner, D. F.; Davis, E. M.; Ennever, P. C.; Granetz, R. S.; Greenwald, M. J.; Irby, J. H.; Leccacorvi, R.; Marmar, E. S.; Parkin, W. C.; Porkolab, M.; Terry, J. L.; Vieira, R. F.; Wolfe, S. M.; Wukitch, S. J.; Alcator C-Mod Team

    2015-11-01

    The Shoelace antenna is a unique device designed to couple to the Quasi-Coherent Mode (QCM, k⊥ ~ 1 . 5 cm-1, 50 < f < 200 kHz) and Weakly-Coherent Mode (WCM, k⊥ ~ 1 . 5 cm-1, 200 < f < 500 kHz), continuous edge fluctuations that sustain high-performance confinement regimes by exhausting impurities. The antenna is used to explore whether modes like the QCM and WCM may be exploited to actively regulate edge transport. In initial experiments, the antenna excited a resonance at the QCM frequency and phase velocity, but transport measurements were unavailable. A subsequent redesign of the winding pitch allows the antenna to be field-aligned while mapping magnetically to the Mirror Langmuir Probe (MLP) on the last-closed flux surface. This has enabled the first measurements of edge transport induced by the antenna-driven fluctuation, which has been further enhanced by quadrupling the antenna source power. This work was supported by U.S. Department of Energy award DE-FC02-99ER54512, using Alcator C-Mod, a DOE SC User Facility.

  10. Effects of the q profile on toroidal rotation in Alcator C-Mod LHCD plasmas

    SciTech Connect

    Rice, J. E.; Mumgaard, R.; Parker, R.; Scott, S.; Shiraiwa, S.; Wallace, G. M.; Bonoli, P. T.; Delgado-Aparicio, L.; Fenzi, C.; Granetz, R. S.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J.; Lee, J. P.; Marmar, E. S.; Wolfe, S. M.

    2016-01-01

    Changes in the core toroidal rotation profiles following injection of lower hybrid (LH) waves have been documented in Alcator C-Mod plasmas. Shot by shot scans of LH input power have been performed at fixed magnetic field and electron density for several plasma currents. For sawtoothing target plasmas, if the input power is low enough that the central safety factor q0 remains below 1, the change in the core rotation is in the counter-current direction, consistent in sign, magnitude and LH power scaling with direct momentum input from the LH waves. If the power level is high enough that there are significant changes to the q profile, including the termination of sawtooth oscillations, the change in the toroidal rotation is in the co-current direction, consistent with changes in the momentum flux through its dependence on the current density profile. The direction of the rotation changes depends on whether q0 is below or above unity, and seemingly not on the magnetic shear, nor the Ohmic confinement regime of the target plasma.

  11. Disruption Mitigation Experiments with Two Gas Jets on Alcator C-Mod, and Implications for ITER

    NASA Astrophysics Data System (ADS)

    Granetz, R. S.; Olynyk, G. M.; Reinke, M. L.; Whyte, D. G.; Terry, J. L.; Lipschultz, B. L.; Combs, S. K.

    2012-10-01

    Until recently, all disruption mitigation experiments with massive gas injection have used one injection location at a time, and measurements have shown that the resulting radiated power is often toroidally asymmetric, which could lead to melting of first wall surfaces in ITER. Therefore, the proposed ITER MGI system has multiple gas jets distributed around the torus, but the effectiveness of this needs to be demonstrated on current machines. On Alcator C-Mod, a 2nd gas jet has been installed 154^o around the torus from the existing gas jet. The hardware components of both gas jets are nominally identical. A toroidally-distributed set of six AXUV detectors has been added to better measure the toroidal peaking factor (TPF) of the radiated power. Experiments have been carried out to characterize the effect of two jets on the radiation TPF by varying the relative timing between the firing of the gas jets shot-to-shot. Measurements of the radiation asymmetry during the pre-thermal quench, thermal quench, and current quench phases will be presented, as well as correlations with the growth rate of n=1 MHD modes. It will also be shown that very slight differences in hardware between the two gas jet systems are important. Implications for the ITER MGI system will be discussed.

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

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

    The efficiency of LHCD on Alcator C-Mod drops exponentially with line average density. At reactor relevant densities (> 1 . 1020 [m-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||-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.

  13. Alcator C-Mod Experiments in Support of the ITER Baseline 15 MA Scenario

    SciTech Connect

    C Kessel, et al

    2013-05-07

    Experiments on Alcator C-Mod have addressed several issues for the ITER 15 MA baseline scenario from 2009-2012. Rampup studies show ICRF can save significant V-s, and that an H-mode in the ramp can be utilized to save 50% more. ICRF modifications to li(1) are minimal, although the Te profile is peaked relative to ohmic in the plasma center, and alter sawtooth onset times. Rampdown studies show H-modes can be routinely sustained, avoiding an OH coil over-current associated with the H-L transition, that fast rampdowns are preferred, the density drops with Ip, and that the H-L transition occurs at Ploss/Pthr,LH ~ 1.0-1.3 at n/nGr ~ 0.85. Flattop plasmas targeting ITER baseline parameters have been sustained for 20 τE or 8-13 τCR, but only reach H98 ~ 0.6 at n/nGr = 0.85, rising to 0.9 at n/nGr = 0.65.

  14. Effects of the q profile on toroidal rotation in Alcator C-Mod LHCD plasmas

    DOE PAGESBeta

    Rice, J. E.; Gao, C.; Mumgaard, R.; Parker, R. R.; Scott, S. D.; Shiraiwa, S.; Wallace, G. M.; Bonoli, P. T.; Delgado-Aparicio, L.; Fenzi, C.; et al

    2016-02-09

    Changes in the core toroidal rotation profiles following injection of lower hybrid (LH) waves have been documented in Alcator C-Mod plasmas. Shot by shot scans of LH input power have been performed at fixed magnetic field and electron density for several plasma currents. For sawtoothing target plasmas, if the input power is low enough that the central safety factor q0 remains below 1, the change in the core rotation is in the counter-current direction, consistent in sign, magnitude and LH power scaling with direct momentum input from the LH waves. If the power level is high enough that there aremore » significant changes to the q profile, including the termination of sawtooth oscillations, the change in the toroidal rotation is in the co-current direction, consistent with changes in the momentum flux through its dependence on the current density profile. Furthermore, the direction of the rotation changes depends on whether q0 is below or above unity, and seemingly not on the magnetic shear, nor the Ohmic confinement regime of the target plasma.« less

  15. Robotic calibration of the motional Stark effect diagnostic on Alcator C-Mod

    SciTech Connect

    Mumgaard, Robert T.; Scott, Steven D.; Ko, Jinseok

    2014-05-15

    The capability to calibrate diagnostics, such as the Motional Stark Effect (MSE) diagnostic, without using plasma or beam-into-gas discharges will become increasingly important on next step fusion facilities due to machine availability and operational constraints. A robotic calibration system consisting of a motorized three-axis positioning system and a polarization light source capable of generating arbitrary polarization states with a linear polarization angle accuracy of <0.05° has been constructed and has been used to calibrate the MSE diagnostic deployed on Alcator C-Mod. The polarization response of the complex diagnostic is shown to be fully captured using a Fourier expansion of the detector signals in terms of even harmonics of the input polarization angle. The system's high precision robotic control of position and orientation allow it to be used also to calibrate the geometry of the instrument's view. Combined with careful measurements of the narrow bandpass spectral filters, this system fully calibrates the diagnostic without any plasma discharges. The system's high repeatability, flexibility, and speed has been exploited to quantify several systematics in the MSE diagnostic response, providing a more complete understanding of the diagnostic performance.

  16. Upgrade to the Gas Puff Imaging Diagnostic that Views Alcator C-Mod's Inboard Edge

    NASA Astrophysics Data System (ADS)

    Sierchio, J. M.; Terry, J. L.

    2012-10-01

    We describe an upgrade of Alcator C-Mod's Gas Puff Imaging system which views the inboard plasma edge and SOL along lines-of-sight that are approximately parallel to the local magnetic field. The views are arranged in a 2D (R,Z) array with ˜2.8 cm radial coverage and ˜2.4 cm poloidal coverage. 23 of 54 available views were coupled via fibers to individual interference filters and PIN photodiode detectors. We are in the process of upgrading the system in order to increase the sensitivity of the system by replacing the PIN photodiodes with a 4x8 array of Avalanche Photo-Diodes (APD). Light from 30 views is coupled to the single-chip APD array through a single interference filter. We expect an improvement in signal-to-noise ratio of more than 10x. The frequency response of the system will increase from ˜400 kHz to 1MHz. The dynamic range of the new system is manipulated by changing the high-voltages on the APDs. Test results of the detectors' channel-to-channel cross-talk, frequency response, and gain curves will be presented, along with schematics of the experimental setup. The upgraded system allows for more study of inboard edge fluctuations, including whether the quasi-coherent fluctuations observed in the outboard edge also exist inboard.

  17. Operation of Alcator C-Mod with high-Z plasma facing components and implications

    SciTech Connect

    Lipschultz, B.; Lin, Y.; Reinke, M.L.; Hubbard, A.; Hutchinson, I.H.; Irby, J.; LaBombard, B.; Marmar, E.S.; Marr, K.; Terry, J.L.; Wolfe, S.M.; Whyte, D.

    2006-05-15

    Studies of potential plasma facing component (PFC) materials for a magnetic fusion reactor generally conclude that tungsten is the best choice due to its low tritium (T) retention, capability to handle high heat fluxes with low erosion, and robustness to nuclear damage and activation. ITER [F. Perkins et al., Nucl. Fusion 39, 2137 (1999)] may operate with all tungsten PFCs to provide the necessary operational experience for a reactor. Alcator C-Mod [I. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] operates with molybdenum (Mo) high-Z PFCs, which have very similar properties to tungsten. The experiments described herein have provided a unique comparison of operation with or without in situ boron coatings applied to the molybdenum PFCs; the latter are likely most relevant to ITER and beyond. ICRF-heated H-modes were readily achieved without boron coatings although the resultant enhancement in energy confinement was typically small (H{sub ITER,89}{approx}1). Molybdenum concentrations, n{sub Mo}/n{sub e}, rise rapidly after the H-mode transition up to 0.1%, cooling the plasma by line radiation, reducing energy confinement, and/or causing a back H/L transition. Surprisingly, the primarily molybdenum PFC surfaces retain 3.5-5.0x10{sup 20} of injected D{sub 2} molecules per discharge, corresponding to 50% of the injected gas. Plasma current disruptions, both randomly occurring over the course of a day, or planned, reduce the retained D long term. After applying boron coatings, n{sub Mo}/n{sub e} was reduced by a factor of 10-20 with H{sub ITER,89} approaching 2. A world-record volume-average plasma pressure of 1.8 atm at 5.4 T was achieved at the ITER normalized {beta}. The effects of each boronization are found to be limited in time, correlated to time-integrated input energy. Intra- and inter-discharge boronization techniques have been developed with the latter being the most successful. This initial study indicates that a low-Z coating over at least a fraction of

  18. Measurements of LHCD current profile and efficiency for simulation validation on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Mumgaard, Robert T.

    2014-10-01

    Lower hybrid current drive (LHCD) is an effective tool to significantly modify the magnetic equilibrium by driving off-axis, non-inductive current. On Alcator C-Mod, an upgraded Motional Stark Effect (MSE) diagnostic enables the current profile to be accurately reconstructed during plasmas with strong LHCD and a hard X-ray camera measures the fast electron Bremsstrahlung profile. LHCD is applied for >4 current relaxation times, producing fully-relaxed magnetic equilibria in plasmas with non-inductive current fraction up to unity at currents up to 1.0 MA. C-Mod has developed an extensive database of LHCD performance, spanning a wide range in plasma current, launched n||, LHCD power, Te and plasma density. This dataset provides a unique platform for validation of LHCD current drive simulations with the plasma shape, density, field and LH frequency range envisioned for ITER and future reactors. In these conditions the measured current drive efficiencies are similar to that assumed for ITER with values up to 0.4*1020A/Wm2 despite being in a weak single-pass absorption regime. The driven current is observed to be off-axis, broadening the current profile, raising q0 above 1, suppressing sawteeth, decreasing/reversing the magnetic shear and sometimes destabilizing MHD modes and/or triggering internal transport barriers. Measurements indicate increased efficiency at increased temperature and plasma current but with a complicated dependence on launched n||. The MSE-constrained reconstructions show a loss in current drive efficiency as the plasma density is increased above =1.0×1020 m-3 consistent with previous observations of a precipitous drop in hard x-ray emission. Additionally, the measured driven current profile moves radially outward as the density is increased. Ray tracing simulations using GENRAY-CQL3D qualitatively reproduce these trends showing the rays make many passes through the plasma at high density and predicting a narrower current and HXR profile

  19. Characterization of impurity confinement on Alcator C-Mod using a multi-pulse laser blow-off system.

    PubMed

    Howard, N T; Greenwald, M; Rice, J E

    2011-03-01

    A new laser blow-off system for use in impurity transport studies on Alcator C-Mod was developed and installed for the 2009 run campaign. Its design included capabilities for multiple impurity injections during a single plasma pulse and remote manipulation of the ablated spot size. The system uses a 0.68 J, Nd:YAG laser operating at up to 10 Hz coupled with the fast beam steering via a 2D piezoelectric mirror mount able to move spot locations in the 100 ms between laser pulses and a remote controllable optical train that allow ablated spot sizes to vary from ∼0.5 to 7 mm. The ability to ablate a wide range in target Z along with Alcator C-Mod's extensive diagnostic capabilities (soft x-ray, vacuum ultraviolet (VUV), charge exchange spectroscopy, etc.) allows for detailed studies of the impurity transport dependencies and mechanisms. This system has demonstrated the achievement of all its design goals including the ability for non-perturbative operation allowing for insight into underlying impurity transport processes. A detailed overview of the laser blow-off system and initial results of operation are presented. This includes an investigation into the characterization of impurity confinement in the I-mode confinement regime recently investigated on C-Mod. PMID:21456742

  20. Characterization of impurity confinement on Alcator C-Mod using a multi-pulse laser blow-off system

    SciTech Connect

    Howard, N. T.; Greenwald, M.; Rice, J. E.

    2011-03-15

    A new laser blow-off system for use in impurity transport studies on Alcator C-Mod was developed and installed for the 2009 run campaign. Its design included capabilities for multiple impurity injections during a single plasma pulse and remote manipulation of the ablated spot size. The system uses a 0.68 J, Nd:YAG laser operating at up to 10 Hz coupled with the fast beam steering via a 2D piezoelectric mirror mount able to move spot locations in the 100 ms between laser pulses and a remote controllable optical train that allow ablated spot sizes to vary from {approx}0.5 to 7 mm. The ability to ablate a wide range in target Z along with Alcator C-Mod's extensive diagnostic capabilities (soft x-ray, vacuum ultraviolet (VUV), charge exchange spectroscopy, etc.) allows for detailed studies of the impurity transport dependencies and mechanisms. This system has demonstrated the achievement of all its design goals including the ability for non-perturbative operation allowing for insight into underlying impurity transport processes. A detailed overview of the laser blow-off system and initial results of operation are presented. This includes an investigation into the characterization of impurity confinement in the I-mode confinement regime recently investigated on C-Mod.

  1. Characterization of neo-classical tearing modes in high-performance I-mode plasmas with ICRF mode conversion flow drive on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lin, Yijun; Granetz, R.; Reinke, M. L.; Rice, J. E.; Wolfe, S. M.; Wukitch, S. J.

    2012-10-01

    Neo-classical tearing modes (NTM's) have been observed on Alcator C-Mod in high performance I-mode plasmas (Bt0˜5.2 T) that are heated by a combination of ICRF D(H) minority heating at 80 MHz and D(He3) mode conversion (MC) heating at 50 MHz. Due to the stabilizing effect by the energetic minority hydrogen ions, long sawtooth (>=40 ms) and large sawtooth crashes (δTe0>=3 keV) are produced in these hot (Te0<=9 keV) plasmas. NTMs with (m=4, n=3), (5, 4), (3, 2) and (2, 1) have been observed. The onset criteria of the (3, 2) NTMs, βp˜0.4-0.7, βN˜1.0-1.4, βN/ρi^*˜200-350, ν NTM (q=3/2)˜0.04-0.25, approximately follow the trend that obtained from DIII-D and ASDEX Upgrade. For the (3, 2) mode, the saturated width Wsat˜0.8-1 cm, which is 3-4 times the ion banana width. Although the NTMs have a small effect on confinement degradation (δβ/β˜ a few percent), they have significant effect on plasma rotation. MC flow drive generates large toroidal rotation above 100 km/sec in L-mode, and when the plasma enters I-mode, plasma rotation is expected to increase significantly due to the additional intrinsic rotation torque from the edge Te pedestal. The appearance of the (3, 2) mode usually rapidly reduces the rotation speed, and the (2, 1) mode, if it occurs, would completely halt the rotation.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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}{sub p} {le} 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{sup +} 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 = {infinity} and the n = 1 marginal stability limits.

  4. Fluctuation statistics in the scrape-off layer of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

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

    2016-05-01

    We study long time series of the ion saturation current and floating potential, sampled by Langmuir probes dwelled in the outboard mid-plane scrape-off layer and embedded in the lower divertor baffle of Alcator C-Mod. A series of ohmically heated L-mode plasma discharges is investigated with line-averaged plasma density ranging from {{\\bar{n}}\\text{e}}/{{n}\\text{G}}=0.15 to 0.42, where n G is the Greenwald density. All ion saturation current time series that are sampled in the far scrape-off layer are characterized by large-amplitude burst events. Coefficients of skewness and excess kurtosis of the time series obey a quadratic relationship and their histograms coincide partially upon proper normalization. Histograms of the ion saturation current time series are found to agree well with a prediction of a stochastic model for the particle density fluctuations in scrape-off layer plasmas. The distribution of the waiting times between successive large-amplitude burst events and of the burst amplitudes are approximately described by exponential distributions. The average waiting time and burst amplitude are found to vary weakly with the line-averaged plasma density. Conditional averaging reveals that the radial blob velocity, estimated from floating potential measurements, increases with the normalized burst amplitude in the outboard mid-plane scrape-off layer. For low density discharges, the conditionally averaged waveform of the floating potential associated with large amplitude bursts at the divertor probes has a dipolar shape. In detached divertor conditions the average waveform is random, indicating electrical disconnection of blobs from the sheaths at the divertor targets.

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

    NASA Astrophysics Data System (ADS)

    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 < τ E ) response of the scrape-off-layer (SOL) plasma to Lower Hybrid 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.

  6. Initial Operation of the Alcator C-Mod ICRF Antennas with High-Z Metal Antenna Guards

    SciTech Connect

    Schilling, G.; Wukitch, S.J.; Lin, Y.; Parisot, A.; Porkolab, M.

    2005-09-26

    The Alcator C-Mod ICRF antennas have been operated with BN antenna guards since 2000. This modification had followed the observation that metallic impurities in the plasma increased with increasing ICRF power. Systematic improvements to the antenna structure have allowed the launched power to be raised to the 6 MW level, with good heating efficiency and few deleterious effects on the plasma. BN is inherently fragile, and disruption mechanical shocks have resulted in fracturing of the antenna tiles, exposing the supporting metal structure. Since the antennas are now electrically in reasonable shape, and one of C-Mod's goals is to study the behavior of plasmas with all-metal plasma facing components, it was decided to replace the BN tiles with molybdenum. High power operation was resumed in March, 2005, and the behavior of the antennas during the present run campaign will be reported.

  7. Development of Absolute Calibration of the Phase Contrast Imaging Diagnostic and Experimental Tests in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Tsujii, N.; Porkolab, M.; Edlund, E. M.; Lin, L.

    2007-11-01

    The Phase Contrast Imaging (PCI) system in Alcator C-Mod is used to measure density perturbations from MHD modes, turbulence and RF waves. Recently, an absolutely calibrated system has been installed. This system consists of a set of transducers which cover frequency from 30 kHz to 200 kHz, and wavenumber from 5.5 cm-1 to 36.6 cm-1. The amplitude and phase of the transducer wavefronts are measured using a calibrated microphone. We will present the system design and modeling of this calibration system. Initial results, including a comparison with experimental measurements will also be discussed, if available.

  8. Observation of Reversed Shear Alfvén Eigenmodes between Sawtooth Crashes in the Alcator C-Mod Tokamak

    NASA Astrophysics Data System (ADS)

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

    2009-04-01

    Groups of frequency chirping modes observed between sawtooth crashes in the Alcator C-Mod tokamak are interpreted as reversed shear Alfvén eigenmodes near the q=1 surface. These modes indicate that a reversed shear q profile is generated during the relaxation phase of the sawtooth cycle. Two important parameters, qmin⁡ and its radial position, are deduced from comparisons of measured density fluctuations with calculations from the ideal MHD code NOVA. These studies provide valuable constraints for further modeling of the sawtooth cycle.

  9. Edge radial electric field structure and its connections to H-mode confinement in Alcator C-Mod plasmas

    SciTech Connect

    McDermott, R. M.; Lipschultz, B.; Hughes, J. W.; Catto, P. J.; Hubbard, A. E.; Hutchinson, I. H.; Granetz, R. S.; Greenwald, M.; LaBombard, B.; Marr, K.; Reinke, M. L.; Rice, J. E.; Whyte, D.

    2009-05-15

    High-resolution charge-exchange recombination spectroscopic measurements of B{sup 5+} ions have enabled the first spatially resolved calculations of the radial electric field (E{sub r}) in the Alcator C-Mod pedestal region [E. S. Marmar, Fusion Sci. Technol. 51, 261 (2006)]. These observations offer new challenges for theory and simulation and provide for important comparisons with other devices. Qualitatively, the field structure observed on C-Mod is similar to that on other tokamaks. However, the narrow high-confinement mode (H-mode) E{sub r} well widths (5 mm) observed on C-Mod suggest a scaling with machine size, while the observed depths (up to 300 kV/m) are unprecedented. Due to the strong ion-electron thermal coupling in the C-Mod pedestal, it is possible to infer information about the main ion population in this region. The results indicate that in H-mode the main ion pressure gradient is the dominant contributor to the E{sub r} well and that the main ions have significant edge flow. C-Mod H-mode data show a clear correlation between deeper E{sub r} wells, higher confinement plasmas, and higher electron temperature pedestal heights. However, improved L-mode (I-mode) plasmas exhibit energy confinement equivalent to that observed in similar H-mode discharges, but with significantly shallower E{sub r} wells. I-mode plasmas are characterized by H-mode-like energy barriers, but with L-mode-like particle barriers. The decoupling of energy and particle barrier formation makes the I-mode an interesting regime for fusion research and provides for a low collisionality pedestal without edge localized modes.

  10. Study of toroidal flow generation by ion cyclotron range of frequency minority heating in the Alcator C-Mod plasma

    NASA Astrophysics Data System (ADS)

    Murakami, S.; Itoh, K.; Zheng, L. J.; Van Dam, J. W.; Bonoli, P.; Rice, J. E.; Fiore, C. L.; Gao, C.; Fukuyama, A.

    2016-01-01

    The averaged toroidal flow of energetic minority ions during ICRF (ion cyclotron range of frequencies) heating is investigated in the Alcator C-Mod plasma by applying the GNET code, which can solve the drift kinetic equation with complicated orbits of accelerated energetic particles. It is found that a co-directional toroidal flow of the minority ions is generated in the region outside of the resonance location, and that the toroidal velocity reaches more than 40% of the central ion thermal velocity (Vtor ˜ 300 km/s with PICRF ˜ 2 MW). When we shift the resonance location to the outside of |r /a |˜0.5 , the toroidal flow immediately inside of the resonance location is reduced to 0 or changes to the opposite direction, and the toroidal velocity shear is enhanced at r/a ˜ 0.5. A radial diffusion equation for toroidal flow is solved by assuming a torque profile for the minority ion mean flow, and good agreements with experimental radial toroidal flow profiles are obtained. This suggests that the ICRF driven minority ion flow is related to the experimentally observed toroidal rotation during ICRF heating in the Alcator C-Mod plasma.

  11. First operation of a high-heat flux, flush mounted ``rail'' Langmuir probe array on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Kuang, Adam Q.; Brunner, Dan; Labombard, Brian; Leccacorvi, Rick; Vieira, Rui

    2015-11-01

    Divertor Langmuir probes are typically built proud of the divertor surface for an accurate measurement of the plasma flux. However, under the high heat flux conditions seen in Alcator C-Mod, proud tungsten probes that present a 10 degree attack angle to the incident heat flux can experience melt damage with less then 1 second plasma exposure time. A similar situation is anticipated for ITER. It is therefore desirable to develop a flush probe system that can both survive reactor-level fluxes and take accurate measurements. A poloidal array of 21 flush-mounted ``rail'' probes have been installed in the C-Mod outer divertor plate, which are toroidally-extended and field-aligned to minimize sheath expansion effects. Initial results indicate that the ``rail'' probes have a well-defined ion saturation current, reporting similar density and temperature measurements as proud probes. However, uncertainty in the projected area becomes significant when the incident magnetic field angle becomes less than ~0.5 degrees. Additionally, because the flush probes are conformal to the divertor surface, they are ideally suited to measure the poloidal distribution of halo currents during disruptions. Supported by USDoE C-Mod award DE-FC02-99ER54512.

  12. Comparisons of Measurements and Gyrokinetic Simulations of Turbulence and Transport in Alcator C-Mod EDA H-Mode Discharges.

    NASA Astrophysics Data System (ADS)

    Sampsell, M. B.; Bravenec, R. V.; Candy, J.; Ernst, D. R.; Alcator C-Mod Team Nevins

    2004-11-01

    Beam-emission spectroscopy (BES) on Alcator C-Mod has observed long wavelength broadband fluctuations and a `quasi-coherent mode' (the latter exclusive to enhanced D_α H modes) in the plasma edge. However, it has not observed broadband fluctuations at the top of the H-mode pedestal or farther in. In an attempt to understand this, we have run the GYRO gyrokinetic code [J. Candy, J. Comput. Phys. 186, 545 (2003)] for this region, applied `synthetic BES' to the fluctuating density output, and compared with the data. The synthetic BES is composed of i) an anti-aliasing filter in GYRO itself, ii) a conversion from density fluctuations to emissivity fluctuations, and iii) a spatial filter to model the finite viewing area of the diagnostic. We find significant attenuation of the density fluctuations. Transport results from GYRO are also compared to data to validate the simulations.

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

    SciTech Connect

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

    2010-08-27

    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 qmin, 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.

  14. Estimation of the ion toroidal rotation source due to momentum transfer from Lower Hybrid waves in Alcator C-Mod

    SciTech Connect

    Lee, J. P.; Wright, J. C.; Bonoli, P. T.; Parker, R. R.; Catto, P. J.; Podpaly, Y. A.; Rice, J. E.; Reinke, M. L.

    2011-12-23

    Significant ion toroidal rotation (50km/s) has been measured by X-Ray spectroscopy for impurities in Alcator C-Mod during lower hybrid (LH) RF power injection. We investigate the relation between the computed toroidal momentum input from LH waves and the measured INITIAL change of ion toroidal rotation when the LH power is turned on. The relation may depend on the plasma current and magnetic configuration. Because of the fast build up time of the electron quasilinear plateau (<1 millisecond), the electron distribution function rapidly reaches steady state in which the electrons transfer momentum to the ions. The LH wave momentum input is computed from the self consistent steady state electron distribution function and a bounce-averaged quasilinear diffusion coefficient that are obtained by iterating a full wave code (TORLH) with a Fokker Plank code (CQL3D)

  15. Measurement of electron temperature fluctuations using a tunable correlation electron cyclotron emission system on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Howard, N. T.; Sung, C.; White, A. E.

    2014-11-01

    A tunable correlation electron cyclotron (CECE) system was recently installed on the Alcator C-Mod tokamak to provide local, quantitative measurement of electron temperature fluctuations in the tokamak core. This system represents a significant upgrade from the original CECE system, expanding the measurement capabilities from 4 to 8 total channels, including 2 remotely tunable YIG filters (6-18 GHz; 200 MHz bandwidth). Additional upgrades were made to the optical system to provide enhanced poloidal resolution and allow for measurement of turbulent fluctuations below kθρs < 0.3. These expanded capabilities allow for single shot measurement of partial temperature fluctuation profiles in the region ρ = 0.7 - 0.9 (square root of normalized toroidal flux) in a wide variety of plasma conditions. These measurements are currently being used to provide stringent tests of the gyrokinetic model in ongoing model validation efforts. Details of the hardware upgrades, turbulent fluctuation measurements, and ongoing comparisons with simulations are presented.

  16. EMC3-EIRENE modelling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

    SciTech Connect

    Lore, Jeremy D.; Reinke, M. L.; LaBombard, Brian; Lipschultz, B.; Churchill, R. M.; Pitts, R. A.; Feng, Y.

    2014-09-30

    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 modelling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Finally, 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.

  17. Response to ""comment on 'magnetic topology effects on alcator C-mod scrape-off layer flow'

    SciTech Connect

    Sinakov, Andrei N; Catto, Peter J

    2008-01-01

    Recent interest in the experimental study of tokamak plasma flow for different magnetic field geometries calls for theoretical understanding of the effects of tokamak magnetic topology changes on the flow. The consequences of total magnetic field reversal and/or X-point reversal on divergence-free plasma flow within magnetic flux surfaces are considered and the results are applied to interpret recent Alcator C-Mod scrape-off layer flow measurements. In his comment to that work, Aydemir 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.

  18. Response to ""Comment on ""Magnetic topology effects on alcator c-mod scrape-off layer flow

    SciTech Connect

    Simakov, Andrei N; Catto, Peter J

    2008-01-01

    Recent interest in the experimental study of tokamak plasma flow for different magnetic field geometries calls for theoretical understanding of the effects of tokamak magnetic topology changes on the flow. The consequences of total magnetic field reversal and/or X-point reversal on divergence-free plasma flow within magnetic flux surfaces are considered and the results are applied to interpret recent Alcator C-Mod scrape-off layer flow measurements. In his comment to that work, Aydemir 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.

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

    NASA Astrophysics Data System (ADS)

    Lore, J. D.; Reinke, M. L.; LaBombard, B.; Lipschultz, B.; Churchill, R. M.; Pitts, R. A.; Feng, Y.

    2015-08-01

    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

  20. Thermal Analysis to Calculate the Vessel Temperature and Stress in Alcator C-Mod Due to the Divertor Upgrade

    SciTech Connect

    Han Zhang, Peter H. Titus, Robert Ellis, Soren Harrison and Rui Vieira

    2012-08-29

    Alcator C-Mod is planning an upgrade to its outer divertor. The upgrade is intended to correct the existing outer divertor alignment with the plasma, and to operate at elevated temperatures. Higher temperature operation will allow study of edge physics behavior at reactor relevant temperatures. The outer divertor and tiles will be capable of operating at 600oC. Longer pulse length, together with the plasma and RF heat of 9MW, and the inclusion of heater elements within the outer divertor produces radiative energy which makes the sustained operation much more difficult than before. An ANSYS model based on ref. 1 was built for the global thermal analysis of C-Mod. It models the radiative surfaces inside the vessel and between the components, and also includes plasma energy deposition. Different geometries have been simulated and compared. Results show that steady state operation with the divertor at 600oC is possible with no damage to major vessel internal components. The differential temperature between inner divertor structure, or "girdle" and inner vessel wall is ~70oC. This differential temperature is limited by the capacity of the studs that hold the inner divertor backing plates to the vessel wall. At a 70oC temperature differential the stress on the studs is within allowable limits. The thermal model was then used for a stress pass to quantify vessel shell stresses where thermal gradients are significant.

  1. Comparison of GS2 Turbulence Simulations with Phase Contrast Imaging in Alcator C-Mod Internal Transport Barriers

    NASA Astrophysics Data System (ADS)

    Long, Andrew; Ernst, Darin

    2005-10-01

    Trapped electron mode (TEM) turbulence arises in gyrokinetic simulations of internal transport barriers in Alcator C-mod experiments [1]. C-mod is equipped with a PCI (phase contrast imaging) diagnostic which measures density fluctuations along 32 vertical chords passing near the magnetic axis. The GS2 density fluctuations are output as an integral along field lines. The GS2 poloidal wavelength spectrum is upshifted relative to the PCI spectrum [1]. To make the comparison more direct, we have modified GS2 [2] to calculate electron density fluctuations at the poloidal angles observed by PCI. The longer wavelength modes are more extended along field lines, so that when viewed off the midplane, they are weighted more strongly. Nonlinear simulations are underway, and results will be presented. [1] D. R. Ernst et al., 20th IAEA Fusion Energy Conf. IAEA-CN-116/TH/4-1. also Phys. Plasmas 11(5) 2637 (2004). http://www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/TH4-1.html [2] W. Dorland et al., Phys. Rev. Lett. 85(26) 5579 (2000). funding: Contract number DE-AC02-76CH03073.

  2. Comparison of electron temperature fluctuations with gyrokinetic sumulations across the ohmic energy confinement transition in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Sung, C.; White, A.; Howard, N.; Mikkelsen, D.; Rice, J.; Reinke, M.; Gao, C.; Ennever, P.; Porkolab, M.; Churchill, R.; Theiler, C.; Hubbard, A.; Greenwald, M.

    2013-10-01

    Long wavelength electron temperature fluctuations (kyρs < 0 . 3) near the edge (r / a ~ 0 . 85) are reduced across the ohmic confinement transition from Linear Ohmic Confinement(LOC) regime to Saturated Ohmic Confinement(SOC) regime in Alcator C-Mod. Linear stability analysis shows that the dominant mode of long wavelength turbulence near the edge is changed from Trapped Electron Mode(TEM) to Ion Temperature Gradient(ITG) mode while the dominant mode is not changed deeper in the core (r / a ~ 0 . 5). This indicates that local turbulence changes near the edge might be responsible for the change of global energy confinement in ohmic plasmas. Further study using nonlinear gyrokinetic simulations is being performed to clarify the relation between the change of local turbulence and global ohmic energy confinement. Through nonlinear gyrokinetic simulation (GYRO), we will investigate the change of fluctuating quantities (T~ , ñ , ϕ~) and their phase relations across ohmic confinement transitions, and relate them to the change of energy transport. A synthetic CECE diagnostic for C-Mod has been developed, and it will be used to validate the gyrokinetic simulations. Research supported by USDoE awards DE-SC0006419, DE-FC02-99ER54512.

  3. Measurements of relativistic emission from runaway electrons in Alcator C-Mod: spectrum, polarization, and spatial structure

    NASA Astrophysics Data System (ADS)

    Granetz, Robert; Mumgaard, Robert

    2014-10-01

    At low densities, runaway electrons (RE's) can be generated during the flattop of Alcator C-Mod discharges with highly relativistic energies, γ >> 1 , allowing careful study under steady conditions. These RE's emit light in a narrow forward-peaked cone which is detected with a number of diagnostics, including spectrometers, a video imaging camera, and polarimetry (using the MSE system), in addition to the standard hard x-ray detectors. These measurements of the relativistic emission can provide information about the RE energy distribution, pitch angle distribution, and spatial distribution. Unlike most other tokamaks, C-Mod's high magnetic field shifts the peak of the continuum emission into the visible, due to the smaller gyroradius and higher gyro-frequency, allowing for excellent spectral coverage with standard spectrometers, and thus detailed comparison to theoretical predictions of synchrotron and bremsstrahlung spectra. Additionally, camera images occasionally show highly structured formations. Profiles of the polarization fraction and polarization angle show radial structure, including a jump of 90° outboard of the magnetic axis, in qualitative agreement with recent theoretical calculations for relativistic electrons in a tokamak field. This work is supported by the U.S. Department of Energy.

  4. Quasi-coherent fluctuations limiting the pedestal growth on Alcator C-Mod: experiment and modelling

    DOE PAGESBeta

    Diallo, A.; Hughes, J. W.; Baek, S-G.; LaBombard, Brian; Terry, J.; Cziegler, I.; Hubbard, A.; Davis, E.; Walk, J.; Delgado-Aparicio, L.; et al

    2015-01-01

    Performance predictions for future fusion devices rely on an accurate model of the pedestal structure. The candidate for predictive pedestal structure is EPED, and it is imperative to test the underlying hypotheses to further gain confidence for ITER projections. Here, we present experimental work testing one of the EPED hypotheses, namely the existence of a soft limit set by microinstabilities such as the kinetic ballooning mode. This work extends recent work on Alactor C-Mod (Diallo et al 2014 Phys. Rev. Lett. 112 115001), to include detailed measurements of the edge fluctuations and comparisons of edge simulation codes and experimental observations.

  5. Edge Temperature Gradient as Intrinsic Rotation Drive in Alcator C-Mod Tokamak Plasmas

    SciTech Connect

    Rice, J. E.; Hughes, J. W.; Podpaly, Y. A.; Reinke, M. L.; Greenwald, M. J.; Hubbard, A. E.; Marmar, E. S.; Whyte, D. G.; Diamond, P. H.; Kosuga, Y.; McDevitt, C. J.; Guercan, Oe. D.; Hahm, T. S.

    2011-05-27

    Intrinsic rotation has been observed in I-mode plasmas from the C-Mod tokamak, and is found to be similar to that in H mode, both in its edge origin and in the scaling with global pressure. Since both plasmas have similar edge {nabla}T, but completely different edge {nabla}n, it may be concluded that the drive of the intrinsic rotation is the edge {nabla}T rather than {nabla}P. Evidence suggests that the connection between gradients and rotation is the residual stress, and a scaling for the rotation from conversion of free energy to macroscopic flow is calculated.

  6. Parallel transport studies of high-Z impurities in the core of Alcator C-Mod plasmasa)

    NASA Astrophysics Data System (ADS)

    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.

    2013-05-01

    Measurements of poloidal variation, ñz/⟨nz⟩, 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 -0.20, 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, -0.050 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, nzZ2/ni≪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, vθ ,z. Artificially modifying main-ion flow in parallel transport simulations is shown to impact both n˜z/⟨nz⟩ and vθ ,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.

  7. PCI measurements of Turbulence and Transport in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Ennever, P.; Porkolab, M.; Dorris, J.; Tsujii, N.; Davis, E.; Alcator C-Mod Team

    2011-10-01

    Measurements of turbulent density fluctuations on C-Mod using Phase Contrast Imaging (PCI) are reported. PCI is an interferometric technique that measures line-integrated density fluctuations at frequencies up to 2 MHz in the wave number range 0.5-30 cm-1. The PCI system on C-Mod consists of a beam that passes vertically through the plasma core and images onto a 1-D array of 32 HgCdTe detectors. Owing to lack of localization along the beam path, the signal includes contributions from the plasma edge as well as the core. To distinguish edge turbulence from that emanating from the core, we compare the spectrum to other edge localized turbulence measurements (ie, GPI, reflectometry). The core turbulence spectrum is modeled by a synthetic PCI diagnostic and nonlinear global GYRO analysis. Fluctuation spectra predicted by GYRO will be compared with those measured with PCI in different modes of plasma operation. Supported by US DoE awards DE-FG02-94-ER54235 and DE-FC02-99-ER54512.

  8. Design and operation of a novel divertor cryopumping system in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Labombard, B.; Beck, B.; Bosco, J.; Childs, R.; Gwinn, D.; Irby, J.; Leccacorvi, R.; Marazita, S.; Mucic, N.; Pierson, S.; Rokhman, Y.; Titus, P.; Vieira, R.; Zaks, J.; Zhukovsky, A.

    2007-11-01

    C-Mod's recently installed upper-divertor cryopump is unique among the world's tokamaks, employing an array of gas-pumping slots that penetrate the upper divertor target. This geometry enables the use of a single toroidal loop of liquid helium, operating in an efficient heat transfer regime with low or no helium flow. A system pumping speed of 9,600 l/sec for D2 gas has been achieved, matching that of a full-scale prototype system. Neutral pressures in the pumping slots during upper-null plasmas (USN) are found to meet or exceed pressures in the lower divertor's private flux region during lower-null (LSN) -- evidence that the pumping-slot geometry is performing as intended. Very high steady-state pumping throughputs (exceeding ˜140 torr-l/s) have been demonstrated in USN. Reliable and efficient operation of the pump has been established, synchronized with the C-Mod shot cycle and consuming 60 to 90 liters of liquid helium during a full day of operation.

  9. Stability of Microturbulent Drift Modes during Internal Transport Barrier Formation in the Alcator C-Mod Radio Frequency Heated H-mode

    SciTech Connect

    M.H. Redi; C.L. Fiore; W. Dorland; D.R. Mikkelsen; G. Rewoldt; P.T. Bonoli; D.R. Ernst; J.E. Rice; S.J. Wukitch

    2003-11-20

    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.

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

    NASA Astrophysics Data System (ADS)

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

    2013-08-01

    The first measurements of long wavelength (kyρs < 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.

  11. Quantitative comparison of experimental impurity transport with nonlinear gyrokinetic simulation in an Alcator C-Mod L-mode plasma

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

    Nonlinear gyrokinetic simulations of impurity transport are compared to experimental impurity transport for the first time. The GYRO code (Candy and Waltz 2003 J. Comput. Phys. 186 545) was used to perform global, nonlinear gyrokinetic simulations of impurity transport for a standard Alcator C-Mod, L-mode discharge. The laser blow-off technique was combined with soft x-ray measurements of a single charge state of calcium to provide time-evolving profiles of this non-intrinsic, non-recycling impurity over a radial range of 0.0 ⩽ r/a ⩽ 0.6. Experimental transport coefficient profiles and their uncertainties were extracted from the measurements using the impurity transport code STRAHL and rigorous Monte Carlo error analysis. To best assess the agreement of gyrokinetic simulations with the experimental profiles, the sensitivity of the GYRO predicted impurity transport to a wide range of turbulence-relevant plasma parameters was investigated. A direct comparison of nonlinear gyrokinetic simulation and experiment is presented with an in depth discussion of error sources and a new data analysis methodology.

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

    SciTech Connect

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

    2009-09-01

    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. 2009 American Institute of Physics.

  13. Surface thermocouples for measurement of pulsed heat flux in the divertor of the Alcator C-Mod tokamak

    SciTech Connect

    Brunner, D.; LaBombard, B.

    2012-03-15

    A novel set of thermocouple sensors has been developed to measure heat fluxes arriving at divertor surfaces in the Alcator C-Mod tokamak, a magnetic confinement fusion experiment. These sensors operate in direct contact with the divertor plasma, which deposits heat fluxes in excess of {approx}10 MW/m{sup 2} over an {approx}1 s pulse. Thermoelectric EMF signals are produced across a non-standard bimetallic junction: a 50 {mu}m thick 74% tungsten-26% rhenium ribbon embedded in a 6.35 mm diameter molybdenum cylinder. The unique coaxial geometry of the sensor combined with its single-point electrical ground contact minimizes interference from the plasma/magnetic environment. Incident heat fluxes are inferred from surface temperature evolution via a 1D thermal heat transport model. For an incident heat flux of 10 MW/m{sup 2}, surface temperatures rise {approx}1000 deg. C/s, corresponding to a heat flux flowing along the local magnetic field of {approx}200 MW/m{sup 2}. Separate calorimeter sensors are used to independently confirm the derived heat fluxes by comparing total energies deposited during a plasma pulse. Langmuir probes in close proximity to the surface thermocouples are used to test plasma-sheath heat transmission theory and to identify potential sources of discrepancies among physical models.

  14. BOUT simulations of the Quasi-Coherent Mode in the EDA Regime of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Xu, X. Q.; Nevins, W. M.; Mazurenko, A.; Porkolab, M.; Mossessian, D.

    2001-10-01

    The Quasi-Coherent (QC) mode is always seen by the phase contrast imaging (PCI) diagnostic during the EDA H-mode(Greenwald, M., et al, Phys. Plasmas 6, 1943 (1999).) in the Alcator C-Mod tokamak. The typical QC mode is characterized by a high frequency ( ~100kHz) ``quasi-coherent'' density fluctuation with high poloidal wavenumber, k_θ~= 5cm-1 (near the X-point region) localized in the edge steep density gradient region (pedestal). In order to understand the QC mode, simulations have been carried out using the BOUT code(Xu, X.Q., et al, J.R., Phys. of Plasmas, Vol.7, 1951 (2000).). The measured dispersion and mode stability is in good agreement with the resistive X-point ballooning mode predicted by the BOUT code. The principal results are: (1) Frequency spectrum of BOUT simulation resembles that of PCI, magnetic probe and reflectometer measurements; (2) Magnetic fluctuations peak near the separatrix and rapidly decay radially; (3) Radial particle flux shows maximum near the X-point region.

  15. Lower Hybrid Wave Induced SOL Emissivity Variation at High Density on the Alcator C-Mod Tokamak

    SciTech Connect

    Faust, I.; Terry, J. L.; Reinke, M. L.; Meneghini, O.; Shiraiwa, S.; Wallace, G. M.; Parker, R. R.; Schmidt, A. E.; Wilson, J. R.

    2011-12-23

    Lower Hybrid Current Drive (LHCD) in the Alcator C-Mod tokamak provides current profile control for the generation of Advanced Tokamak (AT) plasmas. Non-thermal electron bremsstrahlung emission decreases dramatically at n-bar{sub e}>1{center_dot}10{sup 20}[m{sup -3}] for diverted discharges, indicating low current drive efficiency. It is suggested that Scrape-Off-Layer (SOL) collisional absorption of LH waves is the cause for the absence of non-thermal electrons at high density. VUV and visible spectroscopy in the SOL provide direct information on collision excitation processes. Deuterium Balmer-, Lyman- and He-I transition emission measurements were used for initial characterization of SOL electron-neutral collisional absorption. Data from Helium and Deuterium LHCD discharges were characterized by an overall increase in the emissivity as well as an outward radial shift in the emissivity profile with increasing plasma density and applied LHCD power. High-temperature, high-field (T{sub e} = 5keV,B{sub t} = 8T) helium discharges at high density display increased non-thermal signatures as well as reduced SOL emissivity. Variations in emissivity due to LHCD were seen in SOL regions not magnetically connected to the LH Launcher, indicating global SOL effects due to LHCD.

  16. Measurement of electron temperature fluctuations using a tunable correlation electron cyclotron emission system on Alcator C-Mod

    SciTech Connect

    Howard, N. T.; Sung, C.; White, A. E.

    2014-11-15

    A tunable correlation electron cyclotron (CECE) system was recently installed on the Alcator C-Mod tokamak to provide local, quantitative measurement of electron temperature fluctuations in the tokamak core. This system represents a significant upgrade from the original CECE system, expanding the measurement capabilities from 4 to 8 total channels, including 2 remotely tunable YIG filters (6–18 GHz; 200 MHz bandwidth). Additional upgrades were made to the optical system to provide enhanced poloidal resolution and allow for measurement of turbulent fluctuations below k{sub θ}ρ{sub s} < 0.3. These expanded capabilities allow for single shot measurement of partial temperature fluctuation profiles in the region ρ = 0.7 − 0.9 (square root of normalized toroidal flux) in a wide variety of plasma conditions. These measurements are currently being used to provide stringent tests of the gyrokinetic model in ongoing model validation efforts. Details of the hardware upgrades, turbulent fluctuation measurements, and ongoing comparisons with simulations are presented.

  17. EMC3-EIRENE modelling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

    DOE PAGESBeta

    Lore, Jeremy D.; Reinke, M. L.; LaBombard, Brian; Lipschultz, B.; Churchill, R. M.; Pitts, R. A.; Feng, Y.

    2014-09-30

    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 modelling, with the simulation yielding a toroidal asymmetry in the heat flow to the outermore » strike point. Finally, 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.« less

  18. Measurement of electron temperature fluctuations using a tunable correlation electron cyclotron emission system on Alcator C-Mod.

