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

    NASA Astrophysics Data System (ADS)

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

    2015-09-01

    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 Ip. 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/Ip 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/Ip scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/Ip 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.

  3. Perturbative transport modeling and comparison to cold-pulse and heat-pulse propagation experiments in Alcator C-Mod and DIII-D

    NASA Astrophysics Data System (ADS)

    Rodriguez Fernandez, P.; White, A. E.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Howard, N. T.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Petty, C. C.; Rice, J. E.; Alcator C-Mod Team

    2016-10-01

    Possible ``non-local'' transport phenomena are often observed in tokamak plasmas. Different models have been proposed to explain fast responses during perturbative transport experiments, including non-diffusive effects. Specific tools to characterize the dynamic behavior and power balance analysis using TRANSP and the quasi-linear trapped gyro-landau fluid code TGLF have been developed to analyze Alcator C-Mod experiments. Recent results from cold pulse experiments show that fast core temperature increases following edge cold-pulse injections (peak within 10ms , while τE 25ms) are not correlated with the direction of intrinsic rotation, and instead the amplitude of the core response depends on density, plasma current and RF input power. The propagation of the cold pulse can be compared with propagation of heat pulses from sawteeth, and both may be used to probe changes in temperature profile stiffness. A Laser Blow Off (LBO) system is being developed for DIII-D that will allow further validation and cross-machine comparison of cold pulse experiments. LBO at DIII-D will also allow for direct comparisons with ECH perturbative heat pulse experiments. Work supported by US DOE under Grants DE-FC02-99ER54512 (C-Mod) and DE-FC02-04ER54698 (DIII-D) and La Caixa Fellowship.

  4. The Alcator C-Mod FIR Polarimeter

    NASA Astrophysics Data System (ADS)

    Xu, P.; Irby, J. H.; Bosco, J.; Kanojia, A.; Leccacorvi, R.; Marmar, E. S.; Michael, P.; Murray, R.; Rokhman, Y.; Vieira, R.; Brower, D. L.; Ding, W. X.; Mansfield, D. K.

    2009-11-01

    A multi-chord FIR polarimetry diagnostic is being developed for the Alcator C-Mod Tokamak to be used to determine the q-profile and to study density and magnetic field fluctuations. This poloidally viewing system using retro-reflectors on the inner wall will have geometry and fields similar to those planned for ITER. The full optical layout will be discussed, as well as simulations of the expected Faraday and Cotton-Mouton signal levels. Bench test results from a single chord system including all optical components will be presented, and preliminary experimental results from C-Mod will be compared with simulated Faraday rotation angle calculated using Thomson Scattering density profiles and EFIT reconstructions of actual C-Mod plasmas.

  5. Highlights of Recent Alcator C-Mod Research

    NASA Astrophysics Data System (ADS)

    Marmar, Earl

    2012-10-01

    Alcator C-Mod focuses on broad scientific issues with particular emphasis on ITER needs and requests. First results from a high power, magnetic field-aligned ICRF antenna show significant reduction of high Z impurity contamination under various plasma conditions Detailed comparisons of core and edge fluctuations in L-mode and I-mode reveal that core turbulence is reduced in I-mode before the reduction of edge turbulence and the onset of the WCM, in stark contrast to the usual turbulence changes observed in H-mode on other tokamaks. The connection among rotation reversals, non-local heat transport, energy confinement saturation (the transition between the linear Ohmic confinement, and saturated Ohmic confinement, regimes) and changes in underlying turbulence has been demonstrated. Joint pedestal studies among C-Mod, DIII-D and NSTX reveal common features, including an upper limit on pedestal pressure in ELMy H-mode determined by peeling-ballooning modes (PBMs), and pedestal width scaling approximately as βpol^1/2. A novel stochastic model for intermittent SOL plasma fluctuations has been constructed, is in excellent agreement with experiment, and reveals important details on the underlying physics. Design of the first actively heated tokamak tungsten divertor is described.

  6. Volume Recombination in Alcator C-Mod Divertor Plasmas

    NASA Astrophysics Data System (ADS)

    Terry, J. L.

    1997-11-01

    Volume recombination has been predicted(See, for example, A. Loarte, Proc. 12th PSI Conf, J. Nucl. Mater (1996) I9, in press.) to be a significant sink for plasma ions under the detached divertor conditions achieved on many tokamaks. This volume recombination sink was observed initially in Alcator C-Mod and shown to be a major fraction of the ion loss. Signatures of recombination have now been observed on DIII-D(R.C. Isler, et al., paper submitted for publication), Asdex-UG (B. Napiontek, et al. 24th EPS Conf., Berchtesgaden, Germany, 1997, P4.007, in press.), and JET(R.D. Monk, et al. 24th EPS Conf., Berchtesgaden, Germany, 1997, P1.030, in press.). It is important primarily because the recombined atoms are not accelerated through the sheath - thus reducing divertor plate sputtering, and because most of the potential energy of recombination (13.6 eV) is released as radiation before the ion strikes the plate. The Alcator C-Mod measurements show that the recombination occurs in low Te ( ~1 eV), high ne ( ~1× 10^21 m-3) regions, and is significantly larger in detached regions. At the inboard, detached divertor plate the measured volume recombination rate is typically greater than the rate of ion collection at that plate and is about an order of magnitude higher than on the attached, outer plate. These spatially resolved measurements also show that the recombination rate is peaked near the strike point and imply that the recombination is occurring close to the plate surface. The C-Mod observations about the magnitude and spatial distribution of the recombination are consistent with the modelling of similar discharges(F. Wising et al., Contrib. Plasma Phys. 36, p 136 (1996).). The experimental evidence for recombination is found in the deuterium emission spectra from the divertor, in particular in the Balmer- and/or Lyman-series. The spectra show that the dominant recombination mechanism is 3-body recombination into excited states of deuterium and that the populations

  7. Overview of Alcator C-Mod Research

    NASA Astrophysics Data System (ADS)

    Marmar, Earl; Alcator C-Mod Team

    2016-10-01

    C-Mod is the only divertor tokamak in the world capable of operating at B fields up to 8 T, equaling and exceeding that planned for ITER. C-Mod is compact, accessing regimes of extreme edge power density (q|| 1 GW/m2) . surpassing the design for ITER, and approaching the levels envisioned in power plants. C-Mod results are particularly important for providing the physics basis of the high-field, compact tokamak approach. Results of experiments and related modeling span the topics of core transport and turbulence, RF heating and current drive, pedestal physics, scrape-off layer, divertor and plasma-wall interactions. ICRF has been successfully applied to control and reverse accumulation of high Z impurities in the core plasma. For the first time ever, feedback of low Z seeding for divertor power dissipation has been tied directly to real-time plasma power fluxes measured on the high-Z metal PFCs in the divertor, and used to mitigate those fluxes with no degradation of the pedestal pressure or core confinement. The naturally ELM-less I-mode and EDA-H-mode regimes have been extended BT = 8T. I-mode threshold scalings show a weak dependence on B, yielding a significantly broader window for I-mode operation at high field. Quiescence of the high-field side scrape-off layer makes this a potentially attractive location for placement of RF actuators to ameliorate plasma interactions with launchers; the wave physics for penetration and damping, for both ICRF and LHRF appears very favorable for high-field side launch. Supported by USDoE Award DE-FC02-99ER54512.

  8. 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.

  9. 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.

  10. I-mode access and transitions in an expanded operating space on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Hubbard, A. E.; Baek, S.-G.; Churchill, R. M.; Golfinopoulos, T.; Hughes, J. W.; Lin, Y.; Marmar, E. S.; Terry, J. L.; Theiler, C.; Walk, J. R.; White, A. E.; Whyte, D. G.; Wolfe, S. M.; Wukitch, S. J.; Cziegler, I.; Edlund, E.

    2014-10-01

    New experiments in the I-mode regime on Alcator C-Mod have further expanded the wide range of conditions over which it has been studied. I-mode is attractive in that it combines the high energy confinement of H-mode, up to H98 = 1.2, with the low particle/impurity confinement of L-Mode. Accessing and maintaining the regime, without transitions to L-mode or H-mode, is key to its extrapolation. The most robust conditions on C-Mod have been in lower null plasmas with unfavorable drift and BT of 5-6 T, spanning the ITER value. In 2014, experiments at 2.8 T showed that the power range between L-I and I-H transitions was significantly lower, consistent with results on DIII-D and AUG. I-mode was also accessed in near double null. Fueling into I-modes can raise the density, up to 2×1020 m-3. The upper ne increases with input power. Local pedestal parameters and evolution of turbulence at transitions are being assessed with the aim of understanding the physical mechanisms and for extrapolation. Supported by DOE Award DE-FC02-99ER54512.

  11. 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.

  12. Progress in characterizing pedestal stability on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Hughes, J. W.; Churchill, R. M.; Cziegler, I.; Davis, E. M.; Dominguez, A.; Ennever, P.; Ernst, D.; Hubbard, A. E.; Lipschultz, B.; Ma, Y.; Walk, J. R.; Wolfe, S. M.; Snyder, P. B.; Osborne, T.; Xu, X.; Sugiyama, L.

    2011-10-01

    Experimental studies on Alcator C-Mod explore pedestal structure and edge relaxation mechanisms primarily in three high confinement regimes: ELMy H-mode, EDA H-mode and I-mode. Extensive scans of BT, IP, ne, Pnet and shaping parameters have been carried out in these regimes, allowing the characterization of the operating space for ELMs as well as benign pedestal relaxation mechanisms (e.g. quasi- and weakly coherent modes). Well resolved edge profiles and accurate equilibrium reconstructions are produced from typical discharges, enabling evaluation of edge stability using various computational tools, such as ELITE, BOUT++, M3D and GS2. Relationships among dominant edge instabilities, radial transport and pedestal structure will be discussed. Supported by US DoE Award DE-FC02-99-ER54512.

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

    NASA Astrophysics Data System (ADS)

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

    2012-12-01

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

  14. Stationary Density Profiles in Alcator C-mod

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

    In the absence of an internal particle source, plasma turbulence will impose an intrinsic relationship between an inwards pinch and an outwards diffusion resulting in a stationary density profile as determined by the turbulent equipartition (TEP) theory. The Alcator C-mod tokamak utilizes RF heating and current drive so that fueling only occurs in the vicinity of the separatrix. Density is determined from Thomson scattering. Discharges that transition from L-mode to I-mode are seen to maintain a stationary profile. For reversed shear discharges maintained by non-inductive current drive (Vloop 0) a drop of density in the vicinity of the axis is consistent with an observed rise in q, although error in the measurement precludes making this observation definitive.

  15. Development of the Alcator C-Mod FIR Polarimeter

    NASA Astrophysics Data System (ADS)

    Xu, P.; Irby, J. H.; Bosco, J.; Kanojia, A.; Leccacorvi, R.; Marmar, E. S.; Michael, P.; Murray, R.; Vieira, R.; Wolfe, S.; Brower, D. L.; Ding, W. X.; Mansfield, D. K.

    2008-11-01

    A multi-chord FIR polarimetry diagnostic is being developed for the Alcator C-Mod Tokamak to be used to determine the q-profile and to study density and magnetic field fluctuations. This poloidally viewing system using retro-reflectors on the inner wall will have geometry and fields similar to those planned for ITER. The optical layout will be discussed, as well as simulations of the expected Faraday and Cotton-Mouton signal levels, and the plans to integrate these data into EFIT. Details of the hardware being developed and procured including the FIR laser system, the laser power and frequency control system, optical components, detectors, beam position feedback system, and inner wall retro-reflectors and shutter will be presented.

  16. High confinement dissipative divertor operation on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Goetz, J. A.; LaBombard, B.; Lipschultz, B.; Pitcher, C. S.; Terry, J. L.; Boswell, C.; Gangadhara, S.; Pappas, D.; Weaver, J.; Welch, B.; Boivin, R. L.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hubbard, A.; Hutchinson, I.; Irby, J.; Marmar, E.; Mossessian, D.; Porkolab, M.; Rice, J.; Rowan, W. L.; Schilling, G.; Snipes, J.; Takase, Y.; Wolfe, S.; Wukitch, S.

    1999-05-01

    Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] has operated a High-confinement-mode (H-mode) plasma together with a dissipative divertor and low core Zeff. The initially attached plasma is characterized by steady-state enhancement factor, HITER89P [P. N. Yushmanov et al., Nucl. Fusion 30, 1999 (1990)], of 1.9, central Zeff of 1.1, and a radiative fraction of ˜50%. Feedback control of a nitrogen gas puff is used to increase radiative losses in both the core/edge and divertor plasmas in almost equal amounts. Simultaneously, the core plasma maintains HITER89P of 1.6 and Zeff of 1.4 in this nearly 100% radiative state. The power and particle flux to the divertor plates have been reduced to very low levels while the core plasma is relatively unchanged by the dissipative nature of the divertor.

  17. 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.

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

    SciTech Connect

    Niemczewski, Artur 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.

  19. Correlation ECE diagnostic in Alcator C-Mod

    DOE PAGES

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

    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

  20. 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.

  1. 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.

  2. 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.

  3. 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

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

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

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

  5. Highlights of the Alcator C-Mod Research Campaign

    NASA Astrophysics Data System (ADS)

    Greenwald, Martin; Alcator Team

    2011-10-01

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

  6. Identification of Mercier instabilities in Alcator C-Mod tokamak

    SciTech Connect

    In, Y.; Ramos, J. J.; Hastie, R. J.; Catto, P. J.; Hubbard, A. E.; Hutchinson, I. H.; Marmar, E.; Porkolab, M.; Snipes, J.; Wolfe, S.

    2000-12-01

    During current ramp-up discharges, highly localized magnetohydrodynamic (MHD) fluctuations were observed on the electron cyclotron emission diagnostics of Alcator C-Mod tokamak [I. H. Hutchinson , Phys. Plasmas 1, 1511 (1994)]. The electron temperature profile was hollow, while the density profile was weakly decreasing. Assuming that the equilibration time was short enough to quickly thermalize ions the pressure profile was also found to be hollow. Using this pressure profile as an additional constraint to the EFIT program, an equilibrium with reversed shear was constructed having a q(0)>>1. The localized MHD activity was observed near the inner q=5 rational surface in this reconstructed equilibrium, where the Mercier criterion for ideal MHD stability was violated because of the reversed pressure gradient (dp/dr>0), q>1 and moderate shear. When kinetic effects were added, the ideal Mercier mode was finite ion Larmor radius stabilized. However, ion Landau damping was found to be strong enough to drive a kinetic Mercier instability.

  7. Stimulation of MHD Modes in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Golfinopoulos, T.; Granetz, R.; Labombard, B.; Lin, Y.; Parker, R. R.; Sears, J.; Wukitch, S. J.

    2010-11-01

    Active MHD (AMHD) spectroscopy involves stimulating MHD modes by external means to study the modes or diagnose the plasma. In many AMHD experiments, drive frequency is swept across a 100-200 kHz range in which modes are expected; this allows for robust techniques to detect resonant poles in the presence of direct pickup from the driver. However, there is flexibility in the drive mechanism. At Alcator, we have employed a parametric excitation method, amplitude-modulating the ICRF wave (80 MHz) with envelope signals in the AE frequency range (100's kHz). This builds off the ICRF beat technique used in JET in 1996 and ASDEX Upgrade in 2006, but is unique in its use of a single antenna, improving coherence. An advantage of this approach is its ability to couple to the plasma core. It also has high input power, though efficiency is limited by the Manley-Rowe relations. In initial experiments, we excited weak, stable modes in the toroidal Alfvén eigenmode band gap. We plan to explore this and other methods for coupling to various MHD-like modes, especially C-Mod's Quasi Coherent mode.

  8. 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.

  9. H-modes on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Snipes, J. A.; Hubbard, A. E.; Garnier, D. T.; Golovato, S. N.; Granetz, R. S.; Greenwald, M.; Hutchinson, I. H.; Irby, J.; La Bombard, B.; Marmar, E. S.; Niemczewski, A.; O'Shea, P. J.; Porkolab, M.; Stek, P.; Takase, Y.; Terry, J. L.; Watterson, R.; Wolfe, S. M.

    1996-08-01

    H-modes exhibiting improved confinement above the L-mode are achieved in Alcator C-Mod with ICRF and with ohmic heating alone without boronization. Both ELM-free and ELMy H-modes are obtained with total input power from 0.75 to 4.2 MW over a range of densities (0.8 to 0741-3335/38/8/005/img1) and toroidal fields (3 to 8 T). Type III ELMs are often observed to have coherent, high m and n precursor oscillations with frequencies of 100 - 160 kHz. The threshold power required to achieve the H-mode increases with density and toroidal field, in rough agreement with scalings derived from other tokamaks. The power densities and density times toroidal field products are an order of magnitude larger than in other tokamaks, in the range of values expected for ITER. The L - H and H - L transitions occur at approximately the same edge electron temperature. A low density limit to the H-mode is found at about 0741-3335/38/8/005/img2. A high midplane neutral pressure limit of about 0.6 mTorr is also observed.

  10. 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.

  11. 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.

  12. Investigation of Edge Localized Modes in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Terry, J. L.

    2006-10-01

    Characteristics of discrete ELMs produced in Alcator C-Mod discharges of low edge collisionality and high triangularity are examined. These discharges have high values for central Te and ne (reaching 4.5 keV and 2x10^20 m-3 respectively) and good confinement, consistent with ITER98y2 ELMy H-mode scaling. Pedestal temperature heights reach 0.9 keV at densities above 1x10^20 m-3. Studies of the stability of the pedestal profiles to peeling/ballooning modes will be presented. The energy lost per ELM from the H-mode pedestal is ˜10-20% of the pedestal energy. These ELMs exhibit relatively long-lived precursor oscillations, often with two modes of intermediate (n˜10) toroidal mode number present. At the ELM ``crash'' a high frequency (˜0.5 MHz), short-lived magnetic oscillation is initiated, and multiple plasma filament structures are expelled into the Scrape-Off-Layer. The initial ELM filaments, ``primaries'', are large perturbations to the SOL. The perturbation increases the local Dα emission by factors ranging from 1.5 (just outside the LCFS) to ˜100. In the outboard midplane region the primary filaments have radial extents of 0.5-1 cm and typical radial propagation velocities of 1-2 km/s. The poloidal extent of the filaments is greater than the 4.5 cm diagnostic field-of-view. The initial filaments are followed (at intervals of ˜100μs) by multiple, less perturbing ``secondary'' filaments. The radial dynamics of the ELM are also studied at the inboard midplane. The perturbation on the inboard edge appears to be a rapid profile relaxation and recovery. The onset of the inboard profile relaxation is sometimes observed to occur before filaments are seen on the outboard side.

  13. 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.

  14. 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

  15. 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.

  16. Anomalous transport in the H-mode pedestal of Alcator C-Mod discharges

    NASA Astrophysics Data System (ADS)

    Pankin, A. Y.; Hughes, J. W.; Greenwald, M. J.; Kritz, A. H.; Rafiq, T.

    2017-02-01

    Anomalous transport in the H-mode pedestal region of five Alcator C-Mod discharges, representing a collisionality scan is analyzed. The understanding of anomalous transport in the pedestal region is important for the development of a comprehensive model for the H-mode pedestal slope. In this research, a possible role of the drift resistive inertial ballooning modes (Rafiq et al 2010 Phys. Plasmas 17 082511) in the edge of Alcator C-Mod discharges is analyzed. The stability analysis, carried out using the TRANSP code, indicates that the DRIBM modes are strongly unstable in Alcator C-Mod discharges with large electron collisionality. An improved interpretive analysis of H-mode pedestal experimental data is carried out utilizing the additive flux minimization technique (Pankin et al 2013 Phys. Plasmas 20 102501) together with the guiding-center neoclassical kinetic XGC0 code. The neoclassical and neutral physics are simulated in the XGC0 code and the anomalous fluxes are computed using the additive flux minimization technique. The anomalous fluxes are reconstructed and compared with each other for the collisionality scan Alcator C-Mod discharges. It is found that the electron thermal anomalous diffusivities at the pedestal top increase with the electron collisionality. This dependence can also point to the drift resistive inertial ballooning modes as the modes that drive the anomalous transport in the plasma edge of highly collisional discharges.

  17. Investigation of ELMs on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Terry, J. L.

    2005-10-01

    C-Mod typically operates in regimes without large ELMs. Recently, discrete ELMs have been routinely produced by making plasmas with large lower triangularity (i.e. >0.75), compared to the more typical C-Mod values <0.6. The ELM character was substantially modified as the triangularity was reduced, changing from discrete ELMs of ˜60 μsec duration, to H-to-L mode back transitions, lasting ˜4msec. The discrete ELMs are most apparent when the density is just above the low-density H-mode threshold, ne˜8x10^19m-3. Pedestal Te up to 1 keV was measured early in the H-mode phase. The spatial structure and propagation of the discrete ELMs are studied using fast-framing (˜250 kHz) cameras and other high resolution optical diagnostics. The magnetics and the optical diagnostics show a rapidly-growing precursor oscillation (100-200 kHz just prior to the ELM crash) that is localized radially to around the top of the pedestal. Outside the LCFS the enhanced emission from the ELM propagates radially outward with a complicated spatial structure, similar in many respects to `blobs.'

  18. 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.

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

    NASA Astrophysics Data System (ADS)

    Greenwald, Martin

    2013-10-01

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

  20. Helium Diagnostic for Alcator C-Mod Edge Studies

    NASA Astrophysics Data System (ADS)

    Pappas, D. A.; Labombard, B.; Lipschultz, B.; Pitcher, C. S.; Brix, M.; Schweer, B.

    1997-11-01

    We are developing a diagnostic based on the thermal helium beam technique pioneered on TEXTOR [1,2], which is intended to continuously measure electron temperature and density profiles in the edge plasma at the midplane with high radial and temporal resolution (Δr=1 mm, Δt=1 ms). As a He neutral source, we are presently using a simple gas capillary embedded in one of the C-Mod limiters at the midplane, typically 10 to 15 mm away from the location of the plasma separatrix. Four fibers from a tangentially viewing telescope, which are fed to a visible spectrometer, are used to monitor the HeI lines of interest. We are looking at both the red (667.8 nm, 706.5 nm, 728.1 nm) and the green (471.3 nm, 501.5 nm, 504.7 nm) lines [2], whose intensity ratios can be used to obtain electron temperature and density. The experiments are performed under a variety of plasma conditions and the results are compared to those obtained with a scanning probe. [1] B. Schweer, et al., J. Nucl. Mat. 196-198, 174 (1992). [2] M. Brix and B. Schweer, 24^th EPS Conference, Berchtesgaden, Germany, June 1997. * Work Supported by D.o.E. Contract DE-AC02-78ET51013

  1. 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.

  2. 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.

  3. Scaling of Global LHCD Efficiency in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Scott, S.; Bonoli, P.; Mumgaard, R.; Shiraiwa, S.; Wallace, G.; Whyte, D.

    2014-10-01

    A database of global current-drive efficiency by Lower Hybrid waves has been assembled covering nine years of C-Mod operation. Plasma conditions were averaged over 50-ms time slices during equilibrated current-profile time periods, excluding transient events such as Prad spikes. The database comprises 1800 time slices spanning: PLH < 1 . 1 MW, n|| = 1.5-2.3, Ip = 0.3-1.0 MA, nebar = 0.35-1.5 e20. Nine percent of the data points are approximately non-inductive (ΔV/V >0.9), while 17 percent experience low m,n MHD that degrades the LHCD efficiency. During LHCD, a simple Spitzer model is used to estimate the residual inductively-driven current which scales the pre-LH current by the ratio of the loop voltage to the pre-LH loop voltage, correcting also for the change in conductivity. The current-drive efficiency is defined as η = nebar R ILHCD /PLH [1020 MA/m2 MW], where ILHCD is the current driven by LH waves and PLH is the forward-directed LH power. In approximately non-inductive, MHD-free plasmas, the global current drive efficiency shows a striking positive correlation with plasma current, η = 0.065 + 0.40 x Ip [MA], reaching a value of η = 0.47 at Ip =1.02 MA. A positive but weaker correlation between η and Teo does not explain the η dependence on Ip. Preliminary GENRAY/CQL3D simulations at Ip =1.0 MA predict 900 kA of driven current versus 1000 kA observed. Comparisons of η to numerical simulations over a wide parameter range will be discussed. Work supported by DoE Awards DE-FC02-99ER54512 and DE-AC02-09CH11466.

  4. Multi-Channel Validation of Nonlinear Gyrokinetic Simulations in Alcator C-Mod I-mode Plasmas

    NASA Astrophysics Data System (ADS)

    Creely, A. J.

    2016-10-01

    New multi-channel validation of nonlinear gyrokinetic simulations (GYRO) is carried out for I-mode plasmas on Alcator C-Mod, utilizing heat fluxes, profile stiffness, and density and temperature fluctuations. I-mode plasmas are characterized by high energy confinement, similar to H-mode, but with L-mode-like particle confinement, making them favorable for reactors due to natural absence of ELMs, but without impurity accumulation [Whyte NF 2010]. At C-Mod, I-mode plasmas have been obtained across a wide range of plasma currents (Ip = 0.55-1.2MA) and magnetic fields (Bt = 2.8-8.0T). I-mode is also actively studied at ASDEX Upgrade, DIII-D and other tokamaks [Hubbard NF 2016]. Open questions remain regarding core transport in I-mode compared to L and H-mode, making validation studies in I-mode of great interest. Previous work at C-Mod found that ITG/TEM-scale GYRO simulations can match both electron and ion heat fluxes within error bars in I-mode [White PoP 2015], suggesting that multi-scale, cross-scale coupling effects [Howard PoP 2016] may be less important in I-mode than in L-mode. Adding the constraint of experimental perturbative heat diffusivity, however, revealed that ITG/TEM scale simulations do not adequately capture the high profile stiffness in I-mode [Creely NF 2016]. These results motivated more comprehensive comparisons of gyrokinetic simulations with I-mode plasmas. This talk expands upon past I-mode GYRO validation work to simultaneously constrain nonlinear gyrokinetic simulations with experimental electron and ion heat fluxes, electron temperature fluctuations measured with Correlation ECE, density fluctuations measured with Phase Contrast Imaging and reflectometry, and the temperature profile stiffness measured using partial sawtooth heat pulses. This work is supported by the US DOE under Grants DE-SC0006419 and DEFC02-99ER54512-CMOD.

  5. Evidence for Chaotic Edge Turbulence in the Alcator C-Mod Tokamak

    NASA Astrophysics Data System (ADS)

    Zhu, Ziyan; White, Anne; Carter, Troy; Terry, Jim; Baek, Seung Gyou

    2016-10-01

    Turbulence greatly reduces the confinement time of magnetic-confined plasmas; understanding the nature of this turbulence and the associated transport is therefore of great importance. This research seeks to establish whether turbulent fluctuations in Alcator C-Mod are chaotic or stochastic. This has an important impact on transport caused by turbulence in C-Mod: stochastic fluctuations sample all of phase space and can lead to diffusive transport, whereas chaotic fluctuations live in a restricted phase space (e.g. on attractors) and a diffusive description may not be valid. By analyzing the time series from an O-Mode reflectometer, turbulent edge density fluctuations in Ohmic plasmas and L-mode plasmas in the Alcator C-Mod tokamak are shown to be chaotic. Supporting evidence for chaos in the edge region includes: the observation of an exponential power spectra (which is associated with Lorentzian-shaped pulses in the time series) and the location of the signal in the Complexity-Entropy plane (C-H plane) and its corresponding Brandt-Pompe (BP) probability distribution. These analysis techniques will be briefly introduced along with a discussion of the analysis results. Different diagnostic techniques, such as Gas Puff Imaging (GPI), could be used to confirm the results. Work supported by the U.S. Department of Energy Office of Science under Agreement DE-FC02-99ER54512 and DE-FC02-07ER54918:011.

  6. 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

  7. 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

  8. 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.

  9. Analysis of Runaway Electron Synchrotron Radiation in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Tinguely, R. A.; Granetz, R. S.

    2015-11-01

    In Alcator C-Mod, runaway electron (RE) beams can reach energies of ~30 MeV or higher, emitting synchrotron radiation in the visible wavelength range. Two spectrometers, with spectral ranges of 350-1020 nm and resolutions of ~4 nm, have been installed and absolutely calibrated on C-Mod to observe this emission. Recent theoretical work predicts that the RE distribution tends toward a mono-energetic bump as the synchrotron radiation and collisional friction balance the electric force. Our analysis of RE synchrotron spectra shows that it is possible to distinguish the emission of a mono-energetic and mono-pitch beam from that of a distribution of energies and pitch angles (as calculated in). Preliminary results indicate that the mono-energetic bump is formed as predicted, instead of a broader distribution. Supported by US DoE award DE-FC02-99ER54512.

  10. Analysis of Runaway Electron Synchrotron Emission in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Tinguely, A.; Granetz, R.; Stahl, A.

    2016-10-01

    Alcator C-Mod's high magnetic field allows relativistic ``runaway'' electron (RE) synchrotron radiation (SR) to be observed in the visible wavelength range. Our aim is to determine the evolution of the RE energy distribution function, current, and density from measured SR spectra, providing insight into basic plasma physics as well as mitigation for fusion devices. Recent theoretical studies predict that the SR reaction force and collisional friction will balance the electric force, forming a ``bump'' on the tail of the energy distribution. However, both mono-energetic and monotonically-decreasing distributions fit the experimental data equally well. The COllisonal Distribution of Electrons code is applied to C-Mod RE discharges and compared to experiment. In addition, a scan in magnetic field from 2.7 - 8 T explores the importance of SR as a power loss mechanism and limit on the maximum RE energy. Supported by USDoE Award DE-FC02-99ER54512.

  11. 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.

  12. 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.

  13. 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.

  14. Novel energy resolving x-ray pinhole camera on Alcator C-Mod

    SciTech Connect

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

    2012-10-15

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

  15. 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

  16. 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.

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

    SciTech Connect

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

    1993-12-01

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

  18. 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).

  19. 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.

  20. 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.

  1. 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.

  2. Search for gyrokinetic dependencies in helium transport at Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Liao, Kenneth; Rowan, William; Hatch, David; Bespamyatnov, Igor; Horton, Wendell

    2013-10-01

    Helium-3 and helium-4 impurity transport measurements and density profile measurements have been obtained on Alcator C-Mod in a variety of discharge conditions, using the core Charge Exchange Recombination Spectroscopy (CXRS) diagnostic. The helium concentrations range from trace (< 2 %) to large minority (~ 20 %). L-mode, H-mode, and I-mode results are included, with Ohmic heated, ICRF heated, and LH heated plasmas. Helium profiles are observed to vary with plasma current, and also change in time during ICRF shots. Linear and nonlinear gyrokinetic simulations are performed for some shots using the GENE code. Sensitivity scans are done for magnetic shear, impurity density, and other plasma parameters and transport scalings are compared with experimental results. Simulated transport flux is compared with experimentally derived D and v parameters. Supported by USDoE awards DE-FG03-96ER-54373 and DE-FC02-99ER54512.

  3. In-out impurity density asymmetry in the pedestal region of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Churchill, R. M.; Lipschultz, B.; Theiler, C.; the Alcator C-Mod Team

    2013-12-01

    Measurements are presented of the boron (B5+) density in the pedestal region at the low-field side (LFS) midplane and the high-field side (HFS) midplane of Alcator C-Mod. In H-mode plasmas, a large (≳10x) in-out asymmetry in impurity density forms, with larger densities at the HFS. In contrast, there is no impurity density asymmetry in L-mode or I-mode plasmas. A comparison of pedestal parameters in H-mode plasmas shows that the HFS impurity density pedestal width and position remain fairly fixed over a range of plasma conditions, while the LFS pedestal width widens, and the pedestal position shifts in towards the core as the plasma current is lowered, indicating a change in the underlying transport.

  4. Plasma wave simulation based on a versatile FEM solver on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Shiraiwa, Syun'ichi

    2009-11-01

    A new efficient full wave simulation code of the lower hybrid (LH) wave was developed using the finite element method (FEM). A dielectric tensor consisting of the cold plasma contribution and the electron Landau damping (ELD) was used. The non-trivial problem of introducing non-local hot plasma effects into an FEM solver was addressed by iteratively solving the coupled problem of the Maxwell's equations with the convolution integral. With this approach, the EM problem is numerically sparse, and the computational requirements are reduced significantly compared to spectral domain solvers [1]. The simulation of an Alcator C-Mod scale plasma has been done on a desktop computer, suggesting the possibility of an ITER scale plasma simulation. The algorithm was implemented using a general purpose FEM software, COMSOL Multiphysics, and the simulation results of a Maxwellian tokamak plasma showed good agreement with ray tracing calculations in the strong single pass absorption regime. Integration of a Fokker-Planck calculation for a more realistic non-Maxwellian plasma is underway and initial results show reasonable shift of the power absorption towards the plasma edge [2]. Importantly, the FEM approach allows seamless handling of the core, SOL, and antenna regions. This flexibility has been exploited to address issues of antenna-plasma coupling in the LH and ICRF frequency ranges. Techniques to use the FEM package for this purpose were validated by solving the LH grill antenna coupling problem whose analytic solution is known. The code has been applied to a new Alcator C-Mod ICRF antenna to assess the antenna near field pattern [3]. [4pt] [1] J. C. Wright, et. al., Comput. Phys. 4, 545 (2008) [0pt] [2] O. Meneghini, et. al., this conference [0pt] [3] M. Garrett, et. al., this conference

  5. 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.

  6. 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.

  7. Fast Imaging of Filaments in the X-Point Region of Alcator C-MOD

    NASA Astrophysics Data System (ADS)

    Terry, J. L.; Ballinger, S.; Brunner, D.; Labombard, B.; White, A. E.; Zweben, S. J.

    2016-10-01

    A rich variety of field-aligned fluctuations has been revealed using fast imaging of Dα emission from Alcator C-Mod's lower X-point region. Field-aligned filamentary fluctuations are observed along the inner divertor leg, within the Private-Flux-Zone (PFZ), in the Scrape-Off Layer outside the outer divertor leg, and, under some conditions, at or above the X-point. The locations and dynamics of the filaments in these regions are strikingly complex in C-Mod. Changes in the filaments' generation appear to be ordered by plasma density and magnetic configuration. In a Lower Single Null with 0.12 < n /nGreenwald <0.45 and Bx ∇B directed down, filaments typically move up the inner divertor leg toward the X-point. Reversing the field direction results in the appearance of filaments outside of the outer divertor leg. With the divertor targets ``detached'', filaments inside the LCFS are seen. These studies were motivated by observations of filaments in the X-point and PFZ regions in MAST, and comparisons with those observations will be made. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC02-09CH11466.

  8. Access to and performance of I-mode plasmas on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Hubbard, A. E.; Wolfe, S. M.; Baek, S.-G.; Churchill, R. M.; Howard, N.; Hughes, J. W.; Lin, Y.; Marmar, E. S.; Reinke, M. L.; Rice, J. E.; Terry, J. L.; Theiler, C.; Walk, J. R.; White, A. E.; Whyte, D. G.; Wukitch, S. J.; Cziegler, I.

    2013-10-01

    The I-mode regime of operation features an edge thermal transport barrier, without a particle barrier. Stationary conditions are thus achieved without impurity accumulation, and usually without ELMs. In contrast to the EDA H-mode regime on Alcator C-Mod, it is readily accessed at low q95 and low collisionality, both relevant for ITER. Analysis of a dataset of 400 discharges at q95 ~ 3 shows normalized energy confinement in I-modes reaches or exceeds that in most H-modes, up to H98 = 1.2. Confinement and pedestal temperature increase with input power. In some cases I-mode is maintained up to the maximum available power (5 MW ICRF) while in others a transition to H-mode limits the performance. Understanding and extending the conditions for entering and staying in I-mode is thus critical for extrapolation of the regime. Experiments have extended the regime both to lower densities and to higher densities and powers through gas puffing into established I-modes. Results from an expanded database of C-Mod discharges will be presented, along with details of I-mode profiles and fluctuations, including GAMs and a weakly coherent mode, which are providing insights into the physics of the regime. Supported by DOE award DE-FC02-99ER54512.

  9. A new interferometry-based electron density fluctuation diagnostic on Alcator C-Mod.

    PubMed

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

    2012-10-01

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

  10. 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.

  11. Fast imaging of filaments in the X-point region of Alcator C-Mod

    DOE PAGES

    Terry, J. L.; Ballinger, S.; Brunner, D.; ...

    2017-01-27

    A rich variety of field-aligned fluctuations has been revealed using fast imaging of Dα emission from Alcator C-Mod's lower X-point region. Field-aligned filamentary fluctuations are observed along the inner divertor leg, within the Private-Flux-Zone (PFZ), in the Scrape-Off Layer (SOL) outside the outer divertor leg, and, under some conditions, at or above the X-point. The locations and dynamics of the filaments in these regions are strikingly complex in C-Mod. Changes in the filaments’ generation appear to be ordered by plasma density and magnetic configuration. Filaments are not observed for plasmas with n/nGreenwald ≲ 0.12 nor are they observed in Uppermore » Single Null configurations. In a Lower Single Null with 0.12 ≲ n/nGreenwald ≲ 0.45 and Bx∇B directed down, filaments typically move up the inner divertor leg toward the X-point. Reversing the field direction results in the appearance of filaments outside of the outer divertor leg. With the divertor targets “detached”, filaments inside the LCFS are seen. Lastly, these studies were motivated by observations of filaments in the X-point and PFZ regions in MAST, and comparisons with those observations are made.« less

  12. Comparison of Small ELM Characteristics and Regimes in Alcator C-Mod, MAST, and NSTX

    SciTech Connect

    Maingi, Rajesh; Hubbard, A.E.; Meyer, H.; Kirk, A.; Maqueda, R.; Terry, J. L.

    2011-01-01

    We report on the status of a set of ITPA-coordinated experiments between the Alcator C-Mod, MAST and NSTX devices to compare the characteristics and access conditions of discharges with small edge-localized modes (ELMs). The small ELMs in C-Mod, MAST and one of the two small ELM types in NSTX exist when {beta}{sub ped}{sup pol} approached 10-15%, although the lower/upper limits of the operational windows differ. These small ELM regimes appear in diverted configurations very close to balanced double-null in each device. We classify these small ELMs as type II, based on the published characteristics from a number of previous studies. In addition, these type II ELMs in each device had multiple filaments with propagation in the co-I{sub p} or ion diamagnetic drift direction. Moreover, we conclude that these type II ELMs are distinct from the type V ELMs routinely observed in NSTX, which have one or two filaments and propagate in the electron diamagnetic drift direction.

  13. Edge Ion Velocity Measurements with a Novel Doppler Spectrometer at the Alcator C-Mod Tokamak

    NASA Astrophysics Data System (ADS)

    Graf, Alexander; May, Mark; Beiersdorfer, Peter; Terry, Jim

    2006-10-01

    A high throughput, f/# ˜3.1, transmission grating Doppler spectrometer for visible light (3500-6700 å) is currently measuring ion or neutral velocities and temperatures at the Alcator C-Mod tokamak. The ion velocities are measured through the Doppler shift of impurities that are present in the plasma. A line width of as small as 0.4 å(velocity sensitivity of ˜10^5 cm/s) has been measured using calibration lamps. The spectrometer is fiber optically coupled and has access to toroidal and poloidal views. A spectral survey has been done with various views of the C-Mod plasma identifying various intrinsic impurities. The first Doppler measurements of B II were recorded with ˜15 ms per frame. Additional Doppler velocity and temperature measurements in both poloidal and toroidal directions for some of the brighter impurities (e.g. He II and N III), will be given. This work is supported was performed under the auspices of the DoE by UC LLNL under contract W-7405-ENG-48 and also under DoE Coop. Agreement DE-FC02-99ER54512.

  14. Observations of Intrinsic Rotation Reversal Hysteresis in Alcator C-Mod Plasmas

    NASA Astrophysics Data System (ADS)

    Cao, Norman; Rice, John; White, Anne; Baek, Seung; Chilenski, Mark; Creely, Alexander; Ennever, Paul; Hubbard, Amanda; Hughes, Jerry; Irby, Jim; Rodriguez-Fernandez, Pablo; Reinke, Matthew; Diamond, Patrick; Alcator C-Mod Team

    2016-10-01

    Intrinsic core toroidal rotation in Alcator C-Mod L-mode plasmas has been observed to spontaneously reverse direction when the normalized collisionality ν*, evaluated at the profile minimum, passes through a critical value around 0.4. In Ohmic plasmas, the low density linear Ohmic confinement regime exhibits co-current toroidal rotation, and the higher density saturated Ohmic confinement regime exhibits counter-current rotation. The reversal manifests a hysteresis loop in ν*, where the critical collisionalities for the forward and reverse transitions differ by 10-15%. There appears to be memory associated with the rotation state, since reversals which do not begin from fully saturated rotation states do not manifest this hysteresis. In addition, high-k PCI fluctuation ``wings'' (kθρs up to 1) at low density and high current appear only in the co-current rotation state, while density peaking and ``non-local'' heat transport behavior do not appear to change significantly with the rotation state. Results from fluctuation measurements and preliminary transport and stability analyses will also be presented. This work is supported by the US DOE under Grant DE-FC02-99ER54512 (C-Mod).

  15. Advances in lower hybrid current drive technology on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Wallace, G. M.; Shiraiwa, S.; Hillairet, J.; Preynas, M.; Beck, W.; Casey, J. A.; Doody, J.; Faust, I. C.; Fitzgerald, E.; Johnson, D. K.; Kanojia, A. D.; Koert, P.; Lau, C.; Lin, Y.; Leccacorvi, R.; MacGibbon, P.; Meneghini, O.; Murray, R.; Parker, R. R.; Terry, D. R.; Vieira, R.; Wilson, J. R.; Wukitch, S.; Zhou, L.

    2013-07-01

    Lower hybrid current drive (LHCD) is an attractive option for non-inductive tokamak operation due to its high current drive efficiency and ability to drive current off axis. The parameters of the Alcator C-Mod LHCD system (f0 = 4.6 GHz, Bφ ≃ 5.5 T, \\bar{n}_\\rme \\simeq 10^{20}\\,m^{-3} ) are similar to the proposed LHCD system on ITER. This paper will describe improvements in LHCD technology on C-Mod designed to increase single-pass absorption at high \\bar{n}_\\rme , extend pulse length (to >3 s), and increase power delivered to the plasma (to ∼2 MW). Modelling of lower hybrid (LH) wave propagation indicates that the observed loss of LHCD efficiency at higher \\bar{n}_\\rme can be mitigated by enhancing the single pass power absorption through use of an off mid-plane launcher. The four rows of the launcher are located above the mid-plane (with Ip and Bφ both clockwise viewing from the top down) in order to exploit the poloidal upshift of n‖ as rays propagate from the antenna into the plasma. The transmitter protection system (TPS) was redesigned to model the coolant temperature in real time and shut off the klystron beam voltage if the coolant is close to boiling. The TPS upgrade has been installed and operated on C-Mod for pulses up to 4.5 s into dummy loads and 1.0 s into the plasma. A new movable local LH launcher protection limiter was designed to reduce reflection coefficients across a wide range of launcher positions. Finally, a high power waveguide double-stub tuner is under development to provide feedback controlled load matching to reduce power reflected from the antenna under poor coupling conditions.

  16. 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.

  17. 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.

  18. Marginal Stability Studies of Microturbulence Near ITB Onset on Alcator C-Mod

    SciTech Connect

    Baumgaertel, J.; Redi, M.H.; Budny, R.V.; McCune, D.C.; Dorland, W.; Fiore, C.L.

    2005-01-01

    Insight into microturbulence and transport in tokamak plasmas is being sought using linear simulations of drift waves near the onset time of an internal transport barrier (ITB) on Alcator C-Mod. Microturbulence is likely generated by instabilities of drift waves and causes transport of heat and particles. This transport is studied because the containment of heat and particles is important for the achievement of practical nuclear fusion. We investigate nearness to marginal stability of ion temperature gradient (ITG) modes for conditions in the ITB region at the trigger time for ITB formation. Data from C-Mod, analyzed by TRANSP (a time dependent transport analysis code), is read by the code TRXPL and made into input files for the parallel gyrokinetic model code GS2. Temperature and density gradients in these input files are modified to produce new input files. Results from these simulations show a weak ITG instability in the barrier region at the time of onset, above marginal stability; the normalized critical temperature gradient is 80% of the experimental temperature gradient. The growth rate increases linearly above the critical value, with the spectrum of ITG modes remaining parabolic up to a multiplicative factor of 2. The effect of varying density gradients is found to be much weaker and causes the fastest growing drift mode to change from ITG to trapped electron mode character. Simulations were carried out on the NERSC IBM 6000 SP using 4 nodes, 16 processors per node. Predictive simulations were examined for converged instability after 10,000-50,000 timesteps in each case. Each simulation took approximately 30 minutes to complete on the IBM SP.

  19. Marginal Stability of Microturbulence near ITB Onset on Alcator C-Mod

    SciTech Connect

    J.A. Baumgaertel; M.H. Redi; R.V. Budny; D.C. McCune; W. Dorland; C.L. Fiore

    2004-08-23

    Insight into microturbulence and transport in tokamak plasmas is being sought using linear simulations of drift waves near the onset time of an internal transport barrier (ITB) on Alcator C-Mod. Microturbulence is likely generated by instabilities of drift waves and causes transport of heat and particles. This transport is studied because the containment of heat and particles is important for the achievement of practical nuclear fusion. We investigate nearness to marginal stability of ion-temperature-gradient (ITG) modes for conditions in the ITB region at the trigger time for ITB formation. Data from C-Mod, analyzed by TRANSP (a time-dependent transport analysis code), is read by the code TRXPL and made into input files for the parallel gyrokinetic model code GS2. Temperature and density gradients in these input files are modified to produce new input files. Results from these simulations show a weak ITG instability in the barrier region at the time of onset, above marginal stability; the normalized critical temperature gradient is 80% of the experimental temperature gradient. The growth rate increases linearly above the critical value, with the spectrum of ITG modes remaining parabolic up to a multiplicative factor of 2. The effect of varying density gradients is found to be much weaker and causes the fastest growing drift mode to change from ITG to trapped-electron mode character. Simulations were carried out on the NERSC [National Energy Research Supercomputer Center] IBM 6000 SP using 4 nodes, 16 processors per node. Predictive simulations were examined for converged instability after 10,000-50,000 time-steps in each case. Each simulation took approximately 30 minutes to complete on the IBM SP.

  20. 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.

  1. 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.

  2. 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.

  3. Electron heating via mode converted ion Bernstein waves in the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Bonoli, P. T.

    1996-11-01

    Highly localized electron heating (FWHM <~ 0.2 a) via mode converted ion Bernstein waves (IBW) has been demonstrated in the Alcator C-Mod tokamak. These experiments were carried out using 80 MHz fast wave ICRF power at P_rf <~ 2.4 MW. Electron heating at or near the plasma center (r/a <~ 0.3) has been observed in H-(^3He) plasmas at B0 = (6.4 - 7.3)T. In this case the ion-ion hybrid layer is near the plasma center and the fundamental H and ^3He cyclotron resonances are located respectively on the low and high field sides of the tokamak. Off-axis heating (r/a >≈ 0.5) has also been observed in D-(^3He) plasmas at 7.9 T. In this case the ^3He cyclotron resonance is on-axis and the fundamental D resonance and mode conversion layer are on the high field side of the tokamak. The concentration of ^3He in these experiments was in the range n_^3He / ne ~= (0.2 - 0.3) and the location of the mode conversion layer was controlled by changing the ^3He concentration or the toroidal magnetic field. The rf heating profiles were deduced using an rf power modulation technique in which the local electron heating rate was obtained from a ``break in slope'' analysis of the measured electron temperature versus time. Detailed comparisons with 1-D and toroidal full-wave ICRF models (FELICE and FISIC codes) have been carried out. The 1-D predictions for the fractional electron power absorption and damping location are found to be in qualitative agreement with the experiment. However discrepancies have been found between the full-wave toroidal code predictions and experiment. This disagreement is thought to be due to a lack of radial and poloidal resolution in the solution of the mode converted ion Bernstein wave in toroidal geometry and will be discussed. A fast wave current drive package has been modified to study the current generated via the mode converted IBW. Based on these numerical results and the experimental results for power absorption, off-axis current of up to 200 kA is

  4. 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.

  5. 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.

  6. Intrinsic rotation in DIII-D

    SciTech Connect

    DeGrassie, J. S.; Rice, J. E.; Burrell, K. H.; Groebner, R. J.; Solomon, W. M.

    2007-05-15

    In the absence of any auxiliary torque input, the DIII-D plasma consists of nonzero toroidal angular momentum, in other words, it rotates. This effect is commonly observed in tokamaks, being referred to as intrinsic rotation. Measurements of intrinsic rotation profiles have been made in DIII-D [J. Luxon, Nucl. Fusion 42, 614 (2002)] H-mode discharges, with both Ohmic heating (OH) and electron cyclotron heating (ECH) in which there is no auxiliary torque. Recently, the H-mode data set has been extended with the newly configured DIII-D simultaneous co- and counter-directed neutral beam injection (NBI) capability resulting in control of the local torque deposition, where co and counter refer to the direction relative to the toroidal plasma current. Understanding intrinsic rotation is important for projection toward burning plasma performance where any NBI torque will be relatively small. The toroidal velocity is recognizably important regarding issues of stability and confinement. In DIII-D ECH H-modes the rotation profile is hollow, co-directed at large minor radius and depressed, or actually counter-directed, nearer the magnetic axis. This profile varies with the ECH power deposition profile to some extent. In contrast, OH H-modes have a relatively flat co-directed rotation profile. There is a scaling of the DIII-D intrinsic toroidal velocity with W/I{sub p}, as seen in intrinsic rotation in Alcator C-Mod [J. Rice, Nucl. Fusion 39, 1175 (1999)], where W is the total plasma thermal energy and I{sub p} is the magnitude of the toroidal plasma current. This common scaling resulted in a dimensionless similarity experiment between DIII-D and Alcator C-Mod on intrinsic rotation, obtaining a single spatial point match in the toroidal velocity normalized to the ion thermal velocity. The balanced NBI capability in DIII-D is a useful tool to push scaling studies to higher values of the plasma normalized energy, notwithstanding the details of torque deposition for co-NBI versus

  7. Motivation and goals of the new heated outer divertor for Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lipschultz, B.; Doody, J.; Ellis, R.; Granetz, R.; Harrison, S.; Labombard, B.; Vieira, R.; Zhang, H.; Zhou, L.

    2012-10-01

    A precision-aligned, high-temperature outer divertor is being developed for Alcator C-Mod to enhance heatflux handling and to advance our knowledge and experience with high-Z Plasma Facing Components (PFCs) in a reactor-level power density environment. Several departures from the design of the current divertor will be implemented: Instead of 10 toroidal divertor segments that expand toroidally as they heat up, the divertor plate will be toroidally continuous, with no openings or leading edges in the high-heat flux region. It will expand in the radial direction when heated while maintaining good alignment with shallow field line angles (˜ 2 degrees), a requirement for future divertors. Those characteristics will reduce both impurity sources and disruption forces. A second design goal is to be able to control the divertor temperature up to 600^oC by installing heaters in the structure. Given the Arrhenius relation between hydrogen diffusivity and temperature in tungsten (and molybdenum) this will open up a new area of study for tokamaks - exploration of the effect of PFC temperature on fuel retention. Temperature control may also open up a new area of study into the effect of changes in divertor recycling on fueling and core confinement.

  8. Core impurity transport in Alcator C-Mod L-, I- and H-mode plasmas

    NASA Astrophysics Data System (ADS)

    Rice, J. E.; Reinke, M. L.; Gao, C.; Howard, N. T.; Chilenski, M. A.; Delgado-Aparicio, L.; Granetz, R. S.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Lin, Y.; Marmar, E. S.; Mumgaard, R. T.; Scott, S. D.; Terry, J. L.; Walk, J. R.; White, A. E.; Whyte, D. G.; Wolfe, S. M.; Wukitch, S. J.

    2015-03-01

    Core impurity transport has been investigated for a variety of confinement regimes in Alcator C-Mod plasmas from x-ray emission following injection of medium and high Z materials. In ohmic L-mode discharges, impurity transport is anomalous (Deff ≫ Dnc) and changes very little across the LOC/SOC boundary. In ion cyclotron range of frequencies (ICRF) heated L-mode plasmas, the core impurity confinement time decreases with increasing ICRF input power (and subsequent increasing electron temperature) and increases with plasma current. Nearly identical impurity confinement characteristics are observed in I-mode plasmas. In enhanced Dα H-mode discharges the core impurity confinement times are much longer. There is a strong connection between core impurity confinement time and the edge density gradient across all confinement regimes studied here. Deduced central impurity density profiles in stationary plasmas are generally flat, in spite of large amplitude sawtooth oscillations, and there is little evidence of impurity convection inside of r/a = 0.3 when averaged over sawteeth.

  9. Investigating electromagnetic effects on core transport in Alcator C-Mod H-mode discharges

    NASA Astrophysics Data System (ADS)

    Guttenfelder, W.; Howard, N. T.; Irby, J.; Poli, F. M.; White, A. E.; Bergerson, W. F.; Brower, D. L.; Ding, W. X.; Kessel, C. E.; Sung, C.; Wolfe, S. M.; Xu, P.

    2014-10-01

    Understanding the importance of electromagnetic effects on core turbulence and transport is being pursued at Alcator C-Mod, especially for higher performance H-mode plasmas at increasing beta. Previously reported measurements from a line-integrated polarimeter diagnostic reveal broadband, high frequency fluctuations. The presence of these features, absent in core and edge density fluctuation measurements from phase contrast imaging, suggest they may be related to fluctuations in the magnetic field. Such features were observed in a number of H-mode plasmas over a range of normalized beta (βN ~ 1-2) and Greenwald fraction (fGW ~ 0.45-0.85). To investigate the possible influence of electromagnetic effects on core transport and turbulence, gyrokinetic simulations are used to predict microinstability of these discharges, the corresponding relative amplitude of the magnetic fluctuations in comparison to density fluctuations, and the sensitivity of these predictions to variations in beta. Results of both linear and nonlinear simulations and their comparison with transport and turbulence measurements will be presented. This work is supported by US DOE Contracts DE-AC02-09CH11466 and DE-FC02-99ER54512.

  10. Radiation asymmetry and MHD activity in gas jet rapid shutdowns on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Olynyk, Geoffrey; Granetz, Robert; Whyte, Dennis; Alcator C-Mod Team

    2013-10-01

    Radiative rapid shutdown via massive noble gas injection (MGI) is an integral part of the ITER disruption mitigation system (DMS). However, observations have shown that the radiation during MGI rapid shutdowns may be spatially asymmetric, particularly during the initial phase when the plasma's thermal energy is converted to radiation. ITER requires the radiation peaking factor (PF) to be less than approximately 2.0 to 2.5 in this thermal quench (TQ) phase in order to prevent melting of the beryllium wall even in the case of a successful MGI rapid shutdown. We report on observations of rotating MHD modes in single- and multiple-gas-jet rapid shutdowns on Alcator C-Mod, and discuss the role of mode rotation during the TQ in setting the radiation peaking factor. The implications for the ITER DMS are discussed. This work was supported by the United States Department of Energy under Contract No. DE-FC02-99ER54512 and the Natural Sciences and Engineering Research Council of Canada PGS D program.

  11. 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.

  12. Statistics of fluctuation induced transport in the scrape-off layer of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Kube, Ralph; Garcia, Odd Erik; Theodorsen, Audun; Labombard, Brian; Terry, James

    2016-10-01

    The fluctuation induced transport in the scrape-off layer of Alcator C-Mod is investigated in an ohmically heated lower single-null discharge using Mirror Langmuir Probes. The probes are connected to a horizontal scanning probe which dwells at the outboard mid plane limiter radius and to electrodes in the outer divertor baffle. At the limiter radius the electron density, electron temperature and plasma potential are correlated with linear correlation coefficients r of approximately r=0.8. The bursts show a steep rise and a decay on a time scales of approximately 5 and 10 microseconds respectively. Amplitudes of bursts in the density, temperature, and plasma potential time series are correlated with r approximately 0.7-0.8. Conditionally averaged bursts in the radial particle and heat flux time series are less coherent and less reproducible, their amplitudes are correlated to the amplitude of bursts in the density time series with r=0.4. Statistics of the fluctuating plasma parameters at the outer divertor baffle are qualitatively similar to those at outboard midplane. Histograms, as well as statistics for level crossings and excess times spent above a given threshold for the time series compare favorably to a stochastic model for time series of scrape-off layer plasmas.

  13. Lower hybrid current drive experiments on Alcator C-Mod: Comparison with theory and simulationa)

    NASA Astrophysics Data System (ADS)

    Bonoli, P. T.; Ko, J.; Parker, R.; Schmidt, A. E.; Wallace, G.; Wright, J. C.; Fiore, C. L.; Hubbard, A. E.; Irby, J.; Marmar, E.; Porkolab, M.; Terry, D.; Wolfe, S. M.; Wukitch, S. J.; Alcator C-Mod Team; Wilson, J. R.; Scott, S.; Valeo, E.; Phillips, C. K.; Harvey, R. W.