    PubMed

    Howard, N T; Sung, C; White, A E

    2014-11-01

    A tunable correlation electron cyclotron (CECE) system was recently installed on the Alcator C-Mod tokamak to provide local, quantitative measurement of electron temperature fluctuations in the tokamak core. This system represents a significant upgrade from the original CECE system, expanding the measurement capabilities from 4 to 8 total channels, including 2 remotely tunable YIG filters (6-18 GHz; 200 MHz bandwidth). Additional upgrades were made to the optical system to provide enhanced poloidal resolution and allow for measurement of turbulent fluctuations below kθρs < 0.3. These expanded capabilities allow for single shot measurement of partial temperature fluctuation profiles in the region ρ = 0.7 - 0.9 (square root of normalized toroidal flux) in a wide variety of plasma conditions. These measurements are currently being used to provide stringent tests of the gyrokinetic model in ongoing model validation efforts. Details of the hardware upgrades, turbulent fluctuation measurements, and ongoing comparisons with simulations are presented. PMID:25430224

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

    NASA Astrophysics Data System (ADS)

    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.

  20. Validation study of gyrokinetic simulation (GYRO) near the edge in Alcator C-Mod ohmic discharges

    NASA Astrophysics Data System (ADS)

    Sung, C.; White, A.; Howard, N.; Mikkelsen, D.; Holland, C.; Rice, J.; Reinke, M.; Gao, C.; Ennever, P.; Porkolab, M.; Churchill, R.; Theiler, C.; Walk, J.; Hughes, J.; Hubbard, A.; Greenwald, M.

    2014-10-01

    A validation study of local gyrokinetic simulations (GYRO) near the edge region (r / a ~ 0 . 85) has been performed for two C-Mod ohmic discharges, namely one that is in the Linear Ohmic Confinement (LOC) regime and the other one in the Saturated Ohmic Confinement (SOC) regime. Comparing the simulated heat fluxes and synthetic Te fluctuations with the experiments, it is found that GYRO can reproduce the ion heat flux and the Te fluctuation level measured by the Correlation ECE (CECE) diagnostic within their uncertainties, while the simulated electron heat flux is under-predicted. Furthermore, the synthetic Te spectral shape is not matched with the measured spectrum in both LOC/SOC discharges. We have also performed global simulations to consider the interaction of turbulence within the sampling volume of the CECE diagnostic, enabling us to evaluate the importance of global simulations in applying a synthetic CECE diagnostic in this study. The LOC/SOC transition physics will be also explored. Research supported by USDoE Awards DE-SC0006419, DE-FC02-99ER54512.

  1. Upgrade of the Edge Charge Exchange Diagnostic on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    McDermott, Rachael; Lipschultz, Bruce; Marr, Kenneth

    2006-10-01

    The current edge Charge Exchange Spectroscopy system is being upgraded to include both a beam viewing and a background viewing toroidal periscope. The beam viewing periscope will be focused on the center of the DNB and will cover a 4cm radial region at the edge of the plasma starting a few centimeters in from and extending up to 1centimeter outside of the last closed flux surface. The background periscope will view the same radial region but will be displaced toroidally by 36 degrees. Each periscope has 20 chordal views with a radial resolution of 2.5-3mm. The presence of a background periscope obviates the need for a chopped DNB by providing time synchronized background B^+4 spectral data that can be subtracted directly from the active beam-derived B^+4 line-shapes. This system has been designed to work in conjunction with the current poloidal Charge Exchange periscope which has 25 fibers focused in the same region with equivalent radial resolution. The new toroidal system will enable concurrent measurements of the poloidal and toroidal velocity as well as the temperature and density of the B^+5 ions in the edge pedestal region; a measurement that currently does not exist on C-Mod. This information will then be used to calculate radial electric field profiles and study edge physics phenomena.

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

    SciTech Connect

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

    2002-06-18

    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.

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

    SciTech Connect

    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.

    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.

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

    SciTech Connect

    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.

  5. Status of diagnostic development to measure parallel wavenumber of lower hybrid waves on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Baek, S. G.; Wallace, G. M.; Shinya, T.; Shiraiwa, S.; Parker, R. R.; Takase, Y.; Brunner, D.

    2015-12-01

    Recent lower hybrid (LH) current drive experiments on Alcator C-Mod have motivated measurement of the parallel wavenumber of LH waves with an aim to understand the significance of the k|| up-shift mechanisms such as scattering by turbulence or parametric decay instabilities. To this end, a new diagnostic system is under development, consisting of two rows of three RF magnetic loop probes (one row sensitive to B||, the other row B⊥) and three Langmuir probes. These will be mounted on a radially movable probe system on the low field side of the tokamak, which is magnetically mapped to the LH launcher but toroidally separated by about 110 deg from the launcher. This location is expected to be ideal for detecting the parallel wavenumber spectrum of the pump and sideband LH waves up to n|| of 6.5. The use of the loop probes will help unambiguously resolve the polarization of these waves. These loop probes have been developed under the collaboration with the University of Tokyo, and vacuum-compatible versions have recently been fabricated and tested on the bench. To evaluate the phase of the detected waves, the signals at 4.6 GHz will be frequency down-converted to 25 MHz in an intermediate frequency stage, and directly digitized at a sampling rate of 100 MS/sec. This system will output the dominant parallel wavenumber for each frequency, selected by controlling the frequency of a local oscillator in the IF stage. In addition to these loop probes, the Langmuir probes will be used to provide the density and temperature information at the measurement location to perform instability analyses. The Langmuir probes will be also used to examine the sensitivity of the radial measurement location on the strength of the sideband LH waves. Details of this proposed diagnostic system and the latest status will be presented.

  6. Microturbulent Drift Mode Stability before Internal Transport Barrier Formation in the Alcator C-Mod Radio Frequency Heated H-mode

    SciTech Connect

    M.H. Redi; W. Dorland; C.L. Fiore; P.T. Bonoli; M.J. Greenwald; J.E. Rice; J.A. Baumgaertel; T.S. Hahm; G.W. Hammett; K. Hill; D.C. McCune; D.R. Mikkelsen; G. Rewoldt

    2004-09-01

    H-mode experiments on Alcator C-Mod [I.H. Hutchinson, et al., Phys. Plasma 1 (1994) 1511] which exhibit an internal transport barrier (ITB), have been examined with gyrokinetic simulations, near the ITB onset time. Linear simulations support the picture of ion and electron temperature gradient (ITG, ETG) microturbulence driving high {chi}{sub i} and {chi}{sub e}, respectively, and that stable ITG correlates with reduced particle transport and improved ion thermal confinement on C-Mod. In the barrier region ITG is weakly unstable, with a critical temperature gradient higher than expected from standard models. Nonlinear calculations and the role of E x B shear suppression of turbulence outside the plasma core are discussed in light of recent profile measurements for the toroidal velocity. The gyrokinetic model benchmarks successfully against experiment in the plasma core.

  7. A Spatially Resolving X-ray Crystal Spectrometer for Measurement of Ion-temperature and Rotation-velocity Profiles on the AlcatorC-Mod Tokamak

    SciTech Connect

    Hill, K. W.; Bitter, M. L.; Scott, S. D.; Ince-Cushman, A.; Reinke, M.; Rice, J. E.; Beiersdorfer, P.; Gu, M. F.; Lee, S. G.; Broennimann, C. H.; Eikenberry, E. F.

    2009-03-24

    A new spatially resolving x-ray crystal spectrometer capable of measuring continuous spatial profiles of high resolution spectra (λ/dλ > 6000) of He-like and H-like Ar Kα lines with good spatial (~1 cm) and temporal (~10 ms) resolutions has been installed on the Alcator C-Mod tokamak. Two spherically bent crystals image the spectra onto four two-dimensional Pilatus II pixel detectors. Tomographic inversion enables inference of local line emissivity, ion temperature (Ti), and toroidal plasma rotation velocity (vφ) from the line Doppler widths and shifts. The data analysis techniqu

  8. Validation of full-wave simulations for mode conversion of waves in the ion cyclotron range of frequencies with phase contrast imaging in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

    Mode conversion of fast waves in the ion cyclotron range of frequencies (ICRF) is known to result in current drive and flow drive under optimised conditions, which may be utilized to control plasma profiles and improve fusion plasma performance. To describe these processes accurately in a realistic toroidal geometry, numerical simulations are essential. Quantitative comparison of these simulations and the actual experimental measurements is important to validate their predictions and to evaluate their limitations. The phase contrast imaging (PCI) diagnostic has been used to directly detect the ICRF waves in the Alcator C-Mod tokamak. The measurements have been compared with full-wave simulations through a synthetic diagnostic technique. Recently, the frequency response of the PCI detector array on Alcator C-Mod was recalibrated, which greatly improved the comparison between the measurements and the simulations. In this study, mode converted waves for D-3He and D-H plasmas with various ion species compositions were re-analyzed with the new calibration. For the minority heating cases, self-consistent electric fields and a minority ion distribution function were simulated by iterating a full-wave code and a Fokker-Planck code. The simulated mode converted wave intensity was in quite reasonable agreement with the measurements close to the antenna, but discrepancies remain for comparison at larger distances.

  9. Validation of full-wave simulations for mode conversion of waves in the ion cyclotron range of frequencies with phase contrast imaging in Alcator C-Mod

    SciTech Connect

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

    2015-08-15

    Mode conversion of fast waves in the ion cyclotron range of frequencies (ICRF) is known to result in current drive and flow drive under optimised conditions, which may be utilized to control plasma profiles and improve fusion plasma performance. To describe these processes accurately in a realistic toroidal geometry, numerical simulations are essential. Quantitative comparison of these simulations and the actual experimental measurements is important to validate their predictions and to evaluate their limitations. The phase contrast imaging (PCI) diagnostic has been used to directly detect the ICRF waves in the Alcator C-Mod tokamak. The measurements have been compared with full-wave simulations through a synthetic diagnostic technique. Recently, the frequency response of the PCI detector array on Alcator C-Mod was recalibrated, which greatly improved the comparison between the measurements and the simulations. In this study, mode converted waves for D-{sup 3}He and D-H plasmas with various ion species compositions were re-analyzed with the new calibration. For the minority heating cases, self-consistent electric fields and a minority ion distribution function were simulated by iterating a full-wave code and a Fokker-Planck code. The simulated mode converted wave intensity was in quite reasonable agreement with the measurements close to the antenna, but discrepancies remain for comparison at larger distances.

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

    PubMed

    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. PMID:26931846

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

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

    2009-11-26

    Mode converted ion cyclotron wave (ICW) has been observed with phase contrast imaging (PCI) in D-{sup 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.

  13. Application of PILATUS II Detector Modules for High Resolution X-Ray Imaging Crystal Spectrometers on the Alcator C-Mod Tokamak

    SciTech Connect

    M.L. Bitter, Ch. Borennimann, E.F. Eikenberry, K.W. Hill, A. Ince-Chushman, S.G. Lee, J.E. Rice, and S. Scott.

    2007-07-23

    A new type of X-ray imaging crystal spectrometer for Doppler measurements of the radial profiles of the ion temperature and plasma rotation velocity in tokamak plasmas is presently being developed in a collaboration between various laboratories. The spectrometer will consist of a spherically bent crystal and a two-dimensional position sensitive detector; and it will record temporally and spatially resolved X-ray line spectra from highly-charged ions. The detector must satisfy challenging requirements with respect to count rate and spatial resolution. The paper presents the results from a recent test of a PILATUS II detector module on Alcator C-Mod, which demonstrate that the PILATUS II detector modules will satisfy these requirements.

  14. A spatially resolving x-ray crystal spectrometer for measurement of ion-temperature and rotation-velocity profiles on the Alcator C-Mod tokamak

    SciTech Connect

    Hill, K. W.; Bitter, M. L.; Scott, S. D.; Ince-Cushman, A.; Reinke, M.; Rice, J. E.; Beiersdorfer, P.; Gu, M.-F.; Lee, S. G.; Broennimann, Ch.; Eikenberry, E. F.

    2008-10-15

    A new spatially resolving x-ray crystal spectrometer capable of measuring continuous spatial profiles of high resolution spectra ({lambda}/d{lambda}>6000) of He-like and H-like Ar K{alpha} lines with good spatial ({approx}1 cm) and temporal ({approx}10 ms) resolutions has been installed on the Alcator C-Mod tokamak. Two spherically bent crystals image the spectra onto four two-dimensional Pilatus II pixel detectors. Tomographic inversion enables inference of local line emissivity, ion temperature (T{sub i}), and toroidal plasma rotation velocity (v{sub {phi}}) from the line Doppler widths and shifts. The data analysis techniques, T{sub i} and v{sub {phi}} profiles, analysis of fusion-neutron background, and predictions of performance on other tokamaks, including ITER, will be presented.

  15. A new fast two-color interferometer at Alcator C-Mod for turbulence measurements and comparison with phase contrast imaging

    SciTech Connect

    Kasten, C. P. White, A. E.; Irby, J. H.

    2014-04-15

    Accurately predicting the turbulent transport properties of magnetically confined plasmas is a major challenge of fusion energy research. Validation of transport models is typically done by applying so-called “synthetic diagnostics” to the output of nonlinear gyrokinetic simulations, and the results are compared to experimental data. As part of the validation process, comparing two independent turbulence measurements to each other provides the opportunity to test the synthetic diagnostics themselves; a step which is rarely possible due to limited availability of redundant fluctuation measurements on magnetic confinement experiments. At Alcator C-Mod, phase-contrast imaging (PCI) is a commonly used turbulence diagnostic. PCI measures line-integrated electron density fluctuations with high sensitivity and wavenumber resolution (1.6 cm{sup −1}≲|k{sub R}|≲11 cm{sup −1}). A new fast two-color interferometry (FTCI) diagnostic on the Alcator C-Mod tokamak measures long-wavelength (|k{sub R}|≲3.0 cm{sup −1}) line-integrated electron density fluctuations. Measurements of coherent and broadband fluctuations made by PCI and FTCI are compared here for the first time. Good quantitative agreement is found between the two measurements. This provides experimental validation of the low-wavenumber region of the PCI calibration, and also helps validate the low-wavenumber portions of the synthetic PCI diagnostic that has been used in gyrokinetic model validation work in the past. We discuss possibilities to upgrade FTCI, so that a similar comparison could be done at higher wavenumbers in the future.

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

    SciTech Connect

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

    2012-08-15

    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.

  17. Reduced-model (SOLT) simulations of an EDA H-mode shot at Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Russell, D. A.; D'Ippolito, D. A.; Myra, J. R.; Labombard, B.; Terry, J. L.; Zweben, S. J.

    2011-10-01

    Reduced-model scrape-off layer turbulence (SOLT) simulations of an Enhanced D-Alpha (EDA) H-mode observed at C-Mod were conducted to explore observed variations in scrape-off-layer (SOL) width. The amplitude of a mean poloidal flow was varied to control the level of turbulence in the simulation and to reproduce the observed heat flux across the separatrix. SOL width decreased with increasing input power and with increasing separatrix temperature in both experiment and simulation, consistent with the strong temperature dependence of collision-limited parallel heat flux. A persistent quasi-coherent mode (QCM) dominates the SOLT turbulence. The wavelength of the SOLT QCM is comparable to that of the QCM consistently observed on C-Mod during EDA operation. The SOLT QCM consists of a quasi-stationary string of vortices, located just inside the separatrix, poloidally convected by the mean flow and occasionally emitting blobs into the SOL. The mode frequency is dominated by the Doppler shift of this convected pattern. Analysis reveals underlying drift-interchange and Kelvin-Helmholtz instabilities. Supported by USDOE under DE-FG02-97ER54392, DE-AC02-09CH11466, DE-FC02-99ER54512 and S009625-F.

  18. Mean flows and blob velocities in scrape-off layer (SOLT) simulations of an L-mode discharge on Alcator C-Mod

    DOE PAGESBeta

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

    2016-06-10

    Two-dimensional scrape-off layer turbulence (SOLT) code simulations are compared with an L-mode discharge on the Alcator C-Mod tokamak [M. Greenwald, et al., Phys. Plasmas 21, 110501 (2014)]. Density and temperature profiles for the simulations were obtained by smoothly fitting Thomson scattering and mirror Langmuir probe (MLP) data from the shot. Simulations differing in turbulence intensity were obtained by varying a dissipation parameter. Mean flow profiles and density fluctuation amplitudes are consistent with those measured by MLP in the experiment and with a Fourier space diagnostic designed to measure poloidal phase velocity. Blob velocities in the simulations were determined from themore » correlation function for density fluctuations, as in the analysis of gas-puff-imaging (GPI) blobs in the experiment. In the simulations, it was found that larger blobs moved poloidally with the ExB flow velocity, vE , in the near-SOL, while smaller fluctuations moved with the group velocity of the dominant linear (interchange) mode, vE + 1/2 vdi, where vdi is the ion diamagnetic drift velocity. Comparisons are made with the measured GPI correlation velocity for the discharge. The saturation mechanisms operative in the simulation of the discharge are also discussed. In conclusion, it is found that neither sheared flow nor pressure gradient modification can be excluded as saturation mechanisms.« less

  19. Mean Flows and Blob Velocities in Scrape-Off Layer (SOLT) Simulations of an L-mode discharge on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Russell, D. A.; Myra, J. R.; D'Ippolito, D. A.; Labombard, B.; Terry, J. L.; Zweben, S. J.

    2015-11-01

    Two-dimensional scrape-off layer turbulence (SOLT) code simulations are compared with an L-mode discharge on Alcator C-Mod. Density and temperature profiles for the simulations were obtained by smoothly fitting Thomson scatter and mirror Langmuir probe (MLP) data from the shot. Simulations differing in turbulence intensity were obtained by varying a dissipation parameter. Mean flow profiles and density fluctuation amplitudes are consistent with those measured by MLP in the experiment. Blob velocities in the simulations were determined from the correlation function for density fluctuations, as in the analysis of gas-puff-imaging (GPI) blobs in the experiment. In the simulations, it was found that larger blobs moved poloidally with the ExB flow velocity, vE , in the near-SOL, while smaller fluctuations moved with the group velocity of the dominant linear (interchange) mode, vE + 1/2 vdi, where vdi is the ion diamagnetic drift velocity. Comparisons are made with the measured GPI correlation velocity. The saturation mechanisms operative in the simulation of the discharge are explored. Work supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences, under Agreement DE-FC02-99ER54512, Contract DE-AC02-09CH11466, and Princeton Plasma Physics Laboratory Subcontract S013429-U.

  20. Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

    SciTech Connect

    Lore, J. D.; Reinke, M. L.; Lipschultz, B.; Brunner, D.; LaBombard, B.; Terry, J.; Pitts, R. A.; Feng, Y.

    2015-05-15

    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.

  1. Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

    SciTech Connect

    Lore, Jeremy D.; Reinke, M. L.; Brunner, D.; LaBombard, B. A.; Lipschultz, B.; Terry, J. L.; Pitts, R. A.; Feng, Y.

    2015-04-28

    We study 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-modeplasmas. This is in contrast to similar experiments in Ohmically heated L-modeplasmas, 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. In conclusion, 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.

  2. New insights on boundary plasma turbulence and the quasi-coherent mode in Alcator C-Mod using a Mirror Langmuir Probe

    SciTech Connect

    LaBombard, B.; Golfinopoulos, T.; Terry, J. L.; Brunner, D.; Davis, E.; Greenwald, M.; Hughes, J. W.

    2014-05-15

    A new “Mirror Langmuir Probe” diagnostic, combined with a double-coil scanning magnetic probe, is used to interrogate Alcator C-Mod's quasi-coherent mode (QCM) with unprecedented detail. In ohmic EDA H-modes, the QCM is found to reside in a region of positive radial electric field, with a radial width (∼3 mm) that spans open and closed field line regions. Large amplitude, in-phase sinusoidal bursts (∼100 kHz) in density, electron temperature, and plasma potential are observed, with potential lagging density by ∼16°, producing an outward radial transport velocity of ∼10 m/s. Mode propagation corresponds to the sum of local E × B and electron diamagnetic drift velocities. Poloidal magnetic field fluctuations project to current filaments carrying peak current densities of ∼25 A/cm{sup 2}. An evaluation of parallel electron force balance (Ohm's law) over a fluctuation cycle indicates a significant electromotive component. Interchange drive is also a contributor in the current continuity (vorticity) equation. Thus, the QCM is primarily a separatrix-spanning electron drift-wave with interchange and electromagnetic contributions.

  3. Comparison of Edge Turbulence Imaging at Two Different Poloidal Locations in the Scrape-off Layer of Alcator C-Mod

    SciTech Connect

    S.J. Zweben, et. al.

    2013-03-29

    This paper describes 2-D imaging measurements of plasma turbulence made in the scrape-off layer of the Alcator C-Mod tokamak simultaneously at two different poloidal locations, one near the outer midplane and the other near the divertor X-point region. These images were made with radial and poloidal resolution using two gas puff imaging (GPI) diagnostics, which were not directly connected along a B field line. The turbulence correlation structure has a significantly different tilt angle with respect to the local flux surfaces for the midplane and X-regions, and a slightly different ellipticity and size. The time-averaged turbulence velocities can be different in the midplane and Xregions, even within the same flux surface in the same shot, and in most cases the fluctuations in poloidal velocity in these two regions were not correlated. These structures are partially consistent with a magnetic flux tube mapping model, and the velocities are compared with various poloidal flow models.

  4. Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

    DOE PAGESBeta

    Lore, Jeremy D.; Reinke, M. L.; Brunner, D.; LaBombard, B. A.; Lipschultz, B.; Terry, J. L.; Pitts, R. A.; Feng, Y.

    2015-04-28

    We study 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-modeplasmas. This is in contrast to similar experiments in Ohmically heated L-modeplasmas, 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 tomore » 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. In conclusion, 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.« less

  5. Mean flows and blob velocities in scrape-off layer (SOLT) simulations of an L-mode discharge on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

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

    2016-06-01

    Two-dimensional scrape-off layer turbulence (SOLT) code simulations are compared with an L-mode discharge on the Alcator C-Mod tokamak [Greenwald et al., Phys. Plasmas 21, 110501 (2014)]. Density and temperature profiles for the simulations were obtained by smoothly fitting Thomson scattering and mirror Langmuir probe (MLP) data from the shot. Simulations differing in turbulence intensity were obtained by varying a dissipation parameter. Mean flow profiles and density fluctuation amplitudes are consistent with those measured by MLP in the experiment and with a Fourier space diagnostic designed to measure poloidal phase velocity. Blob velocities in the simulations were determined from the correlation function for density fluctuations, as in the analysis of gas-puff-imaging (GPI) blobs in the experiment. In the simulations, it was found that larger blobs moved poloidally with the E × B flow velocity, vE, in the near-SOL, while smaller fluctuations moved with the group velocity of the dominant linear (interchange) mode, vE + 1/2 vdi, where vdi is the ion diamagnetic drift velocity. Comparisons are made with the measured GPI correlation velocity for the discharge. The saturation mechanisms operative in the simulation of the discharge are also discussed. It is found that neither sheared flow nor pressure gradient modification can be excluded as saturation mechanisms.

  6. Experimental Investigation of RF Sheath Rectification in ICRF and LH Heated Plasmas on Alcator C-Mod

    SciTech Connect

    Ochoukov, R.; Whyte, D. G.; Faust, I.; LaBombard, B.; Lipschultz, B.; Meneghini, O.; Wallace, G.; Wukitch, S.; Myra, J.

    2011-12-23

    Radio frequency (RF) rectification of the plasma sheath is being actively studied on C-Mod as a likely mechanism that leads to prohibitively high molybdenum levels in the plasma core of ion cyclotron RF (ICRF) heated discharges. We installed emissive, ion sensitive, Langmuir, and 3-D B-dot probes to quantify the plasma potentials ({Phi}{sub P}) in ICRF and lower hybrid (LH) heated discharges. Two probe sets were mounted on fixed limiter surfaces and one set of probes was mounted on a reciprocating (along the major radius) probe. Initial results showed that RF rectification is strongly dependent on the local plasma density and not on the local RF fields. The RF sheaths had a threshold-like appearance at the local density of {approx}10{sup 16} m-{sup 3}. Radial probe scans revealed that the RF sheaths peaked in the vicinity of the ICRF limiter surface, agreeing with a recent theory. The highest {Phi}{sub P}'s were observed on magnetic field lines directly mapped to the active ICRF antenna. Measurements in LH heated plasmas showed a strong {Phi}{sub P} dependence on the parallel index of refraction n{sub ||} of the launched LH waves: {Phi}{sub P} is greater at lower n{sub ||}. Little dependence was observed on the local plasma density.

  7. Heat-flux footprints for I-mode and EDA H-mode plasmas on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Terry, J. L.; LaBombard, B.; Brunner, D.; Hughes, J. W.; Reinke, M. L.; Whyte, D. G.

    2013-07-01

    IR thermography is used to measure the heat flux footprints on C-Mod's outer target in I-mode and EDA H-mode plasmas. The footprint profiles are fit to a function with a simple physical interpretation. The fit parameter that is sensitive to the power decay length into the SOL, λSOL, is ˜1-3× larger in I-modes than in H-modes at similar plasma current, which is the dominant dependence for the H-mode λSOL. In contrast, the fit parameter sensitive to transport into the private-flux-zone along the divertor leg is somewhat smaller in I-mode than in H-mode, but otherwise displays no obvious dependence on Ip, Bt, or stored energy. A third measure of the footprint width, the "integral width", is not significantly different between H- and I-modes. Also discussed are significant differences in the global power flows of the H-modes with "favorable"∇B drift direction and those of the I-modes with "unfavorable"∇B drift direction.

  8. Simulation of 3D effects on partially detached divertor conditions in NSTX and Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lore, Jeremy

    2014-10-01

    Establishing a validated, predictive capability for the divertor plasma is critical for future fusion reactors, which must operate with detached divertors to reduce peak heat fluxes to the plasma facing components (PFC) and to mitigate net material erosion. This is challenging even for existing 2D codes, and is complicated further by non-axisymmetric effects due to divertor-localized gas injection, 3D magnetic fields, and 3D PFCs, modeling of which requires 3D simulations. New experiments performed on C-Mod at the request of the ITER organization to examine the consequence of localized nitrogen gas injection, show clear toroidal asymmetries in radiated power, impurity radiation, and divertor pressure. The 3D plasma/neutral transport code EMC3-EIRENE has been applied to model these experiments in the first attempt to benchmark the code against tokamak experimental data under detached conditions. The measured pressure modulation and the impurity radiation trends in the edge are qualitatively reproduced by the simulations, which also predict a ~2x modulation in heat flux at the outer strike point. Discrepancies are found in comparison with the measured private region radiation, and the simulations also indicate colder, denser divertor conditions than measured, suggesting that drifts and kinetic corrections may be required for more quantitative agreement. In separate experiments on NSTX, detached divertor plasmas are observed to reattach when 3D fields are applied. Modeling of the NSTX experiments reproduces these trends, with an increased peak heat flux with 3D fields and qualitative agreement of the striated flux patterns. The experimental identification of toroidal asymmetries in detached plasmas highlights the need for reliable 3D models for projecting the impact for ITER and beyond. Support from USDOE DE-AC05-00OR22725, DE-AC02-09CH11466, DE-FC02-99ER54512.

  9. Helium ELMy H-modes in Alcator C-Mod in Support of ITER Helium Operating Phases

    NASA Astrophysics Data System (ADS)

    Kessel, C. E.; Wolfe, S. M.; Chilenski, M. A.; Hughes, J. W.; Lin, Y.; Reinke, M. L.; Wukitch, S. J.; C-Mod Team

    2015-11-01

    ITER will operate helium majority plasmas in its earlier phases to shakedown the facility and provide plasmas in both L-mode and H-mode for commissioning and preparation for DT burning plasma operation. Part of this activity is to produce ELMy H-modes to test ELM mitigation schemes and observe the ELM impacts on the plasma facing components. It is of interest to characterize helium ELMy H-modes on present experiments to provide some basis to project to ITER and anticipate the plasma performance and ability to obtain H-modes with sufficient performance. ELMy H-mode is accessed in C-Mod by using LSN with an elongation of about 1.55, and with high lower triangularity and low upper triangularity. These regimes were produced with 1.5-4.0 MW of ICRF heating, and with H-mode line average densities of 2.0-3.2x1020 /m3, producing higher frequency repetitive to large infrequent ELMs, respectively. The infrequent ELM regime showed a cross between EDA and ELMy H-mode, with the EDA signature of a quasi-coherent mode at about 200 kHz. Tungsten laser blow-off was done. The pedestal features, energy confinement, ELM character, L-H threshold (1.7-2.5 MW) and W confinement will be discussed. Comparisons with deuterium ELMy H-modes will be made. Work supported by DOE DE-AC02-09CH11466 and DE-FC02-99ER54512.

  10. Studies of turbulence and transport in Alcator C-Mod H-mode plasmas with phase contrast imaging and comparisons with GYRO

    SciTech Connect

    Lin, L.; Porkolab, M.; Edlund, E. M.; Rost, J. C.; Fiore, C. L.; Greenwald, M.; Lin, Y.; Tsujii, N.; Wukitch, S. J.; Mikkelsen, D. R.

    2009-01-15

    Recent advances in gyrokinetic simulation of core turbulence and associated transport requires an intensified experimental effort to validate these codes using state of the art synthetic diagnostics to compare simulations with experimental data. A phase contrast imaging (PCI) diagnostic [M. Porkolab, J. C. Rost, N. Basse et al., IEEE Trans. Plasma Sci. 34, 229 (2006)] is used to study H-mode plasmas in Alcator C-Mod [M. Greenwald, D. Andelin, N. Basse et al., Nucl. Fusion 45, S109 (2005)]. The PCI system is capable of measuring density fluctuations with high temporal (2 kHz-5 MHz) and wavenumber (0.5-55 cm{sup -1}) resolution. Recent upgrades have enabled PCI to localize the short wavelength turbulence in the electron temperature gradient range and resolve the direction of propagation (i.e., electron versus ion diamagnetic direction) of the longer wavelength turbulence in the ion temperature gradient (ITG) and trapped electron mode range. The studies focus on plasmas before and during internal transport barrier formation in an enhanced D{sub {alpha}} H-mode plasma assisted with ion cyclotron resonance frequency heating. Nonlinear GYRO simulations have also been performed [J. Candy and R. E. Waltz, Phys. Rev. Lett. 91, 045001 (2003)] and the predicted fluctuation is compared against experimental measurements through a synthetic PCI diagnostic method. The simulated fluctuations from GYRO agree with experimental measurements in the ITG regime. GYRO also shows good agreement in transport predictions with experimental measurements after reducing the ion temperature gradient ({approx}15%) and adding ExB shear suppression, all within the experimental uncertainty.

  11. External excitation of a short-wavelength fluctuation in the Alcator C-Mod edge plasma and its relationship to the quasi-coherent mode

    SciTech Connect

    Golfinopoulos, T.; LaBombard, B.; Parker, R. R.; Burke, W.; Davis, E.; Granetz, R.; Greenwald, M.; Irby, J.; Leccacorvi, R.; Marmar, E.; Parkin, W.; Porkolab, M.; Terry, J.; Vieira, R.; Wolfe, S.

    2014-05-15

    A novel “Shoelace” antenna has been used to inductively excite a short-wavelength edge fluctuation in a tokamak boundary layer for the first time. The principal design parameters, k{sub ⊥}=1.5±0.1 cm{sup −1} and 45Alcator C-Mod, responsible for exhausting impurities in the steady-state, ELM-free Enhanced D{sub α} H-mode. In H-mode, whether or not there is a QCM, the antenna drives coherent, field-aligned perturbations in density, n{sup ~}{sub e}, and field, B{sup ~}{sub θ}, which are guided by field lines, propagate in the electron diamagnetic drift direction, and exhibit a weakly damped (γ/ω{sub 0}∼5%−10%) resonance near the natural QCM frequency. This result is significant, offering the possibility that externally driven modes may be used to enhance particle transport. In L-mode, the antenna drives only a non-resonant B{sup ~}{sub θ} response. The facts that the driven mode has the same wave number and propagation direction as the QCM, and is resonant at the QCM frequency, suggest the antenna may couple to this mode, which we have shown elsewhere to be predominantly drift-mode-like [B. LaBombard et al., Phys. Plasmas 21, 056108 (2014)].

  12. Wide-frequency range, dynamic matching network and power system for the “Shoelace” radio frequency antenna on the Alcator C-Mod tokamak

    SciTech Connect

    Golfinopoulos, Theodore LaBombard, Brian; Burke, William; Parker, Ronald R.; Parkin, William; Woskov, Paul

    2014-04-15

    A wide-frequency range (50–300 kHz) power system has been implemented for use with a new RF antenna – the “Shoelace” antenna – built to drive coherent plasma fluctuations in the edge of the Alcator C-Mod tokamak. A custom, dynamically tunable matching network allows two commercial 1 kW, 50-Ω RF amplifiers to drive the low-impedance, inductive load presented by the antenna. This is accomplished by a discretely variable L-match network, with 81 independently selected steps available for each of the series and parallel legs of the matching configuration. A compact programmable logic device provides a control system that measures the frequency with better than 1 kHz accuracy and transitions to the correct tuning state in less than 1 ms. At least 85% of source power is dissipated in the antenna across the operational frequency range, with a minimum frequency slew rate of 1 MHz/s; the best performance is achieved in the narrower band from 80 to 150 kHz which is of interest in typical experiments. The RF frequency can be run with open-loop control, following a pre-programmed analog waveform, or phase-locked to track a plasma fluctuation diagnostic signal in real time with programmable phase delay; the amplitude control is always open-loop. The control waveforms and phase delay are programmed remotely. These tools have enabled first-of-a-kind measurements of the tokamak edge plasma system response in the frequency range and at the wave number at which coherent fluctuations regulate heat and particle transport through the plasma boundary.

  13. Wide-frequency range, dynamic matching network and power system for the "Shoelace" radio frequency antenna on the Alcator C-Mod tokamak.

    PubMed

    Golfinopoulos, Theodore; LaBombard, Brian; Burke, William; Parker, Ronald R; Parkin, William; Woskov, Paul

    2014-04-01

    A wide-frequency range (50-300 kHz) power system has been implemented for use with a new RF antenna - the "Shoelace" antenna - built to drive coherent plasma fluctuations in the edge of the Alcator C-Mod tokamak. A custom, dynamically tunable matching network allows two commercial 1 kW, 50-Ω RF amplifiers to drive the low-impedance, inductive load presented by the antenna. This is accomplished by a discretely variable L-match network, with 81 independently selected steps available for each of the series and parallel legs of the matching configuration. A compact programmable logic device provides a control system that measures the frequency with better than 1 kHz accuracy and transitions to the correct tuning state in less than 1 ms. At least 85% of source power is dissipated in the antenna across the operational frequency range, with a minimum frequency slew rate of 1 MHz/s; the best performance is achieved in the narrower band from 80 to 150 kHz which is of interest in typical experiments. The RF frequency can be run with open-loop control, following a pre-programmed analog waveform, or phase-locked to track a plasma fluctuation diagnostic signal in real time with programmable phase delay; the amplitude control is always open-loop. The control waveforms and phase delay are programmed remotely. These tools have enabled first-of-a-kind measurements of the tokamak edge plasma system response in the frequency range and at the wave number at which coherent fluctuations regulate heat and particle transport through the plasma boundary. PMID:24784610

  14. Wide-frequency range, dynamic matching network and power system for the "Shoelace" radio frequency antenna on the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Golfinopoulos, Theodore; LaBombard, Brian; Burke, William; Parker, Ronald R.; Parkin, William; Woskov, Paul

    2014-04-01

    A wide-frequency range (50-300 kHz) power system has been implemented for use with a new RF antenna - the "Shoelace" antenna - built to drive coherent plasma fluctuations in the edge of the Alcator C-Mod tokamak. A custom, dynamically tunable matching network allows two commercial 1 kW, 50-Ω RF amplifiers to drive the low-impedance, inductive load presented by the antenna. This is accomplished by a discretely variable L-match network, with 81 independently selected steps available for each of the series and parallel legs of the matching configuration. A compact programmable logic device provides a control system that measures the frequency with better than 1 kHz accuracy and transitions to the correct tuning state in less than 1 ms. At least 85% of source power is dissipated in the antenna across the operational frequency range, with a minimum frequency slew rate of 1 MHz/s; the best performance is achieved in the narrower band from 80 to 150 kHz which is of interest in typical experiments. The RF frequency can be run with open-loop control, following a pre-programmed analog waveform, or phase-locked to track a plasma fluctuation diagnostic signal in real time with programmable phase delay; the amplitude control is always open-loop. The control waveforms and phase delay are programmed remotely. These tools have enabled first-of-a-kind measurements of the tokamak edge plasma system response in the frequency range and at the wave number at which coherent fluctuations regulate heat and particle transport through the plasma boundary.

  15. Characterization of density fluctuations during the search for an I-mode regime on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Marinoni, A.; Rost, J. C.; Porkolab, M.; Hubbard, A. E.; Osborne, T. H.; White, A. E.; Whyte, D. G.; Rhodes, T. L.; Davis, E. M.; Ernst, D. R.; Burrell, K. H.

    2015-09-01

    The I-mode regime, routinely observed on the Alcator C-Mod tokamak, is characterized by an edge energy transport barrier without an accompanying particle barrier and with broadband instabilities, known as weakly coherent modes (WCM), believed to regulate particle transport at the edge. Recent experiments on the DIII-D tokamak exhibit I-mode characteristics in various physical quantities. These DIII-D plasmas evolve over long periods, lasting several energy confinement times, during which the edge electron temperature slowly evolves towards an H-mode-like profile, while maintaining a typical L-mode edge density profile. During these periods, referred to as I-mode phases, the radial electric field at the edge also gradually reaches values typically observed in H-mode. Density fluctuations measured with the phase contrast imaging diagnostic during I-mode phases exhibit three features typically observed in H-mode on DIII-D, although they develop progressively with time and without a sharp transition: the intensity of the fluctuations is reduced; the frequency spectrum is broadened and becomes non-monotonic; two dimensional space-time spectra appear to approach those in H-mode, showing phase velocities of density fluctuations at the edge increasing to about 10 km s-1. However, in DIII-D there is no clear evidence of the WCM. Preliminary linear gyro-kinetic simulations are performed in the pedestal region with the GS2 code and its recently upgraded model collision operator that conserves particles, energy and momentum. The increased bootstrap current and flow shear generated by the temperature pedestal are shown to decrease growth rates, thus possibly generating a feedback mechanism that progressively stabilizes fluctuations.

  16. Recent DIII-D results

    SciTech Connect

    Petersen, P.I.

    1994-07-01

    This paper summarizes the recent DIII-D experimental results and the development of the relevant hardware systems. The DIII-D program focuses on divertor solutions for next generation tokamaks such as International Thermo-nuclear Experimental Reactor (ITER) and Tokamak Physics Experiment (TPX), and on developing configurations with enhanced confinement and stability properties that will lead to a more compact and economical fusion reactor. The DIII-D program carries out this research in an integrated fashion.

  17. Spatial structure of scrape-off-layer filaments near the midplane and X-point regions of Alcator C-Mod

    SciTech Connect

    Terry, J L; Zweben, S J; Umansky, M V; Cziegler, I; Grulke, O; LaBombard, B; Stotler, D P

    2008-05-22

    Movies of edge turbulence at both the outboard midplane and the region outboard of the typical lower X-point location in C-Mod have been obtained using Gas-Puff-Imaging together with fast-framing cameras. Intermittent turbulent structures, typically referred to as blobs or filaments, are observed in both locations. Near the midplane the filaments are roughly circular in cross-section, while in the X-point region they are highly elongated. Filament velocities in this region are {approx}3x faster than the radial velocities at the midplane, in a direction roughly normal to the local flux surfaces. The observations are consistent with the picture that the filaments arise in outboard region and, as a consequence of the rapid parallel diffusion of the potential perturbations, map along field lines. A simulation using the 3D BOUT turbulence code has been made, with the result that reproduces many of the spatial features observed in the experiment.

  18. DIII-D research operations

    SciTech Connect

    Baker, D.

    1993-05-01

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R D; and collaborative efforts.

  19. Nonaxisymmetric field effects on Alcator C-Moda)

    NASA Astrophysics Data System (ADS)

    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.

  20. DIII-D PLASMA CONTROL SIMULATION ENVIRONMENT

    SciTech Connect

    LEUER,J.A; DERANIAN,R.D; FERRON,J.R; HUMPHREYS,D.A; JOHNSON,R.D; PENAFLOR,B.G; WALKER,M.L; WELANDER,A.S; KHAYRUTDINOV,R.R; DOKOUKA,V; EDGELL,D.H; FRANSSON,C.M

    2003-10-01

    OAK-B135 Many advanced have been made to the DIII-D plasma control simulation environment since the previously developed hardware-in-the-loop plasma shape simulation capability was reported. In the present paper they summarize the major improvements to this simulation environment, including, introduction of the non-linear plasma evolution code DINA. Comparisons with DIII-D experimental results are presented. Recent model developments in advanced neoclassical tearing mode (NTM) and resistive wall mode (RWM) control are presented.

  1. Boronization in DIII-D

    SciTech Connect

    Jackson, G.L.; Burrell, K.H.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Hodapp, T.; Kellman, A.G.; Lee, R.; Lippman, S.I.; Phillips, J.; Taylor, T.S.; West, W.P.; Winter, J.; Moyer, R.; Watkins, J.

    1992-05-01

    A thin boron film has been applied to the DIII-D tokamak plasma facing surfaces to reduce impurity influx, particularly oxygen and carbon. A direct result of this surface modification was the observation of a regime of very high energy confinement, VH-mode, with confinement times from 1.5 to 2 times greater than predicted by H-mode scaling relation for the same set of parameters. VH-mode discharges are characterized by low ohmic target densities, low edge neutral pressure, and reduced cycling. These conditions have reduced the collisionality, {nu}*, in the edge region producing a higher edge pressure gradient and a significant bootstrap current, up to 30% of the total current. We will describe the edge plasma properties after boronization including reductions in recycling inferred from measurements of {tau}{sup p}*. In particular we will discuss the edge plasma conditions necessary for access to VH-mode including the boronization process and properties of the deposited film.

  2. Turbulent Impurity Transport Modeling for C-Mod

    NASA Astrophysics Data System (ADS)

    Fu, Xiangrong; Horton, Wendell; Rowan, William; Bespamyatnov, Igor; Benkadda, Sadruddin; Fiore, Catherine

    2012-03-01

    Turbulent particle transport is investigated by analyzing boron impurity transport experiments in the Alcator C-Mod transport experiments with a quasilinear theory. Eigenvalue problems for sets of reduced fluid equations for the multi-component plasmas are solved to get the fluctuating field vector composed of the electric potential φ, the main ion density δni, the impurity density δnz and the ion temperature fluctuation δTi(for ITG). For Alcator C-Mod parameters, we investigate three drift waves models (1) the usual drift waves driven by density gradients, (2)impurity drift waves supported by the impurity density gradients and (3)turbulence driven by ITG mode. With turbulent spectrum obtained from simulations or nonlinear theories, we calculate particle transport coefficients and compare with the experiment and the neoclassical theory. This procedure results in a fast code that could run in real-time on the transport time scale to give the particle fluxes as a function of the state of the plasma. The code may be extended to include multiple modes for a more complete description of plasmas. Examples for the particle fluxes are given for C-Mod in the H modes and newly discovered I modes. Recent experiments reported on LHD are briefly discussed.

  3. Dimensionless size scaling of intrinsic rotation in DIII-D

    NASA Astrophysics Data System (ADS)

    deGrassie, J. S.; Solomon, W. M.; Rice, J. E.; Noterdaeme, J.-M.

    2016-08-01

    A dimensionless empirical scaling for intrinsic toroidal rotation is given: MA˜βNρ* , where MA is the toroidal velocity divided by the Alfvén velocity, βN is the usual normalized β value, and ρ* is the ion gyroradius divided by the minor radius. This scaling describes well experimental data from DIII-D and also some published data from C-Mod and JET. The velocity used in this scaling is in an outer location in minor radius, outside of the interior core and inside of the large gradient edge region in H-mode conditions. This scaling establishes the basic magnitude of the intrinsic toroidal rotation, and its relation to the rich variety of rotation profiles that can be realized for intrinsic conditions is discussed. This scaling has some similarities to existing dimensioned scalings, both the Rice scaling [J. E. Rice et al., Phys. Plasmas 7, 1825 (2000)] and the scaling of Parra et al. [Phys. Rev. Lett. 108, 095001 (2012)]. These relationships are described.

  4. Infrared thermography system on DIII-D

    SciTech Connect

    Petrie, T.W. ); Hill, D.N.; Baptista, J.; Brown, M. )

    1990-10-01

    Six infrared cameras measure temperature changes on the protective graphite armor inside the DIII-D vacuum vessel. Simultaneous time dependent temperature measurements are made on armor tiles located on the centerpost and divertor regions, and on both outboard limiters. The nearly-complete poloidal coverage is useful in measuring both the plasma heat flux distributions inside the vessel and the plasma power balance. Spatial resolution of each camera system is {approx lt}1 cm, while the minimum resolvable time is 125 {mu}sec. Data from the IR TV systems is recorded on video tape, and is post-processed serially, using an image processor with an AT-compatible microcomputer. The processing system controls all VCRs, interprets DIII-D timing pulses, digitizes video data in the pre-determined regions of interest, averages digitized signals to reduce noise, and constructs data files which are then stored as part of the permanent shot record.

  5. Dust Studies in DIII-D Tokamak

    SciTech Connect

    Rudakov, D L; West, W P; Groth, M; Yu, J H; Boedo, J A; Bray, B D; Brooks, N H; Fenstermacher, M E; Hollmann, E M; Hyatt, A W; Krasheninnikov, S I; Lasnier, C J; Moyer, R A; Pigarov, A Y; Smirnov, R; Solomon, W M; Wong, C C

    2008-04-15

    Studies of submicron dust using Mie scattering from Nd:YAG lasers and video data of micron to sub-millimeter sized dust on DIII-D tokamak have provided the first data of dust sources and transport during tokamak discharges. During normal operation on DIII-D dust observation rates are low, a few events per discharge or less. The net carbon content of the dust corresponds to a carbon atom density a few orders of magnitude below the core impurity density. Statistical analysis of Mie data collected over months of operation reveal correlation of increased dust rate with increased heating power and impulsive wall loading due to edge localized modes (ELMs) and disruptions. Generation of significant amounts of dust by disruptions is confirmed by the camera data. However, dust production by disruptions alone is insufficient to account for estimated in-vessel dust inventory in DIII-D. After an extended entry vent, thousands of dust particles are observed by cameras in the first 2-3 plasma discharges. Individual particles moving at velocities up to {approx}300 m/s, breakup of larger particles into pieces, and collisions of particles with walls are observed. After {approx}70 discharges, dust levels are reduced to a few events per discharge. In order to calibrate diagnostics and benchmark modeling, milligram amounts of micron-sized carbon dust have been injected into DIII-D discharges, leading to the core carbon density increase by a factor of 2-3. Following injection, dust trajectories in the divertor are mostly in the toroidal direction, consistent with the ion drag force. Dust from the injection is observed in the outboard midplane by a fast framing camera. The observed trajectories and velocities of the dust particles are in qualitative agreement with modeling by the 3D DustT code.

  6. Dust Studies in DIII-D Tokamak

    SciTech Connect

    Rudakov, D. L.; Yu, J. H.; Boedo, J. A.; Hollmann, E. M.; Krasheninnikov, S. I.; Moyer, R. A.; Pigarov, A. Yu.; Smirnov, R.; West, W. P.; Bray, B. D.; Brooks, N. H.; Hyatt, A. W.; Wong, C. P. C.; Groth, M.; Fenstermacher, M. E.; Lasnier, C. J.; Solomon, W. M.

    2008-09-07

    Studies of submicron dust using Mie scattering from Nd:YAG lasers and video data of micron to sub-millimeter sized dust on DIII-D tokamak have provided the first data of dust sources and transport during tokamak discharges. During normal operation on DIII-D dust observation rates are low, a few events per discharge or less. The net carbon content of the dust corresponds to a carbon atom density a few orders of magnitude below the core impurity density. Statistical analysis of Mie data collected over months of operation reveal correlation of increased dust rate with increased heating power and impulsive wall loading due to edge localized modes (ELMs) and disruptions. Generation of significant amounts of dust by disruptions is confirmed by the camera data. However, dust production by disruptions alone is insufficient to account for estimated in-vessel dust inventory in DIII-D. After an extended entry vent, thousands of dust particles are observed by cameras in the first 2-3 plasma discharges. Individual particles moving at velocities up to {approx}300 m/s, breakup of larger particles into pieces, and collisions of particles with walls are observed. After {approx}70 discharges, dust levels are reduced to a few events per discharge. In order to calibrate diagnostics and benchmark modeling, milligram amounts of micron-sized carbon dust have been injected into DIII-D discharges, leading to the core carbon density increase by a factor of 2-3. Following injection, dust trajectories in the divertor are mostly in the toroidal direction, consistent with the ion drag force. Dust from the injection is observed in the outboard midplane by a fast framing camera. The observed trajectories and velocities of the dust particles are in qualitative agreement with modeling by the 3D DustT code.

  7. PERFORMANCE OF THE DIII-D SYSTEM

    SciTech Connect

    CALLIS,RW; KAJIWARA,K; LOHR,J; GORELOV,YA; PONCE,D

    2003-08-01

    A271 PERFORMANCE OF THE DIII-D SYSTEM. Three 110 GHz gyrotrons with nominal output power of 1 MW each have been installed and are operational on the DIII-D tokamak. All three gyrotrons were built by Communications and Power Industries (CPI). The CPI gyrotrons utilize a single disc CVD (chemical-vapor-deposition) diamond window that employs water cooling around the edge of the disc. Calculations predict that the CVD diamond window should be capable of full 1 MW cw operation, which is supported by IR camera measurements that show the window reaching equilibrium after 2.5 s. All gyrotrons are connected to the tokamak by low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HE{sub 11} mode. Each waveguide system incorporates a two-mirror launcher, which can steer the rf beam poloidally from the center to the outer edge of the plasma. Results obtained using the DIII-D ECH systems will be reported.

  8. DIII-D research program progress

    SciTech Connect

    Stambaugh, R.D.

    1990-11-01

    A summary of highlights of the research on the DIII-D tokamak in the last two years is given. At low q, toroidal beta ({beta}{sub T}) has reached 11%. At high q, {epsilon}{beta}{sub p} has reached 1.8. DIII-D data extending from one regime to the other show the beta limit is at least {beta}{sub T}(%) {ge} 3.5 I/aB (MA, m, T). Prospects for using H-mode in future devices have been enhanced. The discovery of negative edge electric fields and associated turbulence suppression have become part of an emerging theory of H-mode. Long pulse (10 second) H-mode with impurity control has been demonstrated. Radial sweeping of the divertor strike points and gas puffing under the X-point have lowered peak divertor plate heat fluxes a factor of 3 and 2 respectively. T{sub i} = 17 keV has been reached in a hot ion H-mode. Electron cyclotron current drive (ECCD) has produced up to 70 kA of driven current. Program elements now beginning are fast wave current drive (FWCD) and an advanced divertor program (ADP). 38 refs., 10 figs.

  9. DIII-D Advanced Tokamak Research Overview

    SciTech Connect

    V.S. Chan; C.M. Greenfield; L.L. Lao; T.C. Luce; C.C. Petty; G.M. Staebler

    1999-12-01

    This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously {beta}{sub N}H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues.

  10. Disruption mitigation studies in DIII-D

    SciTech Connect

    Taylor, P.L.; Kellman, A.G.; Evans, T.E.

    1999-01-01

    Data on the discharge behavior, thermal loads, halo currents, and runaway electrons have been obtained in disruptions on the DIII-D tokamak. These experiments have also evaluated techniques to mitigate the disruptions while minimizing runaway electron production. Experiments injecting cryogenic impurity killer pellets of neon and argon and massive amounts of helium gas have successfully reduced these disruption effects. The halo current generation, scaling, and mitigation are understood and are in good agreement with predictions of a semianalytic model. Results from killer pellet injection have been used to benchmark theoretical models of the pellet ablation and energy loss. Runaway electrons are often generated by the pellets and new runaway generation mechanisms, modifications of the standard Dreicer process, have been found to explain the runaways. Experiments with the massive helium gas puff have also effectively mitigated disruptions without the formation of runaway electrons that can occur with killer pellets.

  11. The DIII-D cryogenic system upgrade

    SciTech Connect

    Schaubel, K.M.; Laughon, G.J.; Campbell, G.L.; Langhorn, A.R.; Stevens, N.C.; Tupper, M.L.

    1993-10-01

    The original DIII-D cryogenic system was commissioned in 1981 and was used to cool the cryopanel arrays for three hydrogen neutral beam injectors. Since then, new demands for liquid helium have arisen including: a fourth neutral beam injector, ten superconducting magnets for the electron cyclotron heating gyrotrons, and more recently, the advanced diverter cryopump which resides inside the tokamak vacuum vessel. The original cryosystem could not meet these demands. Consequently, the cryosystem was upgraded in several phases to increase capacity, improve reliability, and reduce maintenance. The majority of the original system has been replaced with superior equipment. The capacity now exists to support present as well as future demands for liquid helium at DIII-D including a hydrogen pellet injector, which is being constructed by Oak Ridge National Laboratory. Upgrades to the cryosystem include: a recently commissioned 150 {ell}/hr helium liquefier, two 55 g/sec helium screw compressors, a fully automated 20-valve cryogen distribution box, a high efficiency helium wet expander, and the conversion of equipment from manual or pneumatic to programmable logic controller (PLC) control. The distribution box was designed and constructed for compactness due to limited space availability. Overall system efficiency was significantly improved by replacing the existing neutral beam reliquefier Joule-Thomson valve with a reciprocating wet expander. The implementation of a PLC-based automatic control system has resulted in increased efficiency and reliability. This paper will describe the cryosystem design with emphasis on newly added equipment. In addition, performance and operational experience will be discussed.

  12. Restoration of the DIII-D solenoid

    SciTech Connect

    Anderson, P.M.; Robinson, J.I.; Gonzales, E.; Rolens, G.W.

    1997-11-01

    The DIII-D tokamak has been operated since June 1995 with constrained ohmic heating capability as imposed by the abandonment of half of it`s solenoid system due to a cooling water leak. The solenoid is comprised of A and B windings with separate multiple power leads to each. The cooling water leak occurred in the lead of the B winding. This leak occurred in a remote area under the DIII-D vessel and is believed to be caused by magnetic forces developing cyclic bending loads on the conductor. Visual inspection of the lead using flexible bore scopes indicated that the structural fiberglass overwrap intended to band the supply and return leads into a primary-force canceling group had failed allowing individual conductors to become inadequately supported against bending loads. The overwrap failed as a result of poor epoxy encapsulation of the lead which was manufactured in 1978. Inspection of the A lead confirmed no overwrap failure and that the vacuum encapsulation of the A lead was proper and to specification. In order to continue operations, it was decided to abandon the B winding of the solenoid and operate under reduced (5 V-sec) capability. An in-situ repair approach was mandated by the extensive and lengthy effort required to disassemble, repair, and reassemble the tokamak. Access from outside the tokamak was severely limited. A plan to repair the damaged lead was developed and implemented over a 10 month period. This paper describes the repair of the solenoid lead. A VCR video tape of these remote installation efforts has been assembled and will be shown.

  13. DIII-D tokamak long range plan. Revision 3

    SciTech Connect

    1992-08-01

    The DIII-D Tokamak Long Range Plan for controlled thermonuclear magnetic fusion research will be carried out with broad national and international participation. The plan covers: (1) operation of the DIII-D tokamak to conduct research experiments to address needs of the US Magnetic Fusion Program; (2) facility modifications to allow these new experiments to be conducted; and (3) collaborations with other laboratories to integrate DIII-D research into the national and international fusion programs. The period covered by this plan is 1 November 19983 through 31 October 1998.

  14. Power Accounting in DIII-D

    NASA Astrophysics Data System (ADS)

    Martin, M. J.; Watkins, J. G.; Murphy, C. J.; Evans, T. E.; Jakubowski, M.; Joseph, I.; Lasnier, C. J.

    2007-11-01

    The plasma facing components in a fusion reactor will be exposed to high levels of heat and particle flux. A full accounting of where the plasma energy is deposited is important for designing future fusion reactors. To understand where the power is going, we do a total power accounting in DIII-D for different conditions such as H-mode, ohmic, and ELM-free H-mode plasmas. We use measurements from IR cameras, bolometers, Langmuir probes, and thermocouples to determine the distribution and magnitude of power deposition inside the tokamak as well as the consistency of the different measurements. We compare the thermal response measurements with both a simple thermal diffusion model and a finite element thermal model of the target plate tiles. A basic parameter that relates heat and particle flux at the plasma/materials interface is the power transmission factor. By comparing the particle flux and heat flux, we can measure the sheath factor profile for each of the above conditions.

  15. Density limit studies on DIII-D

    SciTech Connect

    Maingi, R.; Mahdavi, M.A.; Petrie, T.W.

    1998-08-01

    The authors have studied the processes limiting plasma density and successfully achieved discharges with density {approximately}50% above the empirical Greenwald density limit with H-mode confinement. This was accomplished by density profile control, enabled through pellet injection and divertor pumping. By examining carefully the criterion for MARFE formation, the authors have derived an edge density limit with scaling very similar to Greenwald scaling. Finally, they have looked in detail at the first and most common density limit process in DIII-D, total divertor detachment, and found that the local upstream separatrix density (n{sub e}{sup sep,det}) at detachment onset (partial detachment) increases with the scrape-off layer heating power, P{sub heat}, i.e., n{sub e}{sup sep,det} {approximately} P{sub heat}{sup 0.76}. This is in marked contrast to the line-average density at detachment which is insensitive to the heating power. The data are in reasonable agreement with the Borass model, which predicted that the upstream density at detachment would increase as P{sub heat}{sup 0.7}.

  16. Faraday Shield Development on DIII--D

    NASA Astrophysics Data System (ADS)

    Baity, F. W.; Goulding, R. H.; Hoffman, D. J.; Ryan, P. M.; Taylor, D. J.; Callis, R. W.; Pinsker, R. I.; Lindemuth, J. E.; Rosenfeld, J. H.

    1997-11-01

    DIII--D has been the proving ground for a number of innovative Faraday shield developments over the past ten years. The first Faraday shield used had two tiers of copper-plated Inconel rods of circular cross section with 3 mm thick graphite tiles brazed to the plasma-facing side of the front tier. Later antennas used shields with thin coatings of Ti (C,N) and boron carbide. All the coatings proved effective in reducing impurity influx from the antennas during RF operation. There are two shield designs in use currently. One is a single-tier of horizontal Inconel rods with a 6 μm layer of boron carbide applied by physical vapor deposition. The other design has molybdenum rods with a plasma-sprayed boron carbide coating approximately 100 μm thick. Based on comparative performance the thinner coating obtained with physical vapor deposition is preferred for future applicatrions. All Faraday shields have been passively cooled. Future plans call for tests of vanadium elements and of porous-metal helium-cooled elements.

  17. Disruption mitigation studies in DIII-D

    SciTech Connect

    Taylor, P.L.; Kellman, A.G.; Evans, T.E.; Gray, D.S.; Humphreys, D.A.; Hyatt, A.W.; Jernigan, T.C.; Lee, R.L.; Leuer, J.A.; Luckhardt, S.C.; Parks, P.B.; Schaffer, M.J.; Whyte, D.G.; Zhang, J.

    1999-05-01

    Data on the discharge behavior, thermal loads, halo currents, and runaway electrons have been obtained in disruptions on the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. {bold 8}, 2A 441 (1985)]. These experiments have also evaluated techniques to mitigate the disruptions while minimizing runaway electron production. Experiments injecting cryogenic impurity {open_quotes}killer{close_quotes} pellets of neon and argon and massive amounts of helium gas have successfully reduced these disruption effects. The halo current generation, scaling, and mitigation are understood and are in good agreement with predictions of a semianalytic model. Results from {open_quotes}killer{close_quotes} pellet injection have been used to benchmark theoretical models of the pellet ablation and energy loss. Runaway electrons are often generated by the pellets and new runaway generation mechanisms, modifications of the standard Dreicer process, have been found to explain the runaways. Experiments with the massive helium gas puff have also effectively mitigated disruptions without the formation of runaway electrons that can occur with {open_quotes}killer{close_quotes} pellets. {copyright} {ital 1999 American Institute of Physics.}

  18. Performance of spectral MSE diagnostic on C-Mod and ITER

    NASA Astrophysics Data System (ADS)

    Liao, Ken; Rowan, William; Mumgaard, Robert; Granetz, Robert; Scott, Steve; Marchuk, Oleksandr; Ralchenko, Yuri; Alcator C-Mod Team

    2015-11-01

    Magnetic field was measured on Alcator C-mod by applying spectral Motional Stark Effect techniques based on line shift (MSE-LS) and line ratio (MSE-LR) to the H-alpha emission spectrum of the diagnostic neutral beam atoms. The high field of Alcator C-mod allows measurements to be made at close to ITER values of Stark splitting (~ Bv⊥) with similar background levels to those expected for ITER. Accurate modeling of the spectrum requires a non-statistical, collisional-radiative analysis of the excited beam population and quadratic and Zeeman corrections to the Stark shift. A detailed synthetic diagnostic was developed and used to estimate the performance of the diagnostic at C-Mod and ITER parameters. Our analysis includes the sensitivity to view and beam geometry, aperture and divergence broadening, magnetic field, pixel size, background noise, and signal levels. Analysis of preliminary experiments agree with Kinetic+(polarization)MSE EFIT within ~2° in pitch angle and simulations predict uncertainties of 20 mT in | B | and <2° in pitch angle. This material is based upon work supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-FG03-96ER-54373 and DE-FC02-99ER54512.

  19. Results of DIII-D operation with new enabling technologies

    SciTech Connect

    Simonen, T.C.

    1997-03-01

    Recent experiments on DIII-D have been carried out to understand and explore optimized tokamak operating modes by exploiting control of the plasma current and pressure profiles using new RF current drive and divertor technology. DIII-D emphasizes plasma shape and divertor experiments using a digital plasma control system and extensive diagnostics to develop improved understanding and control of transport barriers in high performance plasmas. The emphasis of the program is to extend the duration of high performance operating modes beyond the plasma current relaxation time by using ICRF and ECH current drive. Engineering features of the new RF systems being developed for these experiments as well as new divertor results are described. DIII-D employs multi-element ICRF antennas for fast-wave electron heating and on-axis current drive and is beginning 110 GHz ECH experiments with MW-level gyrotrons for off-axis current drive. DIII-D employs active cryogenic divertor neutral particle pumping for plasma density and plasma pressure profile control. A divertor modification is now being implemented on DIII-D to pump higher triangularity plasmas and to better baffle neutral backflow from the recycling divertor region.

  20. Disruption Mitigation Experiments Carried Out on DIII-D

    NASA Astrophysics Data System (ADS)

    Commaux, N.; Baylor, L. R.; Jernigan, T. C.; Evans, T. E.; Humphreys, D. A.; Parks, P. B.; van Zeeland, M. A.; Wesley, J. C.; Hollmann, E. M.; James, A. N.; Yu, J. H.

    2009-11-01

    Disruptions are an important issue for ITER. Major and vertical displacement disruptions in ITER are predicted to generate multi-MeV runaway electron beams (RE) as well as high energy flux to the plasma facing components and high halo currents in the structures that could potentially damage the machine. To mitigate these phenomena, several techniques have been studied on DIII-D: massive gas injection (MGI), external magnetic perturbation, and injection of large shattered cryogenic pellets. MGI, which is proven to mitigate heat fluxes and halo currents, has been tested on DIII-D in terms of impurity radiation toroidal symmetry and optimization of the gas pulse length. Other techniques expected to mitigate the RE have also been tested on DIII-D: deconfinement of the RE using external magnetic fields and collisionally damping the avalanche multiplication process by rapidly increasing the electron density in the core by several orders of magnitude using a new shotgun pellet injector built by ORNL.

  1. Enhanced DIII-D Data Management Through a Relational Database

    NASA Astrophysics Data System (ADS)

    Burruss, J. R.; Peng, Q.; Schachter, J.; Schissel, D. P.; Terpstra, T. B.

    2000-10-01

    A relational database is being used to serve data about DIII-D experiments. The database is optimized for queries across multiple shots, allowing for rapid data mining by SQL-literate researchers. The relational database relates different experiments and datasets, thus providing a big picture of DIII-D operations. Users are encouraged to add their own tables to the database. Summary physics quantities about DIII-D discharges are collected and stored in the database automatically. Meta-data about code runs, MDSplus usage, and visualization tool usage are collected, stored in the database, and later analyzed to improve computing. Documentation on the database may be accessed through programming languages such as C, Java, and IDL, or through ODBC compliant applications such as Excel and Access. A database-driven web page also provides a convenient means for viewing database quantities through the World Wide Web. Demonstrations will be given at the poster.

  2. Microturbulence in DIII-D tokamak pedestal. I. Electrostatic instabilities

    SciTech Connect

    Fulton, D. P.; Holod, I.; Lin, Z.; Xiao, Y.

    2014-04-15

    Gyrokinetic simulations of electrostatic driftwave instabilities in a tokamak edge have been carried out to study the turbulent transport in the pedestal of an H-mode plasma. The simulations use annulus geometry and focus on two radial regions of a DIII-D experiment: the pedestal top with a mild pressure gradient and the middle of the pedestal with a steep pressure gradient. A reactive trapped electron instability with a typical ballooning mode structure is excited by trapped electrons in the pedestal top. In the middle of the pedestal, the electrostatic instability exhibits an unusual mode structure, which peaks at the poloidal angle θ=±π/2. The simulations find that this unusual mode structure is due to the steep pressure gradients in the pedestal but not due to the particular DIII-D magnetic geometry. Realistic DIII-D geometry appears to have a stabilizing effect on the instability when compared to a simple circular tokamak geometry.

  3. Performance of the DIII-D ECH System

    NASA Astrophysics Data System (ADS)

    Cengher, M.; Lohr, J.; Gorelov, Y. A.; Ponce, D.

    2013-10-01

    The six gyrotron ECH system on DIII-D is performing with good reliability while being upgraded. The total power injected into DIII-D has reached 3.4 MW, with pulse length up to 5 s. The power generated by the individual gyrotrons, the power injected into the tokamak, and the total energy injected into DIII-D will be shown for the present year on a shot-to-shot basis. The efficiency of a new transmission line for the most recently installed gyrotron was measured. This gyrotron injects up to 720 kW of power into DIII-D, for 915 kW of generated power. The polarization was checked and the results are shown in agreement with the computed values. The gyrotron ``Tinman'' was moved to a tank, formerly occupied by the poorly performing ``Han'' gyrotron, which developed an internal water leak. The re-measured transmission efficiency for this line is -0.96 dB. The use of TIMCON event controller to set ECH timing and aiming is expected to lead to a decrease in the time necessary to install the setup for a new shot, eliminate possible operator errors, and provide better coordination with other aspects of the experiment. The data processing includes calculation of the toroidal and poloidal ECH aiming angles and X-mode content for the steerable mirrors that are moved during the plasma shot. Work supported by the US DOE under DE-FC02-04ER54698.

  4. Performance and development of the DIII-D tokamak core

    SciTech Connect

    Anderson, P.M.; Johnson, W.R.; Busath, J.L.; Allen, S.L.

    1998-07-01

    The DIII-D tokamak is an upgrade of the Doublet III configuration which has operated since early 1986. This paper presents recent advances in performance using the upper divertor, fabrication development for vanadium components, operation of the helium leak checking in a high deuterium background, and restoration of the damaged Ohmic heating solenoid.

  5. DIII-D power supply, design, and development

    SciTech Connect

    Nerem, A.

    1995-02-01

    An overview of the DIII-D power supply system with information details concerning the configuration, power ratings, acquisition costs, and cost scaling relevant to the design of ITER and other tokamaks is presented. The power supplies for the DIII-D tokamak were installed and commissioned during the late 1970`s and the beginning of the 1980`s. Several upgrades have been implemented during the last two years to solve increasing reliability problems encountered as the equipment aged, to provide enhanced operational flexibilities, and to enable operation at the higher power levels needed to provide experimental data relevant to the ITER and TPX design activities. These upgrades ranged from redesign of the power supply control systems to the replacement of vacuum circuit breakers which had become unreliable in service. A new interlock and protection system has also been implemented using the latest programmable logic controllers (PLC) and computer technology. These upgrades have been highly successful and are described to provide insight to many issues in the specification of high power converters. Power supply models used in the design of the DIII-D Plasma Control System are also described along with model verification test data. These models are being used in the development of a new advanced plasma control system for the DIII-D tokamak. Recent operational experience and results are presented.

  6. Overview of DIII--D 1997 Experiment Campaign

    NASA Astrophysics Data System (ADS)

    Murakami, M.

    1997-11-01

    A major thrust of the DIII--D experiments in 1997 is to extend and integrate the understanding of fusion science leading to concept improvement, including extending the duration of Advanced Tokamak plasmas toward steady-state. The experiments are being carried out with a modified top divertor capable of pumping and baffling of high triangularity shaped plasmas, with two MW-level-gyrotrons for ECH heating and off-axis current drive, and with improved diagnostics. The DIII--D experimental program contributes to urgently physics R&D for the ITER Engineering Design Activity using its flexibility in plasma shaping and unique diagnostic capability. The key experimental results during the campaign will be summarized.

  7. Stability of TAE modes in DIII-D

    SciTech Connect

    Strait, E.J.; Chu, M.S.; Lao, L.L.; Turnbull, A.D.; Heidbrink, W.W.; Duong, H.H.

    1992-09-01

    TAE modes driven by neutral beam injection have been observed in DIII-D. The measured frequency agrees very well with theoretical predictions for DIII-D discharges. At large amplitude these instabilities can lead to loss of over 50% of the beam power, as well as large loss of non-resonant MeV fusion products. The threshold value of fast ion beta for destabilization and the observed range of unstable mode numbers are in reasonable agreement with predictions for the mode growth rate. Continuum damping dominates at low mode numbers, while damping by electron kinetic effects dominates at high mode numbers. Preliminary experiments suggest that TAB modes can be stabilized by current profile control.

  8. PHYSICS OF ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D

    SciTech Connect

    PETTY,CC; PRATER,R; LUCE,TC; ELLIS,RA; HARVEY,RW; KINSEY,JE; LAO,LL; LOHR,J; MAKOWSKI,MA

    2002-11-01

    OAK A271 PHYSICS OF ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D. Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. The measured ECCD efficiency increases as the deposition location is moved towards the inboard midplane or towards smaller minor radius for both co and counter injection. The measured ECCD efficiency also increases with increasing electron density and/or temperature. The experimental ECCD is compared to both the linear theory (Toray-GA) as well as a quasilinear Fokker-Planck model (CQL3D). The experimental ECCD is found to be in better agreement with the more complete Fokker-Planck calculation, especially for cases of high rf power density and/or loop voltage.

  9. OEDGE Modeling of Divertor Fueling at DIII-D

    NASA Astrophysics Data System (ADS)

    Bray, B. D.; Leonard, A. W.; Elder, J. D.; Stangeby, P. C.

    2015-11-01

    Onion-skin-modeling (OSM) is used to assess the affect of divertor closure on pedestal fueling sources. The OSM includes information from a wide range of diagnostic measurements at DIII-D to constrain the model background plasma for better simulation of neutrals and impurity ions and spectroscopy to compare to the results of the simulation. DIII-D has open lower divertor and closed upper divertor configurations which can be run with similar discharges. Progress toward modeling the pedestal fueling in low density plasmas for these cases will be presented as well as initial comparisons of recent lower single null discharges with the outer leg on the divertor shelf (fully open) and divertor floor (partially open). Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  10. The Multiple Gyrotron System on the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Lohr, J.; Anderson, J.; Brambila, R.; Cengher, M.; Chen, X.; Ellis, R. A.; Grosnickle, W.; Moeller, C.; Prater, R.; Ponce, D.; Riford, L.; Torrezan, A. C.

    2016-01-01

    A major component of the versatile heating systems on the DIII-D tokamak is the gyrotron complex. This system routinely operates at 110 GHz with 4.7 MW-generated rf power for electron cyclotron heating and current drive. The complex is being upgraded with the addition of new depressed collector potential gyrotrons operating at 117.5 GHz and generating rf power in excess of 1.0 MW each. The long-term upgrade plan calls for 10 gyrotrons at the higher frequency being phased in as resources permit, for an injected power near 10 MW. This paper presents a summary of the current status of the DIII-D gyrotron complex, its performance, individual components, testing procedures, operational parameters, plans, and a brief summary of the experiments for which the system is currently being used.

  11. LAUNCHER PERFORMANCE IN THE DIII-D SYSTEM

    SciTech Connect

    KAJIWARA,K; BAXI,CB; LOHR,J; GORELOV,YA; GREEN,MT; PONCE,D; CALLIS,RW

    2003-07-01

    LAUNCHER PERFORMANCE IN THE DIII-D SYSTEM. The thermal performance of three different designs for the steerable mirrors on the ECH launchers installed in the DIII-D tokamak has been evaluated theoretically and experimentally. In each case the disruption forces must be minimized while providing a low loss reflecting surface. One design uses all Glidcop{reg_sign} material, but shaped so that the center is appreciably thicker than the edge. A second design is graphite with a molybdenum surface brazed to the graphite. The latest design is laminated copper/stainless steel construction with a thin copper reflecting surface. All three mirrors employ passive radiative cooling. The mirror temperatures are measured by resistance temperature devices (RTDs) which are attached at the back surfaces of the mirrors. The temperature increases are moderate for the laminated mirror, which has the best overall performance.

  12. PHYSICS OF ELCTRON CYCLOTRON CURRENT DRIVE ON DIII-D

    SciTech Connect

    PETTY,CC; PRATER,R; LUCE,TC; ELLIS,RA; HARVEY,RW; KINSEY,JE; LAO,LL; LOHR,J; MAKOWSKI,MA

    2002-09-01

    OAK A271 PHYSICS OF ELCTRON CYCLOTRON CURRENT DRIVE ON DIII-D. Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. The measured ECCD efficiency increases as the deposition location is moved towards the inboard midplane or towards smaller minor radius for both co and counter injection. The measured ECCD efficiency also increases with increasing electron density and/or temperature. The experimental ECCD is compared to both the linear theory (Toray-GA) as well as a quasilinear Fokker-Planck model (CQL3D). The experimental ECCD is found to be in better agreement with the more complete Fokker-Planck calculation, especially for cases of high rf power density and/or loop voltage. The narrow width of the measured ECCD profile is consistent with only low levels of radial transport for the current carrying electrons.

  13. Density peaking and turbulent pinch in DIII-D discharges

    SciTech Connect

    Estrada-Mila, C.; Candy, J.; Waltz, R. E.

    2006-07-15

    A study of density peaking and particle flow in low confinement (L-mode) DIII-D discharges [G. R. McKee, C. C. Petty, R. E. Waltz et al., Nucl. Fusion 41, 1235 (2001)], using global gyrokinetic simulations, is presented. It is found that under experimental conditions, in particular when realistic collisionality is included, a turbulent pinch driven by electron temperature and density gradients can occur.

  14. Dust Studies in DIII-D and TEXTOR

    SciTech Connect

    Rudakov, D L; Litnovsky, A; West, W P; Yu, J H; Boedo, J A; Bray, B D; Brezinsek, S; Brooks, N H; Fenstermacher, M E; Groth, M; Hollmann, E M; Huber, A; Hyatt, A W; Krasheninnikov, S I; Lasnier, C J; Moyer, R A; Pigarov, A Y; Philipps, V; Pospieszczyk, A; Smirnov, R D; Sharpe, J P; Solomon, W M; Watkins, J G; Wong, C C

    2009-02-17

    Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Direct heating of the dust particles by the neutral beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust. Large flakes or debris falling into the plasma may result in a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micron-size dust in plasma discharges. In DIII-D, a sample holder filled with {approx}30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure ({approx}0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. At the given configuration of the launch, the dust did not penetrate the core plasma and only moderately perturbed the edge plasma, as evidenced by an increase of the edge carbon content.

  15. Electron cyclotron current drive in DIII-D

    SciTech Connect

    Luce, T.C.; Lin-Liu, Y.R.; Lohr, J.M.; Petty, C.C.; Politzer, P.A.; Prater, R.; Harvey, R.W.; Giruzzi, G.; Rice, B.W.

    1999-05-01

    Clear measurements of the localized current density driven by electron cyclotron waves have been made on the DIII-D tokamak. Direct evidence of the current drive is seen on the internal magnetic field measurements by motional Stark effect spectroscopy. Comparison with theoretical calculations in the collisionless limit shows the experimental current drive exceeds the predictions by a substantial amount for currents driven near the half radius. In all cases the experimental current density profile is broader than the predicted one.

  16. Solenoid-free startup experiments in DIII-D

    SciTech Connect

    Leuer, J. A.; Cunningham, G.; Mueller, D.; Brooks, N. H.; Eldietis, N. W.; Humphreys, D. A.; Hyatt, A. W.; Jackson, G. L.; Lohr, J.; Politzer, P. A.; Pinsker, R. I.; Prater, R.; Taylor, P. L.; Walker, M. L.; Budny, R. V.; Gates, D.A.; Nagy, Alex; Hahn, S. H.; Oh, Y. K.; Yoon, S. W.; Yu, J.H.; Murakami, Masanori; Park, J. M.; Sontag, A. C.

    2011-01-01

    A series of DIII-D experiments was performed to investigate the potential for initiating plasma current using only poloidal field coils located outside the DIII-D central solenoid, i.e. 'solenoid-free'. Plasma current to 166 kA was achieved using 2-3MW of electron cyclotron (EC) heating and was limited by coil and power supply constraints. Flux conversion to plasma current was similar to standard DIII-D startup with some degradation at higher plasma current associated with stray fields and vertical stability issues. In preliminary solenoid-free experiments, neutral beam (NB) current drive (CD) levels were small and attributed to reduced CD efficiency associated with low electron temperature produced by the low current, low confinement plasma. Lack of plasma radial position control also contributed to a reduction of NBCD. Similarly, ECCD was small owing to low plasma temperature and outside EC launch which is required in the solenoid-free scenario. Synergistic experiments were carried out using standard solenoid initiated plasmas in order to study noninductive CD in limited, Lmode plasmas, typical of that generated by solenoid-free startup. While substantial noninductive current can be driven, self-sustaining levels of noninductive current have not yet been achieved with our present six-source co-injection NB system combined with EC and fast wave systems. At low plasma current and high levels of localized EC heating, substantial MHD is generated and this was seen to severely limit plasma performance. Although further optimization is possible in the limited plasma regime, full noninductive, steady-state operation may require diverted plasma with H-mode quality confinement. Discharges obtained during the solenoid-free campaign are compared with results of previous DIII-D campaigns aimed at achieving a steady state, noninductive CD solution.

  17. Demonstration of ITER Operational Scenarios on DIII-D

    SciTech Connect

    Doyle, E J; Budny, R V; DeBoo, J C; Ferron, J R; Jackson, G L; Luce, T C; Murakami, M; Osborne, T H; Park, J; Politzer, P A; Reimerdes, H; Casper, T A; Challis, C D; Groebner, R J; Holcomb, C T; Hyatt, A W; La Haye, R J; McKee, G R; Petrie, T W; Petty, C C; Rhodes, T L; Shafer, M W; Snyder, P B; Strait, E J; Wade, M R; Wang, G; West, W P; Zeng, L

    2008-10-13

    The DIII-D program has recently initiated an effort to provide suitably scaled experimental evaluations of four primary ITER operational scenarios. New and unique features of this work are that the plasmas incorporate essential features of the ITER scenarios and anticipated operating characteristics; e.g., the plasma cross-section, aspect ratio and value of I/aB of the DIII-D discharges match the ITER design, with size reduced by a factor of 3.7. Key aspects of all four scenarios, such as target values for {beta}{sub N} and H{sub 98}, have been replicated successfully on DIII-D, providing an improved and unified physics basis for transport and stability modeling, as well as for performance extrapolation to ITER. In all four scenarios normalized performance equals or closely approaches that required to realize the physics and technology goals of ITER, and projections of the DIII-D discharges are consistent with ITER achieving its goals of {ge} 400 MW of fusion power production and Q {ge} 10. These studies also address many of the key physics issues related to the ITER design, including the L-H transition power threshold, the size of ELMs, pedestal parameter scaling, the impact of tearing modes on confinement and disruptivity, beta limits and the required capabilities of the plasma control system. An example of direct influence on the ITER design from this work is a modification of the specified operating range in internal inductance at 15 MA for the poloidal field coil set, based on observations that the measured inductance in the baseline scenario case lay outside the original ITER specification.

  18. RESEARCH PROGRESS AND HARDWARE SYSTEMS AT DIII-D

    SciTech Connect

    PETERSEN,P.I; THE DIII-D TEAM

    2003-10-01

    OAK-B135 During the last two years significant progress has been made in the scientific understanding of DIII-D plasmas. Much of this progress has been enabled by the addition of new hardware systems. The electron cyclotron (EC) system has been upgraded from 3 MW to 6 MW, by adding three 1 MW gyrotrons with diamond windows and three steerable launchers (PPPL). The new gyrotrons have been tested to 1.0 MW for 5 s. The system has been used to control the 3/2 and 2/1 neoclassical tearing modes and to locally heat the plasma and thereby indirectly control the current density. Electron cyclotron current drive ECCD has been used to directly affect the current density. A Li-beam diagnostic has been brought on-line for measuring the edge current density using Zeeman splitting. A set of 12 coils (1-coils), consisting of six picture frame coils each above and below the midplane, with a capability of 7 kA for 10 s has been installed inside the DIII-D vessel. These coils, along with the existing six C-coils, are used to apply non-axisymmetric fields to the plasma for both exciting and controlling plasma instabilities. The DIII-D digital plasma control system is now used to not just control the shape and location of the plasma but also the electron temperature, density, the NTMs, RWMs, plasma beta and disruption mitigation. Plasma disruption experiments are extended to mitigation of real time detected disruptions on DIII-D.

  19. Dust studies in DIII-D and TEXTOR

    SciTech Connect

    Rudakov, D.L.; Litnovsky, A; West, W. P.; Yu, J.H.; Boedo, J.A.; McLean, Adam G

    2009-01-01

    Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicrometre sized dust is routinely observed using Mie scattering from a Nd: Yag laser. The source is strongly correlated with the presence of type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Heating of the dust particles by the neutral beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust; on the other hand, large flakes or debris falling into the plasma may induce a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micrometre-size particles into plasma discharges. In DIII-D, a sample holder filled with 30-40 mg of dust is inserted in the lower divertor and exposed, via sweeping of the strike points, to the diverted plasma flux of high-power ELMing H-mode discharges. After a brief dwell (similar to 0.1 s) of the outer strike point on the sample holder, part of the dust penetrates into the core plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase in the radiated power. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off-layer 0-2 cm radially outside of the last closed flux surface in discharges heated with 1.4 MW of NBI. Launched in this configuration, the dust perturbed the edge plasma, as evidenced by a moderate increase in the edge carbon content, but did not penetrate into the core plasma.

  20. Status of the DIII-D 110 GHz ECH system

    SciTech Connect

    Callis, R.W.; Lohr, J.; O`Neill, R.C.; Tooker, J.F.; Ponce, D.

    1996-06-01

    The DIII-D program is presently commissioning the first NM gyrotron of a planned 3 MW, I 10 GHz electron cyclotron heating (ECH) system for off-axis electron heating and current drive. Advanced tokamak (AT) research in DIII-D and other tokamaks requires the ability to control the current density profile. ECH offers the ability to localize the heating and driven current in a controllable manner and is not dependent upon, the local plasma conditions, so it appears to be an ideal tool for AT research. The planned rf sources for the DIII-D system are I MW state-of-the-art internal mode-converter gyrotrons, with one gyrotron being manufactured by GYCOM, a Russian company, and two gyrotrons being manufactured by CPI (formerly Varian). The GYCOM gyrotron has been tested at the factory to 960 kW, 2 seconds and has been shipped to GA where it is now undergoing initial checkout and testing. The first CPI gyrotron has been assembled and factory tested to 530 kW, 2 seconds and 350 1352 kW, 10 seconds. Both the GYCOM and CPI gyrotrons are limited in pulse length at full power by thermal limits on the output window. The second CPI gyrotron is expected to be ready for testing in April 1996. This paper will report on the initial experiences of using the GYCOM I MW, 110 GHz internal mode- converter gyrotron, at General Atomics, and the observed effects the ECRH power has on the DIII-D plasma.

  1. Development of Negative Triangularity Plasmas in DIII-D

    NASA Astrophysics Data System (ADS)

    Walker, M. L.

    2015-11-01

    We report on development of DIII-D electron cyclotron heated discharges with controlled negative triangularity plasma shapes, to evaluate the effect on electron heat transport in L-mode plasmas, as reported on TCV. Analysis of TCV data found that negative triangularity exerts a stabilizing influence on the trapped electron mode, the dominant instability in the conditions of those experiments. Major objectives of the DIII-D development are producing complementary plasmas, one with negative and one with positive triangularity, approximately symmetric in major radial coordinate and having similar density and current profiles. Major constraints include selection of plasma parameters and toroidal field to optimize fluctuation diagnostic measurements while preventing transition from L- to H-mode and deposition of EC heating power near the q=1 surface to limit sawteeth. Issues discussed are definition of control scenarios by which the pair of shapes are accessed and their resulting controllability under the constraints imposed by DIII-D shaping control. Supported by US DOE under DE-FC02-04ER54698.

  2. Using AORSA to simulate helicon waves in DIII-D

    SciTech Connect

    Lau, C. Blazevski, D.; Green, D. L.; Murakami, M.; Park, J. M.; Jaeger, E. F.; Berry, L. A.; Bertelli, N.; Pinsker, R. I.; Prater, R.

    2015-12-10

    Recent efforts have shown that helicon waves (fast waves at > 20ω{sub ci}) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored, it will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.

  3. Using AORSA to simulate helicon waves in DIII-D

    SciTech Connect

    Lau, Cornwall H; Jaeger, E. F.; Bertelli, Nicola; Berry, Lee Alan; Blazevski, Dan; Green, David L; Murakami, Masanori; Park, J. M.; Pinsker, R. I.; Prater, R.

    2015-01-01

    Recent efforts have shown that helicon waves (fast waves at >20 omega(ci)) may be an attractive option for driving efficient off-axis current drive during non-inductive tokamak operation for DIII-D, ITER and DEMO. For DIII-D scenarios, the ray tracing code, GENRAY, has been extensively used to study helicon current drive efficiency and location as a function of many plasma parameters. The full wave code, AORSA, which is applicable to arbitrary Larmor radius and can resolve arbitrary ion cyclotron harmonic order, has been recently used to validate the ray tracing technique at these high cyclotron harmonics. If the SOL is ignored, it will be shown that the GENRAY and AORSA calculated current drive profiles are comparable for the envisioned high beta advanced scenarios for DIII-D, where there is high single pass absorption due to electron Landau damping and minimal ion damping. AORSA is also been used to estimate possible SOL effects on helicon current drive coupling and SOL absorption due to collisional and slow wave effects.

  4. Interaction of NTM and RMP fields in DIII-D

    NASA Astrophysics Data System (ADS)

    Nazikian, R.; Evans, T. E.; Schaffer, M. J.; Turnbull, A. D.

    2011-10-01

    Edge Localized Mode (ELM) suppression by external Resonant Magnetic Perturbations (RMPs) is sometimes accompanied by the growth of internal neoclassical tearing modes. Both Neoclassical Tearing Modes (NTMs) and RMP fields induce a density pump out effect in DIII-D. A question is whether there is a possible synergy between the externally applied RMP field and internal NTM field that could lead to enhanced stochasticity and/or ELM suppression in DIII-D plasmas. A large scalar database study did not reveal a significant correlation between the NTM level and the degree of density pump out observed during application of RMP fields, however individual discharges appear to show an effect. We will present vacuum field calculations of the superposition of modeled internal NTM fields and RMP fields for n = 2 and n = 3 perturbations in DIII-D in order to better understand the possible interaction of these two sources of magnetic perturbations. Work supported in part by US DOE under DE-AC02-09CH11466 and DE-FC02-04ER54698.

  5. Dust Studies in DIII-D and TEXTOR

    SciTech Connect

    Rudakov, D; Litnovsky, A; West, W; Yu, J; Boedo, J; Bray, B; Brezinsek, S; Brooks, N; Fenstermacher, M; Groth, M; Hollmann, E; Huber, A; Hyatt, A; Krasheninnikov, S; Lasnier, C; Moyer, R; Pigarov, A; Philipps, V; Pospieszezyk, A; Smirnov, R; Sharpe, J; Solomon, W; Watkins, J; Wong, C

    2008-10-14

    Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Energetic plasma disruptions produce significant amounts of dust. However, dust production by disruptions alone is insufficient to account for the estimated in-vessel dust inventory in DIII-D. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by injecting micron-size dust in plasma discharges. In DIII-D, a sample holder filled with {approx}30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure ({approx}0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. Individual dust particles are observed moving at velocities of 10-100 m/s, predominantly in the toroidal direction, consistent with the drag force from the deuteron flow and in agreement with modeling by the 3D DustT code. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. Dust is launched either in the beginning of a discharge or at the initiation of NBI, preferentially in a direction perpendicular to the toroidal magnetic field. At the given configuration of the launch, the dust did not

  6. Collaboration on DIII-D Five Year Plan

    SciTech Connect

    Allen, S

    2003-04-01

    This document summarizes Lawrence Livermore National Laboratory's (LLNL) plan for fusion research on the DIII-D Tokamak, located at General Atomics (GA) in San Diego, California, in the time period FY04-FY08. This document is a companion document to the DIII-D Five-Year Program Plan; which hereafter will be referred to as the ''D3DPP''. The LLNL Collaboration on DIII-D is a task-driven program in which we bring to bear the full range of expertise needed to complete specific goals of plasma science research on the DIII-D facility. This document specifies our plasma performance and physics understanding goals and gives detailed plans to achieve those goals in terms of experimental leadership, code development and analysis, and diagnostic development. Our program is designed to be consistent with the long-term mission of the DIII-D program as documented in the D3DPP. The overall DIII-D Program mission is ''to establish the scientific basis for the optimization of the tokamak approach to fusion energy production''. LLNL Magnetic Fusion Energy (MFE) supports this mission, and we contribute to two areas of the DIII-D program: divertor physics and advanced tokamak (AT) physics. We lead or contribute to the whole cycle of research: experimental planning, diagnostic development, execution of experiments, and detailed analysis. We plan to continue this style in the next five years. DIII-D has identified three major research themes: AT physics, confinement physics, and mass transport. The LLNL program is part of the AT theme: measurement of the plasma current profile, and the mass transport theme: measurement and modeling of plasma flow. In the AT area, we have focused on the measurement and modeling of the current profile in Advanced Tokamak plasmas. The current profile, and it's effect on MHD stability of the high-{beta} ''AT'' plasma are at the heart of the DIII-D program. LLNL has played a key role in the development of the Motional Stark Effect (MSE) diagnostic. Starting

  7. Recent results from the DIII-D tokamak

    SciTech Connect

    Petersen, P.I.

    1998-02-01

    The DIII-D national fusion research program focuses on establishing the scientific basis for optimization of the tokamak approach to fusion energy production. The symbiotic development of research, theory, and hardware continues to fuel the success of the DIII-D program. During the last year, a radiative divertor and a second cryopump were installed in the DIII-D vacuum vessel, an array of central and boundary diagnostics were added, and more sophisticated computer models were developed. These new tools have led to substantial progress in the understanding of the plasma. The authors now have a better understanding of the divertor as a means to manage the heat, particle, and impurity transport pumping of the plasma edge using the in situ divertor cryopumps effectively controls the plasma density. The evolution of diagnostics that probe the interior of the plasma, particularly the motional Stark effect diagnostic, has led to a better understanding of the core of the plasma. This understanding, together with tools to control the profiles, including electron cyclotron waves, pellet injection, and neutral beam injection, has allowed them to progress in making plasma configurations that give rise to both low energy transport and improved stability. Most significant here is the use of transport barriers to improve ion confinement to neoclassical values. Commissioning of the first high power (890 kW) 110 GHz gyrotron validates an important tool for managing the plasma current profile, key to maintaining the transport barriers. An upgraded plasma control system, ``isoflux control,`` which exploits real time MHD equilibrium calculations to determine magnetic flux at specified locations within the tokamak vessel and provides the means for precisely controlling the plasma shape and, in conjunction with other heating and fueling systems, internal profiles.

  8. CONTROL OF NEOCLASSICAL TEARING MODES IN DIII-D

    SciTech Connect

    R.J. LA HAYE; S. GUNTER; D.A. HUMPHREYS; J. LOHR; T.C. LUCE; M.E. MARASCHEK; C.C. PETTY; R.PRATER; J.T. SCOVILLE; E.J. STRAIT

    2001-11-01

    The development of techniques for neoclassical tearing mode (NTM) suppression or avoidance is crucial for successful high beta/high confinement tokamaks. Neoclassical tearing modes are islands destabilized and maintained by a helically perturbed bootstrap current and represent a significant limit to performance at higher poloidal beta. The confinement-degrading islands can be reduced or completely suppressed by precisely replacing the ''missing'' bootstrap current in the island O-point or by interfering with the fundamental helical harmonic of the pressure. Implementation of such techniques is being studied in the DIII-D tokamak [J.L. Luxon, et al., Plasma Phys. and Control. Fusion Research, Vol. 1 (International Atomic Energy Agency, Vienna, 1987) p. 159] in the presence of periodic q = 1 sawtooth instabilities, a reactor relevant regime. Radially localized off-axis electron cyclotron current drive (ECCD) must be precisely located on the island. In DIII-D the plasma control system is put into a ''search and suppress'' mode to make either small rigid radial position shifts of the entire plasma (and thus the island) or small changes in toroidal field (and thus, ECCD location) to find and lock onto the optimum position for complete island suppression by ECCD. This is based on real-time measurements of an m/n = 3/2 mode amplitude dB{sub {theta}}/dt. The experiment represents the first use of active feedback control to provide continuous, precise positioning. An alternative to ECCD makes use of the six toroidal section ''C-Coil'' on DIII-D to provide a large non-resonant static m = 1, n = 3 helical field to interfere with the fundamental harmonic of an m/n = 3/2 NTM. While experiments show success in inhibiting the NTM if a large enough n = 3 field is applied before the island onset, there is a considerable plasma rotation decrease due to n = 3 ''ripple''.

  9. DiMES divertor erosion experiments on DIII-D

    SciTech Connect

    Whyte, D.G.; Brooks, J.N.; Wong, C.P.C.; West, W.P.; Bastasz, R.; Wampler, W.R.; Rubinstein, J.

    1996-06-01

    The DiMES (Divertor Material Evaluation Studies) mechanism allows insertion of material samples to the lower divertor floor of the DIII-D tokamak. The main purpose of these studies is to measure erosion rates and redeposition mechanisms under tokamak divertor plasma conditions in order to obtain a physical understanding of the erosion/redeposition processes and to determine its implications for fusion power plant plasma facing components. Thin metal films of Be, W, V, and Mo, were deposited on a Si depth-marked graphite sample and exposed to the steady-state outer strike point on DIII-D. A variety of surface analysis techniques are used to determine the erosion/redeposition of the metals and the carbon after 5--15 seconds of exposure. These short exposure times ensure controlled exposure conditions and the extensive array of DIII-D divertor diagnostics provide a well characterized plasma for modeling efforts. Erosion rates and redeposition lengths are found to decrease with the atomic number of the metallic species, as expected. Under these conditions, the peak net erosion rate for carbon is {approximately} 4 nm/s, with the erosion following the ion flux profile. Comparisons of the measured carbon erosion with REDEP code calculations show good agreement for both the absolute net erosion rate and its spatial variation. Measured erosion rates of the metals are smaller than predicted for sputtering from a bare metal surface, apparently due to effects of carbon deposition on the metal surface. Visible spectroscopic measurements of singly ionized Be have determined that the erosion process reaches steady-state during the exposure.