    2008-05-01

    Lower hybrid (LH) current drive experiments have been carried out on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] using a radio-frequency system at 4.6GHz. Up to 900kW of LH power has been coupled and driven LH currents have been inferred from magnetic measurements by extrapolating to zero loop voltage, yielding an efficiency of neILHR0/PLH≈2.5±0.2×1019(A/W/m2). We have simulated the LH current drive in these discharges using the combined ray tracing/three-dimensional (r,v⊥,v∥) Fokker-Planck code GENRAY-CQL3D (R. W. Harvey and M. McCoy, in Proceedings of the IAEA Technical Committee Meeting on Simulation and Modeling of Thermonuclear Plasmas, Montreal, Canada, 1992) and found similar current drive efficiencies. The simulated profiles of current density from CQL3D, including both ohmic plus LH drive have been found to be in good agreement with the measured current density from a motional Stark effect diagnostic. Measurements of nonthermal x-ray emission confirm the presence of a significant fast electron population and the three-dimensional (r,v⊥,v∥) electron distribution function from CQL3D has been used in a synthetic diagnostic code to simulate the measured hard x-ray data.

  14. Reduction of Core Turbulence in I-mode Plasmas at Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    White, A. E.; Barnes, M.; Dominguez, A.; Greenwald, M.; Howard, N. T.; Hubbard, A. E.; Hughes, J. W.; Mikkelsen, D. R.; Parra, F. I.; Reinke, M. L.; Sung, C.; Walk, J.; Whyte, D. G.

    2013-10-01

    Core turbulent fluctuations (0.40 < r/a < 0.95) are reduced in the high confinement regime, I-mode, at Alcator C-Mod. Long wavelength density fluctuation levels are observed to decrease from L-mode levels by up to 30%, while long wavelength electron temperature fluctuation levels are observed to decrease by up to 70%. This reduction in core turbulence is correlated with the increases in confinement in I-mode. During some L-I transitions, density fluctuations in the core (r/a ~ 0.5) and near the top of the Te pedestal (r/a ~ 0.95) are both reduced prior to the reduction of low-frequency edge turbulence (r/a ~ 0.99-1.0), and prior to the onset of the edge-localized weakly coherent mode (WCM) (r/a ~ 0.99-1.0). This result contrasts typical observations of confinement transitions (I-mode and H-mode) where changes in edge turbulence occur prior to changes in core turbulence. Supported by USDoE: DE-FC02-99ER54512 and DE-SC0006419.

  15. Reduction of core turbulence in I-mode plasmas in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    White, A. E.; Barnes, M.; Dominguez, A.; Greenwald, M.; Howard, N. T.; Hubbard, A. E.; Hughes, J. W.; Mikkelsen, D. R.; Parra, F. I.; Reinke, M. L.; Sung, C.; Walk, J.; Whyte, D. G.

    2014-08-01

    In this paper, we report observations of reduced core (0.40 < ρ < 0.95) fluctuations in the edge localized mode (ELM)-free high-confinement regime, I-mode, at Alcator C-Mod (Marmar et al2009 Nucl. Fusion 49 104014). Long wavelength (kθρs < 0.5) density fluctuation levels are observed to decrease from L-mode levels by up to 30% in I-mode, while long wavelength (kθρs < 0.3) electron temperature fluctuation levels are observed to decrease by up to 70% in I-mode. This reduction in core turbulence is correlated with the increases in confinement in I-mode compared to L-mode. As the pedestal temperature increases across the L-I transition, core density fluctuations (0.40 < ρ < 0.95) are reduced prior to the onset of the edge-localized (ρ ˜ 0.99-1.0) weakly coherent mode (WCM) and prior to the reduction of low-frequency (ρ ˜ 0.99-1.0) turbulence in the edge/pedestal region. This result helps add to our understanding of the dynamics of confinement transitions such as I-mode and H-mode, where changes in edge turbulence are more typically observed to occur prior to changes in core turbulence.

  16. Alcator C-Mod's Quasi-Coherent Mode Antenna: Experimental Results and Interpretation

    NASA Astrophysics Data System (ADS)

    Golfinopoulos, T.; Labombard, B.; Parker, R. R.; Burke, W.; Davis, E. M.; Granetz, R.; Greenwald, M.; Marmar, E.; Porkolab, M.; Terry, J. L.; Wolfe, S. M.; Woskov, P. P.; Xu, X.

    2012-10-01

    A new ``Shoelace'' antenna has been installed on Alcator C-Mod. Its goal is to interact with edge fluctuations, and particularly the quasi-coherent mode (QCM) associated with the steady-state EDA H-mode. With k=1.5 cm-1 and frequency range, 40

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

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

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

  18. 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.

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

    SciTech Connect

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

    2010-05-15

    The steep gradient edge region and scrape-off-layer (SOL) on the low-field-side of Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] tokamak plasmas are studied using gas-puff-imaging diagnostics. In L-mode plasmas, the region extending approx2 cm inside the magnetic separatrix has fluctuations showing a broad, turbulent spectrum, propagating in the electron diamagnetic drift direction, whereas features in the open field line region propagate in the ion diamagnetic drift direction. This structure is robust against toroidal field strength, poloidal null-point geometry, plasma current, and plasma density. Global parameter dependence of spectral and spatial structure of the turbulence inside the separatrix is explored and characterized, and both the intensity and spectral distributions are found to depend strongly on the plasma density normalized to the tokamak density limit. In H-mode discharges the fluctuations at and inside the magnetic separatrix show fundamentally different trends compared to L-mode, with the electron diamagnetic direction propagating turbulence greatly reduced in ELM-free [F. Wagner et al., Proceedings of the Thirteenth Conference on Plasma Physics and Controlled Nuclear Fusion Research (IAEA, Vienna, 1982), Vol. I, p. 277], and completely dominated by the modelike structure of the quasicoherent mode in enhanced D-alpha regimes [A. E. Hubbard, R. L. Boivin, R. S. Granetz et al., Phys. Plasmas 8, 2033 (2001)], while the normalized SOL turbulence is largely unaffected.

  20. Design, Engineering, and Testing for the Alcator C-Mod Outer Divertor Upgrade

    NASA Astrophysics Data System (ADS)

    Harrison, S.; Vieira, R.; Lipschultz, B.; Ellis, R.; Karnes, D.; Doody, J.; Zhou, L.; Titus, P.; Zhang, H.; Beck, W.; Granetz, R.

    2012-10-01

    Alcator C-mod's major outer divertor upgrade will enable significant advances in our understanding of reactor relevant physics and operations. Two primary features of the new outer divertor are its toroidally continuous design (electrical and mechanical), and ability to be operated up to or independently heated to 600 C. Full control of the divertor PFC temperature from ambient vessel temperature to 600 C, will enable new and important tokamak research into the temperature dependence of fuel retention, PFC deposition and erosion, and divertor recycling. Significant design, analysis, and testing is underway to complete this important and challenging upgrade, which will provide valuable information for ITER and future reactors. Among other aspects of the innovative approach, the divertor plate supports, halo current shunts, and thermal shield assemblies will be discussed. The divertor supports enable pure radial motion of the divertor ring as it expands thermally and robustness to massive disruption induced electro-mechanical loads. Halo current shunts conduct 400kA in an 8T magnetic field and allow for divertor displacement relative to the vessel. Thermal shielding significantly reduces radiation and conduction to surrounding vessel structures.

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

    DOE PAGES

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

    2016-08-17

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

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

    SciTech Connect

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

    2016-08-17

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

  3. The dynamics and structure of edge-localized-modes in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Terry, J. L.; Cziegler, I.; Hubbard, A. E.; Snipes, J. A.; Hughes, J. W.; Greenwald, M. J.; LaBombard, B.; Lin, Y.; Phillips, P.; Wukitch, S.

    2007-06-01

    Characteristics of discrete ELMs produced in Alcator C-Mod discharges of low edge collisionality (0.2 < ν∗ < 1) and large lower triangularity (δlower ∼ 0.75) are examined. The energy lost per ELM from the H-mode pedestal is ∼10% of the pedestal energy. These ELMs exhibit relatively long-lived precursor oscillations, often with two modes of intermediate toroidal mode number present. At the ELM 'crash' multiple plasma filament structures are expelled into the scrape-off-layer. A short-lived high frequency (∼0.5 MHz) magnetic oscillation is initiated at the 'crash'. The initial ELM filaments are large perturbations to the SOL with radial extents of 0.5-1 cm and typical radial propagation velocities of 1 km/s. Velocities of up to 8 km/s have been seen. The poloidal extent of the initial filaments is >4.5 cm. The initial filaments are followed (at intervals of ∼100 μs) by multiple, less perturbing secondary filaments.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

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

  10. Hard x-ray diagnostic for lower hybrid experiments on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Liptac, J.; Parker, R.; Tang, V.; Peysson, Y.; Decker, J.

    2006-10-01

    Alcator C-Mod's lower hybrid current drive (LHCD) system allows the exploration of advanced tokamak (AT) regimes at densities relevant to ITER and fusion reactors. The location of the LHCD is critical to AT performance and may be inferred by measuring the nonthermal bremsstrahlung emission in the hard x-ray (HXR) region. A pinhole camera using an array of 32 CdZnTe detectors is used to image energies in the 20-200keV range. Detectors and pulse processing electronics are integrated into a compact and modular package making extensive use of printed circuit board and surface mount technology. The system also makes use of fast digitization and software signal processing techniques. An ambient environment of neutrons, gammas, and high rf power requires careful shielding. Shielding is studied using the neutron and photon transport code MCNP. The design of the diagnostic is presented along with background measurements in lieu of LHCD fast electrons. Background measurements are then compared to advanced modeling results to predict the power threshold for meaningful HXR data for a H-mode target plasma.

  11. Fluctuating Zonal Flows in I-mode in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Cziegler, Istvan

    2012-10-01

    Velocity fields and density fluctuations of edge turbulence have been studied in I-mode [1] plasmas of Alcator C-Mod, which are characterized by a strong thermal transport barrier in the edge while providing little or no barrier to the transport of both bulk and impurity particles. This allows access to steady state, high performance discharges without explosive edge relaxations or impurity accumulation. The key feature in the I-mode edge seems to be a weakly coherent mode (WCM) at 100-300 kHz, with δf 150 kHz and a poloidal wavenumber k 1.5,-1. Although previous work showed no clear geodesic-acoustic modes (GAM) on C-Mod, using a newly implemented, gas-puff-imaging (GPI) based time-delay-estimate (TDE) velocity inference algorithm, GAM are now shown to be ubiquitous in all I-mode discharges, with the time histories of the GAM and the WCM closely following each other through the entire duration of the regime. The central frequency of the WCM is shown to scale with HITER,98, which itself scales with the depth of the radial electric field well in the edge [2]. Thus, the I-mode presents an example of a plasma state in which quasi-static zonal flows (ZF) and GAM continuously coexist. Using both single- (density) and two-field (density-velocity) bispectral methods, the GAM are shown to be coupled to the WCM and to be responsible for its broad frequency structure. Since the WCM activity is strongly correlated to the I-mode behavior [3], and due to the known dependence of the GAM damping on collisionality [4], the decrease in GAM amplitude, and with it WCM activity, at higher densities offers an explanation for the density limit for I-mode access [3].[4pt] [1] F. Ryter et al, Plasma Phys. Control. Fusion 40 725 (1998)[0pt] [2] R. McDermott et al, Phys. Plasmas 16 056103 (2009)[0pt] [3] D. Whyte et al, Nucl. Fus. 50 105005 (2010)[0pt] [4] S. Novakovskii et al, Phys. Plasmas 4 4272 (1997)

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

    SciTech Connect

    Garnier, Darren Thomas

    1996-06-01

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

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

    SciTech Connect

    Urbahn, John 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.

  14. 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.

  15. 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.

  16. Radial localization of edge modes in Alcator C-Mod pedestals using optical diagnostics

    NASA Astrophysics Data System (ADS)

    Theiler, C.; Terry, J. L.; Edlund, E.; Cziegler, I.; Churchill, R. M.; Hughes, J. W.; LaBombard, B.; Golfinopoulos, T.; the Alcator C-Mod Team

    2017-02-01

    Dedicated experiments in ion cyclotron range heated enhanced D-alpha (EDA) H-mode and I-mode plasmas have been performed on Alcator C-Mod to identify the location of edge fluctuations inside the pedestal and to determine their plasma frame phase velocity. For this purpose, measurements from gas puff imaging (GPI) and gas puff charge exchange recombination spectroscopy (GP-CXRS) have been collected using the same optical views. The data suggest that the EDA H-mode-specific quasi-coherent mode (QCM) is centered near the radial electric field (E r) well minimum and propagates along the ion diamagnetic drift direction in the plasma frame. The weakly coherent mode (WCM) and the geodesic acoustic mode observed in I-mode, on the other hand, are found to be located around the outer shear layer of the E r well. This results in a weak plasma frame phase velocity mostly along the electron diamagnetic drift direction for the WCM. The findings in these EDA H-mode plasmas differ from probe measurements in ohmic EDA H-mode (LaBombard et al 2014 Phys. Plasmas 21 056108), where the QCM was identified as an electron drift-wave located several mm outside the E r well minimum in a region of positive E r. To explore if instrumental effects of the optical diagnostics could be the cause of the difference, a synthetic diagnostic for GPI is introduced. This diagnostic reproduces amplitude ratios and relative radial shifts of the mode profiles determined from poloidally and toroidally oriented optics and, if instrumental effects related to GP-CXRS are also included, indicates that the measured location of the QCM and WCM relative to the E r well reported here is only weakly affected by instrumental effects.

  17. Overview of experimental results and code validation activities at Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Greenwald, M.; Bader, A.; Baek, S.; Barnard, H.; Beck, W.; Bergerson, W.; Bespamyatnov, I.; Bitter, M.; Bonoli, P.; Brookman, M.; Brower, D.; Brunner, D.; Burke, W.; Candy, J.; Chilenski, M.; Chung, M.; Churchill, M.; Cziegler, I.; Davis, E.; Dekow, G.; Delgado-Aparicio, L.; Diallo, A.; Ding, W.; Dominguez, A.; Ellis, R.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fitzgerald, E.; Fredian, T.; Garcia, O. E.; Gao, C.; Garrett, M.; Golfinopoulos, T.; Granetz, R.; Groebner, R.; Harrison, S.; Harvey, R.; Hartwig, Z.; Hill, K.; Hillairet, J.; Howard, N.; Hubbard, A. E.; Hughes, J. W.; Hutchinson, I.; Irby, J.; James, A. N.; Kanojia, A.; Kasten, C.; Kesner, J.; Kessel, C.; Kube, R.; LaBombard, B.; Lau, C.; Lee, J.; Liao, K.; Lin, Y.; Lipschultz, B.; Ma, Y.; Marmar, E.; McGibbon, P.; Meneghini, O.; Mikkelsen, D.; Miller, D.; Mumgaard, R.; Murray, R.; Ochoukov, R.; Olynyk, G.; Pace, D.; Park, S.; Parker, R.; Podpaly, Y.; Porkolab, M.; Preynas, M.; Pusztai, I.; Reinke, M.; Rice, J.; Rowan, W.; Scott, S.; Shiraiwa, S.; Sierchio, J.; Snyder, P.; Sorbom, B.; Soukhanovskii, V.; Stillerman, J.; Sugiyama, L.; Sung, C.; Terry, D.; Terry, J.; Theiler, C.; Tsujii, N.; Vieira, R.; Walk, J.; Wallace, G.; White, A.; Whyte, D.; Wilson, J.; Wolfe, S.; Woller, K.; Wright, G.; Wright, J.; Wukitch, S.; Wurden, G.; Xu, P.; Yang, C.; Zweben, S.

    2013-10-01

    Recent research on the Alcator C-Mod tokamak has focused on a range of scientific issues with particular emphasis on ITER needs and on detailed comparisons between experimental measurements and predictive models. Research on ICRF (ion cyclotron range of frequencies) heating emphasized the origins and mitigation of metallic impurities while work on lower hybrid current drive experiments have focused on linear and nonlinear wave interactions that limit efficiency at high densities in regimes with low single pass absorption. Experiments in core turbulence and transport focused on quantitative, multi-field comparisons between nonlinear gyro-kinetics simulations and experimental measurements of profiles, fluxes and fluctuations. Experiments into self-generated rotation observed spontaneous flow reversal at a critical density identical to the transition density between linear ohmic confinement and saturated ohmic confinement regimes. H-mode studies have measured pedestal widths consistent with kinetic-ballooning-mode-like instabilities, while the pedestal heights quantitatively match the EPED code predictions. Experiments with I-mode have increased the operating window for this promising edge-localized-mode-free regime. Extrapolation of I-mode to ITER suggests that the fusion gain Q ∼ 10 could be possible in ITER. Investigations into the physics and scaling of the power exhaust channel width in attached enhanced D-alpha H-mode and L-mode plasma showed a direct connection between the midplane pressure-folding length and the outer divertor target footprint. The width was found to scale inversely with IP, while being independent of conducted power, BT or q95 and insensitive to the scrape-off layer connection length—a behaviour that suggests critical-gradient physics sets both pressure and heat-flux profiles.

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

    NASA Astrophysics Data System (ADS)

    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.

  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.

  20. 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.

  1. Measurements and Modeling of X-Ray and ECE Spectra During LHCD Experiments on Alcator C-Mod

    SciTech Connect

    Schmidt, A. E.; Bonoli, P. T.; Hubbard, A. E.; Liptac, J.; Parker, R. R.; Wallace, G. M.; Wilson, J. R.; Harvey, R. W.; Smirnov, A. P.

    2007-09-28

    Lower Hybrid Current Drive (LHCD) has been demonstrated on Alcator C-Mod. The LH-driven fast electrons can be detected through hard x-ray Bremsstrahlung emission as well as relativistically downshifted electron cyclotron emission (ECE). C-Mod has a hard x-ray camera with 32 chords, as well as several outboard midplane ECE diagnostics. CQL3D/GENRAY is a modeling package that employs a 3-D Fokker-Planck solver to compute steady-state distribution functions for a given LH N{sub parallel} spectrum and plasma. It also can perform self-consistent synthetic diagnostic calculations. LH phase and power scans have been carried out. Fast electron diffusion time can be inferred from x-ray data. Experimental measurements are compared with synthetic diagnostic modeling to see how changes in phasing and power affect current profile control and to benchmark CQL3D in ITER-relevant regimes.

  2. Lower Hybrid Current Drive Experiments on Alcator C-Mod: Comparison with Theory and Simulation

    NASA Astrophysics Data System (ADS)

    Bonoli, Paul

    2007-11-01

    Recently, lower hybrid current drive (LHCD) experiments have been carried out on Alcator C-Mod using an RF system consisting of 12 klystrons at 4.6 GHz, feeding a 4 x 22 waveguide array. Up to 900 kW of LH power has been coupled in the range1.6 <= n//<= 4), where n// is the parallel refractive index. Driven LH currents have been inferred from magnetic measurements by extrapolating to zero loop voltage, yielding an efficiency of n20ILHR/PLH 0.3 [1]. We have simulated the LH current drive in these discharges using the combined ray tracing / 3D (r, v, v//) Fokker Planck code GENRAY -- CQL3D [2] and found similar current drive efficiencies. Measurements of nonthermal x-ray emission and electron cyclotron emission (ECE) confirm the presence of a significant fast electron population that varies with waveguide phasing and plasma density. Studies are currently underway to investigate the role of fast electron diffusion and full-wave effects such as diffractional broadening in determining the spatial and velocity space structure of the nonthermal electrons. The 3D (r, v, v//) electron distribution function from CQL3D has been used in synthetic diagnostic codes to simulate the measured hard x-ray and ECE emissions. Fast electron diffusion times have been inferred from x-ray data by employing a radial diffusion operator in CQL3D and determining the fast electron diffusivities that are required to reproduce the experimentally observed profiles of hard x-ray emission. Finally, we have been performing full-wave LH field simulations using the massively parallel TORIC --LH solver [3] in order to assess spatial and spectral broadening of the incident wave front that can result from diffraction and wave focusing effects. [1] R. Parker, Bull. Am. Phys. Soc. 51, 20 (2006). [2] R.W. Harvey and M. McCoy, ``The CQL3D Fokker Planck Code,'' Proc. IAEA Tech. Comm. Meeting on Simulation and Modeling of Thermonuclear Plasmas, Montreal, Canada, 1992. [3] J. C. Wright et al., Nucl. Fusion 45

  3. 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.

  4. Experimental Measurements of the Lower Hybrid Electric Field on Alcator C-Mod by Stark Effect Spectroscopy

    NASA Astrophysics Data System (ADS)

    Hillis, D. L.; Mumgaard, R.; Lau, C.; Wallace, G.; Shiraiwa, S.

    2016-10-01

    A new diagnostic was installed on Alcator C-Mod capable of determining both the magnitude and direction of the lower hybrid wave electric field, ELH . The diagnostic, named SELHF (Stark Effect Lower Hybrid Field), simultaneously measures the two orthogonal polarization states of the Dβ spectra by passive optical emission spectroscopy. The ELH vector is then determined by systematically fitting the spectrum to the EZSSS (Explicit Zeeman-Stark Spectra Simulator) code which incorporates a fully quantum mechanical model comprising of the appropriate dynamic electric field and magnetic field operators. The SELHF diagnostic has 27 unique views of the LH launcher and surrounding space, each integrating over a 3 cm in diameter sightline, which is comparable to the waveguide dimension. Two sightlines are simultaneously viewed, yielding four spectra per discharge. In this presentation the diagnostic setup will be given. The methodology behind the spectral modeling and the results of the associated error analysis, yielding the accuracy of the ELH vector information, will be presented. The initial experimental results compared against a 2D cold-plasma model in COMSOL will be discussed. Work supported by DoE Contract No. DE-FC02-99ER54512 on Alcator C-Mod, a Department of Energy Office of Science user facility.

  5. Access to high-confinement regimes on Alcator C-Mod and the complex influence of divertor geometry

    NASA Astrophysics Data System (ADS)

    Hughes, J. W.; Labombard, B.; Brunner, D.; Hubbard, A.; Terry, J.; Rice, J.; Walk, J.; Cziegler, I.; Edlund, E.; Theiler, C.

    2015-11-01

    Placement of X-points and strike points in a diverted tokamak can have a remarkable impact on plasma properties, including thermal and particle confinement. The distinctive divertor of Alcator C-Mod allows substantial variation of divertor leg length, field line attack angle and divertor baffling, allowing us to induce changes in both L-mode confinement and access to both H-mode and I-mode. With the ion ∇B drift directed toward the divertor, scanning the strike point can induce ~ 2 × reductions in H-mode power threshold, and can produce a window for I-mode operation with H98 > 1 . Detailed high-resolution measurements, spanning the last closed flux surface, provide profiles of key quantities (n, T, ϕ) and their gradients, which are of likely importance in determining whether a discharge evolves an edge transport barrier, or remains in an L-mode state. Advances in Langmuir probes have enabled characterization of both radial profiles and fast (< 1 MHz) fluctuations in L-mode as the L-H threshold power is approached. These data allow new tests of models for H-mode access, especially those attempting to explain the non-monotonic density dependence of the H-mode power threshold through changes in transport and/or turbulence. Supported by U.S. Department of Energy award DE-FC02-99ER54512, using Alcator C-Mod, a DOE Office of Science User Facility.

  6. 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.

  7. 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)

  8. 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.

  9. 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.

  10. Profiles of Helium-like Argon Spectra from Alcator C-Mod with High Spatial and Spectral Resolution

    NASA Astrophysics Data System (ADS)

    Hill, K. W.; Bitter, M.; Stratton, B.; Roquemore, L.; Mastrovito, D.; Lee, S. G.; Bak, J. G.; Nam, U. W.; Rice, J. E.; Marmar, E.; Smith, G.

    2003-10-01

    Preliminary measurements of time resolved helium-like argon spectra have been made on Alcator C-Mod with a high resolution x-ray imaging crystal spectrometer, built for use on NSTX. A spherically bent crystal and a 10 cm x 30 cm imaging detector, developed for KSTAR, allow measurement of spectra across the entire plasma profile with a spatial resolution of 1 cm. The estimated resolving power, R = E/dE, for the spectrometer is about 4000. However, in these preliminary measurements R was significantly reduced, possibly due to high count rate effects. Addition of radiation shielding to reduce the background and use of a new detector with better position resolution will enable measurement of profiles of Ti, Te, charge-state equilibrium, and poloidal/toroidal rotation. The spectrometer and initial spectra will be presented, as well as observations regarding background and shielding and system resolution and throughput.

  11. 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.

  12. Correlation Electron Temperature Fluctuation Measurements on Alcator C-Mod and ASDEX Upgrade: Cross Machine Comparisons and Transport Model Validation

    NASA Astrophysics Data System (ADS)

    White, A. E.; Creely, A. J.; Freethy, S.; Cao, N.; Conway, G. D.; Goerler, T.; Happel, T.; Howard, N. T.; Inman, C.; Rice, J. E.; Rodriguez Fernandez, P.; Sung, C.; C-Mod, Alcator; Upgrade, Asdex

    2016-10-01

    Correlation Electron Cyclotron Emission diagnostics have been developed for Alcator C-Mod and ASDEX Upgrade. Measurements of long wavelength (ktheta rhos <0.5) electron temperature fluctuations have been measured in the core plasma (0.5 C-Mod and AUG are presented. This work is supported by the US DOE under Grants DE-SC0006419 and DEFC02-99ER54512-CMOD.

  13. Access conditions, energy and particle confinement of the I-mode regime on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Hubbard, Amanda

    2015-11-01

    Experiments on C-Mod have shown an extended operating range for I-mode at higher magnetic fields, offering options for high-performance, ELM-suppressed operation in future devices. Stationary regimes without significant ELMs are a requirement for ITER and other large burning devices. The I-mode regime offers one potential solution. It features a strong Te and Ti pedestal, up to 1 keV, without a density pedestal. I-mode has been demonstrated on the C-Mod, ASDEX Upgrade and DIII-D tokamaks, over increasingly wide parameter ranges. On C-Mod, global energy confinement is comparable to H-mode, with H98 between 0.7 and 1.2. Scaling of τE with Pheat-0 . 3 is more favorable than H-mode. This lack of saturation and the natural stability to ELMs can now be understood in terms of pedestal stability, with pressure and current gradients well away from stability limits. Impurity confinement τimp is similar in level and scaling to that in L-mode, 15-30 ms for both Ca and Mo, vs 0.1-1 s in H-mode. Key questions for extrapolation to other devices are the conditions for L-I transitions and for avoiding transitions to H-mode. An important new result is that the L-I threshold is independent of field, while the upper range of power for I-mode increases with BT leading to a wider operating space; at 5 T and above, many discharges remain in stationary I-mode with the full heating power of 5 MW. Scaling thresholds with size suggests that I-mode should be obtainable on ITER. Some I-modes have been observed up to 8 T. Another key question for any regime is compatibility with boundary solutions. In usual operation with Bxgrad drift away from the X-point, heat flux is predominantly to the inner divertor leg. Impurity seeding is used to reduce the flux, taking advantage of low τimp. I-modes have now been extended to near-balanced double null. Supported by DOE Award DEFC02- 99ER54512-CMOD.

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

    NASA Astrophysics Data System (ADS)

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

    2013-04-01

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

  15. Edge temperature gradient as intrinsic rotation drive in Alcator C-Mod tokamak plasmas.

    PubMed

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

    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 ∇T, but completely different edge ∇n, it may be concluded that the drive of the intrinsic rotation is the edge ∇T rather than ∇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.

  16. Characterization of the Pedestal in I-Mode Plasmas on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Walk, J. R.; Hughes, J. W.; Terry, J. L.; Hubbard, A. E.; Whyte, D. G.; White, A. E.; Marmar, E. S.; Rice, J. E.; Churchill, R. M.; Theiler, C.; Labombard, B.; Brunner, D.; Reinke, M. L.; Dominguez, A.; Sung, C.

    2012-10-01

    I-mode is an improved energy confinement regime on C-Mod, notably featuring an edge temperature pedestal without the accompanying density pedestal found in H-mode operation. I-mode has been shown to be capable of stationary operation, with the mode sustained for ˜10 energy confinement times, and to have a pedestal naturally lacking large ELMs. Combined with comparable performance to H-mode operation and weaker degradation with input power, this makes I-modes potentially favorable as a reactor regime. A full characterization using the suite of high-resolution edge diagnostics on C-Mod of the pedestal in reversed-field, lower-single-null I-modes (therefore with the ion ∇B drift away from the x-point) is presented here. The presented discharges primarily feature detailed scans of plasma current and ICRF heating power at several density values. The pedestal structure found in these I-modes is also contrasted with previous studies of the pedestal in EDA and ELMy H-modes, particularly potential scalings with βp,ped consistent with the KBM-limited pedestals found in these H-modes.