  10. IMPROVEMENTS TO THE CRYOGENIC CONTROL SYSTEM ON DIII-D

    SciTech Connect

    HOLTROP,K.L; ANDERSON,P.M; MAUZEY,P.S

    2003-10-01

    OAK-B135 The cryogenic facility that is part of the DIII-D tokamak system supplies liquid nitrogen and liquid helium to the superconducting magnets used for electron cyclotron heating, the D{sub 2} pellet injection system, cryopumps in the DIII-D vessel, and cryopanels in the neutral beam injection system. The liquid helium is liquefied on site using a Sulzer liquefier that has a 150 l/h liquefaction rate. Control of the cryogenic facility at DIII-D was initially accomplished through the use of three different programmable logic controllers (PLCs). Recently, two of those three PLCs, a Sattcon PLC controlling the Sulzer liquefier and a Westinghouse PLC, were removed and all their control logic was merged into the remaining PLC, a Siemens T1555. This replacement was originally undertaken because the removed PLCs were obsolete and unsupported. However, there have been additional benefits from the replacement. The replacement of the RS-232 serial links between the graphical user interface and the PLCs with a high speed Ethernet link allows for real-time display and historical trending of nearly all the cryosystem's data. this has greatly increased the ability to troubleshoot problems with the system, and has permitted optimization of the cryogenic system's performance because of the increased system integration. To move the control logic of the Sattcon control loops into the T1555, an extensive modification of the basic PID control was required. These modifications allow for better control of the control loops and are now being incorporated in other control loops in the system.

  11. 3D Equilibrium Reconstructions in DIII-D

    NASA Astrophysics Data System (ADS)

    Lao, L. L.; Ferraro, N. W.; Strait, E. J.; Turnbull, A. D.; King, J. D.; Hirshman, H. P.; Lazarus, E. A.; Sontag, A. C.; Hanson, J.; Trevisan, G.

    2013-10-01

    Accurate and efficient 3D equilibrium reconstruction is needed in tokamaks for study of 3D magnetic field effects on experimentally reconstructed equilibrium and for analysis of MHD stability experiments with externally imposed magnetic perturbations. A large number of new magnetic probes have been recently installed in DIII-D to improve 3D equilibrium measurements and to facilitate 3D reconstructions. The V3FIT code has been in use in DIII-D to support 3D reconstruction and the new magnetic diagnostic design. V3FIT is based on the 3D equilibrium code VMEC that assumes nested magnetic surfaces. V3FIT uses a pseudo-Newton least-square algorithm to search for the solution vector. In parallel, the EFIT equilibrium reconstruction code is being extended to allow for 3D effects using a perturbation approach based on an expansion of the MHD equations. EFIT uses the cylindrical coordinate system and can include the magnetic island and stochastic effects. Algorithms are being developed to allow EFIT to reconstruct 3D perturbed equilibria directly making use of plasma response to 3D perturbations from the GATO, MARS-F, or M3D-C1 MHD codes. DIII-D 3D reconstruction examples using EFIT and V3FIT and the new 3D magnetic data will be presented. Work supported in part by US DOE under DE-FC02-04ER54698, DE-FG02-95ER54309 and DE-AC05-06OR23100.

  12. A design retrospective of the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Luxon, J. L.

    2002-05-01

    The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and RF heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research programme. An integrated picture of the facility and its capabilities is presented.

  13. Plasma rotation and rf heating in DIII-D

    SciTech Connect

    deGrassie, J.S.; Baker, D.R.; Burrell, K.H.

    1999-05-01

    In a variety of discharge conditions on DIII-D it is observed that rf electron heating reduces the toroidal rotation speed and core ion temperature. The rf heating can be with either fast wave or electron cyclotron heating and this effect is insensitive to the details of the launched toroidal wavenumber spectrum. To date all target discharges have rotation first established with co-directed neutral beam injection. A possible cause is enhanced ion momentum and thermal diffusivity due to electron heating effectively creating greater anomalous viscosity. Another is that a counter directed toroidal force is applied to the bulk plasma via rf driven radial current.

  14. Experiments at high elongations in DIII-D

    SciTech Connect

    Lazarus, E.A. ); Turnbull, A.D.; Kellman, A.G.; Ferron, J.R.; Helton, F.J.; Lao, L.L.; Leuer, J.A.; Strait, E.J.; Taylor, T.S. )

    1990-06-01

    In this paper we discuss the limitation to elongation observed in D-shaped plasmas in the DIII-D tokamak. We find that as the triangularity is increased and {ell}{sub i} is decreased that the n = 0 mode takes on an increasingly non-rigid character. Our analysis shows two aspects of the behavior; first, an increasing variation of the m/n = 1/0 component across flux surfaces and second, an increase in the relative amplitude of a m/n = 3/0 component which couples to the m/n = 1/0 component and further destabilizes the mode.

  15. Electron cyclotron current drive in DIII-D

    SciTech Connect

    Luce, T. C.; Lin-Liu, Y. R.; Harvey, R. W.; Giruzzi, G.; Lohr, J. M.; Petty, C. C.; Politzer, P. A.; Prater, R.; Rice, B. W.

    1999-09-20

    Clear measurements of the localized current density driven by electron cyclotron waves have been made on the DIII-D tokamak. Direct evidence of the current drive is seen on the internal magnetic field measurements by motional Stark effect spectroscopy. Comparison with theoretical calculations in the collisionless limit shows the experimental current drive exceeds the predictions by a substantial amount for currents driven near the half radius. In all cases the experimental current density profile is broader than the predicted one. (c) 1999 American Institute of Physics.

  16. An in situ runaway electron diagnostic for DIII-D

    SciTech Connect

    Wurden, G. A. Oertel, J. A.; Evans, T. E.

    2014-11-15

    We are designing a new diagnostic based on laser inverse Compton scattering to study the dynamics of runaway electron formation during killer-pellet triggered disruptions in DIII-D, and their subsequent loss. We can improve the expected S/N ratio by using a high-intensity short-pulse laser combined with gated x-ray imagers. With 80 ps sampling, time-of-flight spatial resolution within the laser chord can be obtained. We will measure the time-resolved spatial profile and energy distribution of the runaway electrons while they are in the core of the tokamak plasma.

  17. Performance characteristics of the DIII-D advanced divertor cryopump

    SciTech Connect

    Menon, M.M.; Maingi, R.; Wade, M.R.; Baxi, C.B.; Campbell, G.L.; Holtrop, K.L.; Hyatt, A.W.; Laughon, G.J.; Makariou, C.C.; Mahdavi, M.A.; Reis, E.E.; Schaffer, M.J.; Schaubel, K.M.; Scoville, J.T.; Smith, J.P.; Stambaugh, R.D.

    1993-10-01

    A cryocondensation pump, cooled by forced flow of two-phase helium, has been installed for particle exhaust from the divertor region of the DIII-D tokamak. The Inconel pumping surface is of coaxial geometry, 25.4 mm in outer diameter and 11.65 m in length. Because of the tokamak environment, the pump is designed to perform under relatively high pulsed heat loads (300 Wm{sup {minus}2}). Results of measurements made on the pumping characteristics for D{sub 2}, H{sub 2}, and Ar are discussed.

  18. Electron cyclotron heating experiments on the DIII-D tokamak

    SciTech Connect

    Prater, R.; Austin, M.E.; Bernabei, S.

    1998-01-01

    Initial experiments on heating and current drive using second harmonic electron cyclotron heating (ECH) are being performed on the DIII-D tokamak using the new 110 GHz ECH system. Modulation of the ECH power in the frequency range 50 to 300 Hz and detection of the temperature perturbation by ECE diagnostics is used to validate the location of the heating. This technique also determines an upper bound on the width of the deposition profile. Analysis of electron cyclotron current drive indicates that up to 0.17 MA of central current is driven, resulting in a negative loop voltage near the axis.

  19. Performance history and upgrades for the DIII-D gyrotron complex

    DOE PAGESBeta

    Lohr, J.; Anderson, J. P.; Cengher, M.; Ellis, R. A.; Gorelov, Y. A.; Kolemen, E.; Lambot, T.; Murakami, D. D.; Myrabo, L.; Noraky, S.; et al

    2015-03-12

    The gyrotron installation on the DIII-D tokamak has been in operation at the second harmonic of the electron cyclotron resonance since the mid-1990s. Prior to that a large installation of ten 60 GHz tubes was operated at the fundamental resonance. The system has been upgraded regularly and is an everyday tool for experiments on DIII-D.

  20. Lawrence Livermore National Laboratory DIII-D cooperation: 1987 annual report

    SciTech Connect

    Allen, S.L.; Calderon, M.O.; Ellis, R.M.; Evans, J.C.; Ferguson, S.W.; Hill, D.N.; James, R.A.; Jenkins, S.L.; Karlsen, C.E.; Kevan, D.

    1988-02-24

    This report summarizes the Lawrence Livermore National Laboratory (LLNL) DIII-D cooperation during FY87. The LLNL participation in DIII-D concentrated on three principal areas: ECH and current-drive physics, divertor and edge physics, and tokamak operations. These topics are dicussed in this report. 27 refs., 11 figs.

  1. ECE RADIOMETER UPGRADE ON THE DIII-D TOKAMAK

    SciTech Connect

    AUSTIN, ME; LOHR, J

    2002-08-01

    OAK A271 ECE RADIOMETER UPGRADE ON THE DIII-D TOKAMAK. The electron cyclotron emission (ECE) heterodyne radiometer diagnostic on DIII-D has been upgraded with the addition of eight channels for a total of 40. The new, higher frequency channels allow measurements of electron temperature into the magnetic axis in discharges at maximum field, 2.15 T. The complete set now extends over the full usable range of second harmonic emission frequencies at 2.0 T covering radii from the outer edge inward to the location of third harmonic overlap on the high field side. Full coverage permits the measurement of heat pulses and magnetohydrodynamic (MHD) fluctuations on both sides of the magnetic axis. In addition, the symmetric measurements are used to fix the location of the magnetic axis in tokamak magnetic equilibrium reconstructions. Also, the new higher frequency channels have been used to determine central T{sub e} with good time resolution in low field, high density discharges using third harmonic ECE in the optically gray and optically thick regimes.

  2. UEDGE code comparisons with DIII-D bolometer DATA

    SciTech Connect

    Daniel, J.M.

    1995-01-01

    This paper describes the work done to develop a bolometer post processor that converts volumetric radiated power values taken from a UEDGE solution, to a line integrated radiated power along chords of the bolometers in the DIII-D tokamak. The UEDGE code calculates plasma physics quantities, such as plasma density, radiated power, or electron temperature, and compares them to actual diagnostic measurements taken from the scrape off layer (SOL) and divertor regions of the DIII-D tokamak. Bolometers are devices measuring radiated power within the tokamak. The bolometer interceptors are made up of two complete arrays, an upper array with a vertical view and a lower array with a horizontal view, so that a two dimensional profile of the radiated power may be obtained. The bolometer post processor stores line integrated values taken from UEDGE solutions into a file in tabular format. Experimental data is then put into tabular form and placed in another file. Comparisons can be made between the UEDGE solutions and actual bolometer data. Analysis has been done to determine the accuracy of the plasma physics involved in producing UEDGE simulations.

  3. FWCD and ECCD experiments on DIII-D

    SciTech Connect

    Prater, R.; Austin, M.E.; Baity, F.W.

    1994-02-01

    Fast wave current drive and electron cyclotron current drive experiments have been performed on the DIII-D tokamak as part of the advanced tokamak program. The goal of this program is to develop techniques for controlling the profile of the current density in order to access regimes of improved confinement and stability. The experiments on fast wave current drive used a four strap antenna with 900 phasing between straps. A decoupler was used to help maintain the phasing, and feedback control of the plasma position was used to keep the resistive loading constant. RF pickup loops demonstrate that the directivity of the antenna is as expected. Plasma currents up to 0.18 MA were driven by 1.5 MW of fast wave power. Electron cyclotron current drive experiments at 60 GHz have shown 0.1 MA of plasma current driven by 1 MW of power. New fast wave and electron cyclotron heating systems are in development for DIII-D, so that the goals of the advanced tokamak program can be carried out.

  4. DIII-D safety procedures and operational experience

    NASA Astrophysics Data System (ADS)

    Petersen, Peter I.; Savercool, Richard L.; Taylor, Peter L.

    1989-06-01

    The DIII-D research tokamak, which has near reactor dimensions, has been operating for approximately ten years. The hazards at the facility include high voltages, high currents, cryogenics, radiation, and normal industrial hazards. Procedures have been developed for operating the device, minimizing both the hazards to personnel and the damage caused by equipment failures, and for logging of failures. The machine and the neutral beamlines have been operated in hydrogen and are now starting operation in deuterium to expand the machine capabilities to reactor relevant regimes. Deuterium operation creates a significant amount of neutrons and therefore a neutron shield is being built around the machine area to reduce personnel radiation to As Low As Reasonably Achievable (ALARA) and to keep radiation at the site boundary at or below a limit of 20 mrem per calendar year. Gamma radiation shielding walls have been in place since the initial machine construction. The safety procedures for the DIII-D facility will be outlined and the accident experience will be discussed together with the neutron radiation monitoring program being developed.

  5. UEDGE code comparisons with DIII-D bolometer data

    SciTech Connect

    Daniel, J.M.

    1994-12-01

    This paper describes the work done to develop a bolometer post processor that converts volumetric radiated power values taken from a UEDGE solution, to a line integrated radiated power along chords of the bolometers in the DIII-D tokamak. The UEDGE code calculates plasma physics quantities, such as plasma density, radiated power, or electron temperature, and compares them to actual diagnostic measurements taken from the scrape off layer (SOL) and divertor regions of the DIII-D tokamak. Bolometers are devices measuring radiated power within the tokamak. The bolometer interceptors are made up of two complete arrays, an upper array with a vertical view and a lower array with a horizontal view, so that a two dimensional profile of the radiated power may be obtained. The bolometer post processor stores line integrated values taken from UEDGE solutions into a file in tabular format. Experimental data is then put into tabular form and placed in another file. Comparisons can be made between the UEDGE solutions and actual bolometer data. Analysis has been done to determine the accuracy of the plasma physics involved in producing UEDGE simulations.

  6. Edge MSE measurements on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Victor, B. S.; Holcomb, C. T.; Allen, S. L.; Meyer, W. H.; Makowski, M. A.

    2015-11-01

    The edge motional Stark effect (MSE) diagnostic on DIII-D has recently been upgraded to provide better constraints on the current density in the outer half radius, including high-spatial resolution measurements in the H-mode pedestal. The channels have been upgraded with new bandpass filters with FWHM of 0.3 nm and >90 % transmission at the central wavelength, and improved detector positioning. A spectrometer has been used to measure the σ and π lines for each of the edge channels to optimize the new filter selection. These upgrades are expected to improve our ability to assess current drive and stability in various DIII-D plasmas. Comparisons will be shown between the Sauter and NEO bootstrap current models and these new measurements in fully non-inductive plasmas. We will present calculations of the ideal MHD βN-limit in various scenarios that use the improved measurements. Finally, changes in the edge pitch angle during ELMs are examined. Supported by US DOE under DE-AC52-07NA27344, DE-FC02-04ER54698, DE-FG02-04ER54761, DE-SC0010661, and DE-AC02-09CH11466.

  7. Structural design of the DIII-D radiative divertor

    SciTech Connect

    Reis, E.E.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Hollerbach, M.A.; Laughon, G.J.; Sevier, D.L.

    1996-10-01

    The divertor of the DIII-D tokamak is being modified to operate as a slot type, dissipative divertor. This modification, called the Radiative Divertor Program (RDP) is being carried out in two phases. The design and analysis is complete and hardware is being fabricated for the first phase. This first phase consists of an upper divertor baffle and cryopump to provide some density control for high triangularity, single or double null discharges. Installation of the first phase is scheduled to start in October, 1996. The second phase provides pumping at all four divertor strike points of double null high triangularity discharges and baffling of the neutral particles from transport back to the core plasma. Studies of the effects of varying the slot length and width of the divertor can be easily accomplished with the design of RDP hardware. Static and dynamic analyses of the baffle structures, new cryopumps, and feedlines were performed during the preliminary and final design phases. Disruption loads and differential thermal displacements must be accommodated in the design of these components. With the full RDP hardware installed, the plasma current in DIII-D will be a maximum of 3.0 MA. Plasma disruptions induce toroidal currents in the cryopump, producing complex dynamic loads. Simultaneously, the vacuum vessel vibrations impose a sinusoidal base excitation to the supports for the cryopump. Static and dynamic analyses of the cryopump demonstrate that the stresses due to disruption and thermal loadings satisfy the stress and deflection criteria.

  8. Control of Neoclassical Tearing Modes in DIII-D

    NASA Astrophysics Data System (ADS)

    Welander, A. S.; Lahaye, R. J.; Penaflor, B. G.; Lohr, J.; Noraky, V.; Prater, R.; Eidietis, N. W.; Humphreys, D. A.; Kolemen, E.; Turco, F.

    2012-10-01

    New techniques have been developed on DIII-D for control of neoclassical tearing modes (NTMs). The NTM is a helical magnetic island formation that can occur on flux surfaces where the safety factor, q is a rational number. An NTM can be suppressed by depositing electron cyclotron current drive (ECCD) on the q-surface by injecting microwave beams into the plasma. On DIII-D, steerable mirrors that reflect these beams into the plasma can be adjusted when the q-surface is moving to keep the ECCD aligned. Accurate tracking is made possible by equilibrium reconstructions that include measurements of the motional Stark effect and by estimating beam refraction. Three different algorithms can be employed to fine-tune alignment when NTMs occur. The first method adjusts ECCD alignment in steps until the island shrinks. The second method sweeps the alignment to find where ECCD has the biggest effect on the NTM. The third method uses temperature measurements by electron cyclotron emission. The gyrotrons are pulsed and the position of the resulting temperature pulses is compared to the position where the rotating NTM causes temperature fluctuations. Recent experimental results and directions toward robust disruption-free control will be presented.

  9. Development of a radiative divertor for DIII-D

    SciTech Connect

    Allen, S.L.; Brooks, N.H.; Campbell, R.B.

    1994-07-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ({approximately}10 cm diameter) radiation zone which results in substantial reduction (3--10) in the divertor heat flux while {delta}{sub E} remains {approximately}2 times ITER-89P scaling. However, ne increases with D{sub 2} puffing, and Z{sub eff} increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity ({delta} {approximately}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented.

  10. New reflectometer systems for the DIII-D tokamak (abstract)

    SciTech Connect

    Doyle, E.J.; Kim, K.W.; Burns, S.; Nguyen, X.; Peebles, W.A.; Rhodes, T.L. )

    1992-10-01

    During a machine vent in December 1991, two new reflectometer systems were successfully installed and tested on the DIII-D tokamak. The first is an {ital X}-mode broadband system primarily intended for density profile measurements, utilizing BWO sources and covering {ital Q} and {ital V} frequency bands (33--50 and 50--75 GHz). The second system is an adaptation of a pre-existing inside launch (high field side) ECRH waveguide to provide an inside launch reflectometer capability at the same frequencies and polarization as an outside launch fixed frequency {ital O}-mode system. The new systems will have a dual role in both directly supporting the DIII-D physics program, and also acting as flexible and adaptable test beds for the development of reactor relevant reflectometer systems, such as required for ITER. Specific examples of planned measurements include investigation of possible in/out plasma asymmetries at the {ital L}--{ital H} transition and ELMs, and demonstration of routine and reliable density profile measurements. It is expected that preliminary data from the inside launch system will be available by the time of the conference. This work is supported by the U. S. Department of Energy under Grant No. DE-FG03-86-ER53225 and General Atomics subcontract SC120536 under DOE Contract No. DE-AC03-89ER51114.

  11. A phase contrast interferometer on DIII-D

    SciTech Connect

    Coda, S.; Porkolab, M.; Carlstrom, T.N.

    1992-04-01

    A novel imaging diagnostic has recently become operational on the DIII-D tokamak for the study of density fluctuations at the outer edge of the plasma. The phase contrast imaging approach overcomes the limitations of conventional scattering techniques in the spectral range of interest for transport-related phenomena, by allowing detection of long wavelength modes (up to 7.6 cm) with excellent spatial resolution (5 mm) in the radial direction. Additional motivation for the diagnostic is provided by wave-plasma interactions during heating and current drive experiments in the Ion Cyclotron range of frequencies. Density perturbations of 4 {times} 10{sup 7} cm{sup {minus}3} with a 1 MHz bandwidth can be resolved. The diagnostic employs a 7.6 cm diameter CO{sub 2} laser beam launched vertically across the plasma edge. An image of the plasma is then created on a 16-element detector array: the detector signals are directly proportional to the density fluctuations integrated along each chord. Wavelengths and correlation lengths can be inferred from the spatial mapping. The phase contrast method and its application to DIII-D are described and tests and first plasma data are presented.

  12. New Pellet Injection Schemes on DIII-D

    SciTech Connect

    Anderson, P.M.; Baylor, L.R.; Combs, S.K.; Foust, C.R.; Jernigan, T.C.; Robinson, J.I.

    1999-11-13

    The pellet fueling system on DIII-D has been modified for injection of deuterium pellets from two vertical ports and two inner wall locations on the magnetic high-field side (HFS) of the tokamak. The HFS pellet injection technique was first employed on ASDEX-Upgrade with significant improvements reported in both pellet penetration and fueling efficiency. The new pellet injection schemes on DIII-D required the installation of new guide tubes. These lines are {approx_equal}12.5 m in total length and are made up of complex bends and turns (''roller coaster'' like) to route pellets from the injector to the plasma, including sections inside the torus. The pellet speed at which intact pellets can survive through the curved guide tubes is limited ({approx_equal}200-300 m/s for HFS injection schemes). Thus, one of the three gas guns on the injector was modified to provide pellets in a lower speed regime than the original guns (normal speed range {approx_equal}500 to 1000 m/s). The guide tube installations and gun modifications are described along with the injector operating parameters, and the latest test results are highlighted.

  13. New Thomson scattering Laser Control for DIII-D

    SciTech Connect

    Makariou, C.C.; Stockdale, R.E.; Carlstrom, T.N.; Hsieh, C.L.; Bramson, G.

    1993-10-01

    A Laser Control system has been built for the DIII-D Scattering Diagnostic. This new system has provided the capability to place the laser probe pulses with one microsecond timing precision throughout the DIII-D shot. The new system fires the eight lasers with a programmable sequence which repeats ever 50 ms. If one wants to probe the plasma at a higher rate to study a fast paced event, the new control circuit can fire all charged lasers in rapid succession (BURST MODE). This burst rate is programmable. The new Laser Control system successfully replaced the previous control scheme which consisted of three VME Motorola 68030 computers (one host under UNIX VME V/68 and two interrupt driven targets under VME Exec. The old system was not successful due to the many VME interrupts needed to service the lasers. The new hardware approach is much more reliable. The old system still controls data acquisition and as a monitoring system since it does not have the burden of controlling the lasers. A brief description of the Thomson Scattering diagnostic will be presented with emphasis in the new upgraded laser firing control system and data acquisition timing control.

  14. Study of the Resistive Wall Mode in DIII--D

    NASA Astrophysics Data System (ADS)

    Garofalo, A. M.; Mauel, M. E.; Navratil, G. A.; Sabbagh, S. A.; Strait, E. J.; La Haye, R. J.; Turnbull, A. D.; DIII-D Team; Rice, B. W.

    1997-11-01

    Stability analysis of DIII--D discharges showed kink mode stabilization by a resistive wall in D-shaped plasmas with βN exceeding the expected no-wall ideal βN limit by a factor of 1.3.(E.J. Strait, et al.), Phys. Rev. Lett. 74, 2483 (1995). We will call this factor a wall stability enhancement factor, E_w: Ew = βN (experiment)/ βN (no-wall limit, MHD model). Recent DIII--D experiments were aimed at achieving a value of Ew > 1.3 in lower single null, JET-like plasmas with B_t=2.0--2.1 T and Ip = 1.8 MA. A lower plasma internal inductance of l_i ~ 0.7 (and thus lower no-wall βN limit) was produced using early neutral beam injection and a fast positive current ramp during beam injection in an H--mode plasma. Preliminary analysis shows a slowly rotating (25 Hz) n = 1 mode growing in a 40 ms time scale just before a β collapse, similar to what was observed in Ref. 2. The results of detailed stability studies using ideal (GATO) and resistive (MARS) MHD codes will be presented.

  15. Stability Modeling of DIII-D Discharges with Transport Barriers

    NASA Astrophysics Data System (ADS)

    Lao, L. L.; Ferron, J. R.; Lin-Liu, Y. R.; Strait, E. J.; Turnbull, A. D.; Taylor, T. S.; Murakami, M.

    1999-11-01

    The stability of DIII--D discharges with transport barriers is systematically studied by modeling the pressure profiles using a hyperbolic tangent representation with various radii, widths, and amplitudes. The q profiles are modeled using a spline representation with varying q(0), q_min, and ρ_q_min. The equilibria are computed using the EFIT and the TOQ codes based on the parameters from a strongly shaped high triangurality DIII--D long pulse high performance discharge. Stability against the ideal low n=1 and 2 modes is evaluated using the GATO code with a conducting wall at 1.5 a. The results show that the stability improves with increasing transport barrier width and radius but varies weakly with q(0). When the transport barriers are L--mode like and have narrow widths in the plasma core, the stability is limited by the n=1 mode. When they are H--mode like and have large widths extending toward the edge, the stability is limited by the n=2 mode.

  16. Implementation of a Microwave Imaging Reflectometer on DIII-D

    NASA Astrophysics Data System (ADS)

    Kriete, D. M.; Tobias, B. J.

    2013-10-01

    The microwave imaging reflectometer (MIR) is a new plasma diagnostic system on DIII-D that will make localized measurements of density fluctuations on a poloidal cross section of the tokamak. The data from these measurements will provide a quantitative picture of plasma turbulence and MHD instabilities. This project's focus is to install the MIR system on DIII-D, perform calibration tests on it, and develop data analysis tools to process MIR data, with a focus on estimating measurement error. Tests include taking dark shots to subtract out passive noise from the measurements and taking plasma shots to better quantify active noise. Synthetic diagnostics based on numerical codes will also be used to evaluate the MIR system. The MIR uses the same optics as the existing electron cyclotron emission imaging (ECEI) system so, after completion, physicists will have access to a 2D image of both density and temperature fluctuations within the plasma. The MIR thus has broad utility across experiments seeking to understand turbulent transport. Supported by the National Undergraduate Fellowship Program in Plasma Physics and Fusion Energy Sciences and the US DOE under DE-FC02-04ER54698 and DE-AC02-09CH11466.

  17. The DIII-D ECH multiple gyrotron control system

    SciTech Connect

    Ponce, D.; Lohr, J.; Tooker, J.F.; Cary, W.P.; Harris, T.E.

    1997-11-01

    DIII-D`s ECH upgrade with 1 MW, 110 GHz gyrotrons is ongoing, and with it, an upgrade of the control system. The ECH Multiple Gyrotron Control System uses software distributed among networked computers, interfaced to a programmable logic controller (PLC), the timing and pulse system, power supplies, vacuum and wave guide controls, and instrumentation. During DIII-D operations, the system will allow a chief and a co-operator to control and monitor a number of gyrotrons from different manufacturers. The software, written using LabVIEW, allows for remote and multiple operator control. Thus any supported computer can become a control station and multiple projects can be simultaneously accommodated. Each operator can be given access to the controls of all gyrotrons or to a subset of controls. Status information is also remotely available. The use of a PLC simplifies the hardware and software design. It reduces interlock and control circuitry, includes monitoring for slow analog signals, and allows one software driver to efficiently interface to a number of systems. In addition, the interlock logic can be easily changed and control points can be forced as needed. The pulse system is designed around arbitrary function generators. Various modulation schemes can be accommodated, including real-time control of the modulation. This discussion will include the hardware and software design of the control system and its current implementation.

  18. Remote collaboration and data access at the DIII-D National Fusion Facility

    SciTech Connect

    Schissel, D.P.

    1998-09-01

    As the number of on-site and remote collaborators has increased, the demands on the DIII-D National Program`s computational infrastructure has become more severe. The Director of the DIII-D Program recognized the increased importance of computers in carrying out the DIII-D mission and in late 1997 formed the Data Analysis Programming Group. Utilizing both software and hardware improvements, this new group has been charged with increasing the DIII-D data analysis throughput and data retrieval rate. Understanding the importance of the remote collaborators, this group has developed a long term plan that will allow for fast 24 hour data access (7x24) with complete documentation and a set of data viewing and analysis tools that can be run either on the collaborators` or DIII-D`s computer systems. This paper presents the group`s long term plan and progress to date.

  19. Molybdenum density profiles on C-Mod using FAC generated cooling curves

    NASA Astrophysics Data System (ADS)

    Reinke, M.

    2005-10-01

    For tokamaks with high-Z plasma facing components, maintaining a low impurity content is necessary to produce high quality, repeatable discharges. A GENeral Impurity Emissivity (GENIE) method is outlined for determining impurity profiles using experimental spectroscopy data, an impurity transport code, and the atomic physics package, Flexible Atomic Code (FAC). Modular programming is emphasized in order to make the method extendable to arbitrary impurities, diagnostic sets and tokamaks. Development of GENIE is ongoing, but a necessary first step is to verify FAC. A testing stage of GENIE that ignores transport is demonstrated and the results are validated against the published molybdenum cooling-curve generated using HULLAC. Bolometry and Thomson scattering data are used to determine molybdenum density profiles on Alcator C-Mod using the Mo cooling-curve. Instances where this method fails are shown as well to illustrate the need for a more advanced version of GENIE that generates and uses charge state distributions that assume transport.

  20. DIII-D Research in Support of ITER

    SciTech Connect

    Strait, E

    2008-10-14

    DIII-D research is providing key information for the design and operation of ITER. Discharges that simulate ITER operating scenarios in conventional H-mode, advanced inductive, hybrid, and steady state regimes have achieved normalized performance consistent with ITER's goals for fusion performance. Stationary discharges with high {beta}{sub N} and 90% noninductive current that project to Q=5 in ITER have been sustained for a current relaxation time ({approx}2.5 s), and high-beta wall-stabilized discharges with fully non-inductive current drive have been sustained for more than one second. Detailed issues of plasma control have been addressed, including the development of a new large-bore startup scenario for ITER. A broad research program provides the physics basis for predicting the performance of ITER. Recent key results include the discovery that the L-H power threshold is reduced with low neutral beam torque, and the development of a successful model for prediction of the H-mode pedestal height in DIII-D. Research areas with the potential to improve ITER's performance include the demonstration of ELM-free 'QH-mode' discharges with both co and counter-injection, and validation of the predicted torque generated by static, non-axisymmetric magnetic fields. New diagnostics provide detailed benchmarking of turbulent transport codes and direct measurements of the anomalous transport of fast ions by Alfven instabilities. DIII-D research also contributes to the basis for reliable operation in ITER, through active control of the chief performance-limiting instabilities. Recently, ELM suppression with resonant magnetic perturbations has been demonstrated at collisionality similar to ITER's, while simultaneous stabilization of NTMs (by localized current drive) and RWMs (by magnetic feedback) has allowed stable operation at high beta and low rotation. In research aimed at improving the lifetime of material surfaces near the plasma, recent experiments have investigated

  1. Hardware and Software Upgrades to DIII-D Main Computer Control System

    SciTech Connect

    Piglowski, D. A.; Penaflor, B.G.; McHarg, JR., B.B.; Greene, K.L.; Coon, R.M.; Phillips, J.C.

    2002-09-01

    The complexities of monitoring and controlling the various DIII-D tokamak systems have always required the aid of high-speed computer resources. Because of recent improvements in computing technology, DIII-D has upgraded both hardware and software for the central DIII-D control system. This system is responsible for coordination of all main DIII-D subsystems during a plasma discharge. The replacement of antiquated older hardware has increased reliability and reduced costs both in the initial procurement and eventual maintenance of the system. As expected, upgrading the corresponding computer software has become the more time consuming and expensive part of this upgrade. During this redesign, the main issues focused on making the most of existing in-house codes, speed with which the new system could be brought on-line, the ability to add new features/enhancements, ease of integration with all DIII-D systems and future portability/upgrades. The resulting system has become a template by which other DIII-D systems can follow during similar upgrade paths; in particular DIII-D's main data acquisition system and neutral beam injection (NBI).

  2. Power deposition on the DIII-D inner wall limiter

    NASA Astrophysics Data System (ADS)

    Stangeby, P. C.; Tsui, C. K.; Lasnier, C. J.; Boedo, J. A.; Elder, J. D.; Kocan, M.; Leonard, A. W.; McLean, A. G.; Pitts, R. A.; Rudakov, D. L.

    2015-08-01

    ITER will use the shaped inner column as startup/rampdown limiter. An exploratory experiment was performed in DIII-D using a similarly shaped inner wall limiter, looking for evidence of a short fall-off feature, λshort, near the last closed flux surface. Measurements were made on the high field side (HFS) using infrared thermography and a Langmuir swing-probe. In some cases clear evidence was found for a narrow feature λshort ∼ ρpolD+ ∼ a few mm. The ratio q||0 short / q||0 long was mostly between 0.1 and 0.6. Measurements made on the low field side and bottom side showed little clear indication of any narrow feature.

  3. Detached divertor operation in DIII-D helium plasmas

    SciTech Connect

    Hill, D. N., LLNL

    1998-05-01

    This paper presents results from operating helium plasmas in DIII-D in which helium gas puffing is used to reduce the peak divertor heat flux by factors of four or more. The threshold density for achieving these conditions is nearly the same as for deuterium plasmas, which is surprising given the fact that lack of chemical sputtering reduces the carbon concentration in the plasma by more than a factor of five. Spectroscopic analysis shows that helium becomes the primary radiation in these plasmas, which is possible because, unlike carbon, it is the primary species present. These plasmas differ from the usual partially detached divertor (PDD) plasmas in that there is no concomitant reduction in target plate ion flux with target plate heat flux in the scrape off later outside the separatrix.

  4. Optimized Baking of the DIII-D Vessel

    SciTech Connect

    P.M. Anderson; A.G. Kellman

    1999-11-01

    The DIII-D tokamak vacuum vessel baking system is used to heat the vessel walls and internal hardware to an average temperature of 350 C to allow rapid conditioning of the vacuum surfaces. The system combines inductive heating and a circulating hot air system to provide rapid heating with temperature uniformity required by stress considerations. In recent years, the time to reach 350 C had increased from 9 hrs to 14 hrs. To understand and remedy this sluggish heating rate, an evaluation of the baking system was recently performed. The evaluation indicated that the mass of additional in-vessel hardware (50% increase in mass) was primarily responsible. This paper reports on this analysis and the results of the addition of an electric air heater and procedural changes that have been implemented. Preliminary results indicate that the time to 350 C has been decreased to 4.5 hours and the temperature uniformity has improved.

  5. Improved timing sequence generator on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Colio, R. A.; Finkenthal, D. F.; Deterly, T. M.

    2011-10-01

    The DIII-D tokamak uses a central clock source and trigger system to synchronize plant operations and diagnostics. The system uses a bi-phase encoding technique to send both clock and trigger signals to remote receivers, and supports both pre-programmed sequences of triggers as well as event-driven triggers. A 1 MHz timebase is used and triggers are encoded as eight-bit hexadecimal words. Currently, the system relies on a cascaded series of CAMAC-based delay generators to produce the trigger sequence. We present a modern and more versatile implementation based on a single FPGA (field programmable gate array) capable of providing clock rates upward of 100 MHz while maintaining compatibility with existing equipment. A proposal for system clock synchronization with GPS for improved precision is also presented. Work supported in part by US DOE under DE-FC02-04ER54698 and the National Undergraduate Fellowship in Fusion Science and Engineering.

  6. Upgraded Calibrations of the Thomson System at DIII-D

    SciTech Connect

    B. Bray; C. Hsieh; T.N. Carlstrom; C.C. Makariou

    2000-08-01

    The DIII-D Thomson system measures electron density and temperature with eight pulsed ND:YAG lasers along three paths through the plasma vessel. The components of the Thomson system are absolutely calibrated so the measurements can be combined into a single profile from a normalized plasma radius ({rho}) of about 0.1 to the edge of the plasma. A monochromator calibration and opto-electronic calibration measure the detectors' absolute sensitivity to background and pulsed light. A Rayleigh scattering calibration and transmission calibrations measure the transmission of light to the detectors. The calibration systems are being upgraded to reduce the effect of systematic errors on the temperature and density measurements. The systematic errors can be checked by a comparison of overlapping channels and estimated from fits to the profiles. The contributions of the systematic uncertainties relative to the statistical uncertainties of the measurement are discussed through simulations and experimental data.

  7. Thomson scattering diagnostic upgrade on DIII-D

    SciTech Connect

    Ponce-Marquez, D. M.; Bray, B. D.; Deterly, T. M.; Liu, C.; Eldon, D.

    2010-10-15

    The DIII-D Thomson scattering system has been upgraded. A new data acquisition hardware was installed, adding the capacity for additional spatial channels and longer acquisition times for temperature and density measurements. Detector modules were replaced with faster transimpedance circuitry, increasing the signal-to-noise ratio by a factor of 2. This allows for future expansion to the edge system. A second phase upgrade scheduled for 2010-2011 includes the installation of four 1 J/pulse Nd:YAG lasers at 50 Hz repetition rate. This paper presents the first completed phase of the upgrade and performance comparison between the original system and the upgraded system. The plan for the second phase is also presented.

  8. Recent results from the DIII-D Tokamak

    SciTech Connect

    Kellman, A.G.

    1995-10-01

    The goal of the DIII-D program is to provide the integrated basis for commercially attractive steady state fusion power plants. Significant progress toward this goal has been achieved, enabled by system improvements including an error field correction coil, an expanded diagnostic set, a digital plasma control system, and high power rf systems. Simultaneous improvements in both the confinement and stability have been achieved during both VH-mode and negative central shear discharges. Fully non-inductive discharges with high bootstrap current fraction have been obtained. The divertor program has demonstrated simultaneous reduction of divertor heat flux and effective particle control using gas puffing and an in-vessel cryopump. Control of the wall particle inventory, He exhaust, and characterization of the scrapeoff layer and divertor plasma have been achieved. Progress has also been made in addressing additional specific needs for ITER: investigation of basic transport scaling, disruption characterization and avoidance, material erosion, and steady state beta limits.

  9. Radiative divertor plasmas with convection in DIII-D

    SciTech Connect

    Leornard, A.W.; Porter, G.D.; Wood, R.D.

    1998-01-01

    The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features.

  10. Microturbulence in DIII-D tokamak pedestal. II. Electromagnetic instabilities

    NASA Astrophysics Data System (ADS)

    Holod, I.; Fulton, D.; Lin, Z.

    2015-09-01

    Gyrokinetic simulations have been used to identify electromagnetic microinstabilities in the H-mode pedestal region of DIII-D shot 131 997 using global gyrokinetic code GTC. It was found that dominant instability at the top of the pedestal is the ion temperature gradient mode (ITG). In the maximum gradient location the most unstable mode is the kinetic ballooning mode (KBM) for the dominant poloidal wavenumber {{k}θ}≈ 1 cm-1. For shorter wavelengths the dominant instability is the trapped-electron mode (TEM). We have demonstrated the ITG-KBM transition at the pedestal top, and TEM-KBM transition in the steep pressure gradient region as plasma pressure increases while gradients remain unchanged.

  11. Improved edge charge exchange recombination spectroscopy in DIII-D

    DOE PAGESBeta

    Chrystal, Colin; Burrell, K. H.; Grierson, Brian A.; Haskey, Shaun R.; Groebner, R. J.; Kaplan, David H.; Briesemeister, Alexis R.

    2016-08-02

    The charge exchange recombination spectroscopy diagnostic on the DIII-D tokamak has been upgraded with the addition of more high radial resolution view chords near the edge of the plasma (r/a > 0.8). The additional views are diagnosed with the same number of spectrometers by placing fiber optics side-by-side at the spectrometer entrance with a precise separation that avoids wavelength shifted crosstalk without the use of bandpass filters. The new views improve measurement of edge impurity parameters in steep gradient, H-mode plasmas with many different shapes. The number of edge view chords with 8 mm radial separation has increased from 16more » to 38.As a result, new fused silica fibers have improved light throughput and clarify the observation of non-Gaussian spectra that suggest the ion distribution function can be non-Maxwellian in low collisionality plasmas.« less

  12. 2-D tomography with bolometry in DIII-D

    SciTech Connect

    Leonard, A.W.; Meyer, W.H.; Geer, B.; Behne, D.M.; Hill, D.N.

    1994-07-01

    We have installed a 48-channel platinum-foil bolometer system on DIII-D achieve better spatial and temporal resolution of the radiated power in diverted discharges. Two 24-channel arrays provide complete plasma coverage with optimized views of the divertor. We have measured the divertor radiation profile for a series of radiative divertor and power balance experiments. We observe a rapid change in the magnitude and distribution of divertor radiation with heavy gas puffing. Unfolding the radiation profile with only two views requires us to treat the core and divertor radiation separately. The core radiation is fitted to a function of magnetic flux and is then subtracted from the divertor viewing chords. The divertor profile is then fit to a 2-D spline as a function of magnetic flux and poloidal angle.

  13. Edge Plasma Effects in DIII-D Impurity Seeded Discharges

    SciTech Connect

    Jackson, G.L.; Boedo, J.A.; Lasnier, C. J.; Leonard, A.W.; McKee, G. R.; Murakami, M; Wade, M.R.; Watkins, J.G.; West, W.P.; Whyte, D.G.

    2002-06-01

    DIII-D, ELMing H-mode radiating mantle discharges have been obtained with electron density near the Greenwald density limit and a large fraction of the input power radiated inside the last closed flux surface, significantly reducing peak divertor heat fluxes. In these ''puff and pump'' discharges, the introduction of argon reduces particle flux to divertor tiles by a factor of 4 while peak heat flux is half of the no impurity value, suggesting that impurity seeding may be a useful control tool to reduce wall heat and particle fluxes in fusion reactors. A robust H-mode transport barrier is maintained and there is little change in the ELM energy or in the ELM frequency.

  14. Remote Experimentation on DIII--D: Operations and Concepts

    NASA Astrophysics Data System (ADS)

    Casper, T. A.; Hill, D. N.; Meyer, W. H.; Moller, J. M.; Stallard, B. W.; Wood, R. D.; Scoville, J. T.; McHarg, B. B.; Operations Group; Operations Group; Colchin, R. J.; Greenwood, D.; Davis, S.; Bly, S.

    1997-11-01

    The DIII--D tokamak in San Diego, CA was operated for a full day of experiments from a Remote Experimental Site at LLNL where a team of researchers directed experimental operations and successfully completed an experiment designed to investigate L--H transition physics and plasma detachment from the divertor. The team at LLNL comprised the physics experiment leader and scientists operating the plasma shape control system, neutral beam heating, diagnostics and intershot data analysis. Researchers at ORNL and PPPL sites also participated in this experiment. Communication among all sites was implemented with four internet-based, audio/video links to provide discussions essential to operations and research in progress. This experiment was completed with an efficiency comparable to that achieved with local control room operations. Details of the remote operation, the concepts involved and its relevance to ITER will be discussed.

  15. The Multiple Gyrotron System on the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Lohr, John; Cengher, Mirela; Doane, John L.; Gorelov, Yuri A.; Moeller, Charles P.; Ponce, Dan; Prater, Ron

    2011-03-01

    The electron cyclotron heating and current drive complex on the DIII-D tokamak presently comprises six gyrotrons injecting rf power from the low field side at 110 GHz, the 2 f ce resonance at the center of the vacuum chamber. Typical injected rf power is 600-650 kW per gyrotron. The launched rf can be directed over ±20° toroidally to create both co- and counter-current drive and scanned over 40° poloidally to permit the injected rf beams to intersect, and be absorbed at, the second harmonic resonance anywhere in the tokamak upper half plane. The elliptical polarization is controlled so that the desired extraordinary or ordinary modes are excited for any injection geometry. The maximum injected energy on a single plasma shot has been 16.6 MJ for six gyrotrons injecting a total of 3.4 MW for 5 seconds.

  16. Horizontal Thomson Scattering Systems for DIII-D and SSPX

    SciTech Connect

    Nilson, D.G.; Hill, D.N.; Wood, R.D.; McClean, H.; Moeller, J.M.; Labik, G.; Carlstron, T.N.; Bray, B.; Hsieh, C.L.

    1999-08-01

    DIII-D--Three of the seven existing core Thomson scattering laser beams were redirected to probe the previously unmeasured central region of the DIII-D plasma. Modifications to the existing collection optics system and support tower were made to inject the lasers and collect scattered light in this new extended region. Stray light levels were reduced to acceptable levels to permit Rayleigh scattering calibration on five of the six new channels, indicating that the new in-vessel dump operates well. Measurements of the plasma temperature and density from the plasma edge to the center are now possible. Peaked density profiles are now observed in this new measurement region. SSPX--We have completed the design and installation of a 10-spatial channel Thomson scattering system to measure the plasma temperature and density profile on SSPX. A single-pulsed YAG laser operating at 0.7 J and 8 ns is used to scatter photons into a 7-element collection optic that provides a spatial resolution of 1.5 cm at the outer plasma edge and 7.5 cm at the inner edge of a .5 m radius spheromak plasma. The collected light is then analyzed by a 4-channel interference filtered polychromator which has been optimized to measure temperatures between 2 eV and 2 keV and densities as low as 1 x 10{sup 12} cm{sup 3}. We use an in-vessel beam dump and a series of entrance and exit baffles to reduce the stray laser light and provide for an absolute density calibration by Rayleigh scattering in argon gas.

  17. The DIII-D Radiative Divertor Project: Status and plans

    SciTech Connect

    Smith, J.P.; Baxi, C.B.; Bozek, A.S.

    1996-10-01

    New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the second and final phase, in 1998. The phased approach enables the continuation of the divertor characterization research in the lower divertor while providing pumping for density control in high triangularity, single- or double-null advanced tokamak discharges. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. By puffing neutral gas into the divertor region, a reduction in the heat flux on the target plates will be be demonstrated without a large rise in core density. This reduction in heat flux is accomplished by dispersing the power with radiation in the divertor region. Experiments and modeling have formed the basis for the new design. The capability of the DIII-D cryogenic system is being upgraded as part of this project. The increased capability of the cryogenic system will allow delivery of liquid helium and nitrogen to three new cryopumps. Physics studies on the effects of slot width and length can be accomplished easily with the design of the Radiative Divertor. The slot width can be varied by installing graphite tiles of different geometry. The change in slot length, the distance from the X-point to the target plate, requires relocating the structure vertically and can be completed in about 6-8 weeks. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Required diagnostic modifications will be minimal for Phase 1, but extensive for Phase 2 installation. These Phase 2 diagnostics will be required to fully diagnose the high triangularity discharges in the divertor slots.

  18. High-beta discharges in the DIII-D tokamak

    SciTech Connect

    Ferron, J.R.; Chu, M.S.; Helton, F.J.; Howl, W.; Kellman, A.G.; Lao, L.L.; Lazarus, E.A.; Lee, J.K.; Osborne, T.H.; Strait, E.J.; Taylor, T.S.; Turnbull, A.D. )

    1990-06-01

    Low-{ital q} ({ital q}{sub 95}{lt}3) double-null divertor discharges with values of the volume-average toroidal beta as high as 9.3% have been operated in the DIII-D tokamak (Fusion Technol. {bold 8}, 441 (1985)). In discharges with {ital q}{sub 95}{approx}5, values of {beta}{sub {ital T}}/({ital I}/{ital aB}) as high as 5 have been obtained. These discharges are shown to be at or below the stability limit to the value of beta for infinite-{ital n}, ideal ballooning modes. The discharges are significantly below the beta limit for ideal,low toroidal mode number kink modes. The kink mode beta limit is shown to be strongly dependent on the radial profiles of plasma pressure and current. The theoretical beta limit in DIII-D is shown to be in the range {beta}{sub {ital T}}/({ital I}/{ital aB})=4 --5 depending on the value of {ital I}/{ital aB}, and this is consistent with the experiment. High-beta discharges have been operated with ion temperature up to 17 keV. Steady-state, high-beta, low-{ital q} operation is demonstrated by a discharge with {ital I}/{ital aB}=2.6, {ital q}{sub 95}=2.7, in which {beta}{sub {ital T}}{gt}7% is maintained for 1.5 sec.

  19. Advanced tokamak research on the DIII-D tokamak

    SciTech Connect

    Chan, V.S.

    1994-01-01

    The objective of the planned research in advanced tokamak development on DIII-D at General Atomics, San Diego, USA. is to establish improved tokamak operation through significant improvements in the stability factor, confinement quality, and bootstrap current fraction using localized radio frequency (rf) current profile control, rf and neutral beam heating for pressure profile control, as well as control of plasma rotation and optimization of the plasma boundary conditions. Recent research results in H-mode confinement, modifications of current profiles to achieve higher confinement and higher {beta}, a new regime of improved confinement (VH-mode), and rf noninductive current drive are encouraging. In this talk, arguments will be presented supporting the need for improved performance in tokamak reactors. Experimentally observed advanced performance regimes on DIII-D will be discussed. Confinement improvement up to H = 4, where H is the ratio of energy confinement time to the ITER89-P scaling H{triple_bond} {Tau}{sub E}/{Tau}{sub E-ITER89-P}, has been achieved. In other discharges {beta}{sub N} = {beta}/(I/aB),[%-m{center_dot}{Tau}/MA] {approx_gt} 6 has been obtained. These values have so far been achieved transiently and independently. Techniques, will be described which can extend the high performance to quasi-steady-state and sustain the high H and {beta}{sub N} values simultaneously. Two high performance regimes, one in first stable regime and the other in second stable regime, have been simulated br self-consistently evolving a magnetohydrodynamic (MHD) equilibrium-transport code. Finally, experimental program plans and outstanding important physics issues will be discussed.

  20. NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK

    SciTech Connect

    WALKER, ML; FERRON, JR; HUMPHREYS, DA; JOHNSON, RD; LEUER, JA; PENAFLOR, BG; PIGLOWSKI, DA; ARIOLA, M; PIRONTI, A; SCHUSTER, E

    2002-10-01

    OAK A271 NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK. The advanced tokamak (AT) operating mode which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. Simultaneous high performance regulation of the plasma boundary and internal profiles requires multivariable control techniques to account for the highly coupled influences of equilibrium shape, profile, and stability control. This paper describes progress towards the DIII-D At mission goal through both significantly improved real-time computational hardware and control algorithm capability.

  1. Enhancing Physics Operations and Increasing Physics Productivity at DIII-D

    NASA Astrophysics Data System (ADS)

    Hyatt, A. W.; Humphreys, D. A.; Johnson, R. D.; Walker, M. L.; Ferron, J. R.; Scoville, J. T.; Wade, M. R.

    2008-11-01

    Having begun operation in 1986, DIII-D is a mature tokamak with highly effective physics operations procedures to plan and execute discharges, monitor machine status, and train personnel. A desire to maximize physics productivity per discharge has driven an ongoing process to enhance physics operations. Developments in modeling and simulation software and data-driven improved fault monitoring allow continuing improvement. Productivity enhancements include use of control design and simulation tools to reduce machine time needed to develop new algorithms, ongoing training programs for physics operators, improved physics and control-related trouble reporting, and deployment of new status reporting displays in the DIII-D Control Room. Use of the DIII-D Plasma Control System at other devices such as NSTX, EAST, and KSTAR have provided solutions for use at DIII-D as well. Results of this ongoing process and lessons learned for next generation devices such as ITER will be discussed.

  2. DIII-D research operations. Annual report, October 1, 1991--September 30, 1992

    SciTech Connect

    Baker, D.

    1993-05-01

    This report discusses the research on the following topics: DIII-D program overview; divertor and boundary research program; advanced tokamak studies; tokamak physics; operations; program development; support services; contribution to ITER physics R&D; and collaborative efforts.

  3. Gyrotron Performance on the 110 GHZ Installation at the DIII-D Tokamak

    SciTech Connect

    Gorelov, I.; Lohr, J.M.; Ponce, D.; Callis, R.W.; Ikezi, H.; Legg, R.A.; Tsimring, S.E.

    1999-06-01

    The 110 GHz gyrotron system on the DIII-D tokamak comprises three different gyrotrons in the 1 MW class. The individual gyrotron characteristics and the operational experience with the system are described.

  4. Enhanced Computational Infrastructure for Data Analysis at the DIII-D National Fusion Facility

    SciTech Connect

    Schissel, D.P.; Peng, Q.; Schachter, J.; Tepstra, T.B.; Casper, T.A.; Freeman, J.; Jong, R.; Keith, K.M.; McHarg, B.B., Jr; Meyer, W.H.; Parker, C.T.; Warner, A.M.

    1999-07-01

    The DIII-D National Team consists of about 120 operating staff and 100 research scientists drawn from 9 U.S. National Laboratories, 19 foreign laboratories, 16 universities, and 5 industrial partnerships. This multi-institution collaboration carries out the integrated DIII-D program mission which is to establish the scientific basis for the optimization of the tokamak approach to fusion energy production. Presently, about two-thirds of the research physics staff are from the national and international collaborating institutions.

  5. ITER startup studies in the DIII-D tokamak

    SciTech Connect

    Jackson, G. L.; Casper, T. A.; Luce, T.C.; Humphreys, D A; Ferron, J.R.; Hyatt, A. W.; Lazarus, Edward Alan; Moyer, R.A.; Petrie, T W; Rudakov, D.L.; West, W. P.

    2008-01-01

    plasma initiation and current ramp up scenario envisioned for ITER has been simulated in DIII-D experiments. These discharges were limited on the low field side (LFS) during the initial current ramp up, as specified for the ITER baseline startup scenario. Initial experiments produced internal inductance (l(i)),higher than the design value for the ITER shaping coils, often leading to vertical instabilities. A modified startup with larger volume was developed to reduce l(i) in the current ramp up. This large-bore scenario, also limiting on the LFS, produced a lower l(i) and avoided the vertical instabilities. Feedback control of l(i), using the ohmic field coil power supply as the actuator, was successfully demonstrated. Such control may be useful in avoiding vertical instabilities and in providing access to sawtooth-free steady state and hybrid scenarios in ITER. Experiments at reduced inductive voltage and with electron cyclotron assist for breakdown and burnthrough have also been carried out. The Corsica equilibrium and transport code has modelled these data to provide validation of transport models used to simulate this phase of ITER discharges in order to yield more accurate extrapolation to ITER scenarios.

  6. Multipoint Thomson scattering diagnostic for DIII-D

    NASA Astrophysics Data System (ADS)

    Hsieh, C. L.; Chase, R.; Deboo, J. C.; Evanko, R. G.; Gohil, P.; Snider, R. T.; Stockdale, R. E.

    1988-02-01

    The multipoint Thomson scattering diagnostic for DIII-D has been in operation since early 1987. Its capability of measuring electron temperatures as low as 50 eV at densities of a few times 10 to the 12th/cu cm and a spatial resolution of 1.4 cm in the plasma edge region has been essential in the study of H-mode plasmas. The major components of the diagnositic system consist of a 10-J ruby laser, many stages of funnel shaped beam baffles, stacks of razor blades for viewing and stray light dumps, two wide-angle lenses for the collection of scattered light through re-entrant window ports with protection shutters, 88 fiber bundles 15 m in length for light transmission to an focal length/1.9 spectrometer, an intensified CCD camera system capable of single photon detection, and a VAX computer for hardware control and data processing. The input end of the fiber bundle is made demountable so the view locations can be reconfigured to optimize the spatial resolution for the plasma region of interest.

  7. Multipoint Thomson scattering diagnostic for DIII-D

    NASA Astrophysics Data System (ADS)

    Hsieh, C. L.; Chase, R.; DeBoo, J. C.; Evanko, R. G.; Gohil, P.; Snider, R. T.; Stockdale, R. E.

    1988-08-01

    The multipoint Thomson scattering diagnostic for DIII-D has been in operation since early 1987. Its capability of measuring electron temperatures as low as 50 eV at densities of a few times 1012 cm-3 and a spatial resolution of 1.4 cm in the plasma edge region has been essential in the study of H-mode plasmas. The major components of the diagnostic system consist of a 10-J ruby laser, many stages of funnel-shaped beam baffles, stacks of razor blades for viewing and stray light dumps, two wide-angle lenses for the collection of scattered light through reentrant window ports with protection shutters, 88 fiber bundles 15 m in length for light transmission to an f/1.9 spectrometer, an intensified CCD camera system capable of single-photon detection, and a VAX computer for hardware control and data processing. The input end of the fiber bundle is made demountable so the view locations can be reconfigured to optimize the spatial resolution for the plasma region of interest.

  8. Emittance Analysis of the DIII-D Neutral Beam Source

    NASA Astrophysics Data System (ADS)

    Lopez, N. A.; Crowley, B.

    2014-10-01

    In a high powered neutral beam system ions are extracted from a low temperature plasma, through apertures in the arc chamber, by application of a potential to an external electrode. It has been determined that to increase the beam energy of the DIII-D neutral beam system beyond 95 keV the accelerator must be reconfigured to avoid excessive electrical breakdown in the grid gaps. Deciding exactly what modifications are to be made requires modeling and experimental effort. A basic problem is to find a geometry with which the extracted beam is intense, low divergence, free of aberrations, and does not strike the focusing electrodes. We present the results of modeling proposed reconfigurations to the accelerator geometry and source conditions. The quality of the beam produced from the various accelerator configurations is quantified through metrics such as the beam emittance and the average divergence per beamlet. By comparing the beam quality and power delivered for each proposed reconfiguration an optimal design is selected and recommended. Work supported in part by the National Undergraduate Fellowship Program in Plasma Physics and Fusion Energy Sciences and the US DOE under DE-FG02-94ER54235, DE-FC02-04ER54698.

  9. Impurity control studies using SOL flow in DIII-D

    SciTech Connect

    Wade, M.R.; Hogan, J.T.; Isler, R.C.

    1998-11-01

    Experiments on DIII-D have demonstrated the efficacy of using induced scrape-off-layer (SOL) flow to preferentially enrich impurities in the divertor plasma. This SOL flow is produced through simultaneous deuterium gas injection at the midplane and divertor exhaust. Using this SOL flow, an improvement in enrichment (defined as the ratio of impurity fraction in the divertor to that in the plasma core) has been observed for all impurities in trace-level experiments (i.e., impurity level is non-perturbative), with the degree of improvement increasing with impurity atomic number. In the case of argon, exhaust gas enrichment using a modest SOL flow is as high as 17. Using this induced SOL flow technique and argon injection, radiative ELMing H-mode plasmas have been produced that combine high radiation losses (P{sub rad}/P{sub input} > 70%), low core fuel dilution (Z{sub eff} < 1.9), and good core confinement ({tau}{sub E} > 1.0 {tau}{sub E},ITER93H).

  10. Large, Shattered Pellets for Disruption Mitigation in DIII-D

    NASA Astrophysics Data System (ADS)

    Jernigan, T. C.; Baylor, L. R.; Combs, S. K.; Commaux, N.; Meitner, S. J.; Hollmann, E. M.; Yu, Y. H.; Humphreys, D. A.; van Zeeland, M. A.; Wesley, J. C.

    2009-11-01

    A new pellet injector, the ``shotgun" pellet injector for disruption mitigation studies, has been installed on the DIII-D tokamak. The large pellets (˜15 mm diam. x 22 mm long, 2.3x10^23 electrons with D2) are shattered on a series of plates and directed toward the plasma magnetic axis. Previous experiments using massive gas injection (MGI) showed that the gas was stopped at the plasma edge and only penetrated diffusively until an MHD event, triggered by the cooling wave, transported some of the ionized gas from the edge to the interior. While the disruption forces and heat load to the first wall were significantly reduced by MGI, the core density achieved was insufficient to achieve suppression of runaway electrons from the avalanche process. Initial experiments with the new injector have demonstrated direct penetration of some of the shattered pieces deep into the plasma. Details of assimilation, mitigation, and density achieved in subsequent experiments will be presented.

  11. Tomographic Reconstruction of Flows in DIII-D

    NASA Astrophysics Data System (ADS)

    Meyer, William; Allen, Steve; Howard, John

    2015-11-01

    The DIII-D flow diagnostic produces video of interference images with horizontal fringes that contain spatial emissivity, flow, and temperature information from the lower divertor. Frames are demodulated and compared against a reference interference image to produce phase and contrast images which are the emissivity weighted flow and temperature integrated along the line-of-site, respectively. Inversion of the flow (phase) images require knowledge of the scalar product of the parallel flow vector, from the equilibrium calculations, and each camera pixel line-of-site. Four response matrices are pre-calculated: the emissivity line integral and the line integral of the scalar product of the lines-of-site with the orthogonal unit vectors of parallel flow. Equilibrium data determines the relative weight of the component matrices used in the final flow matrix. Early reconstructions have shown flow reversal during forward and reverse toroidal field plasmas. Ongoing work is to extract temperature information from the contrast images. Prepared by LLNL under Contract DE-AC52-07NA27344. This material is based upon work supported by the U.S. DOE, Office of Science, Fusion Energy Sciences.

  12. Multipoint Thomson scattering diagnostic for DIII-D

    SciTech Connect

    Hsieh, C.L.; Chase, R.; DeBoo, J.C.; Evanko, R.G.; Gohil, P.; Snider, R.T.; Stockdale, R.E.

    1988-08-01

    The multipoint Thomson scattering diagnostic for DIII-D has been in operation since early 1987. Its capability of measuring electron temperatures as low as 50 eV at densities of a few times 10/sup 12/ cm/sup -3/ and a spatial resolution of 1.4 cm in the plasma edge region has been essential in the study of H-mode plasmas. The major components of the diagnostic system consist of a 10-J ruby laser, many stages of funnel-shaped beam baffles, stacks of razor blades for viewing and stray light dumps, two wide-angle lenses for the collection of scattered light through reentrant window ports with protection shutters, 88 fiber bundles 15 m in length for light transmission to an f/1.9 spectrometer, an intensified CCD camera system capable of single-photon detection, and a VAX computer for hardware control and data processing. The input end of the fiber bundle is made demountable so the view locations can be reconfigured to optimize the spatial resolution for the plasma region of interest.

  13. Multipoint Thomson scattering diagnostic for DIII-D

    SciTech Connect

    Hsieh, C.L.; Chase, R.; DeBoo, J.C.; Evanko, R.G.; Gohil, P.; Snider, R.T.; Stockdale, R.E.

    1988-02-01

    The multipoint Thomson scattering diagnostic for DIII-D has been in operation since early 1987. Its capability of measuring electron temperatures as low as 50 eV at densities of a few times 10/sup 12/ cm/sup -3/ and a spatial resolution of 1.4 cm in the plasma edge region has been essential in the study of H-mode plasmas. The major components of the diagnositic system consist of a 10-J ruby laser, many stages of funnel shaped beam baffles, stacks of razor blades for viewing and stray light dumps, two wide-angle lenses for the collection of scattered light through re-entrant window ports with protection shutters, 88 fiber bundles 15 m in length for light transmission to an /line integral/1.9 spectrometer, an intensified CCD camera system capable of single photon detection, and a VAX computer for hardware control and data processing. The input end of the fiber bundle is made demountable so the view locations can be reconfigured to optimize the spatial resolution for the plasma region of interest. 2 refs., 7 figs.

  14. OEDGE Modeling of Detachment Threshold Experiments on DIII-D

    NASA Astrophysics Data System (ADS)

    Elder, J. D.; Stangeby, P. C.; McLean, A. G.; Leonard, A. W.; Watkins, J. G.

    2015-11-01

    A detachment threshold experiment was performed on DIII-D in which the divertor plasma transitioned from attached to weakly detached at the strike point with minimal changes in upstream parameters. The value of Te at the outer strike point measured by Thompson scattering decreased from ~ 10eV (attached) to ~ 2 eV (weakly detached). Both the Langmuir probes and the divertor Thomson diagnostics recorded increases in the particle flux on the order of a factor of two between these divertor conditions. OEDGE is used to model both of these plasma regimes for both L-mode and H-mode discharges. The behaviour of molecular hydrogen is assessed using OEDGE and possible roles of hydrogen molecules in the detachment process are examined. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-FG02-04ER54578, DE-AC04-94AL85000, DE-AC05-00OR22725, and DE-AC52-07NA27344.

  15. Radiative snowflake divertor studies in DIII-D

    NASA Astrophysics Data System (ADS)

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.

    2015-08-01

    Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D2 seeding. The SF configuration was maintained for many energy confinement times (2-3 s) in H-mode discharges (Ip = 1.2 MA, PNBI = 4- 5 MW, and B × ∇B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary Bp null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

  16. Fast reciprocating Langmuir probe for the DIII-D divertor

    SciTech Connect

    Watkins, J.G.; Hunter, J.; Tafoya, B.; Ulrickson, M.; Watson, R.D.; Moyer, R.A.; Cuthbertson, J.W.; Gunner, G.; Lehmer, R.; Luong, P.; Hill, D.N.; Mascaro, M.; Robinson, J.I.; Snider, R.; Stambaugh, R.

    1997-01-01

    A new reciprocating Langmuir probe was used to measure density and temperature profiles, ion flow, and potential fluctuation levels from the lower divertor floor up to the X point on the DIII-D Tokamak. This probe is designed to make fast (2 kHz swept, 20 kHz Mach, 500 kHz Vfloat) measurements with 2 mm spatial resolution in the region where the largest gradients on the plasma open flux tubes are found and therefore provide the best benchmarks for scrap-off layer and divertor numerical models. Profiles are constructed using the 300 ms time history of the probe measurements during the 25 cm reciprocating stroke. Both single and double null plasmas can be measured and compared with a 20 Hz divertor Thomson scattering system. The probe head is constructed of four different kinds of graphite to optimize the electrical and thermal characteristics. Electrically insulated pyrolytic graphite rings act as a heat shield to absorb the plasma heat flux on the probe shaft and are mounted on a carbon/carbon composite core for mechanical strength. The Langmuir probe sampling tips are made of a linear carbon fiber composite. The mechanical, electrical, data acquisition, and power supply systems will be described. Initial measurements will also be presented. {copyright} {ital 1997 American Institute of Physics.}

  17. Kinetic Alfven eigenmodes in JET and DIII-D

    SciTech Connect

    Jaun, A.; Hellsten, T.; Heidbrink, W.W.; Carolipio, E.

    1996-12-31

    Kinetic effects are studied for global Alfven eigenmodes in realistic tokamak equilibria with finite aspect ratio and plasmas, comparing calculations from the full wave code PENN with experimental measurements. The kinetic plasma model is based on a Larmor radius expansion in toroidal geometry and takes into account the gradients in the equilibrium density and temperatures. It allows for a consistent description of the mode conversion to the kinetic Alfven wave (KAW) and the effect of diamagnetic drifts on electromagnetic waves. Comparisons axe first carried out for a JET discharge, showing that multiple peeks measured in the low frequency Alfven spectrum are the signature of kinetic Alfven eigenmodes (KAE) induced through coupling between a global ellipticity Alfven eigenmode (EAE) and the KAW. In general, series of modes appear in the proximity of global fluid modes, some with a regular spacing in frequency and a very weak Landau damping of {vert_bar}{gamma}/{omega}{vert_bar} {approx_equal} 0.0007. A kinetic analysis of a DIII-D discharge shows that TAE mode wavefields reach the plasma core through electromagnetic drift waves which propagate because of finite temperature gradients in the regions of small k{sub {parallel}}. They can lead to particle diffusion and may explain the large losses of beam ions observed during the TAE instabilities. Comparisons of frequency and eigenmode structure axe carried out for resistive and kinetic models, between the theoretical calculations using the PENN code and the experimental measurements from magnetic probes.

  18. HIGH PERFORMANCE STATIONARY DISCHARGES IN THE DIII-D TOKAMAK

    SciTech Connect

    45th ANNUAL MEETING OF DIVISION OF PLASMA PHYSICS, ALBUQUERQUE, NEW MEXICO, OCTOBER 27-31,2003 AND TO BE PUBLISHED IN PHYS. PLASMAS.

    2003-10-01

    Recent experiments in the DIII-D tokamak [J.L. Luxon, Nucl. Fusion 42,614 (2002)] have demonstrated high {beta} with good confinement quality under stationary conditions. Two classes of stationary discharges are observed--low q{sub 95} discharges with sawteeth and higher q{sub 95} without sawteeth. The discharges are deemed stationary when the plasma conditions are maintained for times greater than the current profile relaxation time. In both cases the normalized fusion performance ({beta}{sub N}H{sub 89P}/q{sub 95}{sup 2}) reaches or exceeds the value of this parameter projected for Q{sub fus} = 10 in the International Thermonuclear Experimental Reactor (ITER) design [R. Aymar, et al., Plasma Phys. Control. Fusion 44, 519 (2002)]. The presence of sawteeth reduces the maximum achievable normalized {beta}, while confinement quality (confinement time relative to scalings) is largely independent of q{sub 95}. Even with the reduced {beta} limit, the normalized fusion performance maximizes at the lowest q{sub 95}. Projections to burning plasma conditions are discussed, including the methodology of the projection and the key physics issues which still require investigation.

  19. Comparison of Sawtooth Phenomenology on TFTR and DIII-D

    SciTech Connect

    B. Rice; E. Fredrickson; J. Manickam; M. Austin; R. Groebner <085>USDOE Office of Energy Research

    1999-06-01

    An experiment to study sawtooth phenomena and to find the threshold for sawtooth stabilization with neutral beam injection heating, as was commonly observed on TFTR, has been done on DIII-D. In the experiments, with co-tangential neutral beam injection at powers of up to 13MW, the sawtooth period was observed to increase to of order 250 msec. Stabilization of the sawteeth for the length of the high power NBI (0.5-0.8 sec) was not observed. The sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. Fast, 2 msec interval, measurements were made of the ion temperature evolution following the sawtooth to document the ion heat pulse characteristics. These data show that the ion heat pulse does not exhibit the very fast, ''ballistic'' behavior seen for the electrons. The current profile and other equilibrium profiles were measured on slower time scales. These results are compared to the data from similar studies carried out on TFTR.

  20. Fast wave current drive in DIII-D

    SciTech Connect

    Petty, C.C.; Callis, R.W.; Chiu, S.C.; deGrassie, J.S.; Forest, C.B.; Freeman, R.L.; Gohil, P.; Harvey, R.W.; Ikezi, H.; Lin-Liu, Y.-R.

    1995-02-01

    The non-inductive current drive from fast Alfven waves launched by a directional four-element antenna was measured in the DIII-D tokamak. The fast wave frequency (60 MHz) was eight times the deuterium cyclotron frequency at the plasma center. An array of rf pickup loops at several locations around the torus was used to verify the directivity of the four-element antenna. Complete non-inductive current drive was achieved using a combination of fast wave current drive (FWCD) and electron cyclotron current drive (ECCD) in discharges for which the total plasma current was inductively ramped down from 400 to 170 kA. For discharges with steady plasma current, up to 110 kA of FWCD was inferred from an analysis of the loop voltage, with a maximum non-inductive current (FWCD, ECCD, and bootstrap) of 195 out of 310 kA. The FWCD efficiency increased linearly with central electron temperature. For low current discharges, the FWCD efficiency was degraded due to incomplete fast wave damping. The experimental FWCD was found to agree with predictions from the CURRAY ray-tracing code only when a parasitic loss of 4% per pass was included in the modeling along with multiple pass damping.

  1. Results from the DIII-D scientific research program

    SciTech Connect

    Taylor, T.S.; Burrell, K.H.; Baker, D.R.

    1998-11-01

    The DIII-D research program is aimed at developing the scientific basis for advanced modes of operation which can enhance the commercial attractiveness of the tokamak as an energy producing system. Features that improve the attractiveness of the tokamak as a fusion power plant include: high power density (which demands high {beta}), high ignition margin (high energy confinement time), and steady state operation with low recirculating power (high bootstrap fraction), as well as adequate divertor heat removal, particle and impurity control. This set of requirements emphasizes that the approach to improved performance must be an integrated approach, optimizing the plasma from the core, through the plasma edge and into the divertor. The authors have produced high performance ELMing H-mode plasmas with {beta}{sub N} H{sub 98y} {approximately} 6 for 5 {tau}{sub E} ({approximately}1 s) and demonstrated that core transport barriers can be sustained for the length of the 5-s neutral beam pulse in L-mode plasmas. They have demonstrated off-axis electron cyclotron current drive for the first time in a tokamak, discovering an efficiency above theoretical expectations. Edge stability studies have shown that the H-mode edge pressure gradient is not limited by ballooning modes; the self-consistent bootstrap provides second stable regime access. Divertor experiments have provided a new understanding of convection and recombination in radiative divertors and have produced enhanced divertor radiation with scrape off layer plasma flows and impurity enrichment.

  2. Electron cyclotron current drive experiments on DIII-D

    SciTech Connect

    James, R.A. ); Giruzzi, G.; Gentile, B. de; Rodriguez, L. ); Fyaretdinov, A.; Gorelov, Yu.; Trukhin, V. ); Harvey, R.; Lohr, J.; Luce, T.C.; Matsuda, K.; Politzer, P.; Prater, R.; Snider, R. (General Atomics, San Di

    1990-05-01

    Electron Cyclotron Current Drive (ECCD) experiments on the DIII-D tokamak have been performed using 60 GHz waves launched from the high field side of the torus. Preliminary analysis indicates rf driven currents between 50 and 100 kA in discharges with total plasma currents between 200 and 500 kA. These are the first ECCD experiments with strong first pass absorption, localized deposition of the rf power, and {tau}{sub E} much longer than the slowing-down time of the rf generated current carriers. The experimentally measured profiles for T{sub e}, {eta}{sub e} and Z{sub eff} are used as input for a 1D transport code and a multiply-ray, 3D ray tracing code. Comparisons with theory and assessment of the influence of the residual electric field, using a Fokker-Planck code, are in progress. The ECH power levels were between 1 and 1.5 MW with pulse lengths of about 500 msec. ECCD experiments worldwide are motivated by issues relating to the physics and technical advantages of the use of high frequency rf waves to drive localized currents. ECCD is accomplished by preferentially heating electrons moving in one toroidal direction, reducing their collisionality and thereby producing a non-inductively driven toroidal current. 6 refs., 4 figs.

  3. Measurement of neutral beam profiles at DIII-D

    SciTech Connect

    Chiu, H.

    1998-06-01

    The neutral beam systems of DIII-D, a National Fusion Facility at General Atomics, are used both for heating the plasma, and as tools for plasma diagnostics. The spatial distribution (profile) and energy of the beam is used in the absolute calibration of both the Charge Exchange Recombination (CER) and Motional Stark Effect (MSE) diagnostics. In the past, the beam spatial profile used in these calibrations was derived from beam divergence calculations and IR camera observations on the tokamak centerpost target tiles. Two experimental methods are now available to better determine the beam profile. In one method, the Doppler shifted D{sub {alpha}} light from the energetic neutrals are measured, and the full-width at half-maximum (FWHM) of the beam can be inferred from the measured divergence of the D{sub {alpha}} light intensity. The other method for determining the beam profile uses the temperature gradients measured by the thermocouples mounted on the calorimeter. A new iterative fitting routine for the measured thermocouple data has been developed to fit theoretical models on the dispersion of the beam. The results of both methods are compared, and used to provide a new experimental verification of the beam profile.

  4. An overview of the DIII-D program

    SciTech Connect

    Luxon, J.L.

    1996-10-01

    The DIII-D program focuses on developing fusion physics in an integrated program of tokamak concept improvement. The intent is both to support the present ITER physics R and D and to develop more efficient concepts for the later phases of ITER and eventual power plants. Progress in this effort can be best summarized by recent results for a diverted deuterium discharge with negative central shear which reached a performance level of Q{sub DT} = 0.32. The ongoing development of the tools needed to carry out this program of understanding and optimization continues to be crucial to its success. Control of the plasma cross-sectional shape and the internal distributions of plasma current, density, and rotation has been essential to optimizing plasma performance. Advanced divertor concepts provide edge power and particle control for future devices such as ITER and provide techniques to help manage the edge power and particle flows for advanced tokamak concepts. New divertor diagnostics and improved modeling are developing excellent divertor understanding. Many of the plasma physics issues being posed by ITER are being addressed. Scrapeoff layer power flow is being characterized to provide an accurate basis for the design of reactor devices. Ongoing studies of the density limit focus on identifying ways in which ITER can achieve the required densities in excess of the Greenwald limit. Better understanding of disruptions is crucial to the design of future reactors.

  5. Multivariable shape control development on the DIII-D tokamak

    SciTech Connect

    Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

    1997-11-01

    In this paper, the authors describe recent work on plasma shape and position control at DIII-D. This control consists of two equally challenging problems--the problem of identifying what the plasma actually looks like in real time, i.e. measuring the parameters to be controlled, and the task of determining the feedback algorithm which best controls these plasma parameters in a multiple input-output system. Recent implementation of the EFIT plasma equilibrium reconstruction algorithm in real time code which produces a new equilibrium estimate every 1.5 ms seems to solve the longstanding problem of obtaining sufficiently accurate plasma shape and position estimation. Stabilization of the open-loop unstable vertical motion is also viewed as a solved problem. The primary remaining problem appears to be how best to command the power supplies to achieve a desired shaping control response. They will describe the effort to understand and apply linearized models of plasma evolution to development and implementation of multivariable plasma controllers.

  6. Wide-angle Tangential Viewing System for DIII-D

    NASA Astrophysics Data System (ADS)

    Lasnier, C. J.; Allen, S. L.; Fenstermacher, M. E.; Hill, D. N.; Weber, T. R.

    2011-10-01

    We are designing a wide-angle tangential viewing system for DIII-D, with co-registered views in the visible and IR. We will examine toroidal and poloidal asymmetries of wall heating and particle flux during ELMs, magnetic perturbations, and disruptions; toroidal and poloidal mode structure of ELMs; poloidal distribution of particle flow velocities, and others. The system will simultaneously view the inner wall, outer wall, and upper and lower divertors, and will have an independent 3X optical zoom capability in visible and IR. Various parts of the image may be viewed at 3X magnification by translating the camera(s) vertically and laterally in the image plane. For IR we have a FLIR SC6000HS 3-5 μm camera, and for visible a Phantom V7.3. Both have high frame rate capability. Visible wavelength and neutral density filters may be selected, or interferometric flow measurement optics may be substituted for the filter system. This system was inspired by a design by CEA Cadarache for JET, and is similar to a system designed by LLNL for ITER upper ports. This work performed under the auspices of the US Department of Energy under ARRAY 2005290 and DE-AC52-07NA27344.

  7. Gyrokinetic simulations of microturbulence in DIII-D tokamak pedestal

    NASA Astrophysics Data System (ADS)

    Holod, Ihor; Fulton, Daniel; Taimourzadeh, Sam; Lin, Zhihong; Nazikian, Raffi; Spong, Donald

    2015-11-01

    The characteristics of H-mode pedestal are generally believed to be constrained by current-driven peeling-ballooning modes and pressure-driven instabilities, such as kinetic ballooning mode (KBM). In this work we use global gyrokinetic code (GTC) to identify and study the edge pressure-driven instabilities in the H-mode pedestal using realistic geometry and plasma profiles of DIII-D shot 131997. In our simulations we observe the KBM mode marginally dominant in the steep gradient region (ψN = 0 . 98), in the range of kθ ~ 1 cm-1 which corresponds to the most unstable mode number in the nonlinearly saturated state. For shorter wavelengths the trapped electron mode becomes dominant since its linear growth rate increases with the mode number, while the KBM gets saturated. In the pedestal top region (ψN = 0 . 95) the ITG dominates. Resonant magnetic perturbations (RMP) are widely applied for ELM mitigation. During RMP suppression, the increase of edge turbulence is often observed. To understand this phenomena we use gyrokinetic simulations to address the direct effect of magnetic perturbations on the microturbulence. Simulations with 3D equilibrium reconstructed by VMEC code have been compared with toroidally averaged equilibrium, using identical pressure profiles. Work supported by DOE grant DE-SC0010416 and by General Atomics subcontract.

  8. INTERMITTENT CONVECTION IN THE BOUNDARY OF DIII-D

    SciTech Connect

    J.A. BOEDO; D.L. RUDAKOV; R.J. COLCHIN; R.A. MOYER; S. KRASHENINNIKOV; D.G. WHYTE; G.R. McKEE; M.J. SCHAFFER; P.C. STANGEBY; W.P. WEST; S.L. ALLEN; A.W. LEONARD

    2002-06-01

    Intermittent plasma objects (IPOs) featuring higher pressure than the surrounding plasma, and responsible for {approx}50% of the E x B{sub T} radial transport, are observed in the scrape-off layer (SOL) and edge of the DIII-D tokamak. The skewness of probe and BES intermittent data suggest IPO formation at or near the last closed flux surface (LCFS) and the existence of hole-IPO pairs. The particle content of the IPOs at the LCFS is linearly dependent on the discharge density, however, when normalized to the local averaged density, it is fairly insensitive to density variations. It is also shown that the IPOs thermalize with the background plasma within 1 cm of the LCFS. The IPOs appear in the SOL of both L and H mode discharges carrying {approx}50% of the total SOL radial E x B{sub T} transport at all radii. However, the total flux and the IPO contribution, are highly reduced in H-mode conditions due to the increased confinement.

  9. ECE Imaging of Broadband Turbulence in DIII-D Plasmas

    NASA Astrophysics Data System (ADS)

    Zemedkun, S. E.; Munsat, T.; Tobias, B. J.; Donier, C. W.; Luhmann, N. C., Jr.

    2013-10-01

    Observations of 2D turbulent structures have been performed with the ECEI instrument on DIII-D in plasmas heated by neutral beam injection (NBI) and electron cyclotron heating (ECH), at a fixed heating power (up to 5 MW). Correlation techniques similar to those used in correlation electron cyclotron emission (CECE) systems are employed, with the advantage that the ECEI system detects a full 2D array of plasma locations; vertical separation is provided by an optical system and horizontal separation is provided by frequency discrimination in the detection electronics. Among the results are 2D images of poloidally-propagating drift waves, and correlation properties of fluctuations (<200 kHz) in the poloidal direction. Observed dispersion relations for two different heating conditions (ECH and NBI) will be presented. Comparison of results with simulations using GEM code will be discussed. In addition to the physics results, the data demonstrates the viability of the ECEI system in the presence of ECH heating. Work supported by the US Department of Energy under DE-FC02-05ER54816, DE-SC0003913, DE-FC02-04ER54698, DE-AC02-09CH11466, and DE-FG02-99ER54531.

  10. Overview of Recent DIII-D Experimental Results

    NASA Astrophysics Data System (ADS)

    Fenstermacher, M. E.; DIII-D Team

    2014-10-01

    Recent DIII-D experiments have added to the ITER physics basis and to physics understanding for extrapolation to future devices. Physics mechanisms contributing to resonant magnetic perturbation ELM suppression and QH-mode were identified. The QH-mode operating space was extended to ITER-relevant parameters and predicted Super-H mode performance was observed at high shaping. Upgraded divertor Thomson data was combined with edge modeling to identify the core density limit at divertor detachment. Pedestal studies were done to determine the role of ν*, Zeff and kinetic ballooning mode instabilities in controlling pedestal structure. Injection of massive high-Z gas dissipates magnetic and kinetic energy of runaway electron beams. 3D magnetics data validate several linear MHD codes, including ability to predict neoclassical tearing viscosity torque. Feedback control of applied 3D fields facilitates access to increased βN values above the no-wall limit. The effect of test blanket module (TBM) fields on fast ion losses and momentum transport, and partial correction of TBM fields at high β was achieved. Density gradient driven trapped electron modes and core ne peaking were controlled by electron cyclotron heating suggesting a possible burn control technique. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  11. Central Thomson scattering upgrade on DIII-D

    SciTech Connect

    Nilson, D.G.; Stallard, B.W.; Carlstrom, T.N.; Hsieh, C.L.; Stockdale, R.E.

    1997-11-01

    The existing 36 channel Thomson scattering system on the DIII-D tokamak measures the plasma temperature and density in the core region. However, with the recent interest in core peaked density profiles, coverage needs to be extended into the magnetic axis. This paper addresses the technical issues involved with extending the viewing region from a major radius of 194 cm to 165 cm. At least one of the existing seven core laser beams will be rerouted to probe the plasma horizontally instead of vertically. To do this, a rigid extension of the existing laser/collection optics tower will be built to route the laser to a nearby tangential port. A fiber bundle array from one of the two existing core plasma collection optics sets will be rotated to allow up to 10 of the 36 core channels to view along this new beam path. A new in-vessel absorbing glass laser dump must be developed since there are no appropriate laser beam exit ports. The close proximity of this laser dump to the viewing region presents stray light issues that must be resolved to allow for an accurate density calibration using Rayleigh scattering in argon gas.

  12. ELM-Induced Plasma Wall Interactions in DIII-D

    SciTech Connect

    Rudakov, D L; Boedo, J A; Yu, J H; Brooks, N H; Fenstermacher, M E; Groth, M; Hollmann, E M; Lasnier, C J; McLean, A G; Moyer, R A; Stangeby, P C; Tynan, G R; Wampler, W R; Watkins, J G; West, W P; Wong, C C; Zeng, L; Bastasz, R J; Buchenauer, D; Whaley, J

    2008-05-14

    Intense transient fluxes of particles and heat to the main chamber components induced by edge localized modes (ELMs) are of serious concern for ITER. In DIII-D, plasma interaction with the outboard chamber wall is studied using Langmuir probes and optical diagnostics including a fast framing camera. Camera data shows that ELMs feature helical filamentary structures localized at the low field side of the plasma and aligned with the local magnetic field. During the nonlinear phase of an ELM, multiple filaments are ejected from the plasma edge and propagate towards the outboard wall with velocities of 0.5-0.7 km/s. When reaching the wall, filaments result in 'hot spots'--regions of local intense plasma-material interaction (PMI) where the peak incident particle and heat fluxes are up to 2 orders of magnitude higher than those between ELMs. This interaction pattern has a complicated geometry and is neither toroidally nor poloidally symmetric. In low density/collisionality H-mode discharges, PMI at the outboard wall is almost entirely due to ELMs. In high density/collisionality discharges, contributions of ELMs and inter-ELM periods to PMI at the wall are comparable. A Midplane Material Evaluation Station (MiMES) has been recently installed in order to conduct in situ measurements of erosion/redeposition at the outboard chamber wall, including those caused by ELMs.

  13. Gamma ray imager on the DIII-D tokamak.

    PubMed

    Pace, D C; Cooper, C M; Taussig, D; Eidietis, N W; Hollmann, E M; Riso, V; Van Zeeland, M A; Watkins, M

    2016-04-01

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1-60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons. PMID:27131674

  14. DIII-D Neutral Beam control system operator interface

    SciTech Connect

    Harris, J.J.; Campbell, G.L.

    1993-10-01

    A centralized graphical user interface has been added to the DIII-D Neutral Beam (NB) control systems for status monitoring and remote control applications. This user interface provides for automatic data acquisition, alarm detection and supervisory control of the four NB programmable logic controllers (PLC) as well as the Mode Control PLC. These PLCs are used for interlocking, control and status of the NB vacuum pumping, gas delivery, and water cooling systems as well as beam mode status and control. The system allows for both a friendly user interface as well as a safe and convenient method of communicating with remote hardware that formerly required interns to access. In the future, to enable high level of control of PLC subsystems, complete procedures is written and executed at the touch of a screen control panel button. The system consists of an IBM compatible 486 computer running the FIX DMACS{trademark} for Windows{trademark} data acquisition and control interface software, a Texas Instruments/Siemens communication card and Phoenix Digital optical communications modules. Communication is achieved via the TIWAY (Texas Instruments protocol link utilizing both fiber optic communications and a copper local area network (LAN). Hardware and software capabilities will be reviewed. Data and alarm reporting, extended monitoring and control capabilities will also be discussed.

  15. Gamma ray imager on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Pace, D. C.; Cooper, C. M.; Taussig, D.; Eidietis, N. W.; Hollmann, E. M.; Riso, V.; Van Zeeland, M. A.; Watkins, M.

    2016-04-01

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1-60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.

  16. X--Point Neutral Density Determination in DIII-D

    NASA Astrophysics Data System (ADS)

    Colchin, R. J.; Maingi, R.; Isler, R. C.; Owen, L. W.; Fenstermacher, M. E.; Carlstrom, T. N.

    1998-11-01

    A method has been developed to measure neutral densities in the divertor and X-point regions of diverted DIII-D plasmas. The method calibrates D_α light intensities from a tangentially-viewing video camera by means of a vertically-viewing photomultiplier. D_α light viewed by the video camera is reconstructed onto a poloidal plane in the region of the lower divertor. Neutral densities are obtained by dividing the D_α light intensities by the electron density and the electron excitation rate coefficients. These coefficients are steep functions of the electron temperature and density, which are measured by the divertor Thomson scattering diagnostic. Data have been analyzed for L-mode plasmas just below the L--H transition at two X-point heights. The neutral density decreases in the vertical direction from 10^13 atoms/cm^3 just above the divertor floor in the private flux region to 10^11 atoms/cm^3 at the X-point. The neutral density just inside the X-point is in the range 10^10--10^11 atoms/cm^3.

  17. SOLPS Modeling of Slot Divertor Configuration on DIII-D

    NASA Astrophysics Data System (ADS)

    Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; Lao, L. L.

    2015-11-01

    A major thrust of the DIII-D boundary/PMI initiative is to develop an advanced divertor configuration for next-step devices, such as FNSF and DEMO. We are adopting an integrated approach by optimizing both divertor structure and magnetic shape. Initial SOLPS modeling was carried out to optimize divertor structure shape to enhance divertor power dissipation, focusing on slot configurations. In particular, four different slot divertor structures, i.e., orthogonal-target slot, slanted-target slot, very narrow slot and v-shaped slot have been analyzed and comparisons made with an open divertor structure. It is found that the slot helps to trap recycling neutrals and impurities thus increasing radiative power dissipation in the divertor, reducing the electron temperature Te and the perpendicular heat flux q⊥ at the target plate. As expected, a narrower slot leads to lower Te and q⊥ than a less narrow one. The v-shaped slot appears to be especially effective at redirecting and concentrating recycling neutrals and impurities near the separatrix, thus promoting detachment at a lower upstream density than the other configurations. Work supported by US DOE under DE-FC02-04ER54698.

  18. The DIII-D 3 MW, 110 GHz ECH System

    SciTech Connect

    Callis, R.W.; Lohr, J.; Ponce, D.; O'Neill, R.C.; Prater, R.; Luce, T.C.

    1999-07-01

    Three 110 GHz gyrotrons with nominal output power of 1 MW each have been installed and are operational on the DIII-D tokamak. One gyrotron is built by Gycom and has a nominal rating of 1 MW and a 2 s pulse length, with the pulse length being determined by the maximum temperature allowed on the edge cooled Boron Nitride window. The second and third gyrotrons were built by Communications and Power Industries (CPI). The first CPI gyrotron uses a double disc FC-75 cooled sapphire window which has a pulse length rating of 0.8 s at 1 MW, 2s at 0.5 MW and 10s at 0.35 MW. The second CPI gyrotron, utilizes a single disc chemical-vapor-deposition diamond window, that employs water cooling around the edge of the disc. Calculation predict that the diamond window should be capable of full 1 MW cw operation. All gyrotrons are connected to the tokamak by a low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HEl 1 mode. Each waveguide system incorporates a two mirror launcher which can steer the rf beam poloidally from the center to the outer edge of the plasma. Central current drive experiments with the two gyrotrons with 1.5 MW of injected power drove about 0.17 MA. Results from using the three gyrotron systems will be reported as well as the plans to upgrade the system to 6 MW.

  19. Operational Performance of the ECH System on DIII-D

    NASA Astrophysics Data System (ADS)

    Cengher, M.; Lohr, J.; Gorelov, Y. A.; Ponce, D.; Moeller, C. P.

    2012-10-01

    The measurement of the rf power in the ECH system on DIII-D is showing the history of the performance for the six 110 GHz, 1 MW class gyrotrons. Four of the six systems show a general trend to higher values for the power injected in the tokamak after improvement of the transmission line, while for the other two systems the lower injected power is explained by operation at lower input power for reliability. The power calibration is based on the measured linearity of the injected power with the gyrotron cavity loading for all 6 systems. Total collector loading was measured versus the beam voltage. The measured transmission loss for 4 of the transmission lines is less than 1.1 dB, close to the theoretical value. The HE11 mode content is over 85% for all the lines. An average gain of 0.035 in the total transmission coefficient in the lines is due to a reduced number of miter bends in the system, reduced waveguide run, and improved angular alignment of the rf beam at the waveguide input. Measurements using a 4-port monitor and a dummy load have shown that the maximum power transmitted to a load corresponds to a maximum in the HE11 mode.

  20. DIII-D contributions towards the scientific basis for sustained burning plasmas

    NASA Astrophysics Data System (ADS)

    Greenfield, C. M.; DIII-D Team

    2011-09-01

    DIII-D is making significant contributions to a scientific basis for sustained burning plasma operation. These include explorations of increasingly reactor-relevant scenarios, studies of key issues for projecting performance, development of techniques for handling heat and particle efflux, and assessment of key issues for the ITER research plan. Advanced scenarios are being optimized in DIII-D via experiments to empirically determine the relationship between transport and the current profile, which in turn can provide essential input to inform improvement of the theory-based models that do not currently capture the observed behaviour. Joint DIII-D/JET ρ* scans in the hybrid regime imply Bohm-like confinement scaling. Startup and shutdown techniques were developed for the restrictive environment of future devices while retaining compatibility with advanced scenarios. Towards the goal of a fully predictive capability, the DIII-D program emphasizes validation of physics-based models, facilitated by a number of new and upgraded diagnostics. Specific areas include transport, rotation, energetic particles and the H-mode pedestal, but this approach permeates the entire research programme. Concerns for heat and particle efflux in future devices are addressed through studies of ELM control, disruption avoidance and mitigation, and hydrogenic retention in DIII-D's carbon wall. DIII-D continues to respond to specific needs for ITER. Recent studies have compared H-mode access in several different ion species, identifying not only isotopic, but density, rotation and geometrical dependences that may guide access to H-mode during ITER's non-activated early operation. DIII-D used an insertable module to simulate the magnetic perturbations introduced by one of ITER's three test blanket module sets, demonstrating that little impact on performance is seen at ITER equivalent levels of magnetic perturbation.

  1. DIII-D research operations. Annual report, October 1, 1992--September 30, 1993

    SciTech Connect

    La Haye, R.J.

    1994-05-01

    The DIII-D tokamak research program is carried out by General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. In doing so, the DIII-D program provides physics and technology R&D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma.

  2. Advanced tokamak physics experiments on DIII-D

    SciTech Connect

    Taylor, T.S.

    1998-12-01

    Significant reductions in the size and cost of a fusion power plant core can be realized if simultaneous improvements in the energy confinement time ({tau}{sub E}) and the plasma pressure (or beta {beta}{sub T} = 2 {mu}{sub 0} < p > /B{sub T}{sup 2}) can be achieved in steady-state conditions with high self driven bootstrap current fraction. In addition, effective power exhaust and impurity and particle control is required. Significant progress has been made in experimentally achieving regimes having the required performance in all of these aspects as well as in developing a theoretical understanding of the underlying physics. The authors have extended the duration of high performance ELMing H-mode plasmas with {beta}{sub N} H{sub iop} {approximately} 10 for 5 {tau}{sub E} ({approximately}1 s) and have demonstrated that core transport barriers can be sustained for the entire 5-s neutral beam duration in L-mode plasmas. Recent DIII-D work has advanced the understanding of improved confinement and internal transport barriers in terms of E x B shear stabilization of micro turbulence. With the aim of current profile control in discharges with negative central magnetic shear, they have demonstrated off-axis electron cyclotron current drive for the first time in a tokamak, finding an efficiency above theoretical expectations. MHD stability has been improved through shape optimization, wall stabilization, and modification of the pressure and current density profiles. Heat flux reduction and improved impurity and particle control have been realized through edge/divertor radiation and understanding and utilization of forced scrape off layer flow and divertor baffling.

  3. The DIII-D 3 MW, 110 GHz ECH system

    SciTech Connect

    Callis, R. W.; Lohr, J.; Ponce, D.; O'Neill, R. C.; Prater, R.; Luce, T. C.