  17. Particle transport and density peaking at low collisionality on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Greenwald, M.; Hughes, J. W.; Mikkelsen, D.; Terry, J.

    2007-11-01

    While H-modes tend to have very flat density profiles, modest density peaking is advantageous for fusion performance. Thus robust pinch mechanisms that could allow operation with peaked profiles, in the absence of any internal particle source, are of considerable interest. Recent experiments on C-Mod^1, at low collisionality, show just such peaking and are quantitatively consistent with earlier results from ASDEX-U^2 and JET^3. By extending the range in machine size, these data break the covariance between collisionality and n/nG, the density normalized to the density limit and strongly support the primary role of collisionality in determining the profile. The implication is that ITER will have density profiles with ne( 0 )/˜1.4-1.6. The C-Mod data also show a small but statistically significant dependence of the peaking factor on the edge safety factor, q. The effect is to increase the peaking by no more than 10% when q is raised from 3.5 to 5.Initial studies of gyrokinetic simulations for these discharges will be shown. ^1M. Greenwald, submitted to Nucl. Fusion, 2007 ^2C. Angioni, et al., PRL 90, 205003, 2003 ^3H.Weisen, Nucl. Fusion 45, L1, 2005

  18. In-situ erosion and deposition measurements of plasma-facing surfaces in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Barnard, Harold S.

    2014-10-01

    The Accelerator Based In-situ Materials Surveillance (AIMS) diagnostic was recently developed to demonstrate the novel application of ion beam analysis (IBA) to in-vessel studies of plasma materials interactions in Alcator C-Mod. The AIMS diagnostic injects a 900 keV deuterium ion beam into the tokamak's vacuum vessel between plasma discharges while magnetic fields are used to steer the ion beam to plasma facing component (PFC) surfaces. Spectroscopic analysis of neutrons and gamma rays from the induced nuclear reactions provides a quantitative, spatially resolved map of the PFC surface composition that includes boron (B) and deuterium (D) content. Since AIMS is sensitive to low-Z elements and C-Mod regularly boronizes PFCs, the evolution of B and D on PFCs can be used to directly study erosion, deposition, and fuel retention in response to plasma operations and wall conditioning processes. AIMS analysis of 18 lower single null I-mode discharges show a net boron deposition rate of 6 +/- 2 nm/s on the inner wall while subsequent inner wall limited discharges and a disruption did not show significant changes in B. Measurements of D content showed relative changes of >2.5 following a similar trend. This suggests high D retention rates and net B deposition rates of ~18 cm/year of plasma exposure are possible and depend strongly on the plasma conditions. Ex-situ IBA was also performed on the same PFCs after removal from C-Mod, successfully validating the AIMS technique. These IBA measurements also show that the B content on the inner wall varied toroidally and poloidally from 0 to 3000 nm, demonstrating the importance of the spatial resolution provided by AIMS and the sensitivity of PFCs to B-field alignment. AIMS upgrades are underway for operation in 2014 and we anticipate new measurements correlating the evolution of PFC surfaces to plasma configuration, RF heating, and current drive scenarios. This work is supported by U.S. DOE Grant No. DE-FG02-94ER54235 and

  19. 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.

  20. 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.

  1. The influence of divertor geometry on access to high confinement regimes on the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Hughes, J. W.; Labombard, B.; Hubbard, A.; Marmar, E.; Terry, J.; Rice, J.; Walk, J.; Whyte, D.; Ma, Y.; Cziegler, I.; Edlund, E.; Theiler, C.

    2014-10-01

    The placement of X-point and strike points in a diverted tokamak can have a remarkable impact on properties of the discharge, including thermal and particle confinement. The distinctive divertor of Alcator C-Mod allows us to demonstrate these effects experimentally, as we vary equilibrium shaping to obtain substantial variation of divertor leg length, field line attack angle and divertor baffling. In response to these changes, we observe differences in both L-mode confinement and access to high-confinement regimes (i.e. ELMy H-mode and I-mode). With the ion grad-B drift directed toward the divertor, scanning the strike point can induce ~2× reductions in H-mode power threshold, and can produce a window for I-mode operation with H98 > 1. Recent experiments seek to explore these effects using improved diagnostics, and to extend them to the case with ion grad-B drift directed away from the divertor. Supported by USDoE award DE-FC02-99ER54512.

  2. Progress in Performance and Understanding of Steady ELM-free I-modes on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Hubbard, A. E.; Whyte, D. G.; Dominguez, A.; Greenwald, M. J.; Howard, N.; Hughes, J. W.; Labombard, B.; Lin, Y.; Lipschultz, B.; Marmar, E. S.; Reinke, M. L.; Rice, J. E.; Terry, J. L.; Walk, J.; White, A. E.; Wukitch, S. J.; Snyder, P.

    2012-10-01

    The I-mode regime of operation has been extended in recent Alcator C-Mod campaigns in duration and robustness, over a wide range of parameters. It features an edge thermal barrier, with L-mode like density profiles and particle transport [1]. I-modes are now routinely maintained in stationary conditions for over 10 τE. They are usually ELM free, a key advantage given the concern over divertor heat pulses on ITER. Instead, a continuous pedestal fluctuation, the ``weakly coherent mode,'' appears to enhance selectively particle over thermal transport. High performance I-modes are usually obtained with unfavorable ion drift direction, and extend to low q95=2.5and νped^*=1.3. τE is in the range of H-mode (H98,y2 up to 1.2), and exhibits less power degradation (W˜P^0.7). Power thresholds for I-mode are higher than typical L-H scalings, and increase with Ip as well as density. The widest power range for I-mode, nearly a factor of two above the L-I threshold, has been obtained in reversed field, lower null discharges at moderate ne. The focus of 2012 experiments is on assessing the density range and dependences, important for extrapolating to ITER, and measuring the pedestal profiles and fluctuations in greater detail.[4pt] [1] Whyte D G et al 2010 Nucl. Fusion 50 105005

  3. Magnetic field amplitude and pitch angle measurements using Spectral MSE on EAST and Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Liao, Ken; Rowan, William; Fu, Jia; Lyu, Bo; Li, Ying-Ying; Marchuk, Oleksandr; Ralchenko, Yuri

    2016-10-01

    Magnetic field amplitude and pitch angle measurements follow from the analysis of the Motional Stark Effect spectrum emitted by high energy neutral beam emission in tokamaks. Here we focus on deriving these quantities on EAST and Alcator C-Mod. These measurements provide a cross check for the polarimetry MSE diagnostics, and also act as a proof of principle for a spectral MSE diagnostic, which could potentially provide real-time measurements of the magnetic field and be used to increase the accuracy of equilibrium reconstruction. Measurement uncertainty is evaluated using the NBASS synthetic diagnostic. The same code allows design of measurements with improved accuracy such as spectral measurements techniques which take advantage of polarization. Accurate fitting of the MSE spectrum requires taking into account non-statistical beam excited state populations. The spectral MSE analysis techniques have applications to measurement of the beam density, which allows for improved analysis of the charge exchange recombination spectroscopy diagnostic. Resolution of beam components improves with increased beam energy and magnetic field, so these techniques have high applicability to future fusion devices. Supported by USDoE Award DE-FG03- 96ER-54373.

  4. 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.

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

    DOE PAGES

    Lore, Jeremy D.; Reinke, M. L.; LaBombard, Brian; ...

    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

  6. High-Performance Computational Modeling of ICRF Physics and Plasma-Surface Interactions in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Jenkins, Thomas; Smithe, David

    2016-10-01

    Inefficiencies and detrimental physical effects may arise in conjunction with ICRF heating of tokamak plasmas. Large wall potential drops, associated with sheath formation near plasma-facing antenna hardware, give rise to high-Z impurity sputtering from plasma-facing components and subsequent radiative cooling. Linear and nonlinear wave excitations in the plasma edge/SOL also dissipate injected RF power and reduce overall antenna efficiency. Recent advances in finite-difference time-domain (FDTD) modeling techniques allow the physics of localized sheath potentials, and associated sputtering events, to be modeled concurrently with the physics of antenna near- and far-field behavior and RF power flow. The new methods enable time-domain modeling of plasma-surface interactions and ICRF physics in realistic experimental configurations at unprecedented spatial resolution. We present results/animations from high-performance (10k-100k core) FDTD/PIC simulations spanning half of Alcator C-Mod at mm-scale resolution, exploring impurity production due to localized sputtering (in response to self-consistent sheath potentials at antenna surfaces) and the physics of parasitic slow wave excitation near the antenna hardware and SOL. Supported by US DoE (Award DE-SC0009501) and the ALCC program.

  7. 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.

  8. 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.

  9. Poloidal variation of high-Z impurity density in Alcator C-Mod ICRF-heated plasmas

    NASA Astrophysics Data System (ADS)

    Reinke, Matthew

    2012-10-01

    The poloidal variation of molybdenum density is measured in the core of ICRF-heated Alcator C-Mod plasmas and found to exhibit strong in/out asymmetries. Existing neoclassical parallel impurity transport theory is extended to include the effects of fast-ions and is shown to agree quantitatively with C-Mod measurements. The flux-surface variation of molybdenum is well described by nz(θ)/=1+nz,c cos(θ)+nz,ssin(θ), where -0.2 < nz,c/ < 0.3 and -0.1 < nz,s/ < 0.1 are observed over a wide range of Ohmic, L/I-mode and EDA H-mode plasmas for r/a < 0.9. The in/out asymmetry, nz,c/, is determined by a combination of centrifugal force due to toroidal rotation, leading to low-field side (LFS) accumulation, and poloidal electric fields sustained by magnetic trapping of cyclotron heated minority ions, leading to high field side (HFS) accumulation. While LFS accumulation due to centrifugal effects has been seen on other tokamaks, this represents the first observation of the effect driven entirely by intrinsic rotation. Scans of the D(H) resonance layer are shown to modify the in/out asymmetry by altering the fast-ion temperature anisotropy, T-/T||, and changing the ICRF power density, PRF/ne, either by ramping down the input power or increasing the density is found to reduce HFS accumulation. Observations of up/down asymmetries nz,s/, of molybdenum density are found to disagree with existing theories in the trace limit, nzZ^2/ni 1, in the collisionless main-ion regime. The link between nz(θ) and poloidal rotation, vθ, is emphasized, as both are assumed to be determined by neoclassical parallel impurity transport, and a more rigorous test of theory which includes matching asymmetries and vθ is discussed. The use of the poloidal variation in nz as a diagnostic for Eθ and T-/T|| as well as the impact of nz,c/ on radial transport are also discussed.

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

    SciTech Connect

    Jablonski, David 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.

  11. 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.

  12. 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.

  13. 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.

  14. Development of a reciprocating probe servomotor control system with real-time feedback on plasma position for the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Brunner, D.; Kuang, A. Q.; Labombard, B.; Burke, W.

    2015-11-01

    Reciprocating probe drives are one of the diagnostic workhorses in the boundary of magnetic confinement fusion experiments. The probe is scanned into an exponentially increasing heat flux, which demands a prompt and precise turn around to maintain probe integrity. A new linear servomotor controlled reciprocating drive utilizing a commercial linear servomotor and drive controller has been developed for the Alcator C-Mod tokamak. The quick response of the controller (able to apply an impulse of 50A in about 1ms) along with real-time plasma measurements from a Mirror Langmuir Probe (MLP) allows for real-time control of the probe trajectory based on plasma conditions at the probe tip. Since the primary concern for probe operation is overheating, an analog circuit has been created that computes the surface temperature of the probe from the MLP measurements. The probe can be programmed to scan into the plasma at various times and then turns around when the computed surface temperature reaches a set threshold, maximizing the scan depth into the plasma while avoiding excessive heating. Design, integration, and first measurements with this new system will be presented. This work was supported by U.S. Department of Energy award DE-FC02-99ER54512, using Alcator C-Mod, A DOE SC User Facility.

  15. 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.

  16. 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.

  17. Plasma heating and generation of energetic ions with novel three-ion ICRF scenarios on Alcator C-Mod and JET tokamak facilities

    NASA Astrophysics Data System (ADS)

    Kazakov, Yevgen

    2016-10-01

    This talk will report the first experimental results of novel three-ion ICRF scenarios (two or more majority ion species and one minority) for plasma heating and generating energetic ions in fusion facilities. The key feature of these scenarios is strong absorption of RF power possible at lower concentrations of minority ions than in two-ion plasmas. Effective plasma heating by injecting a small amount of 3He ions into H-D plasma mixtures with nH /ne 70 % has been successfully demonstrated in Alcator C-Mod and JET tokamaks. In C-Mod, efficient plasma heating was observed for 3He concentrations from 0.4-2%. During the discharges, a strong increase in Alfvén eigenmode activity was found to coincide with the addition of 3He to the H-D plasmas. Even lower 3He concentrations ( 0.2 %) were utilized in recent JET experiments. The potential of the D-(3He) -H scenario for plasma heating and generating MeV-range ions in JET plasmas was confirmed by a set of independent measurements, including stabilization of sawteeth, characteristic γ-ray emission, fast-ion loss detector. Furthermore, toroidal Alfvén eigenmodes with a range of toroidal mode numbers n were detected, which is another indication for the presence of significant population of high-energy 3He ions in a plasma. The discussed mechanism of resonant wave-particle interaction opens up various unexplored opportunities for ICRF system, including new scenarios for plasma heating. Three-ion ICRF scenarios are also relevant for the experimental programme of ITER. The possibility of using intrinsic 9Be impurities as the minority (instead of 3He) was suggested for heating bulk ions in D-T plasmas of JET and ITER, as well as heating trace amounts of 3He and 4He ions in H majority plasmas of ITER. The latest results and simulation comparisons will be presented. On behalf of Alcator C-Mod Team (MIT-PSFC, US) and JET Contributors (Culham, UK). Work supported by the US DOE (C-Mod DE-FC02-99ER54512 and SciDAC DE-FC02-01ER54648

  18. Identification of waves in the lower-hybrid frequency range in the scrape-off layer plasma of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Shinya, Takahiro; Gyou Baek, Seung; Wallace, Gregory M.; Shiraiwa, Syun'ichi; Takase, Yuichi; Parker, Ronald R.; Bonoli, Paul T.; Brunner, Dan; Faust, Ian; LaBombard, Brian L.; Wukitch, Steve

    2017-03-01

    Polarization resolved measurements of the parallel refractive index {{N}\\parallel}\\equiv c{{k}\\parallel}/ω of the driven RF waves in the lower hybrid (LH) range of frequencies are performed using arrays of RF magnetic probes in the scrape-off layer plasma of Alcator C-Mod. The measured {{N}\\parallel} of the RF magnetic field component parallel to the background magnetic field is about  -1.6, which corresponds to the peak of the launched LH {{N}\\parallel} spectrum. Based on the wave dispersion relationship, this wave is identified as the LH slow wave. On the other hand, the RF magnetic field component perpendicular to the magnetic field is found to have a lower {{N}\\parallel} of  -1.2, and is detected only near the last closed flux surface. This wave is identified as the LH fast wave generated by slow-fast wave mode conversion.

  19. 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.

  20. Search for Correlation Between Plasma Rotation and Electron Temperature Gradient Scale Length in LOC/SOC Transition at Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Houshmandyar, Saeid; Rowan, William L.; Phillips, Perry E.; Walk, John R.; Rice, John E.

    2015-11-01

    Understanding the mechanism governing the linear ohmic confinement (LOC) and the transition to saturated ohmic confinement (SOC) has long been a focus of tokamak research. It is commonly accepted that at low density, the confinement is dominated by electron-scale turbulence while at high density, the turbulence is dominated by ion temperature gradient. At Alcator C-Mod, the core rotation reversal was shown to be consistent with this ansatz. However a recent study at AUG suggests that the intrinsic rotation behavior is rather determined by local plasma parameters regardless of the heating method or the confinement regime. Here, we follow this idea and search for dependence of intrinsic rotation on electron temperature gradient scale length, a quantity with a pivotal role in plasma transport. The high-resolution (1 μs, 7mm) electron cyclotron emission diagnostic at C-Mod (FRCECE) coupled with the BT jog technique allows direct LTe measurements. In the BT jog technique, a 1.5% change in the toroidal magnetic field shifts the viewing volume of the ECE by ~ 1 cm, and the ratio of the average of the signal to the change in the signal during its ramp-up yields LTe. Supported by USDoE awards DE-FG03-96ER-54373 and DE-FC02-99ER54512.

  1. Multichannel Transport in L-mode and I-mode Plasmas at Alcator C-Mod and Comparison with Gyrokinetic Simulations

    NASA Astrophysics Data System (ADS)

    White, A. E.

    2012-10-01

    New experiments at Alcator C-Mod are challenging and expanding our understanding of electron, impurity particle, and momentum transport. Understanding these coupled transport channels is important, since alpha particles will slow down mostly on electrons in ITER intrinsically rotating, alpha-heated plasmas. Core density fluctuations are reduced in high performance (H98˜1) I-mode plasmas by up to 30% compared to L-mode. At the L-I transition, the core turbulence changes lead the edge turbulence changes. This is in contrast to observations at L-H transitions, where the edge turbulence changes lead the core turbulence changes. It is also found that L-mode and I-mode plasmas are both ITG dominant, with lower linear ITG growth rates in I-mode. TRANSP analysis indicates that core electron heat transport is reduced in I-mode compared to L-mode, with little change in ion heat transport. This indicates that significant electron heat transport is driven by ITG turbulence in L-mode. However, impurity particle transport is similar in L-mode and I-mode, suggesting that tracking changes in ITG drive alone does not describe as well the transport in this channel. Additionally, in L-mode plasmas, small < 20% changes in line averaged density lead to dramatic differences in the on-axis rotation, due to a hollowing of the radial profile at marginally lower density. This change in the shape of the rotation profile in L-mode plasmas does not appear to occur at the ITG/TEM boundary, contrary to similar phenomena seen in Ohmic plasmas. For the first time, local, long-wavelength density and electron temperature fluctuations in the core plasma (0.4 < r/a < 0.9) at C-Mod have been measured with reflectometry and correlation ECE. These new fluctuation measurements at C-Mod will allow for better testing and validation of transport models. Gyrokinetic simulations are in progress to interpret the new multichannel and multifield turbulence/transport results from C-Mod.

  2. Pedestal and Transport Properties of Steady-state I-mode Plasmas over Expanded Operational Space in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Marmar, Earl

    2011-10-01

    I-mode operation on Alcator C-Mod combines a strong edge thermal transport barrier with L-mode levels of particle and impurity transport, allowing access to very high performance discharges with low pedestal collisionality and central temperatures up to 8 keV, and without large ELMs or other intermittent edge instabilities. In recent campaigns, C-Mod I-modes have been extended to quasi-steady-state, with access in both favorable and unfavorable ion drift directions and typical normalized energy confinement quality factor H98 ~ 1.0 to 1.2. Adding ICRF mode-conversion flow-drive enhances toroidal flow shear near the plasma edge and confinement is further enhanced. I-mode has been maintained with input power up to nearly 2x the I-mode threshold power, with the largest accessible range in closed divertor geometry at modest triangularity. Simple extrapolations at fixed field imply that ITER in unfavorable drift could access I-mode with available power, and stay in I-mode with alpha-dominant heating. Detailed pedestal fluctuation measurements reveal changes in the turbulence, with decreases in the power at some frequencies and size scales, and growth of a weakly coherent mode (WCM) (kθ ~ 1.5 cm-1, δf/f ~.3) which propagates in the electron diamagnetic direction in the plasma frame. The WCM, which has density, temperature and magnetic signatures, appears to play a key role in pedestal density and impurity regulation, and detailed experimental results and associated modeling are presented. The distribution of divertor exhaust power depends on ion drift direction; new measurements of I-mode heat flux footprints on the outer divertor are compared with those in H-mode. Pedestal stability analyses will be shown for I-modes, including some which exhibited small ELMs. Supported by USDOE Award DE-FC02-99-ER54512.

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

    DOE PAGES

    Lore, Jeremy D.; Reinke, M. L.; Brunner, D.; ...

    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

  4. Inboard and outboard radial electric field wells in the H- and I-mode pedestal of Alcator C-Mod and poloidal variations of impurity temperature

    NASA Astrophysics Data System (ADS)

    Theiler, C.; Churchill, R. M.; Lipschultz, B.; Landreman, M.; Ernst, D. R.; Hughes, J. W.; Catto, P. J.; Parra, F. I.; Hutchinson, I. H.; Reinke, M. L.; Hubbard, A. E.; Marmar, E. S.; Terry, J. T.; Walk, J. R.; the Alcator C-Mod Team

    2014-08-01

    We present inboard (HFS) and outboard (LFS) radial electric field (Er) and impurity temperature (Tz) measurements in the I-mode and H-mode pedestal of Alcator C-Mod. These measurements reveal strong Er wells at the HFS and the LFS midplane in both regimes and clear pedestals in Tz, which are of similar shape and height for the HFS and LFS. While the H-mode Er well has a radially symmetric structure, the Er well in I-mode is asymmetric, with a stronger ExB shear layer at the outer edge of the Er well, near the separatrix. Comparison of HFS and LFS profiles indicates that impurity temperature and plasma potential are not simultaneously flux functions. Uncertainties in radial alignment after mapping HFS measurements along flux surfaces to the LFS do not, however, allow direct determination as to which quantity varies poloidally and to what extent. Radially aligning HFS and LFS measurements based on the Tz profiles would result in substantial inboard-outboard variations of plasma potential and electron density. Aligning HFS and LFS Er wells instead also approximately aligns the impurity poloidal flow profiles, while resulting in a LFS impurity temperature exceeding the HFS values in the region of steepest gradients by up to 70%. Considerations based on a simplified form of total parallel momentum balance and estimates of parallel and perpendicular heat transport time scales seem to favor an approximate alignment of the Er wells and a substantial poloidal asymmetry in impurity temperature.

  5. 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.

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

    DOE PAGES

    Russell, D. A.; Myra, J. R.; D'Ippolito, D. A.; ...

    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

  7. 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.

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

    SciTech Connect

    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 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 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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)].

  13. 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.

  14. 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.

  15. Study of Fluctuations with the PCI Diagnostic During the Search for an I-Mode on DIII-D

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

    Recent experiments on DIII-D explored the operational space required to achieve the I-mode regime observed on Alcator C-Mod. This regime features an edge energy transport barrier without an accompanying particle barrier, and broadband instabilities known as WCM. Although none of the DIII-D plasmas exhibited all those characteristics, preceding the H-mode transition, the edge electron temperature slowly and intermittently evolved towards a pedestal-like profile 0.4 keV high while maintaining a typical L-mode edge density profile. During this prelude phase, edge density fluctuations measured with the recently upgraded Phase Contrast Interferometer (PCI) are observed to be qualitatively similar to those seen in H-mode: the intensity of fluctuations is reduced, and the frequency spectrum is broadened in the range of 300 kHz and is non-monotonic. The 2D k-f spectra also approach those typical of an H-mode, showing broad fluctuations in k-space and group velocities increasing to 10 km/s. Several branches are observed in 2D spectra, possibly indicating multiple instabilities at play. Work supported by the US DOE under DE-FG02-04ER54235 and DE-FC02-04ER54698.

  16. 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.

  17. 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.

  18. External Excitation of a Drift-Alfvén Wave Response in the Alcator C-Mod Edge Plasma and its Relationship to the Quasi-Coherent Mode

    NASA Astrophysics Data System (ADS)

    Golfinopoulos, Theodore

    2013-10-01

    Experiments indicate that short-wavelength, k⊥ρs ~ 0 . 1 , drift-Alfvénic turbulence plays an important role in C-Mod edge plasma transport. A Quasi-Coherent Mode (QCM, 50 < f < 150 kHz, k⊥ ~ 1 . 5 cm-1) regulates particle and impurity transport in C-Mod's EDA H-modes. A Weakly Coherent Mode (WCM, 150 < f < 500 kHz, k⊥ ~ 1 . 5 cm-1) plays a similar role in I-mode discharges, suppressing the formation of a density pedestal while maintaining a temperature pedestal. ELMs are not present in either confinement regime. With the idea of exciting, probing, and perhaps exploiting this transport behavior, we have developed a novel antenna system to excite drift-Alfvén-like modes at the outer midplane. A winding with a ``shoelace'' geometry is placed ~ 3 - 5 mm from the LCFS. The principal design parameters, k⊥ = 1 . 5 +/- 0 . 1 cm-1 and 45 < f < 300 kHz, match the QCM and WCM properties, so that the antenna induces parallel currents in the boundary plasma that mimic those observed for the intrinsic modes. Phase-locking to intrinsic modes is also accomplished via a custom circuit. The antenna produces perturbations in density and field comparable to amplitudes of the intrinsic QCM. The plasma response exhibits a resonance near the natural QCM frequency, which generally satisfies the drift wave dispersion relation. While a driven B~θ fluctuation is visible throughout the discharge, the driven ñe is only observed during H-mode, though it precedes the onset of the intrinsic QCM. Like the QCM, the driven mode propagates in the electron diamagnetic drift direction and is approximately field-aligned. Recent mirror probe measurements show the intrinsic QCM structure is predominantly drift-Alfvénic, and we might expect the same of the driven mode. However, the induced perturbation is not global, but is localized to field lines which map to the antenna, suggesting a damped response, and direct measurements of the damping rate indicate γ /ω0 ~ 5 %. If the antenna

  19. 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.

  20. 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.

  1. C-Mod Collaboration Informal Technical Progress Report

    SciTech Connect

    Kenneth W. Gentle

    2007-12-31

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

  2. Progress on the C-Mod FIR Polarimeter System

    NASA Astrophysics Data System (ADS)

    Xu, P.; Irby, J. H.; Bergerson, W. F.; Brower, D. L.; Ding, W. X.; Shiraiwa, S.; Wolfe, S.

    2010-11-01

    A poloidally viewing FIR polarimetry diagnostic is being developed for the Alcator C-Mod Tokamak. The primary diagnostic components are a two-wave FIR laser at 117.73 microns and newly developed detectors whose performance characteristics will be described. Faraday rotation will be used both to refine the q-profile measurement by adding constraints to EFIT , and to study density and magnetic field fluctuations. A three-chord system has been installed, one chord of which is being tested during the FY10 C-Mod campaign. The FIR laser source is affected by both stray magnetic fields and mechanical vibrations present in the experimental cell thereby impacting the measurement. Methods developed to mitigate and correct for these effects will be discussed. Initial Faraday data will be compared with expectations from numerical simulation.

  3. C-Mod MHD stability analysis with LHCD

    NASA Astrophysics Data System (ADS)

    Ebrahimi, Fatima; Bhattacharjee, A.; Delgado, L.; Scott, S.; Wilson, J. R.; Wallace, G. M.; Shiraiwa, S.; Mumgaard, R. T.

    2016-10-01

    In lower hybrid current drive (LHCD) experiments on the Alcator C-Mod, sawtooth activity could be suppressed as the safety factor q on axis is raised above unity. However, in some of these experiments, after applying LHCD, the onset of MHD mode activity caused the current drive efficiency to significantly drop. Here, we study the stability of these experiments by performing MHD simulations using the NIMROD code starting with experimental EFIT equilibria. First, consistent with the LHCD experiment with no signature of MHD activity, MHD mode activity was also absent in the simulations. Second, for experiments with MHD mode activity, we find that a core n=1 reconnecting mode with dominate poloidal modes of m=2,3 is unstable. This mode is a resistive current-driven mode as its growth rate scales with a negative power of the Lundquist number in the simulations. In addition, with further enhanced reversed-shear q profile in the simulations, a core double tearing mode is found to be unstable. This work is supported by U.S. DOE cooperative agreement DE-FC02-99ER54512 using the Alcator C-Mod tokamak, a DOE Office of Science user facility.

  4. EHT Integrator Demonstration at DIII-D

    NASA Astrophysics Data System (ADS)

    Slobodov, Ilia; Miller, Ken; Ziemba, Timothy; Prager, James

    2015-11-01

    Eagle Harbor Technologies, Inc. (EHT) has developed a series of analog integrators for magnetic diagnostics for the fusion science and plasma physics communities. Three varieties of EHT integrators: short pulse, high dynamic range, and long pulse, have all been tested at DIII-D. The EHT short pulse integrators were used to measure the poloidal magnetic field coil, saddle coil, and Rogowski coil. The results were compared to existing diagnostics at DIII-D. The EHT high dynamic range integrator was used to measure the toroidal magnetic field. Due to the high dynamic range, this integrator resolved features in the signal that could not otherwise be observed. Additionally, preliminary testing of the EHT long pulse integrator was conducted. The long pulse integrator meets all the specifications for ITER.

  5. Disruption studies in DIII-D

    SciTech Connect

    Kellman, A.G.; Evans, T.E.; Cuthbertson, J.W.

    1996-09-01

    Characteristics of disruptions in the DIII-D tokamak including the current decay rate, halo current magnitude and toroidal asymmetry, and heat pulse to the divertor are described. Neon and argon pellet injection is shown to be an effective method for mitigating the halo currents and the heat pulse with a 50% reduction in both quantities achieved. The injection of these impurity pellets frequently gives rise to runaway electrons.

  6. Disruption Studies in DIII-D

    SciTech Connect

    Kellman, A.G.; Cuthbertson, J.W.; Luckhardt, S.; Zhang, J.; Lee, R.L.; Leuer, J.A.; Humphreys, D.A.; Hyatt, A.W.; Jahns, G.L.; Wroblewski, D.; Jernigan, T.; Lasnier, C.J.; Scchaffer, M.J.; Taylor, P.L.; Whyte, D.G.

    1997-12-31

    Characteristics of disruptions in the DIII-D tokamak including the current decay rate, halo current magnitude and toroidal asymmetry. and heat pulse to the divertor are described. Neon and argon pellet injection is shown to be an effective method of mitigating the halo currents and the heat pulse with a 50% reduction in both quantities achieved. The injection of these impurity pellets frequently gives rise to runaway electrons.

  7. Overview of Recent DIII-D Experiments

    NASA Astrophysics Data System (ADS)

    Baker, D. R.

    2004-11-01

    The focus of the DIII-D program is to develop the scientific basis for an optimization of fusion energy production, the Advanced Tokamak (AT). Progress was made on several fronts in this effort, and a significant number of experiments are coordinated with the International Tokamak Physics Activity (ITPA), involving both U.S. and international collaborators. As part of the broad DIII-D program, MHD stability of the edge pedestal during ELM-free quiescent H-mode operation on DIII-D has been investigated by detailed edge profile measurements and inductive plasma current ramps. Real time EFITS using MSE predicts the precise location of ECCD required to prevent the initiation of 3/2 NTMs thus allowing beta to be raised above the non-stabilized value. Simultaneous turbulence measurements from a large wavenumber range covering the theoretically predicted wavenumbers of ITG modes, TEM modes, and ETG modes were acquired and are being compared to predictions of turbulence simulation codes. Detailed velocity profile measurements in ECH only H-modes show toroidal rotation well above the diamagnetic drift velocity. Stabilization of RWMs has been studied. ELMs have been stabilized with stochastic edge layers. ^13C experiments showed that nearly all of the carbon deposition is at near the inner strike point in L-mode plasmas.

  8. 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.

  9. Divertor conditions near double null in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Brunner, Dan; Labombard, Brian; Kuang, Adam; Terry, Jim; Mumgaard, Bob; Wolfe, Steve

    2016-10-01

    Many tokamak reactor designs utilize a double-null equilibrium for the boundary plasma because of the expected benefits of heat flux sharing between the two outer divertor leg as well as the attractiveness of the high-field side scrape-off layer plasma in double-null for RF actuators. However, there has been very little reported on boundary plasma conditions near double null, especially at the divertor plate. And, due to the narrow boundary plasma width, there is concern of the precision to which a double-null equilibrium must be controlled to maintain divertor heat flux sharing. To this end, a series of experiments were performed varying the magnetic balance around double null. The magnetic balance between the two nulls was scanned shot-to-shot in L-, I-, and H-mode plasmas. In addition, current and density scans were performed in L-mode plasmas. Results will be presented for relative balances of divertor particle and energy fluxes to the four divertors (inboard/outboard, upper/lower) as well as the sensitivity of changes in divertor conditions to the magnetic balance. Supported by USDoE Award DE-FC02-99ER54512.

  10. Electron Cyclotron Heating on DIII-D

    SciTech Connect

    Prater, R.; Petty, C.C.

    2005-10-15

    Electron cyclotron heating (ECH) has proved to be a very flexible system for heating applications in DIII-D. The outstanding characteristics of ECH - controllable heating location, a high degree of localization of the power, ability to heat without introducing particles, and ability to heat only the electron fluid - have been used in a wide variety of experiments to study wave physics and transport, to control magnetohydrodynamic activity, and to improve discharges. These characteristics along with relatively easy coupling to the plasma make ECH a valuable resource for both heating and instability control in burning plasmas.

  11. ELM Studies on DIII--D

    NASA Astrophysics Data System (ADS)

    Osborne, T. H.; Leonard, A. W.; Porter, G. D.

    1996-11-01

    Three classes of Edge Localized Modes, or ELMs, simply labeled Types I, II, and III, in the DIII--D work, are widely observed. Type I ELMs are distinguished by the fact that their frequency increases with increasing input power. A scaling for the Type I ELM energy loss predicts 3% for ITER. The frequency of Type III ELMs decreases with increasing input power. The energy loss per Type III ELM is a factor of 2 to 4 below that for Type I at the same input power. Experiments on DIII--D suggest that proximity to the H--mode threshold power is the critical parameter for Type III ELMs. In contrast to Type I ELMs, the pressure gradient near the separatrix at a Type III ELM is often well below the ideal ballooning mode limit. Medium n magnetic precursor oscillations are observed with Type III ELMs, while no magnetic precursors are observe with Type I ELMs. Type II ELMs have very high frequency and low energy loss compared to Type I ELMs. Type II ELMs do not require low input power in contrast to Type III ELMs. Type II ELMs are associated with edge second stability in combination with either high q or high β_P.

  12. 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.

  13. 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.

  14. 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.

  15. Advanced Divertor Developments at DIII-D

    NASA Astrophysics Data System (ADS)

    Kolemen, E.; Allen, S. L.; Makowski, M. A.; Soukhanovskii, V. A.; Bray, B. D.; Eldon, D.; Humphreys, D. A.; Johnson, R.; Leonard, A. W.; Liu, C.; Penaflor, B. G.; Petrie, T. W.; McLean, A. G.; Unterberg, E. A.

    2013-10-01

    Novel divertor configurations and control schemes have been implemented at DIII-D to test and optimize heat and particle handling capabilities for advanced tokamaks. The snowflake configuration is stabilized by first calculating the position of the two null-points using real-time equilibrium reconstruction and then regulating the shaping coil currents. Experiments in which the snowflake divertor is stabilized for many confinement times show that it is compatible with high-performance operation and results in greatly reduced divertor heat flux. An advanced divertor control system regulates the gas injection to achieve partial or full detachment by using the divertor temperature measurements from real-time Thomson diagnostics and a line ratio measurement, and adjusts the core and divertor radiation via measurement of the real-time bolometer diagnostics. Prospects of achieving acceptable divertor target heat fluxes for future fusion reactors are analyzed and challenges are presented. Work supported by the US DOE under DE-AC02-09CH11466, DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-AC05-00OR22725.

  16. Current Profile Control in DIII-D

    NASA Astrophysics Data System (ADS)

    Schuster, E. M.; Barton, J. E.; Boyer, M. D.; Wehner, W. P.; Ferron, J. R.; Humphreys, D. A.; Hyatt, A. W.; Jackson, G. L.; Luce, T. C.; Walker, M. L.

    2014-10-01

    Experimental results successfully demonstrate the potential of physics-model-based control for systematic attainment of desired q profiles, with the subsequent benefit of enabling exploration and reproducibility. The control scheme is constructed by embedding a nonlinear, control-oriented, physics-based model of the plasma dynamics into the control design process. This modeling approach combines first-principles laws with empirical correlations obtained from physical observations, which leads to PDE models capturing the high-dimensionality and nonlinearity of the plasma response. Model-based control design includes not only the synthesis of feedback controllers for robust regulation or tracking, but also the determination of optimal feedforward actuator trajectories for a systematic approach to scenario planning. Feedforward+feedback (closed loop) control experiments in DIII-D consistently demonstrate improved current-profile control performance relative to feedforward (open loop) control alone. Supported by the US Department of Energy under DE-SC0001334, DE-SC0010661 and DE-FC02-04ER54698.

  17. DIII-D divertor reflectometer system

    SciTech Connect

    Rhodes, T.L.; Doyle, E.J.; Nguyen, X.V.; Kim, K.W.; Peebles, W.A.; Doane, J.L.

    1997-01-01

    Divertor density profiles, asymmetries, turbulence, and MARFE diagnosis are extremely important and affect the divertor design process for ITER and other future devices. In addition, a functioning divertor density profile system will be essential for the operation of these machines. It is thus critical to prototype and demonstrate diagnostics capable of operating in a divertor environment. To meet these needs a divertor reflectometer system has been designed and installed on DIII-D. The design stresses flexibility, modularity, and simplicity. It consists of a circular, smoothwall, overmoded waveguide followed by a TE{sub 11}{R_arrow}HE{sub 11} mode converter (the HE{sub 11} mode is a low loss Gaussian mode with a very symmetric radiation pattern, optimal for this use) thus allowing use of an arbitrary polarization (f{sub pe},f{sub LH},f{sub RH}). The design provides for testing of a variety of antennas/probing directions including: upward to probe the X-point region, including MARFEs, sideways to probe outboard/inboard divertor legs, and oppositely directed to probe both divertor legs simultaneously. System design, operational considerations, and experimental data are presented. {copyright} {ital 1997 American Institute of Physics.}

  18. Status of DIII-D plasma control

    SciTech Connect

    Walker, M.L.; Ferron, J.R.; Penaflor, B.

    1995-10-01

    A key component of the DIII-D Advanced Tokamak and Radiative Divertor Programs is the development and implementation of methods to actively control a large number of plasma parameters. These parameters include plasma shape and position, total stored energy, density, rf loading resistance, radiated power and more detailed control of the current profile. To support this research goal, a flexible and easily expanded digital control system has been developed and implemented. We have made parallel progress in modeling of the plasma, poloidal coils, vacuum vessel, and power system dynamics and in ensuring the integrity of diagnostic and command circuits used in control. Recent activity has focused on exploiting the mature digital control platform through the implementation of simple feedback controls of more exotic plasma parameters such as enhanced divertor radiation, neutral pressure and Marfe creation and more sophisticated identification and digital feedback control algorithms for plasma shape, vertical position, and safety factor on axis (q{sub 0}). A summary of recent progress in each of these areas will be presented.

  19. VUV Spectroscopy in DIII-D Divertor

    SciTech Connect

    Alkesh Punjabi; Nelson Jalufka

    2004-11-04

    The research carried out on this grant was motivated by the high power emission from the CIV doublet at 155 nm in the DIII-D divertor and to study the characteristics of the radiative divertor. The radiative divertor is designed to reduce the heat load to the target plates of the divertor by reducing the energy in the divertor plasma using upstream scrape-off-layer (SOL) radiation. In some cases, particularly in Partially Detached Divertor (PDD) operations, this emission accounts for more than 50% of the total radiation from the divertor. In PDD operation, produced by neutral gas injection, the particle flow to the target plate and the divertor temperature are significantly reduced. A father motivation was to study the CIV emission distribution in the lower, open divertor and the upper baffled divertor. Two Vacuum Ultra Violet Tangential viewing Television cameras (VUV TTV) were constructed and installed in the upper, baffled and the lower, open divertor. The images recorded by these cameras were then inverted to produce two-dimensional distributions of CIV in the poloidal plane. Results obtained in the project are summarized in this report.

  20. 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.}

  1. New DIII-D tokamak plasma control system

    SciTech Connect

    Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T; Greenfield, C.M.; Pinsker, R.I. ); Lazarus, E.A. JET Joint Undertaking, Abingdon, Oxon )

    1992-09-01

    A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter.

  2. FEEDBACK CONTROL OF THE DIII-D ECH SYSTEM

    SciTech Connect

    J. LOHR; J.R. FERRON; Y.A. GORELOV; K. KAJIWARA; D. PONCE; M.R. WADE

    2002-08-01

    The output power of the DIII-D gyrotron complex has been modulated by the plasma control system using feedback on the difference between a desired electron temperature and the ECE measurement. Operation was stable and permitted control of the flux penetration during initiation of the discharge.

  3. Wall conditioning and plasma surface interactions in DIII-D

    SciTech Connect

    Jackson, G.L.; Petersen, P.I.; Schaffer, M.S.; Taylor, P.L.; Taylor, T.S. ); Doyle, B.L.; Walsh, D.S. ); Hill, D.N. ); Hsu, W.L. ); Winter, J. . Inst. fuer Plasmaphysik)

    1990-09-01

    Wall conditioning is used in DIII-D for both reduction of impurity influxes and particle control. The methods used include: baking, pulsed discharge cleaning, hydrogen glow cleaning, helium and neon glow conditioning, and carbonization. Helium glow wall conditioning applied before every tokamak discharge has been effective in impurity removal and particle control and has significantly expanded the parameter space in which DIII-D operates to include limiter and ohmic H-mode discharges and higher {beta}{sub T} at low q. The highest values of divertor plasma current (3.0 MA) and stored energy (3.6 MJ) and peaked density profiles in H-mode discharges have been observed after carbonization. Divertor physics studies in DIII-D include sweeping the X-point to reduce peak heat loads, measurement of particle and heat fluxes in the divertor region, and erosion studies. The DIII-D Advanced Divertor has been installed and bias and baffle experiments will begin in the fall of 1991. 15 refs., 4 figs.

  4. FIR polarimetry diagnostic for the C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Irby, J. H.; Bergerson, W. F.; Brower, D. L.; Ding, W. X.; Marmar, E. S.; Xu, P.

    2012-02-01

    A three-chord polarimeter on Alcator C-Mod will make measurements of the poloidal magnetic field and plasma fluctuations. The beams from two frequency-offset, 200 mW, FIR lasers operating at 117.73 μm are combined to produce collinear, counter-rotating, circularly polarized beams. The beams are divided into three chords which are directed into the plasma at one toroidal location. Corner cube retro-reflectors mounted on the inside wall return the beam for a double pass. The mixing product of the two beams is detected both before (reference) and after (signal) the plasma using polarization sensitive detectors that produce a beat signal at ~ 4 MHz. During the plasma discharge, the phase delay of the signal mixer, which depends on the Faraday effect, is evaluated with respect to the reference and produces line-integrated information on the poloidal magnetic field. Measurements on C-Mod require the phase error to be at the 0.1 degree level, and great care in the design of optical mounts, polarizers, beam-splitters, focusing optics, and acoustic and magnetic shielding was required. Development of new planar diode Schottky detectors was necessary to provide high sensitivity for a diagnostic that will eventually have at least six chords. Absorption of the FIR laser light by water vapor requires that the entire beam path be purged with dry air. Six retro-reflectors on the inner wall arranged in an ITER-like configuration provide poloidally viewing chords from near the mid-plane to well into the plasma scrape off layer. A pneumatically controlled shutter protects the in-vessel optics during boronizations and during limited discharges that might accelerate damage to the retro-reflector surfaces. Tests indicate there is no measurable signal contamination from the toroidal magnetic field due to the Cotton-Mouton effect. Polarization sensitivity of the wire mesh beamsplitters necessitated system calibration. Good agreement to EFIT reconstructions has been observed along with

  5. The DIII-D Map -- An Area-Preserving Map for Trajectories of Magnetic Field Lines in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Punjabi, Alkesh; Ali, Halima; Boozer, Allen; Evans, Todd

    2007-11-01

    The EFIT data for the DIII-D shot 115467 3000 ms is used to calculate the generating function for an area-preserving map for trajectories of magnetic field lines in the DIII-D. We call this map the DIII-D map. The generating function is a bivariate polynomial in base vectors &1/2circ;cos(θ) and &1/2circ;sin(θ). ψ is toroidal flux and θ is poloidal angle. The generating function is calculated using a canonical transformation from (ψ,θ) to physical coordinates (R,Z) in the DIII-D [1] and nonlinear regression. The equilibrium generating function gives an excellent representation of the equilibrium flux surfaces in the DIII-D. The DIII-D map is then used to calculate effects of the magnetic perturbations in the DIII-D. Preliminary results of the DIII-D map will be presented. This work is supported by US DOE OFES DE-FG02-01ER54624 and DE-FG02-04ER54793. [1] A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys Lett A 364 140--145 (2007).

  6. Fast Wave Current Drive Antenna Performance on DIII-D

    NASA Astrophysics Data System (ADS)

    Mayberry, M. J.; Pinsker, R. I.; Petty, C. C.; Chiu, S. C.; Jackson, G. L.; Lippmann, S. I.; Porkolab, M.; Prater, R.; Baity, F. W.; Goulding, R. H.; Hoffman, D. J.

    1992-01-01

    Fast wave current drive (FWCD) experiments at 60 MHz are being performed on the DIII-D tokamak for the first time in high electron temperature, high β target plasmas. A four-element phased-array antenna is used to launch a directional wave spectrum with the peak n∥ value (≂7) optimized for strong single-pass electron absorption due to electron Landau damping. For this experiment, high power FW injection (2 MW) must be accomplished without voltage breakdown in the transmission lines or antenna, and without significant impurity influx. In addition, there is the technological challenge of impedance matching a four-element antenna while maintaining equal currents and the correct phasing (90°) in each of the straps for a directional spectrum. In this paper we describe the performance of the DIII-D FWCD antenna during initial FW electron heating and current drive experiments in terms of these requirements.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. The DIII-D lithium beam edge diagnostic system (abstract)

    NASA Astrophysics Data System (ADS)

    Thomas, D. M.; Lee, R. L.; Patterson, R. M.; Brooks, N. H.; Robinson, J.; McChesney, J. M.

    1992-10-01

    We are installing a diagnostic system based on a neutral lithium beam to investigate plasma behavior in the edge region of DIII-D discharges [D. M. Thomas et al., Rev. Sci. Instrum. 61, 3040 (1990]. The system will provide neutral equivalent current densities of several mA/cm2 at beam energies from 5 to 30 keV, sufficient to penetrate several centimeters past the last closed flux surface in most of the DIII-D operating regime. Fluorescence of the beam atoms is induced by collisions with plasma particles and is a sensitive measure of the edge density behavior. The emitted 670.8-nm fluorescence is collected and coupled via fiber optics to a multichannel high-speed data acquisition system based on silicon diode detectors. Because of the favorable atomic properties of lithium (i.e., high electron impact excitation cross section, resonance wavelength well separated from Hα) we should be able to study density fluctuations in this region from an analysis of the associated fluctuations in the beam fluorescence. A description of the installed diagnostic, test stand measurements of intrinsic beam fluctuations and any initial operating experience on DIII-D will be presented. This work supported by U. S. Department of Energy Grant No. DE-FG03-90ER5408 and Contract DE-AC03-89ER51114, which support does not constitute an endorsement by DOE of views expressed in this publication.

  15. Upgrade of the DIII-D RF systems

    SciTech Connect

    Callis, R.W.; Cary, W.P.; O`Neill, R.C.

    1995-10-01

    The DIII-D Advanced Tokamak Program requires the ability to modify the current density profile for extended time periods in order to achieve the improved plasma conditions now achieved with transient means. To support this requirement DIII-D has just completed a major addition to its ion cyclotron range of frequency (ICRF) systems. This upgrade project added two new fast wave current drive (FWCD) systems, with each system consisting of a 2 MW, 30 to 120 MHz transmitter, an all ceramic insulated transmission line, and water-cooled four-strap antenna. With this addition of 4 MW of FWCD power to the original 2 MW, 30 to 60 MHz capability, experiments can be performed with centrally localized current drive enhancement. For off-axis current modification, plans are in place to add 110 GHz electron cyclotron heating (ECH) power to DIII-D. Initially, 3 MW of power will be available with plans to increase the power to 6 MW and to 10 MW.

  16. 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

  17. Operation of DIII-D with all-graphite walls

    SciTech Connect

    Holtrop, K.L.; Jackson, G.L.; Kellman, A.G.; Lee, R.L.; Hollerbach, M.A

    1993-10-01

    Recently, additional graphite coverage has been installed in the DIII-D tokamak, increasing the graphite tile coverage from 45% to {approximately}90%. Due to a problem with copper impurity bursts during tokamak discharges, copper foam sheets between the graphite tiles and the Inconel wall have been replaced with GRAFOIL gaskets, which further increases the surface area of the graphite inside the DIII-D torus. Although additional graphite has the potential for improving DIII-D performance, if not properly conditioned it can also degrade tokamak discharges by contributing to increased low Z impurity influx and higher particle fueling. To address these concerns, improvements were implemented both in tile preparation and wall conditioning techniques. All previously installed tiles were grit-blasted with boron-carbide grit to remove co-deposited metal impurities and all tiles were outgassed to 1000{degrees}C prior to installation. The glow discharge system used for wall conditioning was modified to include a large area electrode. The only conditioning techniques used were baking and helium conditioning to avoid producing loosely bound carbon. Previously, H{sup 2} or D{sup 2} Taylor discharge cleaning was used extensively during the vent recovery phase. Recovery from the recent 7-month machine opening was extremely rapid with improved modes of energy confinement (H-mode) transitions observed on the 14th plasma discharge. In addition, very high confinement (VH-mode) discharges were obtained without boronization, which is further indication of the rapid vent recovery. Lower recycling/lower fueling efficiency was also observed during operation with all-graphite walls. We will also discuss the use of Ne and Ar glow, and qualification of the new graphite including outgassing tests of the GRAFOIL material.

  18. Neutral beam current drive scaling in DIII-D

    SciTech Connect

    Porter, G.D.; Bhadra, D.K.; Burrell, K.H.; Callis, R.W.; Colleraine, A.P.; Ferron, J.R.; James, R.A.; Kellman, A.G.; Kim, J.; Matsuoka, M.

    1989-03-01

    Neutral beam current drive scaling experiments have been carried out on the DIII-D tokamak at General Atomics. These experiments were performed using up to 10 MW of 80 keV hydrogen beams. Previous current drive experiments on DIII-D have demonstrated beam driven currents up to 340 kA. In the experiments reported here we achieved beam driven currents of at least 500 kA, and have obtained operation with record values of poloidal beta (epsilon..beta../sub p/ = 1.4). The beam driven current reported here is obtained from the total plasma current by subtracting an estimate of the residual Ohmic current determined from the measured loop voltage. In this report we discuss the scaling of the current drive efficiency with plasma conditions. Using hydrogen neutral beams, we find the current drive efficiency is similar in Deuterium and Helium target plasmas. Experiments have been performed with plasma electron temperatures up to T/sub e/ = 3 keV, and densities in the range 2 /times/ 10/sup 19/m/sup /minus/3/ < n/sub e/ < 4 /times/ 10/sup 19/m/sup /minus/3/. The current drive efficiency (nIR/P) is observed to scale linearly with the energy confinement time on DIII-D to a maximum of 0.05 /times/ 10/sup 20/m/sup /minus/2/ A/W. The measured efficiency is consistent with a 0-D theoretical model. In addition to comparison with this simple model, detailed analysis of several shots using the time dependent transport code ONETWO is discussed. This analysis indicates that bootstrap current contributes approximately 10--20% of the the total current. Our estimates of this effect are somewhat uncertain due to limited measurements of the radial profile of the density and temperatures. 4 refs., 1 fig., 1 tab.

  19. 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.

  20. 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.

  1. Negative ion-based neutral injection on DIII-D

    SciTech Connect

    Stewart, L.D.; Bhadra, D.K.; Colleraine, A.P.; Kim, J.

    1990-01-01

    High energy negative ion-based neutral beam injection is a strong candidate for heating and non-inductive current drive in tokamaks. Many of the questions related to the physics and engineering of this technique remain unanswered. In this paper, we consider the possibility of negative ion-based neutral beam injection on DIII-D. We establish the desired parameter space by examining physics trades. This is combined with potential design constraints and a survey of component technology options to establish an injector concept. Injector performance is estimated assuming particular component technologies, and concept flexibility with respect to incorporating alternate technologies is described. 9 refs., 6 figs., 4 tabs.

  2. Divertor E X B Plasma Convection in DIII-D

    SciTech Connect

    Boedo, J.A.; Schaffer, M.J.; Maingi, M.; Lasnier, C.J.; Watkins, J.G.

    1999-07-01

    Extensive two-dimensional measurements of plasma potential in the DIII-D tokamak divertor region are reported for standard (ion VB{sub T} drift toward divertor X-point) and reversed B{sub T} directions; for low (L) and high (H) confinement modes; and for partially detached divertor mode. The data are consistent with recent computational modeling identifying E x B{sub T} circulation, due to potentials sustained by plasma gradients, as the main cause of divertor plasma sensitivity to B{sub T} direction.

  3. 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.

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

    PubMed

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

    2014-11-01

    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.

  5. 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.

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

    SciTech Connect

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

    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.

  7. Divertor Optimization via Control at DIII-D

    NASA Astrophysics Data System (ADS)

    Kolemen, E.; Allen, S. L.; Makowski, M. A.; Soukhanovskii, V. A.; Bray, B. D.; Humphreys, D. A.; Johnson, R.; Leonard, A. W.; Liu, C.; Penaflor, B. G.; Petrie, T. W.; Eldon, D.; McLean, A. G.; Unterberg, E. A.

    2014-10-01

    DIII-D divertor performance and heat-handling capabilities are optimized using advanced control techniques. The world's first real-time snowflake divertor detection and control system was implemented on DIII-D in order to stabilize and optimize this configuration. A new control system was implemented to regulate and study detachment and radiation, since future fusion reactors will require detached or partially detached plasmas to achieve acceptable divertor target heat fluxes. The algorithm regulates the D2 and impurity gas injection level by using the divertor temperature measurements from real-time Thomson diagnostics to compute the detachment level, and the real-time bolometer diagnostics to determine core and divertor radiation. This control allows the optimization of the detachment and radiation from the core and the divertor to achieve high core performance compatible with reduced heat-flux to the divertor. Work supported by the US DOE under DE-AC02-09CH11466, DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-AC05-00OR22725.

  8. 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.

  9. Model of ELM suppression by RMPs in DIII-D

    NASA Astrophysics Data System (ADS)

    Callen, J. D.; Nazikian, R.; Ferraro, N. M.; Beidler, M. T.; Hegna, C. C.; La Haye, R. J.; Paz-Soldan, C.

    2016-10-01

    Recent DIII-D experiments explored effects of resonant magnetic perturbations (RMPs) near the minimum applied n=2 RMP amplitude required for ELM suppression in ITER-relevant low collisionality pedestals. Comprehensive tokamak forced magnetic reconnection (FMR) theory is used to describe and quantify the many physical processes involved in stages of RMP effects and an ELM crash response that lead to bifurcation into an ELM-suppressed state: 1) in ELMing equilibrium, flow-screening is strong with little magnetic reconnection; 2) the RMP at q=8/2 penetrates via FMR induced by an ELM crash and locks toroidal flow to the lab frame (like error field mode locking); 3) the ELM crash provides a 8/2 seed island (like NTMs) governed by a modified Rutherford equation; 4) if the total 8/2 RMP is large enough the internal tearing response and flow bifurcate and grow; and 5) flutter transport reduces pedestal top gradients which stabilizes P-B modes and hence suppresses ELMs. This analysis is for discharge 158115 in DIII-D; its potential universality is yet to be determined. Work supported by OFES/DOE under DE-FG02-92ER54139, DE-FG02-86ER53218, DE-AC02-09CH11466 and DE-FC02-04ER54698.

  10. 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.

  11. 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.

  12. A fast scanning probe for DIII-D

    NASA Astrophysics Data System (ADS)

    Watkins, J. G.; Salmonson, J.; Moyer, R.; Doerner, R.; Lehmer, R.; Schmitz, L.; Hill, D. N.

    1992-10-01

    A fast reciprocating probe has been developed for DIII-D which can penetrate the separatrix during H mode with up to 5 MW of NBI heating. The probe has been designed to carry various sensor tips into the scrape-off layer at a velocity of 3 m/s and dwell motionless for a programmed period of time. The driving force is provided by a pneumatic cylinder charged with helium to facilitate greater mass flow. The first series of experiments have been done using a Langmuir probe head with five graphite tips to measure radial profiles of ne, Te, φf, ñe, and φ˜f. The amplitude and phase of the fluctuating quantities are measured by using specially constructed vacuum compatible 5-kV coaxial transmission lines which allow us to extend the measurements into the MHz range. TTZ ceramic bearings and fast stroke bellows were also specially designed for the DIII-D probe. Initial measurements will be presented.

  13. 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.

  14. Automated Calculation of DIII-D Neutral Beam Availability

    SciTech Connect

    Phillips, J.C.; Hong, R.M.; Scoville, B.G.