    1999-09-20

    Three 110 GHz gyrotrons with nominal output power of 1 MW each have been installed and are operational on the DIII-D tokamak. One gyrotron is built by Gycom and has a nominal rating of 1 MW and a 2 s pulse length, with the pulse length being determined by the maximum temperature allowed on the edge cooled Boron Nitride window. The second and third gyrotrons were built by Communications and Power Industries (CPI). The first CPI gyrotron uses a double disc FC-75 cooled sapphire window which has a pulse length rating of 0.8 s at 1 MW, 2 s at 0.5 MW and 10 s at 0.35 MW. The second CPI gyrotron, utilizes a single disc chemical-vapor-deposition diamond window, that employs water cooling around the edge of the disc. Calculation predict that the diamond window should be capable of full 1 MW cw operation. All gyrotrons are connected to the tokamak by a low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HE{sub 11} mode. Each waveguide system incorporates a two mirror launcher which can steer the rf beam poloidally from the center to the outer edge of the plasma. Central current drive experiments with the two gyrotrons with 1.5 MW of injected power drove about 0.17 MA. Results from using the three gyrotron systems will be reported as well as the plans to upgrade the system to 6 MW. (c) 1999 American Institute of Physics.

  4. A Fast Visible Camera Divertor-Imaging Diagnostic on DIII-D

    SciTech Connect

    Roquemore, A; Maingi, R; Lasnier, C; Nishino, N; Evans, T; Fenstermacher, M; Nagy, A

    2007-06-19

    In recent campaigns, the Photron Ultima SE fast framing camera has proven to be a powerful diagnostic when applied to imaging divertor phenomena on the National Spherical Torus Experiment (NSTX). Active areas of NSTX divertor research addressed with the fast camera include identification of types of EDGE Localized Modes (ELMs)[1], dust migration, impurity behavior and a number of phenomena related to turbulence. To compare such edge and divertor phenomena in low and high aspect ratio plasmas, a multi-institutional collaboration was developed for fast visible imaging on NSTX and DIII-D. More specifically, the collaboration was proposed to compare the NSTX small type V ELM regime [2] and the residual ELMs observed during Type I ELM suppression with external magnetic perturbations on DIII-D[3]. As part of the collaboration effort, the Photron camera was installed recently on DIII-D with a tangential view similar to the view implemented on NSTX, enabling a direct comparison between the two machines. The rapid implementation was facilitated by utilization of the existing optics that coupled the visible spectral output from the divertor vacuum ultraviolet UVTV system, which has a view similar to the view developed for the divertor tangential TV camera [4]. A remote controlled filter wheel was implemented, as was the radiation shield required for the DIII-D installation. The installation and initial operation of the camera are described in this paper, and the first images from the DIII-D divertor are presented.

  5. Divertor IR thermography on Alcator C-Moda)

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  7. GYRO Simulations of Core Momentum Transport in DIII-D and JET Plasmas

    SciTech Connect

    R.V. Budny; J. Candy; R.E. Waltz; and contributors to the DIII-D and JET-EFDA work programs

    2005-06-27

    Momentum, energy, and particle transport in DIII-D and JET ELMy H-mode plasmas is simulated with GYRO and compared with measurements analyzed using TRANSP. The simulated transport depends sensitively on the nabla(T(sub)i) turbulence drive and the nabla(E(sub)r) turbulence suppression inputs. With their nominal values indicated by measurements, the simulations over-predict the momentum and energy transport in the DIII-D plasmas, and under-predict in the JET plasmas. Reducing |nabla(T(sub)i)| and increasing |nabla(E(sub)r)| by up to 15% leads to approximate agreement (within a factor of two) for the DIII-D cases. For the JET cases, increasing |nabla(T(sub)i)| or reducing |nabla(E(sub)r)| results in approximate agreement for the energy flow, but the ratio of the simulated energy and momentum flows remains higher than measurements by a factor of 2-4.

  8. Divertor heat and particle control experiments on the DIII-D tokamak

    SciTech Connect

    Mahdavi, M.A; Baker, D.R.; Allen, S.L.

    1994-05-01

    In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D{sub 2} gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models.

  9. Performance of V-4Cr-4Ti material exposed to DIII-D tokamak environment

    SciTech Connect

    Tsai, H.; Chung, H.M.; Smith, D.L.

    1997-04-01

    Test specimens made with the 832665 heat of V-4Cr-4Ti alloy were exposed in the DIII-D tokamak environment to support the installation of components made of a V-4Cr-4Ti alloy in the radiative divertor of the DIII-D. Some of the tests were conducted with the Divertor Materials Evaluation System (DiMES) to study the short-term effects of postvent bakeout, when concentrations of gaseous impurities in the DIII-D chamber are the highest. Other specimens were mounted next to the chamber wall behind the divertor baffle plate, to study the effects of longer-term exposures. By design, none of the specimens directly interacted with the plasma. Preliminary results from testing the exposed specimens indicate only minor degradation of mechanical properties. Additional testing and microstructural characterization are in progress.

  10. ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

    SciTech Connect

    HUMPHREYS,DA; FERRON,JR; GAROFALO,AM; HYATT,AW; JERNIGAN,TC; JOHNSON,RD; LAHAYE,RJ; LEUER,JA; OKABAYASHI,M; PENAFLOR,BG; SCOVILLE,JT; STRAIT,EJ; WALKER,ML; WHYTE,DG

    2002-10-01

    A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response.

  11. Development of Burning Plasma and Advanced Scenarios in the DIII-D Tokamak

    SciTech Connect

    Luce, T C

    2004-10-18

    Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q {approx} 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque.

  12. Development of Burning Plasma and Advanced Scenarios in the DIII-D Tokamak

    SciTech Connect

    Luce, T C

    2004-12-01

    Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q {approx} 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque.

  13. Disruption and Runaway Electron Mitigation With MGI in DIII-D

    NASA Astrophysics Data System (ADS)

    Wesley, J. C.; Humphreys, D. A.; Parks, P. B.; Strait, E. J.; Hollmann, E. M.; Antar, G.; Jernigan, T. C.; Combs, S. K.; Groth, M.

    2006-10-01

    Past and ongoing disruption mitigation studies in DIII-D employing massive gas injection (MGI) are reviewed and compared with theoretical expectations. Emphasis in the review will be placed on 1) the gas hydrodynamic delivery considerations that determine the rate of impurity and electron delivery to the plasma edge, 2) the role of MHD instability and internal reconnection in effecting edge-to-core mixing of the edge-deposited impurities, and 3) assessment of the mechanism(s) whereby MGI mitigates divertor energy deposition, reduces halo current magnitude and asymmetry and avoids runaway electron production and/or Coulomb-avalanche multiplication. Selected considerations for application of DIII-D MGI results to ITER (wherein time scales for impurity delivery are relaxed relative to DIII-D and other present experiments) will also be addressed.

  14. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1990--September 30, 1991

    SciTech Connect

    Simonen, T.C.; Evans, T.E.

    1992-03-01

    This report discusses the following topics on Doublet-3 research operations: DIII-D Program Overview; Boundary Plasma Research Program/Scientific Progress; Radio Frequency Heating and Current Drive; Core Physics; DIII-D Operations; Program Development; Support Services; ITER Contributions; Burning Plasma Experiment Contributions; and Collaborative Efforts.

  15. Edge Currents and Stability in DIII-D

    SciTech Connect

    Thomas, D M; Fenstermacher, M E; Finkenthal, D K; Groebner, R J; Lao, L L; Leonard, A W; Mueller, H W; Osborne, T H; Snyder, P B

    2004-12-01

    Understanding the stability physics of the H-mode pedestal in tokamak devices requires an accurate measurement of plasma current in the pedestal region with good spatial resolution. Theoretically, the high pressure gradients achieved in the edge of H-mode plasmas should lead to generation of a significant edge current density peak through bootstrap and Pfirsh-Schl{umlt u}ter effects. This edge current is important for the achievement of second stability in the context of coupled magneto hydrodynamic (MHD) modes which are both pressure (ballooning) and current (peeling) driven. Many aspects of edge localized mode (ELM) behavior can be accounted for in terms of an edge current density peak, with the identification of Type 1 ELMs as intermediate-n toroidal mode number MHD modes being a natural feature of this model. The development of a edge localized instabilities in tokamak experiments code (ELITE) based on this model allows one to efficiently calculate the stability and growth of the relevant modes for a broad range of plasma parameters and thus provides a framework for understanding the limits on pedestal height. This however requires an accurate assessment of the edge current. While estimates of j{sub edge} can be made based on specific bootstrap models, their validity may be limited in the edge (gradient scalelengths comparable to orbit size, large changes in collisionality, etc.). Therefore it is highly desirable to have an actual measurement. Such measurements have been made on the DIII-D tokamak using combined polarimetry and spectroscopy of an injected lithium beam. By analyzing one of the Zeeman-split 2S-2P lithium resonance line components, one can obtain direct information on the local magnetic field components. These values allow one to infer details of the edge current density. Because of the negligible Stark mixing of the relevant atomic levels in lithium, this method of determining j(r) is insensitive to the large local electric fields typically found

  16. Performance of V-4Cr-4Ti material exposed to the DIII-D tokamak environment

    SciTech Connect

    Tsai, H.; Smith, D.L.; Chung, H.M.; Johnson, W.R.; Smith, J.P.; Wampler, W.R.

    1998-03-01

    A series of tests is being conducted in the DIII-D tokamak to determine the effects of environmental exposure on a V-4Cr-4Ti vanadium alloy. These tests are part of the effort to build and install a water-cooled vanadium V alloy structure in the upgrade of the DIII-D radiative divertor. Data from the test series indicate that the performance of the V-4Cr-4Ti alloy would not be significantly affected by environmental exposure. Interstitial absorption by the material appears to be limited to the surface, and neither the tensile nor the impact properties of the material appear to be affected by the exposure.

  17. ELECTRON CYCLOTRON CURRENT DRIVE IN DIII-D: EXPERIMENT AND THEORY

    SciTech Connect

    PRATER,R; PETTY,CC; LUCE,TC; HARVEY,RW; CHOI,M; LAHAYE,RJ; LIN-LIU,Y-R; LOHR,J; MURAKAMI,M; WADE,MR; WONG,K-L

    2003-07-01

    A271 ELECTRON CYCLOTRON CURRENT DRIVE IN DIII-D: EXPERIMENT AND THEORY. Experiments on the DIII-D tokamak in which the measured off-axis electron cyclotron current drive has been compared systematically to theory over a broad range of parameters have shown that the Fokker-Planck code CQL3D provides an excellent model of the relevant current drive physics. This physics understanding has been critical in optimizing the application of ECCD to high performance discharges, supporting such applications as suppression of neoclassical tearing modes and control and sustainment of the current profile.

  18. Soft X-Ray Imaging Design and Analysis Methods on DIII-D

    SciTech Connect

    Shafer, Morgan W; Battaglia, D. J.; Unterberg, Ezekial A; Canik, John; Evans, T. E.; Harris, Jeffrey H; Meitner, Steven J

    2011-01-01

    A new tangential 2D Soft X-Ray Imaging System (SXRIS) is being designed to examine the edge magnetic island structure in the lower X-point region of DIII-D. A synthetic diagnostic calculation coupled to 3D emissivity estimates is used to generate phantom images. Phillips-Tikhonov regularization is used to invert the phantom images for comparison to the original emissivity model. Noise level, island size, and equilibrium accuracy are scanned to assess the feasibility of detecting edge island structures. Models of typical DIII-D discharges indicate integration times > 1 ms with accurate equilibrium reconstruction are needed for small island (< 3 cm) detection.

  19. Study of Aspect Ratio Effects on Kinetic MHD Instabilities in NSTX and DIII-D

    SciTech Connect

    E.D. Fredrickson; W.W. Heidbrink; C.Z. Cheng; N.N. Gorelenkov; E. Belova; A.W. Hyatt; G.J. Kramer; J. Manickam; J. Menard; R. Nazikian; T.L. Rhodes; E. Ruskov

    2004-10-21

    We report general observations of kinetic instabilities on the low aspect-ratio National Spherical Torus Experiment (NSTX) and describe explicit aspect ratio scaling studies of kinetic instabilities using both the NSTX and the DIII-D tokamak. The NSTX and the DIII-D tokamak are nearly ideal for such experiments, having a factor of two difference in major radius but otherwise similar parameters. We also introduce new theoretical work on the physics of kinetic ballooning modes (KBM), toroidal Alfven eigenmodes (TAE), and compressional Alfven eigenmodes (CAE) with applications to NSTX.

  20. Applications of ECH on the DIII-D tokamak and projections for future ECH upgrades

    NASA Astrophysics Data System (ADS)

    Prater, R.; Buttery, R. J.; DeBoo, J.; Ferron, J. R.; Garofalo, A.; Holcomb, C. T.; Jackson, G. L.; La Haye, R. J.; Lohr, J. M.; Luce, T. C.; Petty, C. C.; Politzer, P. A.; Solomon, W. M.; Turco, F.

    2012-09-01

    Electron Cyclotron Heating and Current Drive plays an important role in the DIII-D program. In high performance discharges EC power contributes greatly to MHD stability, and this is particularly important for discharges with low rotational torque applied, as will be the case for ITER. Off-axis EC current drive also plays a key role in the actualization of steady-state scenarios by supporting the desired current profile. In order to carry out these applications at higher beta and higher field, an upgrade of the EC power to 15 MW is needed, and the best gyrotron frequency for the DIII-D program is 117.5 GHz.

  1. Studies of EDA H-mode in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Greenwald, M.; Boivin, R.; Bonoli, P.; Fiore, C.; Goetz, J.; Granetz, R.; Hubbard, A.; Hutchinson, I.; Irby, J.; Lin, Y.; Marmar, E.; Mazurenko, A.; Mossessian, D.; Pedersen, T. Sunn; Rice, J.; Snipes, J.; Schilling, G.; Taylor, G.; Greenwald, M.; Boivin, R.; Bonoli, P.; Fiore, C.; Goetz, J.; Granetz, R.; Hubbard, A.; Hutchinson, I.; Irby, J.; Lin, Y.; Marmar, E.; Mazurenko, A.; Mossessian, D.; Pedersen, T. Sunn; Rice, J.; Snipes, J.; Terry, J.; Wolfe, S.; Wukitch, S.

    2000-05-01

    Studies of the enhanced Dα H-mode (EDA) have been extended to include ohmic plasmas. No clear difference in the EDA/ELMfree boundary or in other phenomenology are seen between ohmic and ICRF-heated plasmas, suggesting that neither the effect of ion tails nor direct RF/edge plasma interaction plays a role in EDA. Edge safety factor (q95) is the principal variable which determines which regime a discharge will be in. When q95 is greater than 4.0 for standard-shaped plasmas, the discharge is almost always EDA, while when it is less than 3.5, the plasma is almost always ELMfree. New edge diagnostics have allowed measurement of pedestal profiles with resolution of the order of 1 mm. Sudden changes in profile widths are not seen when the plasma makes a transition from EDA to ELMfree; however, the widths do vary with the same parameters that determine the EDA/ELMfree boundary. Strong edge-density fluctuations are observed to accompany EDA and may be responsible for the change in particle transport which is observed. The fluctuations have a quasi-coherent component whose frequency varies inversely with the pedestal width as measured by a visible continuum diagnostic.

  2. Studies of EDA H-MODE in Alcator C-MOD

    NASA Astrophysics Data System (ADS)

    Greenwald, M.; Hubbard, A.; Snipes, J.; Boivin, R. L.; Granetz, R.; Hughes, J.; Hutchinson, I.; Irby, J.; Lin, Y.; Marmar, E.; Mazurenko, A.; Mossessian, D.; Pedersen, T.; Rice, J.; Terry, J.; Wolfe, S.

    1999-11-01

    The advantages of EDA H-modes include good energy confinement and no impurity accumulation or large ELMs. In EDA, the edge pressure gradients are at or above the ideal ballooning limit but are not relaxed by type I ELMs; instead a continuous process would seem to be at work. This process is probably related to broadband and quasi-coherent fluctuations which are seen with reflectometry, PCI and magnetic pick-up loops. The coherent component has a frequency on the order of 100 kHz in steady state but shifts to much higher frequencies as the plasma transitions to or from ELMfree. We note that the Doppler shift in the shear layer may make an important contribution to the lab frame frequency. The coherent feature is dominant in the PCI measurements, which show a well defined spatial structure with kR ~ k_θ ~ 6 cm-1. Recent simulations of drift Alfven turbulence(B. Rogers, J. Drake, submitted to Phys. Plasmas (1999)) show a coherent surface mode arise as the pedestal pressure gradient approaches an MHD stability limit.

  3. Observation of ICRF Mode Conversion Plasma Flow Drive on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lin, Yijun

    2008-11-01

    Plasma flow driven by externally launched rf waves could be important in stabilizing micro- and macro-instabilities in tokamaks. We report the first observation of both toroidal (Vφ) and poloidal (Vθ) flows driven via an ICRF mode conversion (MC) process in D(^3He) plasmas. At modest ^3He levels (n3He/ne˜ 8%), in relatively low density plasmas, <= 1.3x10^20m-3, heated with 50 MHz rf power (Bt0˜ 5.1 T), strong Vφ in the co-current direction is observed by high-resolution x-ray spectroscopy. The central Vφ scales with the applied rf power (<= 30 km/s per MW), and is at least a factor of 2 more than the empirically determined intrinsic plasma rotation [1]. The rotation near the plasma center (r/a < 0.3) responds more quickly to the applied rf power than the outer region, indicative of a local flow drive source. Localized poloidal rotation (0.3 <= r/a <= 0.5) in the ion diamagnetic drift direction is observed when Prf >= 1.5 MW and increases with power (˜ 2 km/s at 3 MW). Turbulence spectrum broadening seen by a phase contrast imaging (PCI) system indicates strong flow also exists in the main ions. The mode converted ion cyclotron wave (MC ICW) is observed by PCI and confirmed by 2-D full wave TORIC code simulation. The simulation result shows that due to the up-shifted k||,the MC ICW is strongly damped on ^3He ions in the vicinity of the MC layer, approximately on the same flux surfaces where poloidal flow is observed. The involvement of ion heating and short-wavelength slow wave is consistent with theoretical considerations for efficient rf flow drive. Our experimental results are comparable to the predictions [2], assuming similar ion interaction mechanism for the MC ICW and direct launch ion Bernstein wave. The feasibility of ICRF flow drive on ITER will be discussed. [1] J. E. Rice, et al, Nucl. Fusion 47, 1618 (2007). [2] J. R. Myra and D. A. D'Ippolito, Phys. Plasmas 9, 3867 (2002).

  4. Scaling of the power exhaust channel in Alcator C-Mod

    SciTech Connect

    LaBombard, B.; Terry, J. L.; Hughes, J. W.; Brunner, D.; Payne, J.; Reinke, M. L.; Cziegler, I.; Granetz, R.; Greenwald, M.; Hutchinson, I. H.; Irby, J.; Lin, Y.; Lipschultz, B.; Ma, Y.; Marmar, E. S.; Tsujii, N.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.

    2011-05-15

    Parametric dependences of the heat flux footprint on the outer divertor target plate are explored in EDA H-mode and ohmic L-mode plasmas over a wide range of parameters with attached plasma conditions. Heat flux profile shapes are found to be independent of toroidal field strength, independent of power flow along magnetic field lines and insensitive to x-point topology (single-null versus double-null). The magnitudes and widths closely follow that of the ''upstream'' pressure profile, which are correlated to plasma thermal energy content and plasma current. Heat flux decay lengths near the strike-point in H- and L-mode plasmas scale approximately with the inverse of plasma current, with a diminished dependence at high collisionality in L-mode. Consistent with previous studies, pressure gradients in the boundary scale with plasma current squared, holding the magnetohydrodynamic ballooning parameter approximately invariant at fixed collisionality--strong evidence that critical-gradient transport physics plays a key role in setting the power exhaust channel.

  5. Optical and Mechanical Design of C-Mod Motional Stark Effect Diagnostic

    SciTech Connect

    D.I. Simon; E. Marmar; N.L. Bretz; R. Bravenec; R.F. Parsells

    1999-11-01

    A Motional Stark Effect (MSE) instrument is being installed on the Alcator C-Mod tokamak at MIT. This MSE diagnostic will provide measurements of the spatial profile of the internal poloidal magnetic field. The MSE has its primary collection optics inside the vacuum vessel. The light collected by the internal optics passes through a vacuum window and is relayed to a fiber optic array. The MSE optics are shared by a Beam Emission Spectroscopy (BES) diagnostic which measures electron density fluctuations and their spatial correlations. This optical system requires high throughput and spatial resolution of less than 1 cm at the focal plane in the plasma. The design requirements for the internal optics also include the effects associated with plasma impingement, plasma disruptions, and thermal excursions. The parameters that affect polarization measurement include the location and orientation of optical elements, the choice of substrates and optical materials. These unique design requirements led to a number of interesting optical and mechanical design features which are presented here.

  6. High temperature outgassing tests on materials used in the DIII-D tokamak

    SciTech Connect

    Holtrop, K.L.; Hansink, M.J.

    2006-07-15

    This article is a continuation of previous work on determining the outgassing characteristics of materials used in the DIII-D magnetic fusion tokamak [K. L. Holtrop, J. Vac. Sci. Technol. A 17, 2064 (1999)]. Achievement of high performance plasma discharges in the DIII-D tokamak requires careful control of impurity levels. Among the techniques used to control impurities are routine bakes of the vacuum vessel to an average temperature of 350 deg. C. Materials used in DIII-D must release only very small amounts of impurities (below 2x10{sup -6} mole) at this temperature that could be transferred to the first wall materials and later contaminate plasma discharges. To better study the behavior of materials proposed for use in DIII-D at elevated temperatures, the initial outgassing test chamber was improved to include an independent heating control of the sample and a simple load lock chamber. The goal was to determine not only the total degassing rate of the material during baking, but to also determine the gas species composition and to obtain a quantitative estimate of the degassing rate of each species by the use of a residual gas analyzer. Initial results for aluminum anodized using three different processes, stainless steel plated with black oxide and black chrome, and a commercially available fiber optic feedthrough will be presented.

  7. Modeling of electron cyclotron current drive experiments on DIII-D

    SciTech Connect

    Lin-Liu, Y. R.; Chan, V. S.; Luce, T. C.; Prater, R.; Sauter, O.; Harvey, R. W.

    1999-09-20

    A velocity-space connection formula is proposed to estimate the collisionality effect on electron cyclotron current drive efficiency. The collisionality correction gives modest improvement in agreement between theoretical and recent DIII-D experimental results (c) 1999 American Institute of Physics.

  8. 60 MHz fast wave current drive experiment for DIII-D

    SciTech Connect

    Mayberry, M.J.; Chiu, S.C.; Porkolab, M.; Chan, V.; Freeman, R.; Harvey, R.; Pinsker, R. )

    1989-07-01

    The DIII-D facility provides an opportunity to test fast wave current drive appoach. Efficient FWCD is achieved by direct electron absorption due to Landa damping and transit time magnetic pumping. To avoid competing damping mechamisms we seek to maximize the single-pass asorption of the fast waves by electrons. (AIP)

  9. RMP Enhanced Transport and Rotation Screening in DIII-D Simulations

    SciTech Connect

    Izzo, V; Joseph, I; Moyer, R; Evans, T; Fenstermacher, M; Osborne, T; Lao, L; Snyder, P

    2008-10-13

    The application of resonant magnetic perturbations (RMP) to DIII-D plasmas at low collisionality has achieved ELM suppression, primarily due to a pedestal density reduction. The mechanism of the enhanced particle transport is investigated in 3D MHD simulations with the NIMROD code. The simulations apply realistic vacuum fields from the DIII-D I-coils, C-coils and measure intrinsic error fields to an EFIT reconstructed DIII-D equilibrium, and allow the plasma to respond to the applied fields while the fields are fixed at the boundary, which lies in the vacuum region. A non-rotating plasma amplifies the resonant components of the applied fields by factors of 2-5. The poloidal velocity forms E x B convection cells crossing the separatrix, which push particles into the vacuum region and reduce the pedestal density. Low toroidal rotation at the separatrix reduces the resonant field amplitudes, but does not strongly affect the particle pumpout. At higher separatrix rotation, the poloidal E x B velocity is reduced by half, while the enhanced particle transport is entirely eliminated. A high collisionality DIII-D equilibrium with an experimentally measured rotation profile serves as the starting point for a simulation with odd parity I-coil fields that can ultimately be compared with experimental results. All of the NIMROD results are compared with analytic error field theory.

  10. Recent developments on the high power ECH installation at the DIII-D tokamak

    SciTech Connect

    Lohr, J.; Ponce, D.; Callis, R.W.; Doane, J.L.; Ikezi, H.; Moeller, C.P.

    1998-09-01

    The 110 GHz gyrotron installation on the DIII-D tokamak has been upgraded to three tubes in the megawatt class with plans for further upgrades. The latest addition uses a diamond output window. The report describes the installation, plans, and experimental results to date.

  11. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    SciTech Connect

    Johnson, W.R.; Smith, J.P.; Trester, P.W.

    1997-04-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy.

  12. Issues and Solutions for Implementation of a Nanoparticle Plasma Jet Diagnostic on DIII-D

    NASA Astrophysics Data System (ADS)

    Thompson, J. R.; Bogatu, I. N.

    2014-10-01

    For ITER, runaway electron (RE) beams are considered a critical problem. Moreover, RE beam dynamics involves processes not yet fully understood or precisely diagnosed. FAR-TECH has proposed using a C60/C plasma jet as a novel diagnostic probe for RE beam-plasma interaction on DIII-D. The existing FAR-TECH prototype plasma jet system is expected to deliver up to ~75 mg C60, at ~4 km/s, and within ~1 ms of triggering, resulting in a free and bound electron density ~2.4 × 1021 m3, about 60 times larger than the typical DIII-D pre-disruption operation value. Implementation of a 100 kJ pulsed power plasma jet system is non-trivial, with electromagnetic interference (EMI) and safety being two major issues. Microsecond timescale, high current drivers generate significant EMI from which other DIII D systems need to be shielded. Safety issues associated with high voltage and potential capacitor failure must also be addressed. We will present the status of our investigation into the principle solutions for the critical issues involved in the implementation of FAR-TECH's prototype C60/C plasma jet system on DIII-D. Work supported by US DOE Grant DE-SC0011864.

  13. Investigation of the heat handling capabilities of DIII-D neutral beamline internal components

    SciTech Connect

    Phillips, J.C.; Baxi, C.B.; Hong, R.

    1993-10-01

    The current DIII-D neutral beam system is a nominal five second pulse length upgrade of a previous design rated for only 500 msec operation. While the ion sources are rated for 60 sec operation, in practice pulse lengths are limited both by the beamline internal components ability to handle the fraction of the power which is scraped off, and by the power supplies ability to provide pulse lengths of greater than 5 sec. This paper examines the capability of the existing DIII-D neutral beamline heat removing components both in terms of present and desired operating parameters. To date, at 2.5 MW per ion source, pulses are limited to 3.5 sec by beamline internal components, while at lower ratings of 2.0 MW per ion source, up to 5 sec pulses have been achieved. Recent advances and demonstration of the extraction of 3.5 MW per DIII-D ion source give an even wider window of operating conditions. A full series of beamline thermocouple data has been collected to determine the heat loading and implied surface temperatures for the various DIII-D neutral beamline internal components. These data will be presented along with an analysis of the needs for specific component upgrades, given a desire for 10 sec operation.

  14. Advances in the operation of the DIII-D neutral beam computer systems

    SciTech Connect

    Phillips, J.C.; Busath, J.L.; Penaflor, B.G.; Piglowski, D.; Kellman, D.H.; Chiu, H.K.; Hong, R.M.

    1998-02-01

    The DIII-D neutral beam system routinely provides up to 20 MW of deuterium neutral beam heating in support of experiments on the DIII-D tokamak, and is a critical part of the DIII-D physics experimental program. The four computer systems previously used to control neutral beam operation and data acquisition were designed and implemented in the late 1970`s and used on DIII and DIII-D from 1981--1996. By comparison to modern standards, they had become expensive to maintain, slow and cumbersome, making it difficult to implement improvements. Most critical of all, they were not networked computers. During the 1997 experimental campaign, these systems were replaced with new Unix compliant hardware and, for the most part, commercially available software. This paper describes operational experience with the new neutral beam computer systems, and new advances made possible by using features not previously available. These include retention and access to historical data, an asynchronously fired ``rules`` base, and a relatively straightforward programming interface. Methods and principles for extending the availability of data beyond the scope of the operator consoles will be discussed.

  15. RECENT DEVELOPMENTS ON THE 110 GHZ ELECTRON CYCLOTRON INSTATLLATION ON THE DIII-D TOKAMAK

    SciTech Connect

    PONCE,D; CALLIS,RW; CARY,WP; FERRON,JR; GREEN,M; GRUNLOH,HJ; GORELOV,Y; LOHR,J; ELLIS,RA

    2002-10-01

    OAK A271 RECENT DEVELOPMENTS ON THE 110 GHZ ELECTRON CYCLOTRON INSTALLATION ON THE DIII-D TOKAMAK. Significant improvements are being implement4ed to the capability of the 110 GHz electron cyclotron system on the DIII-D tokamak. Chief among these is the addition of the fifth and sixth 1 MW class gyrotrons, increasing the power available for auxiliary heating and current drive by nearly 60%. These tubes use artificially grown diamond rf output windows to obtain high power with long pulse capability. The beams from these tubes are nearly Gaussian, facilitating coupling to the waveguide. A new fully articulating dual launcher capable of high speed spatial scanning has been designed and tested. The launcher has two axis independent steering for each waveguide. the mirrors can be rotated at up to 100{sup o}/s. A new feedback system linking the DIII-D Plasma Control System (PCS) with the gyrotron beam voltage waveform generators permits real-time feedback control of some plasma properties such as electron temperature. The PCS can use a variety of plasma monitors to generate its control signal, including electron cyclotron emission and Mirnov probes. Electron cyclotron heating and electron cyclotron current drive (ECH and ECCD) were used during this year's DIII-D experimental campaign to control electron temperature, density, and q profiles, induce an ELM-free H-mode, and suppress the m=2/n=1 neoclassical tearing mode. The new capabilities have expanded the role of EC systems in tokamak plasma control.

  16. THERMAL STUDY OF THE DIII-D MACHINE HEAT REMOVAL CAPACITY

    SciTech Connect

    YIP,H; ADERSON,P.M; HOLTROP,K.L; HARRISON,S

    2003-10-01

    OAK-B135 With each plasma shot, the DIII-D tokamak dissipates 0.5 to 1.0 GJ of energy. Plasma shots may occur as frequently as every ten minutes, and the energy is removed in the form of heat by a cooling water system. to remove heat from the machine, cooling water circulates through each major heat source. These sources include the power supplies, motor/generator, rf current drives, neutral beam power supplies, magnetic field coils, and vacuum vessel. The cooling water system consists of isolated primary and secondary cooling loops separated by intermediate heat exchangers. As future DIII-D plans include operation during summer months and longer pulse duration, the cooling system's overall heat removal capability and performance efficiency must be assessed. Temperature and flow data from around the DIII-D facility are collected by a programmable logic controller (PLC); the data are used to analyze the heat generating sources, the heat transfer rate to intermediate heat exchangers, and the ultimate heat rejection to the environment via the cooling towers. A comparison of the original DIII-D machine design versus the actual performance determines the margin of heat removal capacity. projections of the heat removal rate for various longer plasma shots are made. Improvements in design and/or operational procedure will be necessary to attain the desired pulse duration.

  17. Lower hybrid current drive for edge current density modification in DIII-D: Final status report

    SciTech Connect

    Fenstermacher, M.E.; Porkolab, M.

    1993-08-04

    Application of Lower Hybrid (LH) Current Drive (CD) in the DIII-D tokamak has been studied at LLNL, off and on, for several years. The latest effort began in February 1992 in response to a letter from ASDEX indicating that the 2.45 GHz, 3 MW system there was available to be used on another device. An initial assessment of the possible uses for such a system on DIII-D was made and documented in September 1992. Multiple meetings with GA personnel and members of the LH community nationwide have occurred since that time. The work continued through the submission of the 1995 Field Work Proposals in March 1993 and was then put on hold due to budget limitations. The purpose of this document is to record the status of the work in such a way that it could fairly easily be restarted at a future date. This document will take the form of a collection of Appendices giving both background and the latest results from the FY 1993 work, connected by brief descriptive text. Section 2 will describe the final workshop on LHCD in DIII-D held at GA in February 1993. This was an open meeting with attendees from GA, LLNL, MIT and PPPL. Summary documents from the meeting and subsequent papers describing the results will be included in Appendices. Section 3 will describe the status of work on the use of low frequency (2.45 GHZ) LH power and Parametric Decay Instabilities (PDI) for the special case of high dielectric in the edge regions of the DIII-D plasma. This was one of the critical issues identified at the workshop. Other potential issues for LHCD in the DIII-D scenarios are: (1) damping of the waves on fast ions from neutral beam injection, (2) runaway electrons in the low density edge plasma, (3) the validity of the WKB approximation used in the ray-tracing models in the steep edge density gradients.

  18. RECENT DEVELOPMENTS IN ALTERNATIVES TO CAMAC FOR DATA ACQUISITION AT DIII-D

    SciTech Connect

    KELLMAN,D.H; CAMPBELL,G.L; FERRON,J.R; PIGLOWSKI,D.A; AUSTIN,M.E; MCKEE,G.R

    2003-10-01

    OAK-B135 For over twenty years, data acquisition hardware at DIII-D has been based on the CAMAC platform. These rugged and reliable systems, however, are gradually becoming obsolete due to end-of-life issues, ever-decreasing industry support of older hardware, and the availability of modern alternative hardware with superior performance. Efforts are underway at DIII-D to adopt new data acquisition solutions which exploit modern technologies and surpass the limitations of the CAMAC standard. These efforts have involved the procurement and development of data acquisition systems based on the PCI and Compact-PCI platform standards. These systems are comprised of rack-mount computers containing data acquisition boards (digitizers), Ethernet connectivity, and the drivers and software necessary for control. Each digitizer contains analog-to-digital converters, control circuitry, firmware and memory to collect, store, and transfer waveform data acquired using internal or external triggers and clocks. Software has been developed which allows DIII-D computers to program the operational parameters of the digitizers, as well as to upload acquired data into the DIII-D acquisition database. All communication between host computers and the new acquisition systems occurs via standard Ethernet connections, a vast improvement over the slower, serial loop highways used for control and data transfer with CAMAC systems. In addition, the capabilities available in modern integrated and printed circuit manufacture result in digitizers with high channel count and memory density. Cost savings are also realized by utilizing a platform based on standards of the personal computer industry. Details of the new systems at DIII-D are presented, along with initial experience with their use, and plans for future expansion and improvement.

  19. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    SciTech Connect

    Baker, D.

    1993-05-01

    The DIII-D tokamak research program is carried out by, General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data needed by International Thermonuclear Experimental Reactor (ITER) and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics.

  20. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    SciTech Connect

    Simonen, T.C.; Baker, D.

    1993-01-01

    The DIII-D tokamak research program is carried out by General Atomics for the U.S. Department of Energy. The DIII-D is the most flexible and best diagnosed tokamak in the world and the second largest tokamak in the U.S. The primary goal of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated in three major areas: Tokamak Physics, Divertor and Boundary Physics, and Advanced Tokamak Studies.

  1. q-profile measurements with motional stark effect in DIII-D

    SciTech Connect

    Rice, B.W.

    1996-07-01

    Motional Stark Effect (MSE) has emerged as the most reliable technique for the measurement of q profiles in tokamaks equipped with neutral beams. On DIII-D, two separate 8-channel MSE instruments are utilized, giving 16 radial measurements. The viewing geometry for each instrument is designed to optimize radial resolution (1.5 cm < {delta}R < 6 cm) over the measurement range. Dual-frequency polarization modulators combined with lock-in detection provides maximum signal-to-noise with a time response of {approximately}1 ms. Special consideration has gone into the design of the collection optics to minimize undesirable polarization effects on reflectors and Faraday rotation in lenses. Examples of data from negative central magnetic shear, VH-mode, and sawtoothing discharges highlight the capabilities of the DIII-D MSE diagnostic.

  2. Plasma radiometry with 30 chord resolution for fast transients in the DIII-D tokamak

    SciTech Connect

    Gray, D.S.; Hollmann, E.M.; Luckhardt, S.C.; Chalfant, J.; Chousal, L.; Hernandez, R.; Jones, E.; Kellman, A.G.

    2004-10-01

    A diagnostic capable of providing time resolved measurements of plasma radiated power during disruptions and other fast transients, e.g., edge localized modes has been employed in the DIII-D tokamak. The radiation is detected with absolute extreme ultraviolet (AXUV) photodiode arrays. Thirty chords from a single port provide measurements from a full slice of the plasma at one toroidal location. The analog bandwidth is up to 1 MHz for the brightest events, i.e., disruptions. Active cooling of the diode arrays prevents damage during high temperature vessel baking. Effective responsivity values of 0.12-0.18 A/W are taken from previous work on the application of AXUV diodes in DIII-D. The total radiated energy in disruptions typically agrees with bolometer measurements within about 12%.

  3. Arcing and its role in PFC erosion and dust production in DIII-D

    NASA Astrophysics Data System (ADS)

    Rudakov, D. L.; Chrobak, C. P.; Doerner, R. P.; Krasheninnikov, S. I.; Moyer, R. A.; Umstadter, K. R.; Wampler, W. R.; Wong, C. P. C.

    2013-07-01

    Two types of arc tracks are observed on the plasma-facing components (PFCs) in DIII-D. "Unmagnetized" random walk tracks are produced during glow discharges; they are rare and have no importance for PFC erosion but may degrade diagnostic mirrors. "Magnetized" scratch-like type II tracks are produced by unipolar arcs during plasma operations; they are formed by "retrograde BxJ" motion of the cathode spot and are roughly perpendicular to the local magnetic field. Type II arcs cause measurable erosion of graphite, but based on the evidence available they are relatively small contributors to the total erosion of carbon in DIII-D compared to other mechanisms such as physical and chemical sputtering and ablation from leading edges. Erosion by arcing of tungsten films deposited on graphite samples was observed in Divertor Material Evaluation System (DiMES) experiments. New DiMES experiments aimed at time-resolved arc measurements are proposed.

  4. Lower hybrid counter current drive for edge current density modification in DIII-D

    SciTech Connect

    Fenstermacher, M.E.; Nevins, W.M.; Porkolab, M.; Bonoli, P.T.; Harvey, R.W.

    1993-07-01

    Each of the Advanced Tokamak operating modes in DIII-D is thought to have a distinctive current density profile. So far these modes have only been achieved transiently through experiments which ramp the plasma current and shape. Extension of these modes to steady state requires non-inductive current profile control, e.g. with lower hybrid current drive (LHCD). Calculations of LHCD have been done for DIII-D using the ACCOME and CQL3D codes, showing that counter driven current at the plasma edge can cancel some of the undesirable edge bootstrap current and potentially extend the VH-mode. Results are presented for scenarios using 2.45 GHz LH waves launched from both the midplane and off-axis ports. The sensitivity of the results to injected power, n{sub e} and T{sub e}, and launched wave spectrum is also shown.

  5. Lower-hybrid counter current drive for edge current density modification in DIII-D

    SciTech Connect

    Fenstermacher, M.E.; Nevins, W.M. ); Porkolab, M.; Bonoli, P.T. ); Harvey, R.W. )

    1994-10-15

    Each of the Advanced Tokamak operating modes in DIII-D is thought to have a distinctive current density profile. So far these modes have only been achieved transiently through experiments which ramp the plasma current and shape. Extension of these modes to steady state requires non-inductive current profile control, e.g., with lower hybrid current drive (LHCD). Calculations of LHCD have been done for DIII-D using the ACCOME and CQL3D codes, showing that counter driven current at the plasma edge can cancel some of the undesirable edge bootstrap current and potentially extend the VH-mode. Results will be presented for scenarios using 2.45 GHz LH waves launched from both the midplane and off-axis ports. The sensitivity of the results to injected power, [ital n][sub [ital e

  6. Commissioning of the Microwave Imaging Reflectometer (MIR) on DIII-D

    NASA Astrophysics Data System (ADS)

    Muscatello, C. M.; Domier, C. W.; Gamzina, D.; Hu, X.; Luhmann, N. C., Jr.; Ren, X.; Riemenschneider, P.; Spear, A.; Yu, L.; Munsat, T.; Zemedkun, S. E.; Tobias, B. J.

    2013-10-01

    A microwave imaging reflectometer (MIR), capable of simultaneously measuring the poloidal and radial structure of density fluctuations, has been developed for DIII-D and installed in May 2013. The MIR diagnostic concept has undergone numerous technological and system-level upgrades since earlier microwave imaging systems, thereby permitting a higher level of robustness and flexibility. Synthetic diagnostic simulations permit determination of the resolvable wavenumbers and density fluctuations levels. Laboratory qualification tests are performed to characterize the system performance compared to the designed parameters. First plasma results are presented in the form of a brief survey of MIR results collected during several select experiments from the 2013 DIII-D experimental campaign. Work supported by the US Department of Energy under DE-FG02-00ER54531, DE-FC02-04ER54698, DE-SC0003913 DE0FC02-05ER54816, and DE-AC02-09CH11466.

  7. The production and confinement of runaway electrons with impurity killer pellets in DIII-D

    SciTech Connect

    Evans, T.E.; Taylor, P.L.; Whyte, D.G.

    1998-12-01

    Prompt runaway electron bursts, generated by rapidly cooling DIII-D plasmas with argon killer pellets, are used to test a recent knock-on avalanche theory describing the growth of multi-MeV runaway electron currents during disruptions in tokamaks. Runaway current amplitudes, observed during some but not all DIII-D current quenches, are consistent with growth rates predicted by the theory assuming a pre-current quench runaway electron density of approximately 10{sup 15} m{sup {minus}3}. Argon killer pellet modeling yields runaway densities of between 10{sup 15}--10{sup 16} m{sup {minus}3} in these discharges. Although knock-on avalanching appears to agree rather well with the measurements, relatively small avalanche amplification factors combined with uncertainties in the spatial distribution of pellet mass and cooling rates make it difficult to unambiguously confirm the proposed theory with existing data.

  8. Novel current drive experiments on the CDX-U, HIT, and DIII-D Tokamaks

    SciTech Connect

    Ono, M.; Forest, C.B.; Hwang, Y.S.; Armstrong, R.J.; Choe, W.; Darrow, D.S.; Greene, G.; Jones, T. . Plasma Physics Lab.); Jarboe, T.R.; Martin, A.; Nelson, B.A.; Orvis, D.; Painter, C.; Zhou, L.; Rogers, J.A. ); Schaffer, M.J.; Hyatt, A.W.; Pinsker, R.I.; Staebler, G.M.; Stambaugh, R.D.; Strait, E.J.; Greene, K.L.; Leuer, J.A.; Lohr, J.

    1992-01-01

    Two types of novel, non-inductive current drive concepts for starting-up and maintaining tokamak discharges have been developed on the CDX-U, HIT, and DIII-D Tokamaks. On CDX-U, a new, non-inductive current drive technique utilizing fully internally generated pressure driven currents has been demonstrated. The measured current density profile shows a non-hollow profile which agrees with a modeling calculation including helicity conserving non-classical current transport providing the seed current''. Another current drive concept, dc-helicity injection, has been investigated on, CDX-U, HIT and DIII-D. This method utilizes injection of magnetic helicity via low energy electron currents, maintaining the plasma current through helicity conserving relaxiation. In these experiments, non-ohmic tokamak plasmas were formed and maintained in the tens of kA range.

  9. Novel current drive experiments on the CDX-U, HIT, and DIII-D Tokamaks

    SciTech Connect

    Ono, M.; Forest, C.B.; Hwang, Y.S.; Armstrong, R.J.; Choe, W.; Darrow, D.S.; Greene, G.; Jones, T.; Jarboe, T.R.; Martin, A.; Nelson, B.A.; Orvis, D.; Painter, C.; Zhou, L.; Rogers, J.A.; Schaffer, M.J.; Hyatt, A.W.; Pinsker, R.I.; Staebler, G.M.; Stambaugh, R.D.; Strait, E.J.; Greene, K.L.; Leuer, J.A.; Lohr, J.M.

    1992-10-01

    Two types of novel, non-inductive current drive concepts for starting-up and maintaining tokamak discharges have been developed on the CDX-U, HIT, and DIII-D Tokamaks. On CDX-U, a new, non-inductive current drive technique utilizing fully internally generated pressure driven currents has been demonstrated. The measured current density profile shows a non-hollow profile which agrees with a modeling calculation including helicity conserving non-classical current transport providing the ``seed current``. Another current drive concept, dc-helicity injection, has been investigated on, CDX-U, HIT and DIII-D. This method utilizes injection of magnetic helicity via low energy electron currents, maintaining the plasma current through helicity conserving relaxiation. In these experiments, non-ohmic tokamak plasmas were formed and maintained in the tens of kA range.

  10. Handling and archiving of magnetic fusion data at DIII-D

    SciTech Connect

    VanderLaan, J.F.; Miller, S.; McHarg, B.B. Jr.; Henline, P.A.

    1995-10-01

    Recent modifications to the computer network at DIII-D enhance the collection and distribution of newly acquired and archived experimental data. Linked clients and servers route new data from diagnostic computers to centralized mass storage and distribute data on demand to local and remote workstations and computers. Capacity for data handling exceeds the upper limit of DIII-D Tokamak data production of about 4 GBytes per day. Network users have fast access to new data stored on line. An interactive program handles requests for restoration of data archived off line. Disk management procedures retain selected data on line in preference to other data. Redundancy of all components on the archiving path from the network to magnetic media has prevented loss of data. Older data are rearchived as dictated by limited media life.

  11. Study of Current Profile Evolution in Presence of Tearing Modes in DIII-D Hybrid Discharges

    SciTech Connect

    Casper, T; Jayakumar, R; Pearlstein, L; Lodestro, L

    2004-06-25

    An intermediate regime for tokamak operation has been obtained in DIII-D and in other tokamaks in which the inductive flux consumption is reduced and a broad current profile with the safety factor just above or near the sawtoothing limit is obtained and maintained. The DIII-D tokamak was operated in this regime near the no-wall b limit. High stability and good confinement was achieved at a desired level of q{sub 95} {approx} 3 to 4 for durations as long as 35{tau}{sub E}, three times the current-diffusion time. This regime offers the promise of achieving higher fusion gain and yield and/or longer burn duration for ITER.

  12. OBSERVATION OF SUPRATHERMAL ELECTRONS DURING MAGNETIC RECONNECTION AT THE SAWTOOTH INSTABILITY IN DIII-D TOKAMAK

    SciTech Connect

    SAVRUKHIN,RV; STRAIT,EJ

    2002-11-01

    OAK A271 OBSERVATION OF SUPRATHERMAL ELECTRONS DURING MAGNETIC RECONNECTION AT THE SAWTOOTH INSTABILITY IN DIII-D TOKAMAK. Intense bursts of x-ray and electron cyclotron emission are observed during sawtooth instabilities in high-temperature plasmas in the DIII-D tokamak. The bursts are initiated around the X-point of the m = 1, n = 1 magnetic island at the beginning of the sawtooth crash and are displaced to larger radii later during the temperature collapse. Reconstruction of the magnetic configuration using motional Stark effect (MSE) data and numerical simulations indicates that the bursts can be connected with suprathermal electrons (E{sub r} {approx} 30-40 keV) generated during reconnection of the magnetic field around the q = 1 surface.

  13. Reconstruction of Detached Divertor Plasma Conditions in DIII-D Using Spectroscopic and Probe Data

    SciTech Connect

    Stangeby, P; Fenstermacher, M

    2004-12-03

    For some divertor aspects, such as detached plasmas or the private flux zone, it is not clear that the controlling physics has been fully identified. This is a particular concern when the details of the plasma are likely to be important in modeling the problem--for example, modeling co-deposition in detached inner divertors. An empirical method of ''reconstructing'' the plasma based on direct experimental measurements may be useful in such situations. It is shown that a detached plasma in the outer divertor leg of DIII-D can be reconstructed reasonably well using spectroscopic and probe data as input to a simple onion-skin model and the Monte Carlo hydrogenic code, EIRENE. The calculated 2D distributions of n{sub e} and T{sub e} in the detached divertor were compared with direct measurements from the divertor Thomson scattering system, a diagnostic capability unique to DIII-D.