    1999-11-01

    The neutral beam systems for the DIII-D tokamak are an extremely reliable source of auxiliary plasma heating, capable of supplying up to 20 MW of injected power, from eight separate beam sources into each tokamak discharge. The high availability of these systems for tokamak operations is sustained by careful monitoring of performance and following up on failures. One of the metrics for this performance is the requested injected power profile as compared to the power profile delivered for a particular pulse. Calculating this was a relatively straightforward task, however innovations such as the ability to modulate the beams and more recently the ability to substitute an idle beam for one which has failed during a plasma discharge, have made the task very complex. For example, with this latest advance it is possible for one or more beams to have failed, yet the delivered power profile may appear perfect. Availability used to be manually calculated. This paper presents the methods and algorithms used to produce a system which performs the calculations based on information concerning the neutral beam and plasma current waveforms, along with post-discharge information from the Plasma Control System, which has the ability to issue commands for beams in real time. Plots representing both the requested and actual power profiles, along with statistics, are automatically displayed and updated each shot, on a web-based interface viewable both at DIII-D and by our remote collaborators using no-cost software.

  15. 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.

  16. 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.

  17. Simulation of plasma flow in the DIII-D Tokamak

    SciTech Connect

    Porter, G. D., LLNL

    1998-06-19

    The importance of the parallel flow of primary and impurity ions in the Scrape-Off layer (SOL) of divertor tokamaks has been recognized recently. Impurity accumulation on the closed flux surfaces is determined in part by their parallel flow in the SOL. In turn, the parallel transport of the impurity ions is determined in part by drag from the primary ion flow. Measurement of flow in the DIII-D tokamak has begun recently. We describe initial results of modeling plasma ion flow using the 2-D code UEDGE in this paper. We assume the impurity (carbon) arises from chemical and physical sputtering from the walls surrounding the DIII-D plasma. We include six charge states of carbon in our simulations. We make detailed compaison with a multitude of SOL plasma diagnostics, including the flow measurement, to verify the UEDGE physics model. We begin the paper with a brief description of the plasma and neutral models in the UEDGE code in Section 2. We then present initial results of flow simulations and compare them with experimental measurement in Section 3. We conclude with a discussion of the dominant physics processes identified in the modeling in Section 4.

  18. Validation of EFIT++ MHD Equilibrium Reconstructions on DIII-D

    NASA Astrophysics Data System (ADS)

    Cornille, B.; Lanctot, M. J.; Lao, L. L.; Appel, L. C.; Meneghini, O.; Holcomb, C. T.

    2013-10-01

    MHD force balance calculations play a key role in the optimization of transport and stability in tokamaks. In high confinement tokamak plasmas, reconstructions of the MHD plasma equilibrium are needed to resolve key profile features including the edge pressure pedestal and resulting bootstrap current. The EFIT code is the standard tool for calculating MHD force balance in DIII-D and many tokamaks. This code has recently been rebuilt to be machine-independent in order to facilitate cross-machine comparisons. This update, EFIT++, is in its late stages of development and requires validation for widespread use. Benchmarking of EFIT++ against the established EFIT cases including motional Stark effect measurements from DIII-D will be presented. Work supported by the the National Undergraduate Fellowship Program in Plasma Physics and Fusion Energy Sciences and the US Department of Energy under DE-FC02-04ER54698, DE-AC05-06OR23100 and DE-AC52-07NA27344.

  19. 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.

  20. DIII-D First Wall Metal Impurity Migration Trends

    NASA Astrophysics Data System (ADS)

    Chrobak, C. P.; Torreblanca, H.; Holtrop, K.; Thomas, D.; Unterberg, E. A.; Donovan, D. C.; Buchenauer, D.; Stangeby, P.

    2016-10-01

    Metal accumulation on the entire graphite first wall of the DIII-D tokamak was measured using a novel X-Ray Fluorescence technique. The technique is sensitive to and distinguishes between various elements from Al to U. During a two week campaign, the DIII-D tokamak operated with W-coated Mo alloy (TZM) metal tiles in the divertor, arranged in two 5cm wide toroidally uniform rings. These rings provided a large localized source of W to enable the first measurements of whole-vessel high-Z metal migration from a known divertor source. We have also measured deposits of Ni, Cr, Fe, Cu, and Mo originating from various other sources including sputtering from neutral beam drift-duct port walls and RF antenna Faraday shields. The integrated deposition pattern for different elements originating from different sources during these campaigns will be compared. In addition, a detailed description of the intrinsic metal impurities in different unexposed grades of graphite will be presented. Work supported by US DOE under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC05-94AL85000.

  1. Fueling efficiency of pellet injection on DIII-D

    SciTech Connect

    Baylor, L.R.; Jernigan, T.C.; Maingi, R.; Lasnier, C.J.; Ali Mahdavi, M.

    1998-05-01

    Pellet injection has been used on the DIII-D tokamak to study density limits and particle transport in H-mode and inner wall limited L-mode plasmas. These experiments have provided a variety of conditions in which to examine the fueling efficiency of pellets injected into DIII-D plasmas. The fueling efficiency defined as the total increase in number of plasma electrons divided by the number of pellet fuel atoms, is determined by measurements of density profiles before and just after pellet injection. The authors have found that there is a decrease in the pellet fueling efficiency with increased neutral beam injection power. The pellet penetration depth also decreases with increased neutral beam injection power so that, in general, fueling efficiency increases with penetration depth. The fueling efficiency is generally 25% lower in ELMing H-mode discharges than in L-mode due to an expulsion of particles with a pellet triggered ELM. A comparison with fueling efficiency data from other tokamaks shows similar behavior.

  2. EUV Spectroscopy During the DIII-D Tungsten Divertor Campaign

    NASA Astrophysics Data System (ADS)

    Allen, S. L.; Victor, B. S.; Beiersdorfer, P.; Magee, E.; Soukhanovskii, V.; Weller, M.; Loch, S.; Thomas, D.

    2016-10-01

    Two toroidal rings of tungsten-coated tile inserts were installed in the DIII-D lower divertor and a range of L- and H-mode plasma discharges were compared during a dedicated two week run campaign. A high resolution (1340 spectral channels) variable-ruling grating spectrometer viewing the core of the plasma was used to study the spectral region 10-70 Å a second spectrometer viewing 20 - 150 Å was also used. At DIII-D core plasma temperatures 2-3 keV, several emission lines from W38+ through W43+ were identified, including a quasi-continuum feature of W near 50 Å whose structure depends on core Te. Molybdenum (TZM substrate) emissions between 20-30 Å and near 70 Å were also observed. ADAS calculations are used to guide the identification of W emission lines for the measured core plasma Te and ne profiles. The behavior of W emissions during both ``benign'', pellet injection, and impurity accumulation conditions will be presented. Supported by US DOE under DE-AC52-07NA27344, and DE-FC02-04ER54698.

  3. 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.

  4. Current status of DIII-D real-time digital plasma control

    SciTech Connect

    Penaflor, B.G.; Piglowski, D.A.; Ferron, J.R.; Walker, M.L.

    1999-06-01

    This paper describes the current status of real-time digital plasma control for the DIII-D tokamak. The digital plasma control system (PCS) has been in place at DIII-D since the early 1990s and continues to expand and improve in its capabilities to monitor and control plasma parameters for DIII-D fusion science experiments. The PCs monitors over 200 tokamak parameters from the DIII-D experiment using a real-time data acquisition system that acquires a new set of samples once every 60 {micro}s. This information is then used in a number of feedback control algorithms to compute and control a variety of parameters including those affecting plasma shape and position. A number of system related improvements has improved the usability and flexibility of the DIII-D PCS. These include more graphical user interfaces to assist in entering and viewing the large and ever growing number of parameters controlled by the PCS, increased interaction and accessibility from other DIII-D applications, and upgrades to the computer hardware and vended software. Future plans for the system include possible upgrades of the real-time computers, further links to other DIII-D diagnostic measurements such as real-time Thomson scattering analysis, and joint collaborations with other tokamak experiments including the NSTX at Princeton.

  5. Tangles of the ideal separatrix from low mn perturbation in the DIII-D

    NASA Astrophysics Data System (ADS)

    Goss, Talisa; Crank, Willie; Ali, Halima; Punjabi, Alkesh

    2010-11-01

    The equilibrium EFIT data for the DIII-D shot 115467 at 3000 ms is used to construct the equilibrium generating function for magnetic field line trajectories in the DIII-D tokamak in natural canonical coordinates [A. Punjabi, and H. Ali, Phys. Plasmas 15, 122502 (2008); A. Punjabi, Nucl. Fusion 49, 115020 (2009)]. The generating function represents the axisymmetric magnetic geometry and the topology of the DIII-D shot very accurately. A symplectic map for field line trajectories in the natural canonical coordinates in the DIII-D is constructed. We call this map the DIII-D map. The natural canonical coordinates can be readily inverted to physical coordinates (R,φ,Z). Low mn magnetic perturbation with mode numbers (m,n)=(1,1)+(1,-1) is added to the generating function of the map. The amplitude for the low mn perturbation is chosen to be 6X10-4, which is the expected value of the amplitude in tokamaks. The forward and backward DIII-D maps with low mn perturbation are used to calculate the tangles of the ideal separatrix from low mn perturbation in the DIII-D. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

  6. 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.

  7. Transport Simulations of DIII-D Discharges with Impurity Injection

    NASA Astrophysics Data System (ADS)

    Mandrekas, J.; Stacey, W. M.; Murakami, M.

    2001-10-01

    Several recent DIII-D discharges with external impurity injection into L-mode plasmas are analyzed with a coupled main plasma and multi-charge state 1frac 12-D impurity transport code. These discharges exhibit various degrees of confinement improvement, which has been attributed to the synergistic effects of impurity induced enhancement of the E×B shearing rate and reduction of the drift wave turbulence growth rate (M. Murakami, et. al., Nucl. Fusion 41) (2001) 317.. Impurity transport is described by empirical and neoclassical transport models. Both the standard neoclassical theory as well as an enhanced theory which takes into account the effects of external momentum input and radial momentum transport (W.M. Stacey, Phys. Plasmas 8) (2001) 158. have been considered.

  8. Stability in high gain plasmas in DIII-D

    SciTech Connect

    Lazarus, E.A.; Houlberg, W.A.; Murakami, M.; Wade, M.R.

    1996-10-01

    Fusion power gain has been increased by a factor of 3 in DIII-D plasmas through the use of strong discharge shaping and tailoring of the pressure and current density profiles. H-mode plasmas with weak or negative central magnetic shear are found to have neoclassical ion confinement throughout most of the plasma volume. Improved MHD stability is achieved by controlling the plasma pressure profile width. The highest fusion power gain Q (ratio of fusion power to input power) in deuterium plasmas was 0.0015, which extrapolates to an equivalent Q of 0.32 in a deuterium-tritium plasma and is similar to values achieved in tokamaks of larger size and magnetic fields.

  9. Stability in High Gain Plasmas in DIII-D

    SciTech Connect

    Lazarus, E. A.; Hong, R. M.; Navratil, G. A.; Sabbagh, S.; Strait, E. J.; Rice, B. W.; Ferron, J. R.; Greenfield, C. M.; Austin, M. E.; Chan, V. S.; DeBoo, J. C.; Doyle, E. J.; Forest, C. B.; Leonard, A. W.; Schissel, D. P.; Whyte, D. G.

    1997-01-01

    Fusion power gain has been increased by a factor of 3 in DIII-D plasmas through the use of strong discharge shaping and tailoring of the pressure and current density profiles. H-mode plasmas with weak or negative central magnetic shear are found to have neoclassical ion confinement throughout most of the plasma volume. Improved MHD stability is achieved by controlling the plasma pressure profile width. The highest fusion power gain Q (ratio of fusion power to input power) in deuterium plasmas was 0.0015. which extrapolates to an equivalent Q of 0.32 in a deuterium-tritium plasma and is similar to values achieved in tokamaks of larger size and magnetic fields.

  10. 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.

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

    DOE PAGES

    Chrystal, Colin; Burrell, K. H.; Grierson, Brian A.; ...

    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. Soft X-ray Tomography at DIII-D

    NASA Astrophysics Data System (ADS)

    Rinderknecht, H.; Fisher, R. K.; Hollmann, E. M.; Lanctot, M. J.; Volpe, F.

    2007-11-01

    Two new 32 channel SXR pinhole cameras have been recently installed in the DIII-D tokamak. They are sensitive to photons in the 2-20 keV range, but an interchangeable set of diamond filters with five settings allows selection of the range of energies of interest. New tomographic inverters were developed and validated against analytic models and magnetically reconstructed EFIT equilibria. Tomographic inversion techniques suitable for use with the new diagnostic geometry and preliminary inversions of new SXR data will be presented, along with re-analysis of earlier measurements of disruption-generated fast electrons and equilibria. Thanks to a temporal resolution of a few microseconds, progress has also been made in the tomographic reconstruction of rapidly moving, relatively weak emitters such as rotating islands.

  13. MHD Equilibrium Reconstruction in the DIII-D Tokamak

    SciTech Connect

    Lao, L.L.; St John, H.E.; Peng, Q.; Ferron, J.R.; Strait, E.J.; Taylor, T.S.; Meyer, W.H.; Zhang, C.; You, K.I.

    2005-10-15

    Physics elements and advances crucial for the development of axisymmetric magnetohydrodynamic equilibrium reconstruction to support plasma operation and data analysis in the DIII-D tokamak are reviewed. A response function formalism and a Picard linearization scheme are used to efficiently combine the equilibrium and the fitting iterations and search for the optimum solution vector. Algorithms to incorporate internal current and pressure profile measurements, topological constraints, and toroidal plasma rotation into the equilibrium reconstruction are described. Choice of basis functions and boundary conditions essential for accurate reconstruction of L- and H-mode equilibrium plasma boundary and current and pressure profiles is discussed. The computational structure used to efficiently integrate these elements into the equilibrium reconstruction code EFIT is summarized.

  14. Determination of Edge Current in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Hudson, B.; Petty, C. C.; Burrell, K. H.; Lao, L. L.; Snyder, P. B.; Thomas, D. M.; Allen, S. L.; Holcomb, C. T.; Makowski, M. A.

    2008-11-01

    A statistical study of the measurement requirements necessary to infer the current profile in the pedestal of the DIII-D tokamak via motional Stark effect (MSE) and lithium beam diagnostics has been performed. Error in the edge magnetic probes and flux loops, Thomson scattering, and MSE and lithium beam polarimetry was systematically varied to determine the uncertainty in inference of the current profile via magnetic reconstructions using EFIT. It will be shown that edge MSE or lithium beam data accurate to 0.1 deg is sufficient to resolve the peak edge current density to within 20%. This accuracy in the current profile is needed for validation of the peeling-ballooning stability model. Future work using a Hall probe to directly measure the magnetic field components in the plasma edge and SOL to further constrain edge current reconstruction will be presented.

  15. Modeling of detachment experiments at DIII-D

    DOE PAGES

    Canik, John M.; Briesemeister, Alexis R.; Lasnier, C. J.; ...

    2014-11-26

    Edge fluid–plasma/kinetic–neutral modeling of well-diagnosed DIII-D experiments is performed in order to document in detail how well certain aspects of experimental measurements are reproduced within the model as the transition to detachment is approached. Results indicate, that at high densities near detachment onset, the poloidal temperature profile produced in the simulations agrees well with that measured in experiment. However, matching the heat flux in the model requires a significant increase in the radiated power compared to what is predicted using standard chemical sputtering rates. Lastly, these results suggest that the model is adequate to predict the divertor temperature, provided thatmore » the discrepancy in radiated power level can be resolved.« less

  16. Modeling of detachment experiments at DIII-D

    SciTech Connect

    Canik, John M.; Briesemeister, Alexis R.; Lasnier, C. J.; Leonard, A. W.; Lore, J. D.; McLean, A. G.; Watkins, J. G.

    2014-11-26

    Edge fluid–plasma/kinetic–neutral modeling of well-diagnosed DIII-D experiments is performed in order to document in detail how well certain aspects of experimental measurements are reproduced within the model as the transition to detachment is approached. Results indicate, that at high densities near detachment onset, the poloidal temperature profile produced in the simulations agrees well with that measured in experiment. However, matching the heat flux in the model requires a significant increase in the radiated power compared to what is predicted using standard chemical sputtering rates. Lastly, these results suggest that the model is adequate to predict the divertor temperature, provided that the discrepancy in radiated power level can be resolved.

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

    SciTech Connect

    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 16 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.

  18. Intershot Analysis of Flows in DIII-D

    NASA Astrophysics Data System (ADS)

    Meyer, W. H.; Allen, S. L.; Samuell, C. M.; Howard, J.

    2016-10-01

    Analysis of the DIII-D flow diagnostic data require demodulation of interference images, and inversion of the resultant line integrated emissivity and flow (phase) images. 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 inversion matrix. Serial processing has been used for the lower divertor viewing flow camera 800x600 pixel image. The full cross section viewing camera will require parallel processing of the 2160x2560 pixel image. We will discuss using a Posix thread pool and a Tesla K40c GPU in the processing of this data. 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.

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

    NASA Astrophysics Data System (ADS)

    Chrystal, C.; Burrell, K. H.; Grierson, B. A.; Haskey, S. R.; Groebner, R. J.; Kaplan, D. H.; Briesemeister, A.

    2016-11-01

    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 16 to 38. 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.

  20. 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.

  1. Diagnostics for the DIII-D radiative divertor

    SciTech Connect

    Nilson, D.G.; Brooks, N.H.; Smith, J.P.; Snider, R.T.

    1995-10-01

    This paper reviews the design of new diagnostics and the modifications to existing diagnostics needed to carry out radiative divertor experiments in DIII-D following installation in late 1996 of a set of baffle structures that will restrict the backflow to the core plasma of neutral deuterium atoms and impurity gases. The divertor slots formed by the new baffle structures will inhibit the easy view of the divertor legs and target plates that the open divertor geometry in DIII-D currently affords. We review a basic set of diagnostics that are needed to demonstrate the reduction of divertor heat loading and radiative dissipation of energy within the divertor. This will include IR cameras, bolometry, foil bolometers, and Langmuir probes. Within the limits of available funding, we will implement a supplemental set of instruments which provide a more detailed understanding of the underlying physical processes. Many existing diagnostics require only re-aiming to provide proper coverage of the initial 23 cm long divertor plasma configuration (X- point to floor distance). Other diagnostics need extensive reconfiguration using in-vessel fiber-optic bundles or high power laser mirrors. The new divertor baffle panels provide a protective shelf for diagnostic hardware mounted underneath them, but the water cooling channels in the panels limit the permissible size of through holes and, thereby, restrict the available views of under-the- baffle diagnostics. The successful resolution of the design and implementation of these diagnostic modifications is dependent on a strong coordination between GA and its many diagnostic collaborators.

  2. Fast wave current drive on DIII-D

    SciTech Connect

    deGrassie, J.S.; Petty, C.C.; Pinsker, R.I.

    1995-07-01

    The physics of electron heating and current drive with the fast magnetosonic wave has been demonstrated on DIII-D, in reasonable agreement with theoretical modeling. A recently completed upgrade to the fast wave capability should allow full noninductive current drive in steady state advanced confinement discharges and provide some current density profile control for the Advanced Tokamak Program. DIII-D now has three four-strap fast wave antennas and three transmitters, each with nominally 2 MW of generator power. Extensive experiments have been conducted with the first system, at 60 MHz, while the two newer systems have come into operation within the past year. The newer systems are configured for 60 to 120 MHz. The measured FWCD efficiency is found to increase linearly with electron temperature as {gamma} = 0.4 {times} 10{sup 18} T{sub eo} (keV) [A/m{sup 2}W], measured up to central electron temperature over 5 keV. A newly developed technique for determining the internal noninductive current density profile gives efficiencies in agreement with this scaling and profiles consistent with theoretical predictions. Full noninductive current drive at 170 kA was achieved in a discharge prepared by rampdown of the Ohmic current. Modulation of microwave reflectometry signals at the fast wave frequency is being used to investigate fast wave propagation and damping. Additionally, rf pick-up probes on the internal boundary of the vessel provide a comparison with ray tracing codes, with dear evidence for a toroidally directed wave with antenna phasing set for current drive. There is some experimental evidence for fast wave absorption by energetic beam ions at high cyclotron harmonic resonances.

  3. Plasma flow in the DIII-D divertor

    SciTech Connect

    Boedo, J.A.; Porter, G.D.; Schaffer, M.J.

    1998-07-01

    Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor.

  4. 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.

  5. 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.

  6. Real time software for the control and monitoring of DIII-D system interlocks

    SciTech Connect

    Broesch, J.D.; Penaflor, B.G.; Coon, R.M.; Harris, J.J.; Scoville, J.T.

    1996-10-01

    This paper describes the real time, multi-tasking, multi-user software and communications of the E-Power Supply System Integrated Controller (EPSSIC) for the DIII-D tokamak. EPSSIC performs the DIII-D system wide go/no-go determination for the plasma sequencing. This paper discusses the data module handling, task work load balancing, and communications requirements. Operational experience with the new EPSSIC and recent improvements to this system are also described.

  7. 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.

  8. Long-Pulse Integrator Testing with DIII-D Magnetic Diagnostics

    NASA Astrophysics Data System (ADS)

    Slobodov, Ilia; Miller, Kenneth; Ziemba, Timothy; Prager, James; Carscadden, John; Hanson, Eric

    2016-10-01

    Eagle Harbor Technologies (EHT), Inc. has developed a high-gain integrator for magnetic diagnostics that meets ITER specifications including integration time and integration error limits. EHT has conducted testing of this long-pulse integrator at DIII-D with existing DIII-D magnetic probes. The EHT long-pulse integrator was operated for several hours up to a full day. During a single period of EHT integrator operation, DIII-D was pulsed multiple times. The multiple pulses from the DIII-D magnetic diagnostics can be clearly resolved in the integrator signal output. The results are compared to DIII-D measurements. EHT also operated the long pulse integrator in High Dynamic Range Mode (HDRM), which effectively allows for a dramatic increase in measurement bit depth for higher resolution signal acquisition with the same diagnostic and digitizers presently available on DIII-D. Additionally, EHT has tested a new microprocessor and FPGA-based digitizer, which can be included on the integrator PCB, for a single board magnetic diagnostic solution.

  9. Global stability and operational regimes of ignitor, ITER, and alcator C-mod

    SciTech Connect

    Detragiache, P.; Bombarda, F.; Coppi, B.; Migliuolo, S.

    1996-12-31

    One of the primary requirements for an experiment capable of approaching ignition conditions is the macroscopic stability of the plasma column. For two proposed ignition experiments (Ignitor and ITER) we analyze numerically (using both the PEST and DCON codes, involving the ideal MHD linearized approximation) the stability of their confinement configuration for different plasma pressure and current profiles. Since in both cases the aspect ratio is tight (2.8 and 2.9, respectively), when the unwinding parameter q({Psi}) < 1 for r {approx_lt} a/2, n = 1 modes, for which the m = 1 and m = 2 poloidal harmonics prevail, involve a major part of the plasma column (q({Psi}) = 2 for r{sub 2} {approx} 3a/4). For profiles that are nearly flat up to the q(r{sub 1}) = 1 surface and are represented by p = p{sub 0}(1 - {Psi}{sup 3}){sup 3/2}, where {Psi} is the (normalized) poloidal flux variable, pressure driven modes, dominated by the m = 2 harmonic, may be found, depending on the value of {beta}{sub p} = 2 {mu}{sub 0}(p)/B{sub p}{sup 2}. These modes are unstable for the reference parameters of operation of ITER (B{sub T}{congruent} 5.7 {Tau}, n{sub 0} {congruent} 1.5 x 10{sup 20} m{sup -3}, {Tau}{sub 0} {congruent} 20 keV), but are stable in the case of Ignitor B{sub T}{congruent}13 {Tau}, n{sub 0} {congruent} 12 keV.

  10. ICRF Mode Conversion Flow Drive Experiments on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lin, Y.; Reinke, M. L.; Rice, J. E.; Wukitch, S. J.; Granetz, R.; Greenwald, M.; Hubbard, A. E.; Marmar, E. S.; Podpaly, Y. A.; Porkolab, M.; Tsujii, N.; Wolfe, S.

    2011-12-01

    We have carried out a detailed study of the dependence of ICRF mode conversion flow drive (MCFD) on plasma and RF parameters. The flow drive efficiency is found to depend strongly on the 3He concentration in D(3He) plasmas, a key parameter separating the ICRF minority heating regime and mode conversion regime. At +90 ° antenna phasing (waves in the co-Ip direction) and dipole phasing, the driven flow is in the co-Ip direction, and the change of the rotation velocity increases with both PRF and Ip, and scales unfavorably vs. plasma density and antenna frequency. When MCFD is applied to I-mode plasmas, the plasma rotation increases until the onset of MHD modes triggered by large sawtooth crashes. Very high performance I-mode plasmas with HITER98,y2˜1.4 and Te0˜8 keV have been obtained in these experiments.

  11. Light impurity transport in I-mode in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Rowan, W. L.; Bespamyatnov, I. O.; Hatch, D. R.; Horton, W. L.; Liao, K. T.

    2014-10-01

    The I-mode hallmarks are H-mode-like electron temperature pedestal and energy confinement simultaneous with L-mode-like density pedestal and particle confinement. The I-mode is observed over a wide range of plasma parameters and is robust. As might be expected from the particle confinement observation, accumulation of naturally-occuring impurities is reduced compared to H-mode. Heavy impurity measurements confirm the observation of L-mode-like particle confinement. In the results reported here for light impurities, I-mode impurity profiles are compared with both H- and L- mode profiles for helium and boron with an emphasis on core confinement. We search for the dependence of the impurity density gradient scale lengths on the main ion density scale length and temperature scale length, the Zeff the collisionality, as well as on total radiation loss, stored energy, and global confinement. The results allow comparisons among discharge modes, as well as comparison to other devices, and turbulence predictions. The data analysis is compared with gyrokinetic simulations using the GENE code. Supported by USDoE award DE-FG03-96ER-54373.

  12. The physics mechanisms of the weakly coherent mode in the Alcator C-Mod Tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Z. X.; Xu, X. Q.; Gao, X.; Hubbard, A. E.; Hughes, J. W.; Walk, J. R.; Theiler, C.; Xia, T. Y.; Baek, S. G.; Golfinopoulos, T.; Whyte, D.; Zhang, T.; Li, J. G.

    2016-12-01

    The weakly coherent mode (WCM) in I-mode has been studied by a six-field two-fluid model based on the Braginskii equations under the BOUT++ framework for the first time. The calculations indicate that a tokamak pedestal exhibiting a WCM is linearly unstable to drift Alfven wave (DAW) instabilities and the resistive ballooning mode. The nonlinear simulation shows promising agreement with the experimental measurements of the WCM. The shape of the density spectral and location of the spectral peak of the dominant toroidal number mode n = 20 agrees with the experimental data from reflectometry. The simulated mode propagates in electron diamagnetic direction is consistent with the results from the magnetic probes in the laboratory frame, a large ratio of particle to heat diffusivity is consistent with the distinctive experimental feature of I-mode, and the value of the simulated χe at the edge is in the range of experimental errors of χeff from the experiment. The prediction of the WCM shows that free energy is mainly provided by the electron pressure gradient, which gives guidance for pursuing future I-mode studies.

  13. Electron Temperature Gradient Scale Measurements in ICRF Heated Plasmas at Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Houshmandyar, Saeid; Phillips, Perry E.; Rowan, William L.; Howard, Nathaniel T.; Greenwald, Martin

    2016-10-01

    It is generally believed that the temperature gradient is a driving mechanism for the turbulent transport in hot and magnetically confined plasmas. A feature of many anomalous transport models is the critical threshold value (LC) for the gradient scale length, above which both the turbulence and the heat transport increases. This threshold is also predicted by the recent multi-scale gyrokinetic simulations, which are focused on addressing the electron (and ion) heat transport in tokamaks. Recently, we have established an accurate technique (BT-jog) to directly measure the electron temperature gradient scale length (LTe =Te / ∇T) profile, using a high-spatial resolution radiometer-based electron cyclotron emission (ECE) diagnostic. For the work presented here, electrons are heated by ion cyclotron range of frequencies (ICRF) through minority heating in L-mode plasmas at different power levels, TRANSP runs determine the electron heat fluxes and the scale lengths are measured through the BT-jog technique. Furthermore, the experiment is extended for different plasma current and electron densities by which the parametric dependence of LC on magnetic shear, safety factor and density will be investigated. This work is supported by U.S. DoE OFES, under Award No. DE-FG03-96ER-54373.

  14. Central Thomson Scattering Diagnostic for DIII--D

    NASA Astrophysics Data System (ADS)

    Bray, B.; Carlstrom, T. N.; Hsieh, C.; Marakiou, C. C.; Nilson, D.

    1998-11-01

    The Thomson diagnostic on DIII--D has eight YAG lasers operating at 20 Hz each and two vertical beam paths covering most of the plasma including the boundary and divertor regions. In certain instances such as high performance discharges with an internal transport barrier, the measurement can miss the plasma center by about 10--20 cm. In order to cover this central region of increasing importance, we plan to install a horizontal beam path with up to 3 lasers diverted from the existing system, a laser dump inside the machine vessel, and a maximum of 12 viewing channels. Most system hardware including lasers and polychromators will be made sharable so the spatial and temporal resolution can be arranged according to the requirements of a specific experiment. Upgrades are also planned for control and data analysis to replace older computer hardware and software. This new software will have more inspection functions to maintain the quality of the data and provide the flexibility that the plasma experiment requires. The plans for the expansion and progress will be presented.

  15. 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.

  16. 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.

  17. 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.

  18. Visible spectroscopy in the DIII-D divertor

    SciTech Connect

    Brooks, N.H.; Fehling, D.; Hillis, D.L.; Klepper, C.C.; Naumenko, N.; Tugarinov, S.; Whyte, D.G.

    1996-06-01

    Spectroscopy measurements in the DIII-D divertor have been carried out with a survey spectrometer which provides simultaneous registration of the visible spectrum over the region 400--900 nm with a resolution of 0.2 nm. Broad spectral coverage is achieved through use of a fiberoptic transformer assembly to map the curved focal plane of a fast (f/3) Rowland-circle spectrograph into a rastered format on the rectangular sensor area of a two-dimensional CCD camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (< 10 ms) may be obtained by selecting for readout just a small number of the twenty spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges.

  19. 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.

  20. Snowflake Divertor Configuration Studies in DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; 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.

    2014-10-01

    Recent DIII-D studies show that the snowflake (SF) divertor enables significant manipulation of divertor heat transport for power exhaust in attached and radiative divertor conditions, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Results include: 1) Increased scrape-off layer (SOL) width suggesting enhanced divertor heat transport; 2) Direct measurements of divertor null-region poloidal beta βp >> 1 in support of the theoretically proposed instability mechanism leading to fast convective plasma redistribution, especially efficient during ELMs, and contribution to 1); 3) Weak effect on pedestal profile and stability resulting in essentially unchanged ELM regime; 4) Reduction of Type-I ELM energy loss; 5) In radiative SF divertor regimes with D2 seeding, a significant reduction of peak heat fluxes between and during ELMs, as in standard H-modes. Work supported by the US Department of Energy under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698, and DE-AC04-94AL85000.

  1. Snowflake Divertor Configuration Studies in DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

    Experiments in DIII-D show the snowflake divertor (SFD) configuration is compatible with high performance operation (H98 y 2 >= 1) and results in greatly reduced divertor heat flux between and during edge localized modes (ELMs). The SFD was sustained for many energy confinement times using the standard poloidal field shaping coils in 3-5 MW neutral beam injection-heated discharges. Pedestal and divertor effects resulting from a large region of reduced poloidal magnetic field in the SFD are measured and studied using the 2D multi-fluid code UEDGE. The pedestal pressure appeared to be unchanged, while the energy loss per ELM was reduced by 50%. Partial detachment of the SFD was observed at higher ne, with an expanded divertor radiation zone and peak ELM heat flux reduced by up to 80%. Work supported by the US Department of Energy under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698, DE-FG02-07ER54917, and DE-AC04-94AL85000.

  2. 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.

  3. 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.

  4. 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.

  5. Overview of Recent DIII-D Experimental Results

    NASA Astrophysics Data System (ADS)

    Fenstermacher, Max

    2015-11-01

    Recent DIII-D experiments have added to the ITER physics basis and to physics understanding for extrapolation to future devices. ELMs were suppressed by RMPs in He plasmas consistent with ITER non-nuclear phase conditions, and in steady state hybrid plasmas. Characteristics of the EHO during both standard high torque, and low torque enhanced pedestal QH-mode with edge broadband fluctuations were measured, including edge localized density fluctuations with a microwave imaging reflectometer. The path to Super H-mode was verified at high beta with a QH-mode edge, and in plasmas with ELMs triggered by Li granules. ITER acceptable TQ mitigation was obtained with low Ne fraction Shattered Pellet Injection. Divertor ne and Te data from Thomson Scattering confirm predicted drift-driven asymmetries in electron pressure, and X-divertor heat flux reduction and detachment were characterized. The crucial mechanisms for ExB shear control of turbulence were clarified. In collaboration with EAST, high beta-p scenarios were obtained with 80 % bootstrap fraction, high H-factor and stability limits, and large radius ITBs leading to low AE activity. Work supported by the US Department of Energy under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  6. 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.

  7. 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.

  8. Overview of recent DIII-D experimental results

    NASA Astrophysics Data System (ADS)

    Fenstermacher, M. E.; LLNL; DIII-D Team

    2016-10-01

    Recent DIII-D experiments have contributed to the ITER physics basis and to physics understanding for extrapolation to future devices. Resonant Magnetic Perturbation ELM suppression was extended to lower shaping as a guide to joint experiments showing first ELM suppression in ASDEX-U. The physics model of suppression was validated in low torque ITER baseline plasmas. Toroidal variation of density gradients and turbulence was documented during RMP. Plasma rotation was predicted from measured scaling of intrinsic torque and momentum transport. Runaway electron plateau dissipation using SPI was demonstrated including new understanding of synchrotron and collisional damping effects. Good coupling of high frequency RF from a helicon antenna during H-mode was observed. Upstream pedestal density at divertor detachment decreased with increasing divertor closure. Bifurcations to detachment in H-mode with increasing density were reproduced with UEDGE including drifts. Sources, SOL transport and core accumulation of tungsten from toroidally continuous divertor target tiles were identified. Work supported by the US Department of Energy under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  9. 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.

  10. DIII-D dust particulate characterization (June 1998 Vent)

    SciTech Connect

    Carmack, W.J.

    1999-01-01

    Dust is a key component of fusion power device accident source term. Understanding the amount of dust expected in fusion power devices and its physical and chemical characteristics is needed to verify assumptions currently used in safety analyses. An important part of this safety research and development work is to characterize dust from existing experimental tokamaks. In this report, the authors present the collection, data analysis methods used, and the characterization of dust particulate collected from various locations inside the General Atomics DIII-D vacuum vessel following the June 1998 vent. The collected particulate was analyzed at the Idaho National Engineering and Environmental Laboratory (INEEL). Two methods were used to collect particulate with the goal of preserving the particle size distribution and physical characteristics of the particulate. Choice of collection technique is important because the sampling method used can bias the particle size distribution collected. Vacuum collection on substrates and adhesion removal with metallurgical replicating tape were chosen as non-intrusive sampling methods. Seventeen samples were collected including plasma facing surfaces in lower, upper, and horizontal locations, surfaces behind floor tiles, surfaces behind divert or tiles, and surfaces behind ceiling tiles. The results of the analysis are presented.

  11. DIII-D Dust Particulate Characterization (June 1998 Vent)

    SciTech Connect

    Carmack, William Jonathan

    1999-01-01

    Dust is a key component of fusion power device accident source term. Understanding the amount of dust expected in fusion power devices and its physical and chemical characteristics is needed to verify assumptions currently used in safety analyses. An important part of this safety research and development work is to characterize dust from existing experimental tokamaks. In this report, we present the collection, data analysis methods used, and the characterization of dust particulate collected from various locations inside the General Atomics DIII-D vacuum vessel following the June 1998 vent. The collected particulate was analyzed at the Idaho National Engineering and Environmental Laboratory (INEEL). Two methods were used to collect particulate with the goal of preserving the particle size distribution and physical characteristics of the particulate. Choice of collection technique is important because the sampling method used can bias the particle size distribution collected. Vacuum collection on substrates and adhesion removal with metallurgical replicating tape were chosen as non-intrusive sampling methods. Seventeen samples were collected including plasma facing surfaces in lower, upper, and horizontal locations, surfaces behind floor tiles, surfaces behind divertor tiles, and surfaces behind ceiling tiles. The results of the analysis are presented.

  12. 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).

  13. Upgraded divertor Thomson scattering system on DIII-D

    NASA Astrophysics Data System (ADS)

    Glass, F.; Carlstrom, T. N.; Du, D.; McLean, A. G.; Taussig, D. A.; Boivin, R. L.

    2016-11-01

    A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (Te in the range of 0.5 eV-2 keV, ne in the range of 5 × 1018-1 × 1021 m3) for both low Te in detachment and high Te measurement up beyond the separatrix.

  14. 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.

  15. Plasma disruption prediction using machine learning methods: DIII-D

    NASA Astrophysics Data System (ADS)

    Lupin-Jimenez, L.; Kolemen, E.; Eldon, D.; Eidietis, N.

    2016-10-01

    Plasma disruption prediction is becoming more important with the development of larger tokamaks, due to the larger amount of thermal and magnetic energy that can be stored. By accurately predicting an impending disruption, the disruption's impact can be mitigated or, better, prevented. Recent approaches to disruption prediction have been through implementation of machine learning methods, which characterize raw and processed diagnostic data to develop accurate prediction models. Using disruption trials from the DIII-D database, the effectiveness of different machine learning methods are characterized. Developed real time disruption prediction approaches are focused on tearing and locking modes. Machine learning methods used include random forests, multilayer perceptrons, and traditional regression analysis. The algorithms are trained with data within short time frames, and whether or not a disruption occurs within the time window after the end of the frame. Initial results from the machine learning algorithms will be presented. Work supported by US DOE under the Science Undergraduate Laboratory Internship (SULI) program, DE-FC02-04ER54698, and DE-AC02-09CH11466.

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

    NASA Astrophysics Data System (ADS)

    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.

  17. The strongest magnetic barrier in the DIII-D tokamak and comparison with the ASDEX UG

    NASA Astrophysics Data System (ADS)

    Ali, Halima; Punjabi, Alkesh

    2013-05-01

    Magnetic perturbations in tokamaks lead to the formation of magnetic islands, chaotic field lines, and the destruction of flux surfaces. Controlling or reducing transport along chaotic field lines is a key challenge in magnetically confined fusion plasmas. A local control method was proposed by Chandre et al. [Nucl. Fusion 46, 33-45 (2006)] to build barriers to magnetic field line diffusion by addition of a small second-order control term localized in the phase space to the field line Hamiltonian. Formation and existence of such magnetic barriers in Ohmically heated tokamaks (OHT), ASDEX UG and piecewise analytic DIII-D [Luxon, J.L.; Davis, L.E., Fusion Technol. 8, 441 (1985)] plasma equilibria was predicted by the authors [Ali, H.; Punjabi, A., Plasma Phys. Control. Fusion 49, 1565-1582 (2007)]. Very recently, this prediction for the DIII-D has been corroborated [Volpe, F.A., et al., Nucl. Fusion 52, 054017 (2012)] by field-line tracing calculations, using experimentally constrained Equilibrium Fit (EFIT) [Lao, et al., Nucl. Fusion 25, 1611 (1985)] DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. This second-order approach is applied to the DIII-D tokamak to build noble irrational magnetic barriers inside the chaos created by the locked resonant magnetic perturbations (RMPs) (m, n)=(3, 1)+(4, 1), with m and n the poloidal and toroidal mode numbers of the Fourier expansion of the magnetic perturbation with amplitude δ. A piecewise, analytic, accurate, axisymmetric generating function for the trajectories of magnetic field lines in the DIII-D is constructed in magnetic coordinates from the experimental EFIT Grad-Shafranov solver [Lao, L, et al., Fusion Sci. Technol. 48, 968 (2005)] for the shot 115,467 at 3000 ms in the DIII-D. A symplectic mathematical map is used to integrate field lines in the DIII-D. A numerical algorithm [Ali, H., et al., Radiat. Eff. Def. Solids Inc. Plasma Sc. Plasma Tech. 165, 83

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Upgrade of the DIII-D vacuum vessel protection system

    SciTech Connect

    Hollerbach, M.A.; Lee, R.L.; Smith, J.P.; Taylor, P.L.

    1993-10-01

    An upgrade of the General Atomics DIII-D tokamak armor protection system has been completed. The upgrade consisted of armoring the outer wall and the divertor gas baffle with monolithic graphite tiles and cleaning the existing floor, ceiling, and inner wall tiles to remove any deposited impurity layer from the tile surfaces. The new tiles replace the graphite tiles used as local armor for neutral beam shine through, three graphite poloidal back-up limiter bands, and miscellaneous Inconel protection tiles. The total number of tiles increased from 1636 to 3200 and corresponding vessel coverage from 40% to 90%. A new, graphite armored, toroidally continuous, gas baffle between the outer wall and the biased divertor ring was installed in order to accommodate the cryocondensation pump that was installed in parallel with the outer wall tiles. To eliminate a source of copper in the plasma, GRAFOIL gaskets replaced the copper felt metal gaskets previously used as a compliant heat transfer interface between the inertially cooled tiles and the vessel wall. GRAFOIL, an exfoliated, flexible graphite material from Union Carbide, Inc., was used between each tile and the vessel wall and also between each tile and its hold-down hardware. Testing was performed to determine the mechanical compliance, thermal conductance, and vacuum characteristics of the GRAFOIL material. To further decrease the quantity of high Z materials exposed to the plasma, the 1636 existing graphite tiles were identified, removed, and grit blasted to eliminate a thin layer of deposited metals which included nickel, chromium, and molybdenum. Prior to any processing, a selected set of tiles was tested for radioactivity, including tritium contamination. The tiles were grit blasted in a negative-pressure blasting cabinet using 37 {mu}m boron carbide powder as the blast media and dry nitrogen as the propellant.

  3. Measurements of helicon antenna coupling in DIII-D

    NASA Astrophysics Data System (ADS)

    Pinsker, R. I.; Moeller, C. P.; Degrassie, J. S.; Petty, C. C.; Anderson, J. P.; Torreblanca, H.; Porkolab, M.; Lau, C.; Watkins, J. G.; Zeng, L.

    2016-10-01

    In preparation for 1-MW-level helicon wave experiments, good coupling efficiency of a low-power 12-element phased-array antenna at 476 MHz has been found in several plasma regimes in the DIII-D tokamak. The antenna, a traveling-wave structure of the `comb-line' type, is designed to excite helicons (fast waves in the LHRF) at a nominal n| | of 3. This structure is a low-power prototype (operating at up to 0.4 kW) of a 30-element structure intended for operation at the 1 MW level, which is in the design stage. The dependence of wave coupling on the antenna/plasma distance was found to fit well with a simple model with one adjustable parameter in stationary regimes. In ELMing H-mode discharges that are calculated to have complete first-pass absorption of the coupled waves, strong coupling is found even between ELMs, which supports the design of the high-power antenna. To facilitate quantitative modeling, SOL density profiles were measured with a profile reflectometer, and the density adjacent to the antenna was measured with a fixed Langmuir probe. Future experiments using the high-power antenna will permit measurement of the non-inductive current drive efficiency using helicon waves in high-beta discharges. Supported by US DOE under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC04-94AL85000, DE-FG02-08ER54984.

  4. Upgrades to the C-Mod FIR Polarimeter

    NASA Astrophysics Data System (ADS)

    Abraham, Sameer; Irby, Jim; Watterson, Reich; Vieira, Rui; Leccacorvi, Rick; Parkin, William; Murray, Rick; Marmar, Earl

    2016-10-01

    The 3-Chord FIR Polarimeter presently deployed on C-Mod is capable of responding to both fast changes in the plasma equilibrium and high frequency fluctuations. Two FIR lasers locked together with a slight frequency offset provide a signal IF at 4 MHz, which allows for the fast response of the system. Recently implemented upgrades including relocation of the laser table from the C-Mod experimental cell to a more shielded location, the design and installation of a humidity controlled beam-line to convey the FIR beams across the cell, and improved collimation optics will be discussed. Results from initial testing of the system during C-Mod operation, as well as fluctuation data from the most recent and previous campaigns will be presented and compared. Supported by USDoE Award DE-FC02-99ER54512.

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

    SciTech Connect

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

    1999-08-01

    Recently a number of enhancements to the computer hardware infrastructure have been implemented at the DIII-D National Fusion Facility. Utilizing these improvements to the hardware infrastructure, software enhancements are focusing on streamlined analysis, automation, and graphical user interface (GUI) systems to enlarge the user base. The adoption of the load balancing software package LSF Suite by Platform Computing has dramatically increased the availability of CPU cycles and the efficiency of their use. Streamlined analysis has been aided by the adoption of the MDSplus system to provide a unified interface to analyzed DIII-D data. The majority of MDSplus data is made available in between pulses giving the researcher critical information before setting up the next pulse. Work on data viewing and analysis tools focuses on efficient GUI design with object-oriented programming (OOP) for maximum code flexibility. Work to enhance the computational infrastructure at DIII-D has included a significant effort to aid the remote collaborator since the DIII-D National Team consists of scientists from 9 national laboratories, 19 foreign laboratories, 16 universities, and 5 industrial partnerships. As a result of this work, DIII-D data is available on a 24 x 7 basis from a set of viewing and analysis tools that can be run either on the collaborators' or DIII-Ds computer systems. Additionally, a Web based data and code documentation system has been created to aid the novice and expert user alike.

  6. 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.

  7. Recent edge CXRS measurements in I-mode and ELMy H-mode plasmas on C-Mod

    NASA Astrophysics Data System (ADS)

    Theiler, C.; Churchill, R. M.; Hubbard, A.; Hughes, J. W.; Lipschultz, B.; Marmar, E.; Reinke, M. L.; Terry, J. L.; Walk, J. R.; Whyte, D.; Diallo, A.

    2012-10-01

    A high-resolution charge exchange recombination spectroscopy (CXRS) diagnostic is installed on Alcator C-Mod to measure edge profiles of B^5+ density, temperature, and flows and to deduce radial electric field profiles. As donor particles for CX, both hydrogen from a modulated diagnostic neutral beam and deuterium from a local gas puff are employed. While the former technique often suffers from low signal intensities near the LCFS, the latter can be complicated by contamination of the signal by molecular lines and cross-section effects. After discussing how these challenges are overcome in data analysis, we present recent measurements in regimes which have not extensively been investigated previously with edge CXRS on C-Mod. The first one is the I-mode regime, which is characterized by H-mode like energy confinement and L-mode like particle transport. Focusing on I-modes with unfavorable ion B x∇B drift, we study profile scalings with plasma parameters such as density, plasma current and heating power. The second regime is ELMy H-mode, where we explore the capabilities of the CXRS diagnostic to measure profile evolutions during and in between ELMs.

  8. 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.

  9. Outgassing tests on materials used in the DIII-D magnetic fusion tokamak

    SciTech Connect

    Holtrop, K.L.; Hansink, M.; Kellman, A.G.

    1998-12-01

    In order to achieve high performance plasma discharges in the DIII-D magnetic fusion tokamak, impurity levels must be carefully controlled. Since first wall materials can desorb volatile impurities during these discharges, it is important to characterize and control the outgassing of these materials. An outgassing chamber was built to measure the outgassing properties of various materials used in the DIII-D vessel. The results of pump-down tests performed on ATJ graphite, thin Grafoil {reg_sign} gaskets, and MgO coaxial cables will be presented. In addition to pumpdown tests it was desired to study the behavior of the materials at temperatures up to 400 C, which is the maximum temperature to which the DIII-D vessel is baked. The station was modified to include independent heating control of the sample and a simple load-lock chamber.

  10. 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.

  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. 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

    2005-05-05

    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-Schlueter 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 [1]. 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 [2]. 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 [3,4] 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 scale lengths 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. [5,6]. 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

  13. 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

  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. Upgrade Plans for the C-Mod FIR Polarimeter

    NASA Astrophysics Data System (ADS)

    Watterson, R.; Garnier, D.; Irby, J.; Brower, D. L.; Xu, P.; Bergerson, W. F.; Ding, W. X.; Guttenfelder, W.; Marmar, E. S.

    2014-10-01

    The 3-chord FIR polarimeter presently deployed on C-Mod is capable of responding to both fast changes in the plasma equilibrium and high frequency fluctuations. It operates under ITER-like plasma conditions and magnetic fields, and uses an optical layout similar to that proposed for ITER. The details of this system and some results from the C-Mod 2012 campaign will be presented, along with the design of the upgrade that is now being implemented. The new system will provide horizontal chords near the mid-plane and low loss etalon windows to improve both the signal level and our ability to study magnetic fluctuations. The laser table has been relocated from the C-Mod cell to a shielded and climate controlled location, and improvements have been made to its acoustic isolation. New collimation optics, and a beam-line needed to convey the FIR beams into the tokamak port have been designed. Improvements to the detector electronics will also be discussed, as will initial testing of the laser system and reference detectors during C-Mod operation. Supported by USDoE Award DE-FC02-99ER54512.

  16. Spatial and temporal analysis of DIII-D 3D magnetic diagnostic data

    NASA Astrophysics Data System (ADS)

    Strait, E. J.; King, J. D.; Hanson, J. M.; Logan, N. C.

    2016-11-01

    An extensive set of magnetic diagnostics in DIII-D is aimed at measuring non-axisymmetric "3D" features of tokamak plasmas, with typical amplitudes ˜10-3 to 10-5 of the total magnetic field. We describe hardware and software techniques used at DIII-D to condition the individual signals and analysis to estimate the spatial structure from an ensemble of discrete measurements. Applications of the analysis include detection of non-rotating MHD instabilities, plasma control, and validation of MHD stability and 3D equilibrium models.

  17. Spatial and temporal analysis of DIII-D 3D magnetic diagnostic data.

    PubMed

    Strait, E J; King, J D; Hanson, J M; Logan, N C

    2016-11-01

    An extensive set of magnetic diagnostics in DIII-D is aimed at measuring non-axisymmetric "3D" features of tokamak plasmas, with typical amplitudes ∼10(-3) to 10(-5) of the total magnetic field. We describe hardware and software techniques used at DIII-D to condition the individual signals and analysis to estimate the spatial structure from an ensemble of discrete measurements. Applications of the analysis include detection of non-rotating MHD instabilities, plasma control, and validation of MHD stability and 3D equilibrium models.

  18. 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.

  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. Spatial and temporal analysis of DIII-D 3D magnetic diagnostic data

    DOE PAGES

    Strait, E. J.; King, J. D.; Hanson, J. M.; ...

    2016-08-11

    An extensive set of magnetic diagnostics in DIII-D is aimed at measuring non-axisymmetric "3D" features of tokamak plasmas, with typical amplitudes ~10-3 to 10-5 of the total magnetic field. We describe hardware and software techniques used at DIII-D to condition the individual signals and analysis to estimate the spatial structure from an ensemble of discrete measurements. Lastly, applications of the analysis include detection of non-rotating MHD instabilities, plasma control, and validation of MHD stability and 3D equilibrium models.

  1. 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.

  2. Performance of the ECH Transmission Lines and Launchers in DIII-D

    SciTech Connect

    Kajiwara, K.; Mui, A.; Baxi, C.B.; Lohr, J.; Gorelov, I.A.; Green, M.T.; Ponce, D.; Callis, R.W.

    2005-09-26

    The efficiency of the transmission line for the 110 GHz ECH system was measured in DIII-D using a low power rf source. The measured efficiency was about 10% lower than expected from theoretical analysis of the components. The launcher temperature increase during rf pulses was measured and the peak mirror surface temperature was inferred from a simulation.

  3. 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.

  4. 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.

  5. New Plasma Discharge Development Tools for the DIII-D Plasma Control System

    NASA Astrophysics Data System (ADS)

    Welander, A. S.; Eidietis, N. W.; Humphreys, D. A.; Hyatt, A. W.; Leuer, J. A.; Walker, M. L.

    2011-10-01

    A new set of discharge design tools has been implemented under the GA Tokamak System Toolbox (TokSys). A new equilibrium design tool enables development of target equilibria, and upgraded simulation tools enhance testing of new control algorithms for devices that share the DIII-D Plasma Control System (including DIII-D, NSTX, EAST, KSTAR and others). Such tools will be needed for high power devices such as ITER, which require extensive commissioning of discharges to minimize disruptions and maximize the scientific return. Control verification by simulation will enable ITER to focus on exploring the unknown while minimizing risks from the known. The DIII-D simulation simserver is a comprehensive simulation of the tokamak including power supplies, conductors, plasma, diagnostics, and actuators, which can be connected to the actual control system. It has been used extensively to test implementations and study multi-algorithm integrated control performance in DIII-D and other devices. Work supported by the US DOE under DE-FC02-04ER54698.

  6. Comparison of Aspect Ratio Effects on Neoclassical Tearing Modes Between DIII-D and NSTX

    NASA Astrophysics Data System (ADS)

    La Haye, R. J.; Buttery, R. J.; Gerhardt, S. P.; Sabbagh, S. A.; Brennan, D. P.

    2010-11-01

    Experimental data is analyzed in which m/n=2/1 neoclassical tearing modes self-stabilized; this ``marginal point'' is valuable for evaluating the relative importance of the terms in the balanced Modified Rutherford equation. DIII-D and NSTX have similar cross-sectional area and shape except for the large difference in aspect ratio. The aspect ratio effects for NTMs explicitly occur in the MRE in the destabilizing helically perturbed bootstrap current term, in the small island stabilizing effects, and in the stabilizing curvature term. The marginal island width on NSTX at q95˜8 is about three times the ion banana width. This agrees with the q95˜4 data on DIII-D but is a larger multiple than for DIII-D at q95˜7. The balance in the MRE indicates that while the stabilizing effect of the curvature term in DIII-D is negligible, it is important in NSTX. The temporal behavior of the mode suggests NSTX operates closer to marginal classical tearing stability, but benefits from the stabilizing effect of curvature.

  7. 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.

  8. Robust Control of the Spatial Current Profile in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Barton, J.; Schuster, E.; Walker, M. L.; Humphreys, D. A.

    2011-10-01

    Advanced tokamak operating scenarios, characterized by large noninductively driven plasma currents, typically require active regulation of a specific current density profile. Non-model-based control of the q profile has been tested at DIII-D. However, some present limitations of the controller motivate the design of a model-based controller that accounts for the dynamics of the whole q profile in response to the control actuators. A control-oriented model of the current profile evolution in DIII-D was recently developed and used to design feedforward control schemes. In order to reject the effects of external disturbances to the system, a feedback control input needs to be added to the feedforward input. In this work, we report on the design of a robust feedback controller, on the implementation of the combined model-based feedforward + feedback controller in the DIII-D Plasma Control System, and on the experimental validation of the combined controller in the DIII-D tokamak. Supported by the NSF CAREER award program ECCS-0645086 and the US DOE under DE-FG02-09ER55064 and DE-FC02-04ER54698.

  9. Calculation of the inboard magnetic footprint in the DIII-D from the low mn magnetic perturbation

    NASA Astrophysics Data System (ADS)

    Zhao, Michael; Punjabi, Alkesh; Ali, Halima

    2010-11-01

    The equilibrium EFIT data for the DIII-D shot 115467 at 3000 ms is used to construct the equilibrium generating function for magnetic field line trajectories in the DIII-D tokamak in natural canonical coordinates [1,2]. The generating function represents the axisymmetric magnetic geometry and the topology of the DIII-D shot very accurately [1,2]. A symplectic map for field line trajectories in the natural canonical coordinates in the DIII-D is constructed. We call this map the DIII-D map. The natural canonical coordinates can be readily inverted to physical coordinates (R,φ,Z) [1,2]. Low mn magnetic perturbation with mode numbers (m,n)=(1,1)+(1,-1) is added to the generating function of the map. The amplitude for the low mn perturbation is chosen to be 6X10-4, which is the expected value of the amplitude in tokamaks. The forward DIII-D map and its continuous analog are used to calculate the inboard magnetic footprint from the low mn perturbation in the DIII-D. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

  10. Calculation of the outboard magnetic footprint in the DIII-D from the low mn magnetic perturbation

    NASA Astrophysics Data System (ADS)

    Robinson, Quentin; Ali, Halima; Punjabi, Alkesh

    2010-11-01

    The equilibrium EFIT data for the DIII-D shot 115467 at 3000 ms is used to construct the equilibrium generating function for magnetic field line trajectories in the DIII-D tokamak in natural canonical coordinates [1,2]. The generating function represents the axisymmetric magnetic geometry and the topology of the DIII-D shot very accurately [1,2]. A symplectic map for field line trajectories in the natural canonical coordinates in the DIII-D is constructed. We call this map the DIII-D map. The natural canonical coordinates can be readily inverted to physical coordinates (R,φ,Z) [1,2]. Low mn magnetic perturbation with mode numbers (m,n)=(1,1)+(1,-1) is added to the generating function of the map. The amplitude for the low mn perturbation is chosen to be 6X10-4, which is the expected value of the amplitude in tokamaks. The backward DIII-D map and its continuous analog are used to calculate the outboard magnetic footprint from the low mn perturbation in the DIII-D. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

  11. 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.