  14. Edge density modification with rf on TFTR and DIII-D

    SciTech Connect

    Hanson, G.R.; England, A.C.; Wilgen, J.B.

    1995-09-01

    Modification of the electron density profile in front of rf antennas has been observed on TFTR and DIII-D using reflectometers installed in-antenna to ensure localization to the antenna environment. The modification of the edge density gradient has two components: a flattening of the gradient in the private flux zone, and an increase in the edge density inside the last closed flux surface. In general, these modifications result in a significant decrease in the electron density over most of the private flux zone with a sharp rise in the density near the last closed flux surface. Data from TFTR is used to infer the dependence of this edge modification with antenna strap phasing and rf power. Initial results from DIII-D showing edge modification will also be discussed. Antenna modeling has shown that this modification in the private flux zone has only small effects on the antenna loading and launched spectrum.

  15. Scintillator-based diagnostic for fast ion loss measurements on DIII-D.

    PubMed

    Fisher, R K; Pace, D C; García-Muñoz, M; Heidbrink, W W; Muscatello, C M; Van Zeeland, M A; Zhu, Y B

    2010-10-01

    A new scintillator-based fast ion loss detector has been installed on DIII-D with the time response (>100 kHz) needed to study energetic ion losses induced by Alfvén eigenmodes and other MHD instabilities. Based on the design used on ASDEX Upgrade, the diagnostic measures the pitch angle and gyroradius of ion losses based on the position of the ions striking the two-dimensional scintillator. For fast time response measurements, a beam splitter and fiberoptics couple a portion of the scintillator light to a photomultiplier. Reverse orbit following techniques trace the lost ions to their possible origin within the plasma. Initial DIII-D results showing prompt losses and energetic ion loss due to MHD instabilities are discussed. PMID:21033833

  16. TRANSPORT OF ELM ENERGY AND PARTICLES INTO THE SOL AND DIVERTOR OF DIII-D

    SciTech Connect

    LEONARD,AW; OSBORNE,TH; FENSTERMACHER,ME; GROEBNER,RJ; GROTH,M; LASNIER,CJ; MAHDAVI,MA; PETRIE,TW; SNYDER,PB; WATKINS,JG; ZENG,L

    2002-11-01

    A271 TRANSPORT OF ELM ENERGY AND PARTICLES INTO THE SOL AND DIVERTOR OF DIII-D. The reduction in size of Type I edge localized models (ELMs) with increasing density is explored in DIII-D for the purpose of studying the underlying transport of ELM energy. The separate convective and conductive transport of energy due to an ELM is determined by Thomson scattering measurements of electron density and temperature in the pedestal. The conductive transport from the pedestal during an ELM decreases with increasing density, while the convective transport remains nearly constant. The scaling of the ELM energy loss is compared with an edge stability model. The role of the divertor sheath in limiting energy loss from the pedestal during an ELM is explored. Evidence of outward radial transport to the midplane wall during an ELM is also presented.

  17. Operational upgrades to the DIII-D 60 GHz electron cyclotron resonant heating system

    SciTech Connect

    Harris, T.E.; Cary, W.P.

    1993-10-01

    One of the primary components of the DIII-D radio frequency (rf) program over the past seven years has been the 60 GHz electron cyclotron resonant heating (ECRH) system. The system now consists of eight units capable of operating and controlling eight Varian VGE-8006 60 GHz, 200 kW gyrotrons along with their associated waveguide components. This paper will discuss the operational upgrades and the overall system performance. Many modifications were instituted to enhance the system operation and performance. Modifications discussed in this paper include an improved gyrotron tube-fault response network, a computer controlled pulse-timing and sequencing system, and an improved high-voltage power supply control interface. The discussion on overall system performance will include operating techniques used to improve system operations and reliability. The techniques discussed apply to system start-up procedures, operating the system in a conditioning mode, and operating the system during DIII-D plasma operations.

  18. Dimensionally similar discharges with central rf heating on the DIII-D tokamak

    SciTech Connect

    Petty, C.C.; Luce, T.C.; Pinsker, R.I.

    1993-04-01

    The scaling of L-mode heat transport with normalized gyroradius is investigated on the DIII-D tokamak using central rf heating. A toroidal field scan of dimensionally similar discharges with central ECH and/or fast wave heating show gyro-Bohm-like scaling both globally and locally. The main difference between these restats and those using NBI heating on DIII-D is that with rf heating the deposition profile is not very sensitive to the plasma density. Therefore central heating can be utilized for both the low-B and high-B discharges, whereas for NBI the power deposition is decidedly off-axis for the high-B discharge (i.e., high density).

  19. A decade of DIII-D research. Final report for the period of work, October 1, 1989--September 30, 1998

    SciTech Connect

    1999-03-01

    During the ten-year DIII-D tokamak operating period of 1989 through 1998, major scientific advances and discoveries were made and facility upgrades and improvements were implemented. Each year, annual reports as well as journal and international conference proceedings document the year-by-year advances (summarized in Section 7). This final contract report, provides a summary of these historical accomplishments. Section 2 encapsulates the 1998 status of DIII-D Fusion Science research. Section 3 summarizes the DIII-D facility operations. Section 4 describes the major upgrades to the DIII-D facility during this period. During the ten-year period, DIII-D has grown from predominantly a General Atomics program to a national center for fusion science with participants from over 50 collaborating institutions and 300 users who spend more than one week annually at DIII-D to carry out experiments or data analysis. In varying degrees, these collaborators participate in formulating the research program directions. The major collaborating institution programs are described in Section 6.

  20. Spatial dependence of the sheath power transmission factor in DIII-D

    SciTech Connect

    Futch, A.H.; Hill, D.N.; Porter, G.D. ); Matthews, G.F. ); Buchenauer, D. )

    1991-02-01

    The spatial dependence of the power transmission factor, d, associated with an ion-electron pair passing through the sheath at the DIII-D divertor plate has been determined by sweeping the edge plasma across Langmuir probe detectors. Our results show that d decreases from the classically expected value of eight near the low density edge of the scrape-off-layer plasma to values less than unity at the peak of the profile. 13 refs., 16 figs.

  1. Nonlinear MHD simulations of Quiescent H-mode plasmas in DIII-D

    NASA Astrophysics Data System (ADS)

    Liu, F.; Huijsmans, G. T. A.; Loarte, A.; Garofalo, A. M.; Solomon, W. M.; Snyder, P. B.; Hoelzl, M.; Zeng, L.

    2015-09-01

    In the Quiescent H-mode (QH-mode) regime, the edge harmonic oscillation (EHO), thought to be a saturated kink-peeling mode (KPM) driven unstable by current and rotation, is found in experiment to provide sufficient stationary edge particle transport to avoid the periodic expulsion of particles and energy by edge localized modes (ELMs). In this paper, both linear and nonlinear MHD modelling of QH-mode plasmas from the DIII-D tokamak have been investigated to understand the mechanism leading to the appearance of the EHO in QH-mode plasmas. For the first time nonlinear MHD simulations with low-n modes both with ideal wall and resistive wall boundary conditions have been carried out with the 3D non-linear MHD code JOREK. The results show, in agreement with the original conjectures, that in the non-linear phase, kink peeling modes are the main unstable modes in QH-mode plasmas of DIII-D and that the kink-peeling modes saturate non-linearly leading to a 3D stationary state. The characteristics of the kink-peeling modes, in terms of mode structure and associated decrease of the edge plasma density associated with them, are in good agreement with experimental measurements of the EHO in DIII-D. The effect of plasma resistivity, the role of plasma parallel rotation as well as the effect of the conductivity of the vacuum vessel wall on the destabilization and saturation of kink-peeling modes have been evaluated for experimental QH-mode plasma conditions in DIII-D.

  2. First results on fast wave current drive in advanced tokamak discharges in DIII-D

    SciTech Connect

    Prater, R.; Cary, W.P.; Baity, F.W.

    1995-07-01

    Initial experiments have been performed on the DIII-D tokamak on coupling, direct electron heating, and current drive by fast waves in advanced tokamak discharges. These experiments showed efficient central heating and current drive in agreement with theory in magnitude and profile. Extrapolating these results to temperature characteristic of a power plant (25 keV) gives current drive efficiency of about 0.3 MA/m{sup 2}.

  3. Power deposition by mode converted electron Bernstein waves in the DIII-D heat pinch experiments

    SciTech Connect

    Forest, C.B.; Harvey, R.W.; Smirnov, A.P.

    2001-05-01

    Mode converted electron Bernstein waves are shown to play an important role in power deposition in the off-axis electron cyclotron heating experiments on the DIII-D tokamak in which the effect known as the 'heat pinch' was inferred. Ray tracing shows that the mode converted Bernstein modes (generated when the launched X mode reflects from an outboard upper hybrid layer) are damped in the central region where the previous analysis had assumed no power was being deposited.

  4. Noninductive current drive experiments on DIII-D, and future plans

    SciTech Connect

    Prater, R.; Austin, M.E.; Baity, F.W.

    1994-02-01

    Experiments on DIII-D (and other tokamaks) have shown that improved performance can follow from optimization of the current density profile. Increased confinement of energy and a higher limit on beta have both been found in discharges in which the current density profile is modified through transient means, such as ramping of current or elongation. Peaking of the current distribution to obtain discharges with high internal inductance {ell}{sub i} has been found to be beneficial. Alternatively, discharges with broader profiles, as in the VH-mode or with high beta poloidal, have shown improved performance. Noninductive current drive is a means to access these modes of improved confinement on a steady state basis. Accordingly, experiments on noninductive current drive are underway on the DIII-D tokamak using fast waves, electron cyclotron waves. Recent experiments on fast wave current drive have demonstrated the ability to drive up to 180 kA of noninductive current using 1.5 MW of power at 60 MHz, including the contribution from 1 MW of ECCD and the bootstrap current. Higher power rf current drive systems are needed to strongly affect the current profile on DIII-D. An upgrade to the FWCD system is underway to increase the total power to 6 MW, using two additional antennas and two new 30 to 120 MHz transmitters. Additionally, a 1 MW prototype ECH system at 110 GHz is being developed (with eventual upgrade to 10 MW). With these systems, noninductive current drive at the 1 MA level will be available for experiments on profile control in DIII-D.

  5. Detailed Measurements of ECCD Efficiency on DIII-D for Comparison with Theory

    SciTech Connect

    Petty, C.C.; Prater, R.; Lohr, J.; Luce, T.C.; Ellis,III, R.A., Harvey, R.W.; Kinsey, J.E.; Lao, L.L.; Makowski, M.A.

    2002-05-01

    Recent experiments on the DIII-D tokamak have focused on determining the effect of trapped particles on the electron cyclotron current drive (ECCD) efficiency. Localized power deposition and current drive were observed for both L-mode and ELMing H-mode plasmas. The ECCD efficiency was measured for both co and counter injection. Overall, the experimental ECCD efficiency was in good agreement with the CQL3D Fokker-Planck model over a wide range of injection angles and plasma parameters.

  6. System control and data acquisition of the two new FWCD RF systems at DIII-D

    SciTech Connect

    Harris, T.E.; Allen, J.C.; Cary, W.P.; Petty, C.C.; Pinsker, R.I.; Ferguson, S.W.

    1995-12-31

    The Fast Wave Current Drive (FWCD) system at DIII-D has increased its available radio frequency (RF) power capabilities with the addition of two new high power transmitters along with their associated transmission line systems. A Sun Sparc-10 workstation, functioning as the FWCD operator console, is being used to control transmitter operating parameters and transmission line tuning parameters, along with acquiring data and making data available for integration into the DIII-D data acquisition system. Labview, a graphical user interface application, is used to manage and control the above processes. This paper will discuss the three primary branches of the FWCD computer control system: transmitter control, transmission line tuning control, and FWCD data acquisition. The main control program developed uses VXI, GPIB, CAMAC, Serial, and Ethernet protocols to blend the three branches together into one cohesive system. The control of the transmitters utilizes VXI technology to communicate with the transmitter`s digital interface. A GPIB network allows for communication with various instruments and CAMAC crate controllers. CAMAC crates are located at each phase-shifter/stub-tuner station and are used to digitize transmission line parameters along with transmission line fault detection during RF transmission. The phase-shifter/stub-tuner stations are located throughout the DIII-D facility and are controlled from the FWCD operator console via the workstation`s Serial port. The Sun workstation has an Ethernet connection allowing for the utilization of the DIII-D data acquisition Open System architecture and of course providing communication with the rest of the world.

  7. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    SciTech Connect

    Johnson, W.R.; Smith, J.P.; Stambaugh, R.D.

    1996-04-01

    V-4Cr-4-Ti alloy has been recently selected for use in the manufacture of a portion of the DIII-D Radiative Divertor modification, as part of an overall DIII-D vanadium alloy deployment effort developed by General Atomics (GA) in conjunction with the Argonne and Oak Ridge National Laboratories (ANL or ORNL). The goal of this work is to produce a production-scale heat of the alloy and fabricate it into product forms for the manufacture of a portion of the Radiative Divertor (RD) for the DIII-D tokamak, to develop the fabrications technology for manufacture of the vanadium alloy radiative Divertor components, and to determine the effects of typical tokamak environments in the behavior of the vanadium alloy. The production of a {approx}1300-kg heat of V-4Cr-4Ti alloy is currently in progress at Teledyne Wah Chang of Albany, oregon (TWCA) to provide sufficient material for applicable product forms. Two unalloyed vanadium ingots for the alloy have already been produced by electron beam melting of raw processes vanadium. Chemical compositions of one ingot and a portion of the second were acceptable, and Charpy V-Notch (CVN) impact test performed on processed ingot samples indicated ductile behavior. Material from these ingots are currently being blended with chromium and titanium additions, and will be vacuum-arc remelted into a V-4Cr-4Ti alloy ingot and converted into product forms suitable for components of the DIII-D RD structure. Several joining methods selected for specific applications in fabrication of the RD components are being investigated, and preliminary trials have been successful in the joining of V-alloy to itself by both resistance and inertial welding processes and to Inconel 625 by inertial welding.

  8. Recent improvements to the DIII-D neutral beam instrumentation and control system

    SciTech Connect

    Kellman, D.H.; Hong, R.

    1997-11-01

    The DIII-D neutral beam (NB) instrumentation and control (I and C) system provides for operational control and synchronization of the eight DIII-D neutral beam injection systems, as well as for pertinent data acquisition and safety interlocking. Recently, improvements were made to the I and C system. With the replacement of the NB control computers, new signal interfacing was required to accommodate the elimination of physical operator panels, in favor of graphical user interface control pages on computer terminal screens. The program in the mode control (MC) programmable logic controller (PLC), which serves as a logic-processing interface between the NB control computers and system hardware, was modified to improve the availability of NB heating of DIII-D plasmas in the event that one or more individual beam systems suddenly become unavailable while preparing for a tokamak experimental shot sequences. An upgraded computer platform was adopted for the NB control system operator interface and new graphical user interface pages were developed to more efficiently display system status data. A failure mode of the armor tile infrared thermometers (pyrometers), which serve to terminate beam pulsing if beam shine-through overheats wall thermal shielding inside the DIII-D tokamak, was characterized such that impending failures can be detected and repairs effected to mitigate beam system down-time. The hardware that controls gas flow to the beamline neutralizer cells was upgraded to reduce susceptibility to electromagnetic interference (EMI), and interlocking was provided to terminate beam pulsing in the event of insufficient neutralizer gas flow. Motivation, implementation, and results of these improvements are presented.

  9. 2 MW 110 GHz ECH heating system for DIII-D

    SciTech Connect

    Moeller, C.; Prater, R.; Callis, R.; Remsen, D.; Doane, J.; Cary, W.; Phelps, R.; Tupper, M.

    1990-09-01

    A 2 MW 110 GHz ECH system using Varian 0.5 MW gyrotrons is under construction for use on the DIII-D tokamak by late 1991. Most of the components are being design and fabricated at General Atomics, including the gyrotron tanks, superconducting magnets, and transmission line. These components are intended for operation with 10 second pulses and, in the future, with 1 MW gyrotrons. 6 refs., 5 figs.

  10. SYSTEM PERFORMANCE AND EXPERIMENTS WITH THE 110 GHZ MICROWAVE INSTALLATION ON THE DIII-D TOKAMAK

    SciTech Connect

    J.M. LOHR; F.W. BAITY,JR.; G.C. BARBER; R.W. CALLIS; I. GORELOV; C.M. GREENFIELD; R.A. LEGG; T.C. LUCE; C.C. PETTY; D. PONCE; R. PRATER

    2000-09-01

    A powerful microwave system operating at the second harmonic of the electron cyclotron frequency has been commissioned on the DIII-D tokamak. The primary mission of the microwave system is to permit current profile control leading to the improved performance of advanced tokamak operation in quasi-steady state. Initial performance tests and experiments on current drive both near and away from the tokamak axis and on transport have been performed.

  11. System control and data acquisition of the two new FWCD RF systems at DIII-D

    SciTech Connect

    Harris, T.E.; Allen, J.C.; Cary, W.P. Petty, C.C.

    1995-10-01

    The Fast Wave Current Drive (FWCD) system at DIII-D has increased its available radio frequency (RF) power capabilities with the addition of two new high power transmitters along with their associated transmission line systems. A Sun Sparc-10 workstation, functioning as the FWCD operator console, is being used to control transmitter operating parameters and transmission line tuning parameters, along with acquiring data and making data available for integration into the DIII-D data acquisition system. Labview, a graphical user interface application, is used to manage and control the above processes. This paper will discuss the three primary branches of the FWCD computer control system: transmitter control, transmission line tuning control, and FWCD data acquisition. The main control program developed uses VXI, GPIB, CAMAC, Serial, and Ethernet protocols to blend the three branches together into one cohesive system. The control of the transmitters utilizes VXI technology to communicate with the transmitter`s digital interface. A GPIB network allows for communication with various instruments and CAMAC crate controllers. CAMAC crates are located at each phase-shifter/stub-tuner station and are used to digitize transmission line parameters along with transmission line fault detection during RF transmission. The phase-shifter/stub-tuner stations are located through out the DIII-D facility and are controlled from the FWCD operator console via the workstation`s Serial port. The Sun workstation has an Ethernet connection allowing for the utilization of the DIII-D data acquisition {open_quotes}Open System{close_quotes} architecture and of course providing communication with the rest of the world.

  12. Suppression of type-I ELMs with reduced RMP coil set on DIII-D

    NASA Astrophysics Data System (ADS)

    Orlov, D. M.; Moyer, R. A.; Evans, T. E.; Paz-Soldan, C.; Ferraro, N. M.; Nazikian, R.; deGrassie, J. S.; Grierson, B. A.; Eldon, D.; Fenstermacher, M. E.; King, J. D.; Logan, N. C.; Lanctot, M. J.; Maingi, R.; Snyder, P. B.; Strait, E. J.; Wingen, A.

    2016-03-01

    Recent experiments on DIII-D have demonstrated that having a toroidally-monochromatic spectral content of edge-resonant magnetic perturbations (RMPs) is not a necessary condition for suppression of edge localized modes (ELMs). Robust ELM suppression has been reproducibly obtained on DIII-D during experiments in which various non-axisymmetric coil loops were turned off pseudo-randomly producing a variety of n=1 , n= 2 , and n= 3 spectral contributions. It was shown that RMP ELM suppression could be achieved with as few as 5 out of 12 internal coil loops (I-coils) on DIII-D at similar coil currents and with good plasma confinement. Linear MHD plasma response (m3dc1, ipec, mars) and vacuum (surfmn, trip3d) modelling have been performed in order to understand the effects of the perturbation spectrum on the plasma response and ELM suppression. The results suggest that reduction of the dominant n= 3 perturbation field is compensated by increased n= 2 field in the plasma that may lead to RMP ELM suppression at lower levels of n= 3 perturbative magnetic flux from the I-coils. These results provide additional confidence that ITER may be capable of RMP ELM suppression in the event of multiple internal coil failures.

  13. Fast wave current drive experiment on the DIII-D tokamak

    SciTech Connect

    Petty, C.C.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.; Harvey, R.W.; Lohr, J.; Luce, T.C.; Mayberry, M.J.; Prater, R. ); Porkolab, M. ); Baity, F.W.; Goulding, R.H.; Hoffman, J.D. ); James, R.A. ); Kawash

    1992-06-01

    One method of radio-frequency heating which shows theoretical promise for both heating and current drive in tokamak plasmas is the direct absorption by electrons of the fast Alfven wave (FW). Electrons can directly absorb fast waves via electron Landau damping and transit-time magnetic pumping when the resonance condition {omega} {minus} {kappa}{sub {parallel}e}{upsilon}{sup {parallel}e} = O is satisfied. Since the FW accelerates electrons traveling the same toroidal direction as the wave, plasma current can be generated non-inductively by launching FW which propagate in one toroidal direction. Fast wave current drive (FWCD) is considered an attractive means of sustaining the plasma current in reactor-grade tokamaks due to teh potentially high current drive efficiency achievable and excellent penetration of the wave power to the high temperature plasma core. Ongoing experiments on the DIII-D tokamak are aimed at a demonstration of FWCD in the ion cyclotron range of frequencies (ICRF). Using frequencies in the ICRF avoids the possibility of mode conversion between the fast and slow wave branches which characterized early tokamak FWCD experiments in the lower hybrid range of frequencies. Previously on DIII-D, efficient direct electron heating by FW was found using symmetric (non-current drive) antenna phasing. However, high FWCD efficiencies are not expected due to the relatively low electron temperatures (compared to a reactor) in DIII-D.

  14. Data Analysis Software Tools for Enhanced Collaboration at the DIII-D National Fusion Facility

    SciTech Connect

    Schachter, J.; Peng, Q.; Schissel, D.P.

    1999-07-01

    Data analysis at the DIII-D National Fusion Facility is simplified by the use of two software packages in analysis codes. The first is GAP1otObj, an IDL-based object-oriented library used in visualization tools for dynamic plotting. GAPlotObj gives users the ability to manipulate graphs directly through mouse and keyboard-driven commands. The second software package is MDSplus, which is used at DIED as a central repository for analyzed data. GAPlotObj and MDSplus reduce the effort required for a collaborator to become familiar with the DIII-D analysis environment by providing uniform interfaces for data display and retrieval. Two visualization tools at DIII-D that benefit from them are ReviewPlus and EFITviewer. ReviewPlus is capable of displaying interactive 2D and 3D graphs of raw, analyzed, and simulation code data. EFITviewer is used to display results from the EFIT analysis code together with kinetic profiles and machine geometry. Both bring new possibilities for data exploration to the user, and are able to plot data from any fusion research site with an MDSplus data server.

  15. Finding evidence for density fluctuation effects on electron cyclotron heating deposition profiles on DIII-D

    SciTech Connect

    Brookman, M. W. Austin, M. E.; Petty, C. C.

    2015-12-10

    Theoretical work, computation, and results from TCV [J. Decker “Effect of density fluctuations on ECCD in ITER and TCV,” EPJ Web of Conf. 32, 01016 (2012)] suggest that density fluctuations in the edge region of a tokamak plasma can cause broadening of the ECH deposition profile. In this paper, a GUI tool is presented which is used for analysis of ECH deposition as a first step towards looking for this broadening, which could explain effects seen in previous DIII-D ECH transport studies [K.W. Gentle “Electron energy transport inferences from modulated electron cyclotron heating in DIII-D,” Phys. Plasmas 13, 012311 (2006)]. By applying an FFT to the T{sub e} measurements from the University of Texas’s 40-channel ECE Radiometer, and using a simplified thermal transport equation, the flux surface extent of ECH deposition is determined. The Fourier method analysis is compared with a Break-In-Slope (BIS) analysis and predictions from the ray-tracing code TORAY. Examination of multiple Fourier harmonics and BIS fitting methods allow an estimation of modulated transport coefficients and thereby the true ECH deposition profile. Correlations between edge fluctuations and ECH deposition in legacy data are also explored as a step towards establishing a link between fluctuations and deposition broadening in DIII-D.

  16. Upgrades and Additions for the ECH System on DIII-D

    NASA Astrophysics Data System (ADS)

    Gorelov, Y. A.; Lohr, J.; Cengher, M.; Ponce, D.

    2013-10-01

    Six MW-class, 110 GHz gyrotrons have been in routine operation on DIII-D since 2008. One of these gyrotrons, which had low rf production and higher than normal collector power loading, failed due to a collector water leak. Nevertheless, the number of 110 GHz gyrotrons remained the same, as the first new 110 GHz CPI gyrotron with depressed collector potential design was installed and used in the 2013 experimental campaign. The DIII-D ECH transmission line system now comprises seven evacuated transmission lines up to 80 meters in length with transmission efficiencies from 69%-79% and four dual launchers. New stands are being fabricated and installed for two additional depressed collector gyrotrons, one with designed power of 1.2 MW at 110 GHz and the other with 1.5 MW at 117.5 GHz. One gyrotron was relocated to accommodate the new additions. High voltage power supplies, the water-cooling system and new waveguide lines for these gyrotrons are being built. One of the 110 GHz 1.0 MW gyrotrons in DIII-D was used as a source for heat exchanger tests. The rf beam was routed to a mobile test unit (MTU) trailer and shows expected expansion of beam radius vs distance from the waveguide end. Experiments were completed using 50-500 kW injected into the MTU lab at pulse lengths from 5-300 ms. Work supported by the US DOE under DE-FC02-04ER54698.

  17. Effect of thermoelectric current splitting on the magnetic topology in DIII-D

    SciTech Connect

    Wingen, A.; Spatschek, K. H.; Evans, T. E.

    2011-04-15

    An extended model is proposed to describe the magnetic topology during appearance of edge localized modes (ELMs). It is applied to an ELMing H-mode in a lower single null discharge at DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. The process of flux tube formation is discussed based on a previously proposed two-step model. Large currents, as found in measurements in DIII-D, are assumed running through newly formed large flux tubes. Two different realizations of the current distribution within the tubes are compared, namely a single filament in each tube and a scenario where the current in each tube is split into subfilaments. The latter scenario is shown to be the more realistic distribution because it leads to much better agreement with infrared camera observations. It is demonstrated that stripe patterns in the divertor heat flux produced by an ELM in the DIII-D tokamak can be reproduced numerically by taking into account the magnetic perturbation caused by the thermoelectric current subfilaments.

  18. ITER Baseline Scenario with ECCD Applied to Neoclassical Tearing Modes in DIII-D

    NASA Astrophysics Data System (ADS)

    Welander, A. G.; La Haye, R. J.; Lohr, J. M.; Humphreys, D. A.; Prater, R.; Paz-Soldan, C.; Kolemen, E.; Turco, F.; Olofsson, E.

    2015-11-01

    The neoclassical tearing mode (NTM) is a magnetic island that can occur on flux surfaces where the safety factor q is a rational number. Both m/n=3/2 and 2/1 NTM's degrade confinement, and the 2/1 mode often locks to the wall and disrupts the plasma. An NTM can be suppressed by depositing electron cyclotron current drive (ECCD) on the q-surface by injecting microwave beams into the plasma from gyrotrons. Recent DIII-D experiments have studied the application of ECCD/ECRH in the ITER Baseline Scenario. The power required from the gyrotrons can be significant enough to impact the fusion gain, Q in ITER. However, if gyrotron power could be minimized or turned off in ITER when not needed, this impact would be small. In fact, tearing-stable operation at low torque has been achieved previously in DIII-D without EC power. A vision for NTM control in ITER will be described together with results obtained from simulations and experiments in DIII-D under ITER like conditions. Work supported by the US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466, DE-FG02-04ER54761.

  19. Off-axis Neutral Beam Current Drive for Advanced Scenario Development in DIII-D

    SciTech Connect

    Murakami, M; Park, J; Petty, C; Luce, T; Heidbrink, W; Osborne, T; Wade, M; Austin, M; Brooks, N; Budny, R; Challis, C; DeBoo, J; deGrassie, J; Ferron, J; Gohil, P; Hobirk, J; Holcomb, C; Hollmann, E; Hong, R; Hyatt, A; Lohr, J; Lanctot, M; Makowski, M; McCune, D; Politzer, P; Prater, R; John, H S; Suzuki, T; West, W; Unterberg, E; Van Zeeland, M; Yu, J

    2008-10-13

    Modification of the two existing DIII-D neutral beam lines is proposed to allow vertical steering to provide off-axis neutral beam current drive (NBCD) as far off-axis as half the plasma radius. New calculations indicate very good current drive with good localization off-axis as long as the toroidal magnetic field, B{sub T}, and the plasma current, I{sub p}, are in the same direction (for a beam steered downward). The effects of helicity can be large: e.g., ITER off-axis NBCD can be increased by more than 20% if the B{sub T} direction is reversed. This prediction has been tested by an off-axis NBCD experiment using reduced size plasmas that are vertically shifted with the existing NBI on DIII-D. The existence of off-axis NBCD is evident in sawtooth and internal inductance behavior. By shifting the plasma upward or downward, or by changing the sign of the toroidal field, measured off-axis NBCD profiles, determined from MSE data, are consistent with predicted differences (40%-45%) arising from the NBI orientation with respect to the magnetic field lines. Modification of the DIII-D NB system will strongly support scenario development for ITER and future tokamaks as well as providing flexible scientific tools for understanding transport, energetic particles and heating and current drive.

  20. Extending DIII-D Neutral Beam Modulated Operations with a Camac Based Total on Time Interlock

    SciTech Connect

    Baggest, D.S.; Broesch, J.D.; Phillips, J.C.

    1999-11-01

    A new total-on-time interlock has increased the operational time limits of the Neutral Beam systems at DIII-D. The interlock, called the Neutral Beam On-Time-Limiter (NBOTL), is a custom built CAMAC module utilizing a Xilinx 9572 Complex Programmable Logic Device (CPLD) as its primary circuit. The Neutral Beam Injection Systems are the primary source of auxiliary heating for DIII-D plasma discharges and contain eight sources capable of delivering 20MW of power. The delivered power is typically limited to 3.5 s per source to protect beam-line components, while a DIII-D plasma discharge usually exceeds 5 s. Implemented as a hardware interlock within the neutral beam power supplies, the NBOTL limits the beam injection time. With a continuing emphasis on modulated beam injections, the NBOTL guards against command faults and allows the beam injection to be safely spread over a longer plasma discharge time. The NBOTL design is an example of incorporating modern circuit design techniques (CPLD) within an established format (CAMAC). The CPLD is the heart of the NBOTL and contains 90% of the circuitry, including a loadable, 1 MHz, 28 bit, BCD count down timer, buffers, and CAMAC communication circuitry. This paper discusses the circuit design and implementation. Of particular interest is the melding of flexible modern programmable logic devices with the CAMAC format.

  1. Customizable scientific web-portal for DIII-D nuclear fusion experiment

    NASA Astrophysics Data System (ADS)

    Abla, G.; Kim, E. N.; Schissel, D. P.

    2010-04-01

    Increasing utilization of the Internet and convenient web technologies has made the web-portal a major application interface for remote participation and control of scientific instruments. While web-portals have provided a centralized gateway for multiple computational services, the amount of visual output often is overwhelming due to the high volume of data generated by complex scientific instruments and experiments. Since each scientist may have different priorities and areas of interest in the experiment, filtering and organizing information based on the individual user's need can increase the usability and efficiency of a web-portal. DIII-D is the largest magnetic nuclear fusion device in the US. A web-portal has been designed to support the experimental activities of DIII-D researchers worldwide. It offers a customizable interface with personalized page layouts and list of services for users to select. Each individual user can create a unique working environment to fit his own needs and interests. Customizable services are: real-time experiment status monitoring, diagnostic data access, interactive data analysis and visualization. The web-portal also supports interactive collaborations by providing collaborative logbook, and online instant announcement services. The DIII-D web-portal development utilizes multi-tier software architecture, and Web 2.0 technologies and tools, such as AJAX and Django, to develop a highly-interactive and customizable user interface.

  2. Suppression of type-I ELMs with reduced RMP coil set on DIII-D

    DOE PAGESBeta

    Orlov, D. M.; Moyer, R. A.; Evans, T. E.; Paz-Soldan, Carlos; Ferraro, N. M.; Nazikian, R.; deGrassie, John S.; Grierson, Brian A.; Eldon, D.; Fenstermacher, Max E.; et al

    2016-02-19

    Recent experiments on DIII-D have demonstrated that having a toroidally-monochromatic spectral content of edge-resonant magnetic perturbations (RMPs) is not a necessary condition for suppression of edge localized modes (ELMs). Robust ELM suppression has been reproducibly obtained on DIII-D during experiments in which various non-axisymmetric coil loops were turned off pseudo-randomly producing a variety of n=1 , n=2, and n=3 spectral contributions. It was demonstrated that RMP ELM suppression could be achieved with as few as 5 out of 12 internal coil loops (I-coils) on DIII-D at similar coil currents and with good plasma confinement. Linear MHD plasma response (M3DC1, IPEC,more » MARS) and vacuum (SURFMN, TRIP3D) modelling have been performed in order to understand the effects of the perturbation spectrum on the plasma response and ELM suppression. The results suggest that reduction of the dominant n=3 perturbation field is compensated by increased n=2 field in the plasma that may lead to RMP ELM suppression at lower levels of n=3 perturbative magnetic flux from the I-coils. In conclusion, these results provide additional confidence that ITER may be capable of RMP ELM suppression in the event of multiple internal coil failures.« less

  3. DESIGN, FABRICATION, INSTALLATION AND TESTING OF IN-VESSEL CONTROL COILS FOR DIII-D

    SciTech Connect

    ANDERSON, PM; BAXI, CB; KELLMAN, AG; REIS, EE; ROBINSON, JI

    2002-10-01

    OAK A271 DESIGN, FABRICATION, INSTALLATION AND TESTING OF IN-VESSEL CONTROL COILS FOR DIII-D. Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. Recently, these coils have also demonstrated significant benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modeling has shown that substantial performance improvements can be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and have been tested successfully. installation of a set of twelve internal coils and magnetic sensors in the DIII-D tokamak is to be completed in December 2002. The design requirement for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils was in joining of both the copper conductor and the stainless tube without overheating the polyamide insulator.

  4. Modeling of Steady-State Non-Inductive ITB Discharges with Application to DIII-D

    NASA Astrophysics Data System (ADS)

    St John, H. E.; Lao, L. L.; Murakami, M.; Kinsey, J. E.

    2001-10-01

    Establishment of near steady-state high-performance discharges with internal transport barriers in the electron and ion heat and the toroidal momentum channels is investigated using the GLF23 and Weiland confinement models. A combination of neutral beam and electron cyclotron heating and current drive is used to optimally shape the current profile for near non-inductive steady-state operation. The GLF23 and Weiland confinement models have had some success in modeling DIII-D discharges and consequently represents our best choice for DIII-D AT scenario development at this time. By starting the modeling with actual high-performance DIII-D discharges, we expect to obtain experimentally realized results. The stability of our simulations is monitored with the BALOO and GATO codes and rf heating and current drive is modelled with TORAY-GA. This computationally instensive modeling approach requires concurrent computing methods in order to be used routinely. We discuss our efforts to date in producing a parallel computational transport environment.

  5. Modification of Sawteeth Periods By Trapped Fast Ions in DIII-D

    NASA Astrophysics Data System (ADS)

    Choi, M.; Chan, V. S.; Chu, M. S.; Lao, L. L.; Turnbull, A. D.

    2006-10-01

    The main auxiliary heating methods for ITER are neutral beam and ion cyclotron wave heating. Sawtooth physics is very important in optimizing the heating efficiency for ITER. This requires understanding of the interaction between fast ions and fast Alfvén wave (FW) on MHD stability. Experimentally, the DIII-D discharges have demonstrated strong acceleration of deuterium beam ions above the injected beam energy from measurements of enhanced neutron emissions during FW heating. Theory predicts that high pressure from fast ions in the center of plasma may act as a stabilizing kinetic effect on ideal internal kink mode. However, the DIII-D experimental results showed that sawteeth characteristics strongly depend on a combination of plasma and wave conditions. We apply a Monte-Carlo orbit code (ORBIT-RF) and ideal MHD code (GATO) to model existing DIII-D experiments and explore the triggering and stabilization mechanisms for sawteeth. The analytical model by Bussac and Porcelli will be compared with NOVA-K calculations.

  6. TESTS AND PERFORMANCE ON THE SIX GYROTRON SYSTEM ON THE DIII-D TOKAMAK

    SciTech Connect

    LOHR,J; GORELOV,YA; KAJIWARA,K; PONCE,D; CALLIS,RW; FERRON,JR; GREENFIELD,CM; LAHAYE,RJ; PINSKER,RI; PRATER,R; WADE,MR; ELLIS,RA

    2002-09-01

    A271 TESTS AND PERFORMANCE ON THE SIX GYROTRON SYSTEM ON THE DIII-D TOKAMAK. The DIII-D gyrotron complex for electron cyclotron heating (ECH) and electron cyclotron current drive (ECCD) has been expanded to include five operational gyrotrons with a sixth being commissioned. The generated rf power exceeds 4.0 MW and the transmission lines deliver about 80% of this power to the tokamak. Among the experiments performed during the most recent campaign, the installation has been used to achieve stabilization of the m/n=2/1 and 3/2 neoclassical tearing modes, to control the rate of current penetration early in the discharge, and to study ECCD contributions to advanced tokamak discharges with high bootstrap fractions. Observations of the thermal performance of CVD diamond output windows have shown good agreement with theoretical predictions, but have revealed a number of interesting phenomena connected with impurities, including points of visible light emission at hot spots identified by infrared measurements. One window was cleaned in situ by alumina grit blasting and Raman spectra verified the removal of some surface contamination. A significant new capability is the simultaneous control of the output power of the entire array of gyrotrons by the DIII-D plasma control system. This allows a predetermined electron temperature evolution to be followed at a specific location in the plasma and opens a new group of experimental possibilities leading to the achievement of higher levels of tokamak performance.

  7. 3 MW, 110 GHz ECH system for the DIII-D tokamak

    SciTech Connect

    Callis, R.W.; Lohr, J.; Ponce, D.; Harris, T.E.; O`Neill, R.C.; Remsen, D.B.; Prater, R.; Luce, T.C.

    1998-07-01

    To support the Advanced Tokamak (AT) operating regimes in the DIII-D tokamak, methods need to be developed to control the current and pressure profiles across the plasma discharge. In particular, AT plasmas require substantial off-axis current in contrast to normal tokamak discharges where the current peaks on-axis. An effort is under way to use Electron Cyclotron Current Drive (ECCD) as a method of sustaining the off-axis current in AT plasmas. The first step in this campaign is the installation of three megawatts of electron cyclotron heating power. This involves the installation of three rf systems operating at 110 GHz, the second harmonic resonance frequency on DIII-D, with each system generating nominally 1 MW. The three systems will use one GYCOM (Russian) gyrotron and two CPI (formerly Varian) gyrotrons, all with windowless evacuated corrugated low loss transmission lines. The first two of three 1 MW ECH systems is operating routinely at DIII-D with injected power at 110 GHz of approximately 1.5 MW with good power accountability. Transport experiments using modulated ECH have been performed confirming the power deposition location. On-axis and off-axis current drive experiments have been successfully performed with on-axis ECCD currents of 170 kA being observed.

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

    NASA Astrophysics Data System (ADS)

    Sung, C.; White, A. E.; Mikkelsen, D. R.; Greenwald, M.; Holland, C.; Howard, N. T.; Churchill, R.; Theiler, C.

    2016-04-01

    Long wavelength turbulent electron temperature fluctuations (kyρs < 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 (kyρs ≲ 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)].

  9. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1993--September 30, 1994

    SciTech Connect

    Lohr, J.

    1995-07-01

    The DIII-D tokamak research program is managed by General Atomics (GA) for the US Department of Energy (DOE). Major program participants include GA, Lawrence Livermore National Laboratory (LLNL), Oak Ridge National Laboratory (ORNL), and the University of California together with several other national laboratories and universities. The DIII-D is a moderate sized tokamak with great flexibility and extremely capable subsystems. The primary goal of the DIII-D tokamak research program is to provide data for development of a conceptual physics blueprint for a commercially attractive fusion power plant. In so doing, the DIII-D program provides physics and technology R&D output to aid the International Thermonuclear Experimental Reactor (ITER) and the Princeton Tokamak Physics Experiment (TPX) projects. Specific DIII-D objectives include the achievement of steady-state plasma current as well as the demonstration of techniques for radio frequency heating, divertor heat removal, particle exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion in plasmas with high beta and with high confinement. The long-range plan is organized with two principal elements, the development of an advanced divertor and the development of advanced tokamak concepts. These two elements have a common goal: an improved demonstration reactor (DEMO) with lower cost and smaller size than present DEMO concepts. In order to prepare for this long-range development, in FY94 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak studies, and Tokamak Physics.

  10. Particle and Energy Transport in the SOL of DIII-D and NSTX

    SciTech Connect

    Boedo, J; Maqueda, R; Rudakov, D; McKee, G; Kugel, H; Maingi, R; Crocker, N; Moyer, R; Soukhanovskii, V; Menard, J; Watkins, J; Zweben, S; D'Ippolito, D; Evans, T; Fenstermacher, M; Groth, M; Hollmann, E; Lasnier, C; Myra, J; Roquemore, L; West, W; Zeng, L

    2006-10-09

    The far scrape-off layer (SOL) radial transport and plasma-wall contact is mediated by intermittent and ELM-driven transport. Experiments to characterize the intermittent transport and ELMs have been performed in both DIII-D and NSTX under similar conditions. Both intermittent transport and ELMs are comprised of filaments of hot, dense plasma (n{sub e} {approx} 1 x 10{sup 13} cm{sup -3}, T{sub e} {approx} 400 eV) originating at the edge, transport both particles and heat into the SOL by convection, increasing wall interaction and causing sputtering and impurity release. Both intermittent filaments and ELMs leave the pedestal region at speeds of {approx}0.5-3 km/s, losing heat and particles by parallel transport as they travel through the SOL. The intermittency shows many similarities in NSTX and DIII-D, featuring similar size (2-5 cm), large convective radial velocity, ''holes'' inside and peaks outside the LCFS which quickly decay and slow down with radius. Whereas in DIII-D the intermittency decays in both intensity and frequency in H-mode, it chiefly decays in frequency in NSTX. In the low collisionality (v* = {pi}R{sub q{sub 95}}/{lambda}C) (v* {approx} 0.1, N{sub G} {approx} 0.3) case, the ELMs impact the walls quite directly and account for {approx}90% of the wall particle flux, decreasing to {approx}30% at (v* {approx} 1.0, N{sub G} > 0.6).

  11. Implementation of a quasi-realtime display of DIII-D neutral beam heating waveforms

    SciTech Connect

    Phillips, J.C.

    1993-10-01

    The DIII-D neutral beam system employs eight 80 keV ion sources mounted on four beamlines to provide plasma heating to the DIII-D tokamak. The neutral beam system is capable of injecting over 20 MW of deuterium power with flexibility in terms of timing and modulation of the individual neutral beams. To maintain DIII-D`s efficient tokamak shot cycle and make informed control decisions, it is important to be able to determine which beams fired, and exactly when, by the time the tokamak shot is over. Previously this information was available in centralized form only after a several minute wait. A cost-effective alternative to the traditional eight-channel storage oscilloscope has been implemented using off the shelf PC hardware and software. The system provides a real time display of injected neutral beam accelerator voltages and tokamak plasma current, as well an a summation waveform indicative of the total injected power as a function of time. The hardware consists of a Macintosh Centris 650 PC with a Motorola 68040 microprocessor. Data acquisition is accomplished using a National Instrument`s 16-channel analog to digital conversion board for the Macintosh. The color displays and functionality were developed using National Instruments` LabView environment. Because the price of PCs has been decreasing rapidly and their capabilities increasing, this system is far less expensive than an eight-channel storage oscilloscope. As a flexible combination of PC and software, the system also provides much more capability than a dedicated oscilloscope, acting as the neutral beam coordinator`s logbook, recording comments and availability statistics. Data such as shot number and neutral beam parameters are obtained over the local network from other computers and added to the display. Waveforms are easily archived to disk for future recall. Details of the implementation will be discussed along with samples of the displays and a description of the system`s function and capabilities.

  12. LONG PULSE ADVANCED TOKAMAK DISCHARGES IN THE DIII-D TOKAMAK

    SciTech Connect

    P.I. PETERSEN

    2002-06-01

    One of the main goals for the DIII-D research program is to establish an advanced tokamak plasma with high bootstrap current fraction that can be sustained in-principle steady-state. Substantial progress has been made in several areas during the last year. The resistive wall mode stabilization has been done with spinning plasmas in which the plasma pressure has been extended well above the no-wall beta limit. The 3/2 neoclassical tearing mode has been stabilized by the injection of ECH into the magnetic islands, which drives current to substitute the missing bootstrap current. In these experiments either the plasma was moved or the toroidal field was changed to overlap the ECCD resonance with the location of the NTMs. Effective disruption mitigation has been obtained by massive noble gas injection into shots where disruptions were deliberately triggered. The massive gas puff causes a fast and clean current quench with essentially all the plasma energy radiated fairly uniformly to the vessel walls. The run-away electrons that are normally seen accompanying disruptions are suppressed by the large density of electrons still bound on the impurity nuclei. Major elements required to establish integrated, long-pulse, advanced tokamak operations have been achieved in DIII-D: {beta}{sub T} = 4.2%, {beta}{sub p} = 2, f{sub BS} = 65%, and {beta}{sub N}H{sub 89} = 10 for 600 ms ({approx} 4{tau}{sub E}). The next challenge is to integrate the different elements, which will be the goal for the next five years when additional control will be available. Twelve resistive wall mode coils are scheduled to be installed in DIII-D during the summer of 2003. The future plans include upgrading the tokamak pulse length capability and increasing the ECH power, to control the current profile evolution.

  13. DIII-D electron cyclotron heating 2 MW upgrade project. Final report, FY1989--FY1997

    SciTech Connect

    Callis, R.W.

    1997-08-01

    The 2 MW, 110 GHz ECH system was based on the General Atomics Proposal to the Department of Energy: DIII-D Fusion Research Program Vol. I Technical, and Vol. II Cost (GACP-72-166, July 1987 and revised). This proposal was reviewed in August 1987 by a senior technical review committee, who recommended to vigorously pursue increasing the ECH power to 6 MW. The realization of the higher frequency and power ECH on DIII-D was recognized by the committee to be important, not only for the DIII-D program, but also for future devices and the whole ECH area. Subsequently, an engineering cost and schedule review was conducted by DOE-OAK which confirmed the GA costs and schedules and recommended proceeding directly to 10 MW. However, because of budgetary constraints, in the April 1988 Field Task Proposal submission, GA proposed a phased ECH approach, Phase I being 2 MW and Phase II increasing the power to 10 MW. After review, DOE instructed GA to initiate the prototype 2 MW, 110 GHz program. The contract to procure four 500 kW, 110 GHz, 10 s gyrotrons from Varian Associates was initiated in April 1989 with final delivery by November 1990. Because of difficulties in spreading the energy of the electron beam over the collector area, the testing of the first gyrotron delayed its delivery until February 1991. The second gyrotron was able to operate for 1 s at 500 kW and 2 s at 300 kW, but failed when the cavity suffered thermal damage.

  14. 4 MW upgrade to the DIII-D fast wave current drive system

    SciTech Connect

    deGrassie, J.S.; Pinsker, R.I.; Cary, W.P.

    1993-10-01

    The DIII-D fast wave current drive (FWCD) system is being upgraded by an additional 4 MW in the 30 to 120 MHz frequency range. This capability adds to the existing 2 MW 30 to 60 MHz system. Two new ABB transmitters of the type that are in use on the ASDEX-Upgrade tokamak in Garching will be used to drive two new water-cooled four-strap antennas to be installed in DIII-D in early 1994. The transmission and tuning system for each antenna will be similar to that now in use for the first 2 MW system on DIII-D, but with some significant improvements. One improvement consists of adding a decoupler element to counter the mutual coupling between the antenna straps which results in large imbalances in the power to a strap for the usual current drive intrastrap phasing of 90{degrees}. Another improvement is to utilize pressurized, ceramic-insulated transmission lines. The intrastrap phasing will again be controlled in pairs, with a pair of straps coupled in a resonant loop configuration, locking their phase difference at either 0 or 180{degrees}, depending upon the length of line installed. These resonant loops will incorporate a phase shifter so that they will be able to be tuned to resonance at several frequencies in the operating band of the transmitter. With the frequency change capability of the ABB generators, the FWCD frequency will thus be selectable on a shot-to-shot basis, from this preselected set of frequencies. The schedule is for experiments to begin with this added 4 MW capability in mid-1994. The details of the system are described.

  15. Fabrication and installation of the DIII-D radiative divertor structures

    SciTech Connect

    Hollerbach, M.A.; Smith, J.P.

    1997-11-01

    Phase 1A of the Radiative Divertor Program (RDP) is now installed in the DIII-D tokamak located at General Atomics. This hardware was added to enhance both the Divertor and Advanced Tokamak research elements of the DIII-D program. This installation consists of a divertor baffle enveloping a cryocondensation pump at the upper outer divertor target of DIII-D. The divertor baffle consists of two toroidally continuous Inconel 625 water-cooled rings and a toroidal array of discontinuous radiatively-cooled plates. The water-cooled rings are each comprised of four quadrants, mechanically formed, chem.-milled, and resistance and TIG welded Inconel 625 panels. The supports attaching the panels to the vessel wall are designed to accommodate the differential thermal expansion between the rings and vessel during bake and to react the electromagnetic loads induced during disruptions. They are made from either Inconel 625 or Inconel 718 depending on the stress levels predicted in Finite Element Analysis. Gas seals are designed to limit the leakage from the baffle chamber back to the core plasma to 2,500 {ell}/s and incorporate plasma sprayed alumina to minimize currents flowing through them. The bulk of the water-cooled ring fabrication was performed by a vendor, however, the final machining of penetrations in the conical ring for diagnostic access was performed in-house using a unique machining configuration. This configuration, and the machining of the diagnostic cutouts is described. Graphite tiles were machined from ATJ graphite to form a smooth plasma-facing surface. The installation of all divertor components required only four weeks.

  16. Multi-megawatt 110 GHz ECH system for the DIII-D tokamak

    SciTech Connect

    Callis, R.W.; Lohr, J.; O`Neill, R.C.; Ponce, D.; Prater, R.

    1997-11-01

    Two 110 GHz gyrotrons with nominal output power of 1 MW each have been installed on the DIII-D tokamak. The first 110 GHz gyrotron built by Gycom has a nominal rating of 1 MW and a 2s pulse length, with the pulse length being determined by the maximum temperature allowed on the edge cooled boron nitride window. This gyrotron was first operated into the DIII-D tokamak in late 1996. The second gyrotron was built by Communications and Power Industries (CPI) was commissioned during the spring of 1997. The CPI gyrotron uses a double disc FC-75 cooled sapphire window which has a pulse length rating of 0.8s at 1 MW, 2s at 0.5 MW and 10s at 0.2 MW. Both gyrotrons are connected to the tokamak by a low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HE(11) mode. Using short pulse lengths to avoid breakdown inside the air filled waveguide, the microwave beam has been measured inside the DIII-D vacuum vessel using a paper target and an IR camera. The resultant microwave beam was found to be well focused with a spot size of approximately 8 cm. The beam can be steered poloidially from the center to the outer edge of the plasma. The initial operation of the Gycom gyrotron with about 0.5 MW delivered to a low density plasma for 0.5 s showed good central electron heating, with peak temperature in excess of 10 keV. A third gyrotron, being built by CPI, will be installed later this year. Progress with the first CPI tube will also be discussed and future plans for the ECH installation and physics experiments will be presented.

  17. Global Geodesic Acoustic Modes Driven by Energetic Particles in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Nazikian, R.; Fu, G. Y.; Gorelenkov, N. N.; Kramer, G. J.; Austin, M. E.; Berk, H. L.; Heidbrink, W. W.; McKee, G. R.; Shafer, M. W.; Strait, E. J.; van Zeeland, M. A.

    2009-11-01

    Intense axisymmetric oscillations driven by suprathermal passing ions injected in the direction counter to the toroidal plasma current are observed in the DIII-D tokamak. Strong bursting and frequency chirping coincide with large (10-15%) drops in the neutron emission, suggesting that the mode is very effective in displacing beam ions from the plasma core. BES measurements of density fluctuations indicate an outward propagating mode of large radial extent. The large density to temperature ratio of the mode confirms a dominant compressional contribution to the pressure perturbation, indicative of the Geodesic Acoustic Mode (GAM).

  18. DIII-D Edge Plasma, Disruptions, and Radiative Processes. Final Report

    SciTech Connect

    Boedo, J. A.; Luckhardt, S.C.; Moyer, R. A.

    2001-01-01

    The scientific goal of the UCSD-DIII-D Collaboration during this period was to understand the coupling of the core plasma to the plasma-facing components through the plasma boundary (edge and scrape-off layer). To achieve this goal, UCSD scientists studied the transport of particles, momentum, energy, and radiation from the plasma core to the plasma-facing components under normal (e.g., L-mode, H-mode, and ELMs), and off-normal (e.g., disruptions) operating conditions.

  19. Analysis of plasma coupling with the prototype DIII-D ICRF antenna

    SciTech Connect

    Ryan, P.M.; Hoffman, D.J.; Bigelow, T.S.; Baity, F.W.; Gardner, W.L.; Mayberry, M.J.; Rothe, K.E.

    1988-01-01

    Coupling to plasma in the H-mode is essential to the success of future ignited machines such as CIT. To ascertain voltage and current requirements for high-power second harmonic heating (2 MW in a 35- by 50-cm port), coupling to the DIII-D tokamak with a prototype compact loop antenna has been measured. The results show good loading for L-mode and limiter plasmas, but coupling 2 MW to an H-mode plasma demands voltages and currents near the limit of present technology. We report the technological analysis and progress that allow coupling of these power densities. 5 refs., 4 figs.

  20. Radial transport and electron-cyclotron-current drive in the TCV and DIII-D tokamaks.

    PubMed

    Harvey, R W; Sauter, O; Prater, R; Nikkola, P

    2002-05-20

    Calculation of electron-cyclotron-current drive (ECCD) with the comprehensive CQL3D Fokker-Planck code for a TCV tokamak shot gives 550 kA of driven toroidal current, in marked disagreement with the 100-kA experimental value. Published ECCD efficiencies calculated with CQL3D in the much larger, higher-confinement DIII-D tokamak are in excellent agreement with experiment. The disagreement is resolved by including in the calculations electrostatic-type radial transport at levels given by global energy confinement in tokamaks. The radial transport of energy and toroidal current are in agreement. PMID:12005571

  1. Current driven due to localized electron power deposition in DIII-D

    SciTech Connect

    Harvey, R.W.; Lin-Liu, Y.R.; Luce, T.C.; Prater, R.; Sauter, O.; Smirnov, A.P.

    1999-05-01

    Due to spatial localization of electron cyclotron wave injection in DIII-D, electrons heated in an off-axis region must toroidally transit the tokamak 25--50 times before re-entering the heating region. This distance is of the order of the mean free path. The effect of such RF localization is simulated with a time-dependent Fokker-Planck code which is 2D-in-velocity, 1D-in-space-along-B, and periodic in space. An effective parallel electric field arises to maintain continuity of the driven current. Somewhat surprisingly, the localized current drive efficiency remains equal to that for a uniform medium.

  2. Recent results from the 60 GHz inside launch ECH system on the DIII-D tokamak

    SciTech Connect

    Luce, T.C.; Petty, C.C.; Prater, R.; Harvey, R.W.; Lin-Liu, Y.R.; Lohr, J.; Matsuda, K.; Moeller, C.P. ); James, R.A. )

    1992-11-01

    Electron cyclotron heating (ECH) is a useful tool in global transport and local confinement studies. Operational experience with the inside launch ECH system on DIII-D shows that reliable operations are possible with power densities up to 0.7 GW/m[sup 2] in vacuum waveguide. Global confinement is roughly predicted by the Rebut-Lallia or ITER-89P scaling law, but direct analysis indicates a nearly linear scaling with toroidal field not found in these scaling laws. Local transport studies with off-axis heating clearly show inward transport in the electron fluid. This implies that diffusive and critical gradient models cannot completely describe plasma transport.

  3. Recent results from the 60 GHz inside launch ECH system on the DIII-D tokamak

    SciTech Connect

    Luce, T.C.; Petty, C.C.; Prater, R.; Harvey, R.W.; Lin-Liu, Y.R.; Lohr, J.; Matsuda, K.; Moeller, C.P.; James, R.A.

    1992-11-01

    Electron cyclotron heating (ECH) is a useful tool in global transport and local confinement studies. Operational experience with the inside launch ECH system on DIII-D shows that reliable operations are possible with power densities up to 0.7 GW/m{sup 2} in vacuum waveguide. Global confinement is roughly predicted by the Rebut-Lallia or ITER-89P scaling law, but direct analysis indicates a nearly linear scaling with toroidal field not found in these scaling laws. Local transport studies with off-axis heating clearly show inward transport in the electron fluid. This implies that diffusive and critical gradient models cannot completely describe plasma transport.

  4. Generation of Localized Noninductive Current by Electron Cyclotron Waves on the DIII-D Tokamak

    SciTech Connect

    Luce, T. C.; Lin-Liu, Y. R.; Harvey, R. W.; Giruzzi, G.; Politzer, P. A.; Rice, B. W.; Lohr, J. M.; Petty, C. C.; Prater, R.

    1999-11-29

    Localized currents due to electron cyclotron current drive have been measured for the first time in experiments on the DIII-D tokamak. The location of driven current in the plasma has been varied from near the center of the tokamak out to half of the minor radius. The measured current drive efficiency agrees with quasilinear Fokker-Planck calculations near the center and exceeds the predicted value with increasing minor radius. Reduction of the trapped electron fraction due to finite collisionality is a leading candidate to explain the discrepancy. (c) 1999 The American Physical Society.

  5. Features and Initial Results of the DIII-D Advanced Tokamak Radiative Divertor

    SciTech Connect

    R.C. O'Neill; A.S. Bozek; M.E. Friend; C.B. Baxi; E.E. Reis; M.A. Mahdavi; D.G. Nilson; S.L. Allen; W.P. West

    1999-11-01

    The Radiative Divertor Program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryobelium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma. The baffles, which consist of water-cooled panels that allow for the attachment of tiles of various sizes and shapes, house gas puff systems. The intent of the puffing systems is to inject gas in and around the divertor to minimize the heat flux on specific areas on the divertor and its components. The reduction of the heat flux on the divertor minimizes the impurities that are generated from excess heat on divertor components, specifically tiles. Experiments involving the gas puff systems and the divertor structures have shown the heat flux can be spread over a large area of the divertor, reducing the peak heat flux in specific areas. The three diverters also incorporate a variety of diagnostic tools such as halo current monitors, magnetic probes and thermocouples to monitor certain plasma characteristics as well as determine the effectiveness of the cryopumps and baffle configurations. The diverters were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo currents and toroidal induced currents. Incorporated also

  6. Rapid Inward Impurity Transport during Impurity Pellet Injection on the DIII-D Tokamak

    SciTech Connect

    Evans, T.E.; Hyatt, A.W.; Lee, R.L.; Kellman, A.G.; Parks, P.B.; Stockdale, R.; Taylor, P.L.; Whyte, D.G.; Jernigan, T.C.

    1998-11-01

    Neon killer pellets are injected into the DIII-D tokamak plasma in order to radiatively quench the plasma{close_quote}s stored energy and mitigate disruption effects. Inward radial transport on the time scale of the pellet ablation ({le}1 ms) results in central deposition of the neon inside the ablation penetration radius of the pellet, causing effective radiative energy dissipation. This result is in contrast to the radially outward deposition measured for fueling (hydrogenic) pellets. The observed magnitudes of magnetic fluctuations ({delta}B/B{approximately}0.2{percent}) are shown to be capable of causing the radial transport. {copyright} {ital 1998} {ital The American Physical Society }

  7. Measurements of carbon and tungsten erosion/deposition in the DIII-D divertor

    SciTech Connect

    Bastasz, R.; Wampler, W.R.; Cuthbertson, J.W.; Buchenauer, D.A.; Brooks, N.; Junge, R.; West, W.P.; Wong, C.P.C.

    1994-08-01

    Net erosion/deposition rates of carbon and tungsten were measured at the outer strike point of the divertor plasma on the floor of the DIII-D tokamak during deuterium H-mode operation at a peak power deposition of about 40 W/cm{sup 2}. For carbon, net erosion rates of up to 4 nm/s were found. For a tungsten film, no appreciable erosion was detected. However, measurements of deposited tungsten on adjacent carbon surfaces indicated a net W erosion rate of 0.06 nm/s.

  8. Environmental Assessment for the proposed modification and continued operation of the DIII-D facility

    SciTech Connect

    1995-07-01

    The EA evaluates the proposed action of modifying the DIII-D fusion facility and conducting related research activities at the GA San Diego site over 1995-1999 under DOE contract number DE-ACO3-89ER51114. The proposed action is need to advance magnetic fusion research for future generation fusion devices such as ITER and TPX. It was determined that the proposed action is not a major action significantly affecting the quality of the human environment according to NEPA; therefore a finding of no significant impact is made and an environmental impact statement is not required.

  9. Spatially resolved measurements of two-dimensional turbulent structures in DIII-D plasmas

    NASA Astrophysics Data System (ADS)

    Zemedkun, S. E.; Che, S.; Chen, Y.; Domier, C. W.; Luhmann, N. C.; Munsat, T.; Parker, S. E.; Tobias, B.; Wan, W.; Yu, L.

    2015-12-01

    Two-dimensional observations of spatially coherent electron temperature fluctuations at drift-wave scales (k ˜ 1 cm-1) have been made using the electron cyclotron emission imaging diagnostic on the DIII-D tokamak. These measurements enable the extraction of spectral properties, including poloidal dispersion relations. Temperature fluctuation levels are found to be Te ˜/⟨Te⟩=1.2 % , and the phase velocity of the fluctuations is found to be constant across frequencies, consistent with modes having real frequencies low compared to the rotation-induced Doppler shifts. Comparisons with radially global linear gyrokinetic simulations suggest that the observed modes may be trapped electron modes.

  10. Correlation Ece Measurements of Turbulent Electron Temperature Fluctuations in Diii-D

    NASA Astrophysics Data System (ADS)

    White, A. E.; Peebles, W. A.; Rhodes, T. L.; Wang, G.; Schmitz, L.; Carter, T. A.; Hillesheim, J. C.; Doyle, E. J.; Zeng, L.; Holland, C. H.; McKee, G. R.; Staebler, G. M.; Waltz, R. E.; Candy, J.; Deboo, J. C.; Petty, C. C.; Burrell, K. H.

    2011-02-01

    This paper describes measurements of long wavelength, turbulent electron temperature fluctuations in the core plasma of the DIII-D tokamak made with a correlation electron cyclotron emission (CECE) radiometer-based diagnostic. Experimental and simulation results indicate that long wavelength electron temperature fluctuations (1) are similar in amplitude and spectrum to density fluctuations, (2) can be associated with both ITG and TEM turbulence, (3) exhibit changes in the relative fluctuation level that correlate with changes in electron thermal transport, and (4) are correlated, but out of phase, with density fluctuations measured simultaneously with reflectometry.

  11. HIGH POWER LONG PULSE PERFORMANCE OF THE DIII-D GYROTRON INSTALLATION

    SciTech Connect

    J. LOHR; Y.A. GORELOV; R.W. CALLIS; H.J. GRUNLOH; J.J. PEAVY; R.I. PINSKER; D. PONCE; R. PRATER; R.A. ELLIS,III

    2002-05-01

    At DIII-D, five 110 GHz gyrotrons are operating routinely for 2.0 s pulses at generated power levels {ge}750 kW per gyrotron. A sixth gyrotron is being installed, which should bring the generated power level to >4 MW and the injected power to about 3.0 MW. The output power now can be modulated by the plasma control system to fix T{sub e} at a desired value. The system is being used as a tool for control of current diffusion, for current profile control and other experiments leading to advanced tokamak operation.

  12. Experiments on ion cyclotron damping at the deuterium fourth harmonic in DIII-D

    SciTech Connect

    Pinsker, R.I.; Petty, C.C.; Baity, F.W.; Bernabei, S.; Greenough, N.; Heidbrink, W.W.; Mau, T.K.; Porkolab, M.

    1999-05-01

    Absorption of fast Alfven waves by the energetic ions of an injected beam is evaluated in the DIII-D tokamak. Ion cyclotron resonance absorption at the fourth harmonic of the deuteron cyclotron frequency is observed with deuterium neutral beam injection (f = 60 MHz, B{sub T} = 1.9 T). Enhanced D-D neutron rates are evidence of absorption at the Doppler-shifted cyclotron resonance. Characteristics of global energy confinement provide further proof of substantial beam acceleration by the rf. In many cases, the accelerated deuterons cause temporary stabilization of the sawtooth (monster sawteeth), at relatively low rf power levels of {approximately}1 MW.

  13. TOROID ROTATION IN NEUTRAL BEAM HEATED DISCHARGES IN DIII-D

    SciTech Connect

    J.S. deGRASSIE; D.R. BAKER; K.H. BURRELL; P. GOHIL; C.M. GREENFIELD; R.J. GROEBNER; D.M. THOMAS

    2002-08-01

    It is known that the toroidal angular momentum and the ion thermal energy are correlated in tokamak discharges heated by neutral beam injection. Here, data from ten years of measurements on DIII-D are considered, for representative discharges from all types and all conditions. The ratio of simple replacement times for momentum and energy is found to order this correlation indicating that these times are approximately equal, across the minor radius. Representative discharges of several types are discussed in more detail, as well as transport analysis results for the momentum and thermal ion diffusivities.

  14. LAUNCHER PERFORMANCE AND THERMAL CAPABILITY OF THE DIII-D ECH SYSTEM

    SciTech Connect

    KAJIWARA,K; LOHR,J; GORELOV,I.A; GREEN,M.T; PONCE,D; CALLIS,R.W; ELLIS,R.A

    2003-10-01

    OAK-B135 The temperatures of components of DIII-D ECH launchers were observed during 2003 tokamak operation. The injected power was typically 500-700 kW and the pulse length was typically 2s. Plasma shots were performed at intervals of about 17 min from 9 a.m. to 5 p.m. The temperatures of a movable mirror, a fixed mirror and a launcher reached an equilibrium after about six hours of repetitive pulsing. The saturation temperature depends to some extent on the plasma stored energy. However, even in high {beta} plasma, the temperatures plateaued at acceptable values.

  15. CONTROL OF THE INJECTED RF BEAM IN THE DIII-D ECH SYSTEM

    SciTech Connect

    K. KAJIWARA; C.B. BAXI; J.L. DOANE; R.E. ELLIS; M.E. FRIEND; M. GREEN; Y.A. GORELOV; J. LOHR; C.P. MOELLER; R.I. PINSKER; D. PONCE; R. PRATER

    2002-08-01

    The DIII-D ECH system includes three launcher assemblies each of which can accommodate the rf beams from two gyrotrons. The launchers use four different designs for the mirrors which focus and direct the beams into the tokamak. The designs use molybdenum brazed to graphite, thin Glidcop or variable thickness Glidcop. A fourth design with laminated Glidcop/stainless steel construction has been operated, but no thermal data are available. All the mirrors operate without active cooling. This paper presents preliminary analyses and measurements of the thermal performance of the three designs for which data have been obtained.

  16. Performance Assessment of the C-Mod Multi-Spectral Line Polarization MSE (MSE-MSLP) Diagnostic

    NASA Astrophysics Data System (ADS)

    Scott, Steven; Mumgaard, Robert; Khoury, Matthew

    2015-11-01

    The accuracy of the Alcator C-Mod Motional Stark Effect (MSE) diagnostic is limited primarily by partially polarized background light that varies rapidly both in time (1 ms) and space - factor 10 variations are observed between adjacent spatial channels. ITER is likely to operate in a similar regime. Visible Bremsstrahlung, divertor molecular D2 emission, and glowing invessel structures generate unpolarized light that becomes partially polarized upon reflection. Because all three sources are broadband, the background light can be measured in real-time at wavelengths close to the MSE spectrum, thereby allowing the background to be interpolated in wavelength rather than in time. A 10-spatial-channel, 4-wavelength MSE-MSLP system has been developed using polarization polychromators that measure simultaneously the MSE pi- and sigma- lines as well as two nearby wavelengths that were chosen to avoid both the MSE spectrum and all known impurity lines on each sightline. Initial performance evaluation indicates that the background channel measurements faithfully track the background light in the pi- and sigma- lines. The improvement in accuracy of pitch-angle measurements and increased diagnostic flexibility over a wide range of plasma conditions will be reported. This work is supported by USDoE awards DE-FC02-99ER54512 and DE-AC02-09CH11466.

  17. High-performance finite-difference time-domain simulations of C-Mod and ITER RF antennas

    NASA Astrophysics Data System (ADS)

    Jenkins, Thomas G.; Smithe, David N.

    2015-12-01

    Finite-difference time-domain methods have, in recent years, developed powerful capabilities for modeling realistic ICRF behavior in fusion plasmas [1, 2, 3, 4]. When coupled with the power of modern high-performance computing platforms, such techniques allow the behavior of antenna near and far fields, and the flow of RF power, to be studied in realistic experimental scenarios at previously inaccessible levels of resolution. In this talk, we present results and 3D animations from high-performance FDTD simulations on the Titan Cray XK7 supercomputer, modeling both Alcator C-Mod's field-aligned ICRF antenna and the ITER antenna module. Much of this work focuses on scans over edge density, and tailored edge density profiles, to study dispersion and the physics of slow wave excitation in the immediate vicinity of the antenna hardware and SOL. An understanding of the role of the lower-hybrid resonance in low-density scenarios is emerging, and possible implications of this for the NSTX launcher and power balance are also discussed. In addition, we discuss ongoing work centered on using these simulations to estimate sputtering and impurity production, as driven by the self-consistent sheath potentials at antenna surfaces.

  18. Aspect ratio effects on neoclassical tearing modes from comparison between DIII-D and National Spherical Torus Experiment

    NASA Astrophysics Data System (ADS)

    La Haye, R. J.; Buttery, R. J.; Gerhardt, S. P.; Sabbagh, S. A.; Brennan, D. P.

    2012-06-01

    Neoclassical tearing mode islands are sustained by helically perturbed bootstrap currents arising at finite beta from toroidal effects that trap a fraction of the particles in non-circulating orbits. DIII-D and NSTX are here operated with similar shape and cross-sectional area but almost a factor of two difference in inverse aspect ratio a /R. In these experiments, destabilized n =1 tearing modes were self-stabilized (reached the "marginal point") by reducing neutral-beam power and thus beta. The measure of the marginal island gives information on the small-island stabilizing physics that in part (with seeding) governs onset. The marginal island width on NSTX is found to be about three times the ion banana width and agrees with that measured in DIII-D, except for DIII-D modes closer to the magnetic axis, which are about two times the ion banana width. There is a balance of the helically perturbed bootstrap term with small island effects with the sum of the classical and curvature terms in the modified Rutherford equation for tearing-mode stability at the experimental marginal point. Empirical evaluation of this sum indicates that while the stabilizing effect of the curvature term is negligible in DIII-D, it is important in NSTX. The mode temporal behavior from the start of neutral-beam injection reduction also suggests that NSTX operates closer to marginal classical tearing stability; this explains why there is little hysteresis in beta between mode onset, saturation, and self-stabilization (while DIII-D has large hysteresis in beta). NIMROD code module component calculations based on DIII-D and NSTX reconstructed experimental equilibria are used to diagnose and confirm the relative importance of the stabilizing curvature effect, an advantage for low aspect ratio; the relatively greater curvature effect makes for less susceptibility to NTM onset even if the classical tearing stability index is near marginal.

  19. Aspect ratio effects on neoclassical tearing modes from comparison between DIII-D and National Spherical Torus Experiment

    SciTech Connect

    La Haye, R. J.; Buttery, R. J.; Gerhardt, S. P.; Sabbagh, S. A.; Brennan, D. P.

    2012-06-15

    Neoclassical tearing mode islands are sustained by helically perturbed bootstrap currents arising at finite beta from toroidal effects that trap a fraction of the particles in non-circulating orbits. DIII-D and NSTX are here operated with similar shape and cross-sectional area but almost a factor of two difference in inverse aspect ratio a/R. In these experiments, destabilized n=1 tearing modes were self-stabilized (reached the 'marginal point') by reducing neutral-beam power and thus beta. The measure of the marginal island gives information on the small-island stabilizing physics that in part (with seeding) governs onset. The marginal island width on NSTX is found to be about three times the ion banana width and agrees with that measured in DIII-D, except for DIII-D modes closer to the magnetic axis, which are about two times the ion banana width. There is a balance of the helically perturbed bootstrap term with small island effects with the sum of the classical and curvature terms in the modified Rutherford equation for tearing-mode stability at the experimental marginal point. Empirical evaluation of this sum indicates that while the stabilizing effect of the curvature term is negligible in DIII-D, it is important in NSTX. The mode temporal behavior from the start of neutral-beam injection reduction also suggests that NSTX operates closer to marginal classical tearing stability; this explains why there is little hysteresis in beta between mode onset, saturation, and self-stabilization (while DIII-D has large hysteresis in beta). NIMROD code module component calculations based on DIII-D and NSTX reconstructed experimental equilibria are used to diagnose and confirm the relative importance of the stabilizing curvature effect, an advantage for low aspect ratio; the relatively greater curvature effect makes for less susceptibility to NTM onset even if the classical tearing stability index is near marginal.

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

    NASA Astrophysics Data System (ADS)

    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.

  1. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1990--September 30, 1991. Magnetic Fusion Research Program

    SciTech Connect

    Simonen, T.C.; Evans, T.E.

    1992-03-01

    This report discusses the following topics on Doublet-3 research operations: DIII-D Program Overview; Boundary Plasma Research Program/Scientific Progress; Radio Frequency Heating and Current Drive; Core Physics; DIII-D Operations; Program Development; Support Services; ITER Contributions; Burning Plasma Experiment Contributions; and Collaborative Efforts.

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

    NASA Astrophysics Data System (ADS)

    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.

  3. Neoclassical tearing modes in DIII-D and calculations of the stabilizing effects of localized electron cyclotron current drive

    SciTech Connect

    Prater, R.; Bernabei, S.; Harvey, R. W.; La Haye, R. J.; Lin-Liu, Y. R.; Lohr, J.; Perkins, F. W.; Wong, K.-L.

    1999-09-20

    Neoclassical tearing modes are found to limit the achievable beta in many high performance discharges in DIII-D. Electron cyclotron current drive within the magnetic islands formed as the tearing mode grows has been proposed as a means of stabilizing these modes or reducing their amplitude, thereby increasing the beta limit by a factor around 1.5. Some experimental success has been obtained previously on Asdex-U. Here we examine the parameter range in DIII-D in which this effect can best be studied. (c) 1999 American Institute of Physics.

  4. Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code

    SciTech Connect

    Evans, T.E.; Mahdavi, M.A.; Sager, G.T.; West, W.P.; Fenstermacher, M.E.; Meyer, W.H.; Porter, G.D.

    1995-07-01

    A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DIII-D divertor is discussed. MCI simulation results are compared to experimental DIII-D carbon measurements.

  5. Thermal analysis and design of polarizer mirrors for waveguides in the DIII-D tokamak

    SciTech Connect

    Baxi, C.B.; Doane, J.L.; Sevier, D.L.

    1992-07-01

    A new high power electron cyclotron heating (ECH) system designed to operate at 110 GHz with a power output of 2 MW has been introduced on DIII-D. All components of the system are capable of handling a 10 second pulse at an interval of 10 minutes. Transmission of ECH power from the the source (a millimeter-wave gyrotron) to the plasma through waveguide miter bends may change the polarization and rotate the polarization major axis. Polarizing elements are therefore required to correct for the effect of transmission lines and also to generate proper polarization for coupling into the plasma. Rotating mirrors with different rectangular grooved gratings in two successive miter bends can generate the required wide range of elliptical polarizations. Peak heat fluxes due to ohmic losses in these mirrors are several MW/m{sub 2} for a 0.5 MW gyrotron power. The complex distribution of losses in the grooves requires a detailed thermal stress analysis to ensure that temperature and stress limits are not exceeded. The desired pulse length is 10 sec, with a cooling time of 10 min between pulses. The temperature rise in the polarizing mirrors must be limited to less than 300{degree}C to prevent thermal fatigue and outgassing in the vacuum lines. This paper presents an analysis for the polarizing mirrors for the DIII-D ECH system.

  6. Localized measurement of short wavelength plasma fluctuations with the DIII-D phase contrast imaging diagnostic

    SciTech Connect

    Dorris, J. R.; Rost, J. C.; Porkolab, M.

    2009-02-15

    A novel rotating mask system has been designed and implemented on the DIII-D phase contrast imaging (PCI) diagnostic to produce the first spatially localized PCI measurements of a tokamak plasma. The localization technique makes use of the variation in the magnetic field component perpendicular to the viewing chord as a function of chord height. This new capability provides measurements in the range of 2DIII-D PCI at wave numbers up to 40 cm{sup -1} to probe electron scale turbulence in the plasma core.

  7. Performance and data analysis aspects of the new DIII-D monostatic profile reflectometer systema)

    NASA Astrophysics Data System (ADS)

    Zeng, L.; Peebles, W. A.; Doyle, E. J.; Rhodes, T. L.; Crocker, N.; Nguyen, X.; Wannberg, C. W.; Wang, G.

    2014-11-01

    A new frequency-modulated profile reflectometer system, featuring a monostatic antenna geometry (using one microwave antenna for both launch and receive), has been installed on the DIII-D tokamak, providing a first experimental test of this measurement approach for profile reflectometry. Significant features of the new system are briefly described in this paper, including the new monostatic arrangement, use of overmoded, broadband transmission waveguide, and dual-polarization combination/demultiplexing. Updated data processing and analysis, and in-service performance aspects of the new monostatic profile reflectometer system are also presented. By using a raytracing code (GENRAY) to determine the approximate trajectory of the probe beam, the electron density (ne) profile can be successfully reconstructed with L-mode plasmas vertically shifted by more than 10 cm off the vessel midplane. Specifically, it is demonstrated that the new system has a capability to measure ne profiles with plasma vertical offsets of up to ±17 cm. Examples are also presented of accurate, high time and spatial resolution density profile measurements made over a wide range of DIII-D conditions, e.g., the measured temporal evolution of the density profile across a L-H transition.

  8. Theory and Simulation Basis for Magnetohydrodynamic Stability in DIII-D

    SciTech Connect

    Turnbull, A.D.; Brennan, D.P.; Chu, M.S.; Lao, L.L.; Snyder, P.B.

    2005-10-15

    Theory and simulation have provided one of the critical foundations for many of the significant achievements in magnetohydrodynamic (MHD) stability in DIII-D over the past two decades. Early signature achievements included the validation of tokamak MHD stability limits, beta and performance optimization through cross-section shaping and profiles, and the development of new operational regimes. More recent accomplishments encompass the realization and sustainment of wall stabilization using plasma rotation and active feedback, a new understanding of edge stability and its relation to edge-localized modes, and recent successes in predicting resistive tearing and interchange instabilities. The key to success has been the synergistic tie between the theory effort and the experiment made possible by the detailed equilibrium reconstruction data available in DIII-D and the corresponding attention to the measured details in the modeling. This interaction fosters an emphasis on the important phenomena and leads to testable theoretical predictions. Also important is the application of a range of analytic and simulation techniques, coupled with a program of numerical tool development. The result is a comprehensive integrated approach to fusion science and improving the tokamak approach to burning plasmas.

  9. Prospects for Edge Current Density Determination Using Li beam on DIII-D

    SciTech Connect

    D.M. Thomas; A.S. Bozek; T.N. Carlstrom; D.K. Finkenthal; R. Jayakumar; M.A. Makowski; D.G. Nilson; T.H. Osborne; B.W. Rice; R.T. Snider

    2000-08-01

    The specific size and structure of the edge current profile has important effects on the MHD stability and ultimate performance of many advanced tokamak (AT) operating modes. This is true for both bootstrap and externally driven currents that may be used to tailor the edge shear. Absent a direct local measurement of j(r), the best alternative is a determination of the poloidal field. Measurements of the precision (0.1-0.01{sup o} in magnetic pitch angle and 1-10 ms) necessary to address issues of stability and control and provide constraints for EFIT are difficult to do in the region of interest ({rho} = 0.9-1.1). Using Zeeman polarization spectroscopy of the 2S-2P lithium resonance line emission from the DIII-D LIBEAM, measurements of the various field components may be made to the necessary precision in exactly the region of interest to these studies. Because of the negligible Stark mixing of the relevant atomic levels, this method of determining j(r) is insensitive to the large local electric fields typically found in enhanced confinement (H-mode) edges, and thus avoids an ambiguity common to Motional Stark Effect (MSE) measurements of B. Key issues for utilizing this technique include good beam quality, an optimum viewing geometry, and a suitable optical pre-filter to isolate the polarized emission line. A prospective diagnostic system for the DIII-D AT program will be described.

  10. DIII-D Thomson Scattering Diagnostic Data Acquisition, Processing and Analysis Software

    SciTech Connect

    Middaugh, K.R.; Bray, B.D.; Hsieh, C.L.; McHarg, B.B., Jr.; Penaflor, B.G.

    1999-06-01

    One of the diagnostic systems critical to the success of the DIII-D tokamak experiment is the Thomson scattering diagnostic. This diagnostic is unique in that it measures local electron temperature and density: (1) at multiple locations within the tokamak plasma; and (2) at different times throughout the plasma duration. Thomson ''raw'' data are digitized signals of scattered light, measured at different times and locations, from the laser beam paths fired into the plasma. Real-time acquisition of this data is performed by specialized hardware. Once obtained, the raw data are processed into meaningful temperature and density values which can be analyzed for measurement quality. This paper will provide an overview of the entire Thomson scattering diagnostic software and will focus on the data acquisition, processing, and analysis software implementation. The software falls into three general categories: (1) Set-up and Control: Initializes and controls all Thomson hardware and software, synchronizes with other DIII-D computers, and invokes other Thomson software as appropriate. (2) Data Acquisition and Processing: Obtains raw measured data from memory and processes it into temperature and density values. (3) Analysis: Provides a graphical user interface in which to perform analysis and sophisticated plotting of analysis parameters.

  11. Active and passive spectroscopic imaging in the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Van Zeeland, M. A.; Yu, J. H.; Brooks, N. H.; Heidbrink, W. W.; Burrell, K. H.; Groebner, R. J.; Hyatt, A. W.; Luce, T. C.; Pablant, N.; Solomon, W. M.; Wade, M. R.

    2010-04-01

    Wide-angle, 2D imaging of Doppler-shifted, Balmer alpha (Dα) emission from high energy injected neutrals, charge exchange recombination (CER) emission from neutral beam interaction with thermal ions and fully stripped impurity ions and visible bremsstrahlung (VB) from the core of DIII-D plasmas has been carried out. Narrowband interference filters were used to isolate the specific wavelength ranges of visible radiation for detection by a tangentially viewing, fast-framing camera. Measurements of the Dα emission from fast neutrals injected into the plasma from the low field side reveal the vertical distribution of the beam, its divergence and the variation in its radial penetration with density. Modeling of this emission using both a full Monte Carlo collisional radiative code as well as a simple beam attenuation code coupled to Atomic Data and Analysis Structure emissivity lookup tables yields qualitative agreement, however the absolute magnitudes of the emissivities in the predicted distribution are larger than those measured. Active measurements of carbon CER brightness are in agreement with those made independently along the beam midplane using DIII-D's multichordal, CER spectrometer system, confirming the potential of this technique for obtaining 2D profiles of impurity density. Passive imaging of VB, which can be inverted to obtain local emissivity profiles, is compared with measurements from both a calibrated filter/photomultiplier array and the standard multichordal CER spectrometer system.

  12. Regimes of improved confinement and stability in DIII-D obtained through current profile modifications

    SciTech Connect

    Lao, L.L.; Ferron, J.R.; Taylor, T.S.; Chan, V.S.; Osborne, T.H.; Burrell, K.H.; Chu, M.S.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Jackson, G.L.; St. John, H.; Strait, E.J.; Thompson, S.J.; Turnbull, A.D.; Doyle, E.J.; Rettig, C.; James, R.; Wroblewski, D.; Lazarus, E.A.; Zohm, H.

    1992-09-01

    Several regimes of improved confinement and stability have been obtained in recent experiments in the DIII-D tokamak by dynamically varying the toroidal current density profile to transiently produce a poloidal magnetic field profile with more favorable confinement and stability properties. A very peaked current density profile with high plasma internal inductance, {ell}{sub i}, is produced either by a rapid change in the plasma poloidal cross section or by a rapid change in the total plasma current. Values of thermal energy confinement times nearly 1.8 times the JET/DIII-D ELM-free H-mode thermal confinement scaling are obtained. The confinement enhancement factor over the ITER89-P L-mode confinement scaling, H, is as high as 3. Normalized toroidal beta, {beta}{sub N}, greater than 6%-m-T/MA and values of the product {beta}{sub N}H greater than 15 have also been obtained. Both the confinement and the maximum achievable {beta} vary with {ell}{sub i} and decrease as the current profile relaxes. For strongly shaped H-mode discharges, in addition to the current density profile peakedness, as measured by {ell}{sub i} other current profile parameters, such as its distribution near the edge region, may also affect the confinement enhancement.

  13. Regimes of improved confinement and stability in DIII-D obtained through current profile modifications

    SciTech Connect

    Lao, L.L.; Ferron, J.R.; Taylor, T.S.; Chan, V.S.; Osborne, T.H.; Burrell, K.H.; Chu, M.S.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Jackson, G.L.; St. John, H.; Strait, E.J.; Thompson, S.J.; Turnbull, A.D. ); Doyle, E.J.; Rettig, C. ); James, R.; Wroblewski, D. (Lawrence Livermore National Lab., CA (United Sta

    1992-09-01

    Several regimes of improved confinement and stability have been obtained in recent experiments in the DIII-D tokamak by dynamically varying the toroidal current density profile to transiently produce a poloidal magnetic field profile with more favorable confinement and stability properties. A very peaked current density profile with high plasma internal inductance, [ell][sub i], is produced either by a rapid change in the plasma poloidal cross section or by a rapid change in the total plasma current. Values of thermal energy confinement times nearly 1.8 times the JET/DIII-D ELM-free H-mode thermal confinement scaling are obtained. The confinement enhancement factor over the ITER89-P L-mode confinement scaling, H, is as high as 3. Normalized toroidal beta, [beta][sub N], greater than 6%-m-T/MA and values of the product [beta][sub N]H greater than 15 have also been obtained. Both the confinement and the maximum achievable [beta] vary with [ell][sub i] and decrease as the current profile relaxes. For strongly shaped H-mode discharges, in addition to the current density profile peakedness, as measured by [ell][sub i] other current profile parameters, such as its distribution near the edge region, may also affect the confinement enhancement.

  14. Stability of negative central magnetic shear discharges in the DIII-D tokamak

    SciTech Connect

    Strait, E.J.; Chu, M.S.; Ferron, J.R.

    1996-12-01

    Discharges with negative central magnetic shear (NCS) hold the promise of enhanced fusion performance in advanced tokamaks. However, stability to long wavelength magnetohydrodynamic modes is needed to take advantage of the improved confinement found in NCS discharges. The stability limits seen in DIII-D experiments depend on the pressure and current density profiles and are in good agreement with stability calculations. Discharges with a strongly peaked pressure profile reach a disruptive limit at low beta, {beta}{sub N} = {beta} (I/aB){sup -1} {le} 2.5 (% m T/MA), caused by an n = 1 ideal internal kink mode or a global resistive instability close to the ideal stability limit. Discharges with a broad pressure profile reach a soft beta limit at significantly higher beta, {beta}{sub N} = 4 to 5, usually caused by instabilities with n > 1 and usually driven near the edge of the plasma. With broad pressure profiles, the experimental stability limit is independent of the magnitude of negative shear but improves with the internal inductance, corresponding to lower current density near the edge of the plasma. Understanding of the stability limits in NCS discharges has led to record DIII-D fusion performance in discharges with a broad pressure profile and low edge current density.

  15. Coherent Modes in the Edge of DIII--D H--modes

    NASA Astrophysics Data System (ADS)

    Moyer, R. A.; Lehmer, R.; Boedo, J. A.; Rhodes, T. L.; Rettig, C. L.; Doyle, E. J.; Watkins, J. G.; Groebner, R. J.; Burrell, K. H.

    1997-11-01

    Electrostatic fluctuations in the edge of DIII--D H--modes are often characterized by the presense of a large amplitude, coherent or quasi-coherent mode (f=50--60 kHz; δf=5--20 kHz). This mode is similar to coherent modes in the edge of ohmic discharges in TEXT(Tsui, et al., Phys. Rev. Lett. 70) (1993) 1565. and has also been seen in the edge of H--modes in PBX-M.(Tynan et al., Phys. Plasmas 1) (1994) 3301. The mode is localized to a narrow radial extent just inside the separatrix. In ohmic H--modes, the ``mode'' is a quasi-coherent, localized burst of turbulence. In beam heated discharges, the mode is coherent and dominates the power spectra for both density and potential fluctuations after onsetting 10--30 ms into the H--mode phase. In DIII--D, this mode directly drives particle and convective heat fluxes. Mode characteristics and plasma conditions at onset of the mode will be compared with theoretical predictions of such modes.

  16. Neutron Detection With Ultra-Fast Digitizer and Pulse Identification Techniques on DIII-D

    NASA Astrophysics Data System (ADS)

    Zhu, Y. B.; Heidbrink, W. W.; Piglowski, D. A.

    2013-10-01

    A prototype system for neutron detection with an ultra-fast digitizer and pulse identification techniques has been implemented on the DIII-D tokamak. The system consists of a cylindrical neutron fission chamber, a charge sensitive amplifier, and a GaGe Octopus 12-bit CompuScope digitizer card installed in a Linux computer. Digital pulse identification techniques have been successfully performed at maximum data acquisition rate of 50 MSPS with on-board memory of 2 GS. Compared to the traditional approach with fast nuclear electronics for pulse counting, this straightforward digital solution has many advantages, including reduced expense, improved accuracy, higher counting rate, and easier maintenance. The system also provides the capability of neutron-gamma pulse shape discrimination and pulse height analysis. Plans for the upgrade of the old DIII-D neutron counting system with these techniques will be presented. Work supported by the US Department of Energy under SC-G903402, and DE-FC02-04ER54698.

  17. Comparison of Experimental Fluctuation and Turbulence Measurements with Theory and Simulation at DIII-D

    SciTech Connect

    Rhodes, T.L.; McKee, G.R.; Politzer, P.A.; Ross, D.W.

    2005-10-15

    Considerable research at DIII-D has been aimed at detailed comparisons of a variety of experimental fluctuation and turbulence measurements to turbulence simulations and theory. The goals of such comparisons are to improve the understanding of turbulence and transport as well as to test and provide feedback to the theory and simulations. Progress in this area will lead to confidence in the extrapolation of predictions to next-step fusion devices and, potentially, to improved control of transport. This paper summarizes some of the more recent and significant results of comparisons of experiment to theory and simulation that have been performed at DIII-D. These comparisons cover a range of plasma conditions (ohmic, L-mode, and impurity enhanced confinement), physical phenomena [transport, avalanches, zonal flows, and geodesic acoustic modes (GAMs)], and measurements (fluctuation levels, fluctuation spectra, radial correlation lengths, heat transport, and poloidal fluctuation velocity). Results reviewed here include comparisons between experimental turbulent radial correlation lengths and nonlinear turbulence simulations, measurements showing GAM activity (a type of zonal flow) similar to predictions, long-range or avalanche-type behavior with significant heat transport similar to that seen in nonlinear simulations, and reduction of turbulence with an enhancement of confinement during impurity injection similar to theory and simulation.

  18. Recent Results on Coupling Fast Waves to High Performance Plasmas on DIII-D

    SciTech Connect

    Pinsker, R. I.; Luce, T. C.; Politzer, P. A.; Porkolab, M.; Goulding, R. H.; Hanson, G. R.; Ryan, P. M.; Hosea, J. C.; Nagy, A.; Wilson, J. R.; Maggiora, R.; Milanesio, D.; Zeng, L.

    2011-12-23

    Fast Waves (FWs) at 60 MHz and 90 MHz are used in DIII-D for central electron heating and current drive. Coupling of FWs to high-performance discharges is limited by low antenna loading in these regimes. To extend the application of high-power FWs to such regimes, methods of increasing the antenna loading in these regimes are needed. A systematic study of loading enhancement techniques has been carried out in DIII-D, including reduction of the antenna/plasma distance, gas puffing into the far scrape-off layer (SOL), and control of other parameters that affect the particle balance in the far SOL. Quantitative understanding of the physics of the loading resistance and its dependence on edge density profiles is demonstrated. The core FW heating efficiency appeared to be {approx}100% in the Advanced Inductive regime, consistent with the high first-pass direct electron absorption of {approx}75% that is predicted by the ray-tracing code GENRAY in this high electron beta regime.

  19. High Frequency ELM Pacing by Lithium Pellet Injection on DIII-D

    NASA Astrophysics Data System (ADS)

    Bortolon, A.; Maingi, R.; Mansfield, D. K.; Nagy, A.; Roquemore, A. L.; Lunsford, R.; Jackson, G. L.; Osborne, T. H.; Parks, P. B.

    2015-11-01

    Full-shot, high-frequency pacing of edge localized modes (ELM) by lithium pellet injection has been demonstrated in DIII-D. A Lithium Granule Injector (LGI), recently installed on DIII-D to study pacing efficiency dependence on granule size and velocity, was tested in different ELMy scenarios (βN = 1.2-2.0) injecting granules of nominal diameter 0.3-0.9 mm, with injection speed 50-120 m/s and injection rates up to 500 Hz. Robust ELM pacing was documented on time windows up to 3.5 s, with triggering efficiency close to 100% obtained with 0.9 mm diameter granules, lower with smaller sizes and weakly dependent on granule velocity. Paced ELM frequencies up to 100 Hz were achieved, with a 2-5 fold increase over the natural ELM frequency and a consequent reduction of divertor peak heat flux. Overall, LGI high frequency pacing appeared to be compatible with high plasma performance, in terms of global confinement and pedestal characteristics. Work supported by the US Department of Energy under DE-AC02-09CH11466, DE-FC02-04ER54698.

  20. Spectroscopic measurements and modeling of tungsten erosion in the DIII-D divertor

    NASA Astrophysics Data System (ADS)

    Abrams, T. D.; Ding, R.; Guo, H. Y.; Leonard, A. W.; Thomas, D. M.; Allen, S. L.; McLean, A. G.; Briesemeister, A. R.; Unterberg, E. A.; Chrobak, C.; Doerner, R. P.; Rudakov, D. L.; Elder, J. D.; Stangeby, P. C.; Wampler, W. R.; Watkins, J. G.

    2015-11-01

    In situ time-resolved measurements of the gross W erosion rate have been performed in DIII-D by monitoring W/I (400.9 nm) emission in the divertor via a filtered camera and high-resolution spectrometer. The erosion rate of a thin W coating on DiMES, inferred via the S/XB method, was found to be ~ 0.7 nm/s during deuterim L-mode exposure, in fair agreement with post-mortem IBA analysis but lower than REDEP/WBC modeling. During H-mode He bombardment of W disks, average erosion rates of ~ 2.9 nm/s and ~ 9.0 nm/s were estimated during the inter-ELM and intra-ELM phases, using ne and Te from divertor Thomson scattering and Langmuir probes. Results will also be presented from additional W erosion experiments in preparation for the DIII-D mini-campaign to measure high-Z transport in the edge plasma. Comparisons will be made with ERO modeling Supported by US DOE DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC05-00OR22725, DE-SC0001961, DE-AC04-94AL85000.