  12. 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.

  13. 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.

  14. Faraday-effect polarimeter diagnostic for internal magnetic field fluctuation measurements in DIII-D

    NASA Astrophysics Data System (ADS)

    Chen, J.; Ding, W. X.; Brower, D. L.; Finkenthal, D.; Muscatello, C.; Taussig, D.; Boivin, R.

    2016-11-01

    Motivated by the need to measure fast equilibrium temporal dynamics, non-axisymmetric structures, and core magnetic fluctuations (coherent and broadband), a three-chord Faraday-effect polarimeter-interferometer system with fast time response and high phase resolution has recently been installed on the DIII-D tokamak. A novel detection scheme utilizing two probe beams and two detectors for each chord results in reduced phase noise and increased time response [δb ˜ 1G with up to 3 MHz bandwidth]. First measurement results were obtained during the recent DIII-D experimental campaign. Simultaneous Faraday and density measurements have been successfully demonstrated and high-frequency, up to 100 kHz, Faraday-effect perturbations have been observed. Preliminary comparisons with EFIT are used to validate diagnostic performance. Principle of the diagnostic and first experimental results is presented.

  15. Faraday-effect polarimeter diagnostic for internal magnetic field fluctuation measurements in DIII-D.

    PubMed

    Chen, J; Ding, W X; Brower, D L; Finkenthal, D; Muscatello, C; Taussig, D; Boivin, R

    2016-11-01

    Motivated by the need to measure fast equilibrium temporal dynamics, non-axisymmetric structures, and core magnetic fluctuations (coherent and broadband), a three-chord Faraday-effect polarimeter-interferometer system with fast time response and high phase resolution has recently been installed on the DIII-D tokamak. A novel detection scheme utilizing two probe beams and two detectors for each chord results in reduced phase noise and increased time response [δb ∼ 1G with up to 3 MHz bandwidth]. First measurement results were obtained during the recent DIII-D experimental campaign. Simultaneous Faraday and density measurements have been successfully demonstrated and high-frequency, up to 100 kHz, Faraday-effect perturbations have been observed. Preliminary comparisons with EFIT are used to validate diagnostic performance. Principle of the diagnostic and first experimental results is presented.

  16. 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.

  17. 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.

  18. Power Accountability with 1 MW 110 GHz Gyrotron System on the DIII--D Tokamak

    NASA Astrophysics Data System (ADS)

    Zhang, Daqing; Lohr, John; Tooker, J. W.; Ponce, Dan; Callis, R. W.

    1996-11-01

    A new gyrotron system (110 GHz, 1 MW, 2 sec) has been built for the DIII--D tokamak. Before the ECH physics experiments were carried out on DIII--D, the power produced by the gyrotron was measured carefully as well as the power absorbed by parts of the system such as in the gyrotron window, mirror optics unit, miter bends, and dummy loads, for different pulse durations. The maximum output power achieved up to now is 885 kW for 500 msec pulse length. The gyrotron generation efficiency is 37%, very close to the optimum operation parameters, and the whole system's efficiency is 29%. The output mode and frequency of the gyrotron was also measured. Detailed experimental results will be presented.

  19. Gyrokinetic Simulations of Electrostatic Turbulence Reduction due to Radial Electric Field Shear in DIII-D

    NASA Astrophysics Data System (ADS)

    Taimourzadeh, Sam; Holod, Ihor; Lin, Zhihong; Nazikian, Raffi

    2015-11-01

    It has been demonstrated that edge localized modes (ELMs) can be fully suppressed in DIII-D H-mode plasmas with the application of resonant magnetic perturbations (RMPs), and that there is a corresponding reduction of pedestal gradients, changes in rotation, and changes in the radial electric field (Er) profile. However, with the application of RMPs there is also an increase in short wavelength, electrostatic turbulence on top of the pedestal, as observed with BES, DBS, and other fluctuation diagnostics. The effects of Er shear on this turbulence, is investigated using gyrokinetic simulations via the gyrokinetic toroidal code (GTC) for in DIII-D shot 158103, at times 3750 ms (RMP on, ELM active) and 3050 ms (RMP on, ELM suppressed).

  20. Gyrokinetic Simulations of RMP Effects on DIII-D Edge Turbulence

    NASA Astrophysics Data System (ADS)

    Taimourzadeh, Sam; Holod, Ihor; Lin, Zhihong; Nazikian, Raffi; Wingen, Andreas

    2016-10-01

    It has been demonstrated that edge localized modes (ELMs) can be fully suppressed in DIII-D H-mode plasmas with the application of resonant magnetic perturbations (RMPs), and that there is a corresponding reduction of pedestal gradients, changes in rotation, and changes in the radial electric field (Er) profile. However, with the application of RMPs there is also an increase in long wavelength, electrostatic turbulence on top of the pedestal, as observed with BES, DBS, and other fluctuation diagnostics. Using the Gyrokinetic Toroidal Code (GTC), DIII-D shots 158103, at times 3750 ms (ELMing w/ RMP) and 3050 ms (ELM suppressed w/ RMP), and shot 158104.1350 (ELMing) are investigated, and a link between increased pedestal top turbulence, during the ELM suppressed phase, and a shift in the Er profile is demonstrated.

  1. Study on Sawtooth and ELM activities in DIII-D and KSTAR Plasmas

    NASA Astrophysics Data System (ADS)

    Bak, J.-G.; Kim, H. S.; Lee, S. G.; Lee, K. D.; Ko, W. H.; Kim, J.; Jeon, Y. M.; Kim, W. C.; Bae, Y. S.; Strait, E. J.; La Haye, R. J.; Buttery, R. J.; Wade, M. R.; Park, J. K.; Hanson, J. M.

    2012-10-01

    Sawtooth precursor oscillations (SPOs) are studied in neutral beam heated plasmas on DIII-D and KSTAR. The characteristics of the SPO (5-20 kHz, m/n=1/1) are investigated using magnetic sensors along with electron cyclotron emission (ECE) and soft x-ray diagnostics. In addition, the Type I edge localized mode (ELM) precursors (8-40 kHz, n=2,3) are detected before the ELM burst in neutral beam heated plasmas. The characteristics of the ELM precursors are investigated by using magnetic sensor data. In this work, the experimental investigations of the SPOs and ELM precursors in DIII-D and KSTAR plasmas will be presented.

  2. 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.

  3. Exposures of tungsten nanostructures to divertor plasmas in DIII-D

    SciTech Connect

    Rudakov, D. L.; Wong, C. P. C.; Doerner, R. P.; Wright, G. M.; Abrams, T.; Baldwin, M. J.; Boedo, J. A.; Briesemeister, A. R.; Chrobak, C. P.; Guo, H. Y.; Hollmann, E. M.; McLean, A. G.; Fenstermacher, M. E.; Lasnier, C. J.; Leonard, A. W.; Moyer, R. A.; Pace, D. C.; Thomas, D. M.; Watkins, J. G.

    2016-01-22

    Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with little obvious damage except in the areas where unipolar arcing occurred. In conclusion, arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.

  4. 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.

  5. 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.

  6. TSC plasma halo simulation of a DIII-D vertical displacement episode

    SciTech Connect

    Sayer, R.O.; Peng, Y.K.M.; Jardin, S.C.; Kellman, A.G.; Wesley, J.C.

    1993-01-01

    A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench, and current quench. Inclusion of a 1-to 4-eV halo surrounding the main plasma was found to be necessary to match simulation and experimental results for plasma current decay, trajectory, toroidal and poloidal vessel currents, and magnetic probe and flux loop values for the entire VDE.

  7. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

    SciTech Connect

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

    2016-09-14

    A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

  8. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

    NASA Astrophysics Data System (ADS)

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

    2016-12-01

    A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

  9. 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.

  10. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

    DOE PAGES

    Guo, H. Y.; Hill, D. N.; Leonard, A. W.; ...

    2016-09-14

    A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, whichmore » we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less

  11. Equilibrium reconstructions using imaging MSE on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Victor, B. S.; Holcomb, C. T.; Allen, S. L.; Thorman, A.; Howard, J.

    2016-10-01

    For the first time, an imaging motional Stark effect (IMSE) diagnostic has been successfully implemented on DIII-D. This poster presents initial results on incorporating these measurements into EFIT equilibrium reconstructions. IMSE provides a 2D image of the polarization angle as a function of major radius and vertical position. The single, wider bandpass filter used with IMSE allows for polarization angle measurements across multiple neutral beams and with variable beam voltage. The traditional MSE systems on DIII-D only work with one beam at a constant voltage. In addition to 2D data with improved spatial resolution, IMSE has provided the lowest major radius internal plasma magnetic measurements on DIII-D, producing new equilibrium constraints. Advanced tokamak scenarios with qmin > 1.5 are studied in detail with the new constraints provided by the IMSE system. The additional constraints are being used to assess ideal, resistive, and energetic particle stability in these scenarios. Supported by US DOE under DE-AC52-07NA27344, DE-FC02-04ER54698, and DE-FG02-04ER54761.

  12. 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.

  13. Backstepping Control of the Current Profile in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Boyer, M. D.; Barton, J.; Schuster, E.; Walker, M. L.; Humphreys, D. A.

    2011-10-01

    Control of the spatial profile of the plasma current in tokamaks has been demonstrated to be a key condition for advanced scenarios with improved confinement and steady-state operation. Non-model-based controllers tested at DIII-D have shown limitations, motivating the design of model-based controllers that account for the dynamics of the q profile. In this work, we utilize a control-oriented model of the current profile evolution in DIII-D to design a backstepping boundary control law for regulating the current profile around a desired feed-forward trajectory. The control scheme makes use of the total plasma current, total power, and line averaged density as actuators. A simulation study is done to test the control law against uncertainties in the model parameters and initial conditions, as well as input disturbances. Finally, the implementation of the controller in the DIII-D plasma control system is discussed and experimental results are presented. Supported by the NSF CAREER award program ECCS-0645086 and the US DOE under DE-FG02-09ER55064 and DE-FC02-04ER54698.

  14. Analysis and Simulation of ITER Steady-State Discharges on DIII-D

    NASA Astrophysics Data System (ADS)

    Diem, S. J.; Murakami, M.; Park, J. M.; Sontag, A. C.

    2013-10-01

    One of the primary goals of the ITER project is to demonstrate a reactor scale steady-state operation for future tokamaks. This is a challenging task which requires simultaneous operation with fully noninductive current drive, a fusion gain of Q >= 5 and IBS for discharges approximately 3000s in length. Previously, DIII-D has demonstrated fully noninductive scenario in ITER-like shaped plasmas at relatively high q95 ~ 6 . 5 and moderate βN ~ 3 but with low fusion performance (G =βNH89 /q952 ~ 0 . 15). Recent high qmin experiment and modeling indicate that the goal of G = 0 . 3 predicted for Q = 5 operation on ITER can be achieved noninductively at reduced q95 and at higher βN. An optimum choice of q95 and βN for the ITER steady-state mission will be discussed based on the experimental scaling from ITER demonstration discharges on DIII-D, as well as predictive FASTRAN scenario modeling using TGLF coupled to the Integrated Plasma Simulator. FASTRAN is a new iterative numerical procedure that integrates a variety of models (transport, heating, CD, equilibrium and stability) and has been shown to reproduce most features of DIII-D high beta discharges with a stationary current profile. ORNL is managed by UT-Battelle, LLC for the US DOE under DE-AC02-05ER22725 and DE-FC02-04ER54698.

  15. 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.

  16. Influence of the Resonant Magnetic Perturbation on the Plasma Boundary in DIII-D

    SciTech Connect

    Jakubowski, M. W.; Evans, T.E.; Fenstermacher, M. E.; Lasnier, C. J.; Schmitz, O.; Wolf, R. C.; Baylor, Larry R; Boedo, J.A.; Burrell, K. H.; Frerichs, H.; DeGrassie, J. S.; Gohil, P.; Mordijck, S.; Moyer, R.A.; Leonard, A. W.; Reiter, D.; Samm, U.; Schaffer, M. J.; Osborne, T. H.; Unterberg, Ezekial A; Watkins, J. G.

    2010-01-01

    Stochastic boundaries in fusion devices have been investigated in tokamaks, stellarators and reversed field pinch experiments for many years. However, since edge localized modes (ELMs) have been successfully eliminated in H-mode plasmas at the DIII-D tokamak [1,2] with small, edge resonant magnetic perturbations, they have become a widely investigated topic in tokamaks. In DIII-D stochastic boundaries are produced by coils external to the plasma. The magnetic field there consists of field lines with very different connection lengths, which produces a three dimensional, heterogenous structure of stochastic volume. The most obvious manifestation of the perturbed plasma edge is the strike line splitting observable in heat and particle fluxes, which changes with collisionality. The interaction of the magnetic perturbation and the magnetic equilibrium is of a resonant nature and the structure of the stochastic volume is a strong function of q(95). This is observed as a modulation of e.g. electron temperature as measured by ECE or Thomson scattering. In this work we summarize recent experimental findings on properties of the stochastic boundary in DIII-D. (C) 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

  17. Measuring the effect of divertor closure on detachment in DIII-D

    NASA Astrophysics Data System (ADS)

    Moser, Auna; Leonard, A. W.; Petrie, T. W.; Sang, C. F.; Allen, S. L.; McLean, A. G.; Fenstermacher, M. E.; Joseph, I.; Lasnier, C. J.; Makowski, M. A.; Watkins, J. G.; Briesemeister, A. R.

    2015-11-01

    Recent experiments compared the open lower divertor and semi-closed upper divertor in DIII-D to measure the effect of divertor closure on detachment onset and heat flux control, extending past work showing reduced core fueling with the more-closed upper DIII-D divertor. Experiments were performed to determine the extent to which closure may facilitate detachment at collisionalities more relevant to future devices. This work builds on previous experiments that quantified effects of divertor magnetic geometry, including connection length, ∇B-drift direction, incidence angle, and flux expansion; efforts were made to match these parameters while comparing single null configurations in the upper and lower divertor in order to isolate the effects of closure. Experimental measurements coupled with simulation results will help weigh the benefits of a more-closed divertor in facilitating detachment and reducing heat flux against the constraints imposed on the magnetic geometry by a more-closed divertor tile structure, aiding in the design of a future advanced divertor for DIII-D. Supported by the US Department of Energy under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, and DE-AC05-00OR22725.

  18. High Harmonic Fast Wave Damping on an Ion Beam: NSTX and DIII-D Regimes Compared

    NASA Astrophysics Data System (ADS)

    Pinsker, R. I.; Choi, C. C.; Petty, C. C.; Porkolab, M.; Wilson, J. R.; Murakami, M.; Harvey, R. W.

    2004-11-01

    Both NSTX and DIII-D use the combination of fast Alfven waves (FW) and neutral beam injection (NBI) for central electron heating and current drive. Damping of the fast wave on the beam ions at moderate to high harmonics (4th--20th) of the beam ion cyclotron frequency represents a loss process. In DIII-D current drive experiments at low density in which 4th and 8th harmonics were compared, damping at the 8th harmonic damping was much weaker than at the 4th [1]. However, recent simulations have predicted that in higher density and higher beam power regimes (of interest to the Advanced Tokamak program) the beam ion absorption will transition to the unmagnetized ion regime, where the damping is significant and essentially independent of harmonic number. In the present work, the transition from magnetized to unmagnetized ion regimes for the NSTX and DIII-D HHFW experiments is studied theoretically, with a combination of simple semi-analytic models and numerical models. \\vspace0.25 em [1] C.C. Petty, et al., Plasma Phys. and Contr. Fusion 43, 1747 (2001).

  19. Confinement Optimization by Controlling q-shear in DIII-D Steady-State Discharges

    NASA Astrophysics Data System (ADS)

    Murakami, M.; Park, J. M.; Yoshida, M.; Grierson, B. A.; McKee, G. R.; Holcomb, C. T.; Meneghini, O.; Petty, C. C.

    2016-10-01

    Recent experiment in DIII-D to study effects of varying shear of safety factor (q) on transport and fluctuations showed that negative shear (NS) discharges formed an internal transport barrier inside ρ = 0.5, resulting in substantial improvement in ion confinement in the core, and reduced amplitudes of broad turbulent (low- to mid-k) spectra compared with those for the positive shear (PS) discharges. Previous DIII-D experiments under similar conditions but with higher on-axis NBI powers and reversed-BT direction (▽B drift toward upper single null divertor rather than lower single null divertor), showed longer confinement improvement periods with broad NS q-profile. Measured transport characteristics are analyzed in both experiments and compared with predictions of theory-based transport models using the integrated plasma simulation frameworks (OMFIT/IPS). Based on such validated transport models, predictions of prolonged confinement improvement with better sustained ITB using higher powers of off-axis NBI and ECCD available in DIII-D will be discussed. Work supported by the US DOE under DE-AC05-00OR22725 and DE-FC02-04ER54698.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Particle Dynamics in the Magnet Region of the DIII-D Neutral Beam System

    NASA Astrophysics Data System (ADS)

    Blackford, C. A.; Crowley, B. J.; Rauch, J. M.; Scoville, J. T.

    2016-10-01

    The Neutral Beam system on the DIII-D tokamak consists of eight ion sources on four beam lines using the Common Long Pulse Source (CLPS) developed at Berkeley in the 1980s. This poster presents the results of modeling efforts aimed to understand the anomalous power deposition in the bending magnet region of the neutral beam system at DIII-D. The code tracks individual particles in 3D electric and magnetic fields. In these fields, the particles intercept solid boundaries and deposit power on the magnet pole shields as well as various collimators, the ion dump, and the beam dump calorimeter. This code allows investigation of phenomena including the non-uniformity of the magnetic field, the space charge effects on the neutral beam, and deviant ion trajectories within the system. These phenomena are expected to contribute largely to the power deposition within the bending magnet region. Results of the analysis may lead to adjustments that could increase the efficiency of the neutral beam system at the DIII-D facility. Work supported by U.S. DOE under DE-FC02-04ER54698.

  7. Summary of RF power measurements on the ECH System on DIII-D

    NASA Astrophysics Data System (ADS)

    Cengher, M.; Lohr, J.; Doane, J. L.; Gorelov, Y. A.; Moeller, C. P.; Deboo, J. C.; Leonard, A. W.; Ponce, D.

    2009-11-01

    For the six 110 GHz, 1 MW class gyrotrons of the ECH system on DIII-D at GA, the power injected at the tokamak by the electron cyclotron heating (ECH) system is measured for every pulse using calorimetry. The relationship between generated rf power and component cooling was determined originally by the manufacturer with 1% total power accountability. Transmission line efficiencies have been measured directly at DIII-D for high power. The injected power calculations resulting from combining the generated power measurements corrected for transmission loss have been verified using synchronous detection of ECE signals of plasma heating using modulated rf injection. Direct measurements of the injected power at the tokamak for pulses up to 5 s in length are being developed. The result from the following techniques will be described: measurements of the response of the DIII-D diagnostic bolometers to rf injection, heating of a leaky waveguide gap monitor, detailed rf power accounting in dummy loads and waveguides, and rf pickoff in a 4 port miter.

  8. Finding evidence for density fluctuation effects on electron cyclotron heating deposition profiles on DIII-D

    NASA Astrophysics Data System (ADS)

    Brookman, M. W.; Austin, M. E.; Petty, C. C.

    2015-12-01

    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 Te 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.

  9. 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.

  10. Far-infrared polarimetry diagnostic for measurement of internal magnetic field dynamics and fluctuations in the C-MOD Tokamak (invited).

    PubMed

    Bergerson, W F; Xu, P; Irby, J H; Brower, D L; Ding, W X; Marmar, E S

    2012-10-01

    A laser-based (2.55 THz) mulitchord polarimeter is now operational on Alcator C-Mod and is used to make measurements of the internal magnetic field structure as well as plasma fluctuations. The polarimeter is designed to measure the Faraday effect for high-field (up to 8.3 T) and high-density (up to 5 × 10(20) m(-3)) ITER relevant plasma conditions. Initial 3 chord tests are consistent with magnetic equilibrium reconstructions and indicate no measurable contamination from the toroidal magnetic field due to the Cotton-Mouton effect or misalignment. Time response of <1 μs enables the measurement of fast equilibrium temporal dynamics as well as high-frequency fluctuations.

  11. Far-infrared polarimetry diagnostic for measurement of internal magnetic field dynamics and fluctuations in the C-MOD Tokamak (invited)a)

    NASA Astrophysics Data System (ADS)

    Bergerson, W. F.; Xu, P.; Irby, J. H.; Brower, D. L.; Ding, W. X.; Marmar, E. S.

    2012-10-01

    A laser-based (2.55 THz) mulitchord polarimeter is now operational on Alcator C-Mod and is used to make measurements of the internal magnetic field structure as well as plasma fluctuations. The polarimeter is designed to measure the Faraday effect for high-field (up to 8.3 T) and high-density (up to 5 × 1020 m-3) ITER relevant plasma conditions. Initial 3 chord tests are consistent with magnetic equilibrium reconstructions and indicate no measurable contamination from the toroidal magnetic field due to the Cotton-Mouton effect or misalignment. Time response of <1 μs enables the measurement of fast equilibrium temporal dynamics as well as high-frequency fluctuations.

  12. An area-preserving mapping in natural canonical coordinates for magnetic field line trajectories in the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Punjabi, Alkesh

    2009-11-01

    The new approach of integrating magnetic field line trajectories in natural canonical coordinates (Punjabi and Ali 2008 Phys. Plasmas 15 122502) in divertor tokamaks is used for the DIII-D tokamak (Luxon and Davis1985 Fusion Technol. 8 441). The equilibrium EFIT data (Evans et al 2004 Phys. Rev. Lett. 92 235003, Lao et al 2005 Fusion Sci. Technol. 48 968) for the DIII-D tokamak shot 115467 at 3000 ms is used to construct the equilibrium generating function (EGF) for the DIII-D in natural canonical coordinates. The EGF gives quite an accurate representation of the closed and open equilibrium magnetic surfaces near the separatrix, the separatrix, the position of the X-point and the poloidal magnetic flux inside the ideal separatrix in the DIII-D. The equilibrium safety factor q from the EGF is somewhat smaller than the DIII-D EFIT q profile. The equilibrium safety factor is calculated from EGF as described in the previous paper (Punjabi and Ali 2008 Phys. Plasmas 15 122502). Here the safety factor for the open surfaces in the DIII-D is calculated. A canonical transformation is used to construct a symplectic mapping for magnetic field line trajectories in the DIII-D in natural canonical coordinates. The map is explored in more detail in this work, and is used to calculate field line trajectories in the DIII-D tokamak. The continuous analogue of the map does not distort the DIII-D magnetic surfaces in different toroidal planes between successive iterations of the map. The map parameter k can represent effects of magnetic asymmetries in the DIII-D. These effects in the DIII-D are illustrated. The DIII-D map is then used to calculate stochastic broadening of the ideal separatrix from the topological noise and field errors, the low mn, the high mn and peeling-ballooning magnetic perturbations in the DIII-D. The width of the stochastic layer scales as 1/2 power of amplitude with a maximum deviation of 6% from the Boozer-Rechester scaling (Boozer and Rechester 1978 Phys

  13. 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.

  14. 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.

  15. 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.

  16. Recent Results from the C-Mod Polarimeter

    NASA Astrophysics Data System (ADS)

    Xu, P.; Irby, J. H.; Bergerson, W. F.; Brower, D. L.; Ding, W. X.; Marmar, E. S.; Delgado-Aparicio, L.; Ernst, D.; Hughes, J. W.; Mumgaard, R.; Parker, R.; Scott, S.; Shiraiwa, S.; Wallace, G. M.; White, A. E.; Wolfe, S. M.

    2013-10-01

    The C-Mod 3 chord FIR polarimeter, with a 2 MHz bandwidth, is capable of responding to both fast changes in the plasma equilibrium and high frequency fluctuations. It operates under ITER-like plasma conditions and magnetic fields, and uses an optical layout and FIR sources very similar to those proposed for the ITER polarimeter. Results from the polarimeter as a function of plasma density and current will be discussed, as well as the effects of lower hybrid power levels, phasing and plasma density on the current drive efficiency. The possible identification of some broadband fluctuations as primarily magnetic in nature, and gyrokinetic simulation results from the modeling of these fluctuations will also be presented. Estimates of the localization of this mode will be described. The polarimeter response to low frequency MHD modes will be compared with results from the Fast Two-Color-Interferometer. USDoE contract DE-FC02-99ER54512.

  17. 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.

  18. 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.

  19. MHD modeling of DIII-D QH-mode discharges and comparison to observations

    NASA Astrophysics Data System (ADS)

    King, Jacob

    2016-10-01

    MHD modeling of DIII-D QH-mode discharges and comparison to observations Nonlinear NIMROD simulations, initialized from a reconstruction of a DIII-D QH-mode discharge with broadband MHD, saturate into a turbulent state, but do not saturate when flow is not included. This is consistent with the experimental results of the quiescent regime observed on DIII-D with broadband MHD activity [Garofalo et al., PoP (2015) and refs. within]. These ELM-free discharges have the normalized pedestal-plasma confinement necessary for burning-plasma operation on ITER. Relative to QH-mode operation with more coherent MHD activity, operation with broadband MHD tends to occur at higher densities and lower rotation and thus may be more relevant to ITER. The nonlinear NIMROD simulations require highly accurate equilibrium reconstructions. Our equilibrium reconstructions include the scrape-off-layer profiles and the measured toroidal and poloidal rotation profiles. The simulation develops into a saturated turbulent state and the n=1 and 2 modes become dominant through an inverse cascade. Each toroidal mode in the range of n=1-5 is dominant at a different time. The perturbations are advected and sheared apart in the counter-clockwise direction consistent with the direction of the poloidal flow inside the LCFS. Work towards validation through comparison to magnetic coil and Doppler reflectometry measurements is presented. Consistent with experimental observations during QH-mode, the simulated state leads to large particle transport relative to the thermal transport. Analysis shows that the phase of the density and temperature perturbations differ resulting in greater convective particle transport relative to the convective thermal transport. This work supported by the U.S. Department of Energy Office of Science and the SciDAC Center for Extended MHD Modeling under Contract Numbers DE-FC02-06ER54875, DE-FC02-08ER54972 and DE-FC02-04ER54698.

  20. 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).

  1. ITER test blanket module error field simulation experiments at DIII-D

    NASA Astrophysics Data System (ADS)

    Schaffer, M. J.; Snipes, J. A.; Gohil, P.; de Vries, P.; Evans, T. E.; Fenstermacher, M. E.; Gao, X.; Garofalo, A. M.; Gates, D. A.; Greenfield, C. M.; Heidbrink, W. W.; Kramer, G. J.; La Haye, R. J.; Liu, S.; Loarte, A.; Nave, M. F. F.; Osborne, T. H.; Oyama, N.; Park, J.-K.; Ramasubramanian, N.; Reimerdes, H.; Saibene, G.; Salmi, A.; Shinohara, K.; Spong, D. A.; Solomon, W. M.; Tala, T.; Zhu, Y. B.; Boedo, J. A.; Chuyanov, V.; Doyle, E. J.; Jakubowski, M.; Jhang, H.; Nazikian, R. M.; Pustovitov, V. D.; Schmitz, O.; Srinivasan, R.; Taylor, T. S.; Wade, M. R.; You, K.-I.; Zeng, L.; DIII-D Team

    2011-10-01

    Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δv/v ~ 60% via non-resonant braking. Changes to global Δn/n, Δβ/β and ΔH98/H98 were ~3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

  2. 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.

  3. 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.

  4. ITER Test Blanket Module Error Field Simulation Experiments at DIII-D

    SciTech Connect

    Schaffer, M. J.; Testa, D.; Snipes, J. A.; Gohil, P.; De Vries, P.; Evans, T. E.; Fenstermacher, M. E.; Gao, X.; Garofalo, A.; Gates, D.A.; Greenfield, C. M.; Heidbrink, W.; La Haye, R.; Liu, S.; Loarte, A.; Nave, M. F. F.; Oyama, N.; Osakabe, M.; Park, J. K.; Ramasubramanian, N.; Reimerdes, H.; Saibene, G.; Saimi, A.; Shinohara, K.; Spong, Donald A; Solomon, W. M.; Tala, T.; Zhu, Y. B.; Zhai, K.; Boedo, J.; Chuyanov, V.; Doyle, E. J.; Jakubowski, M. W.; Jhang, H.; Nazikian, Raffi; Pustovitov, V. D.; Schmitz, O.; Sanchez, Raul; Srinivasan, R.; Taylor, T. S.; Wade, M.; You, K. I.; Zeng, L.

    2011-01-01

    Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Delta upsilon/upsilon similar to 60% via non-resonant braking. Changes to global Delta n/n, Delta beta/beta and Delta H(98)/H(98) were similar to 3 times smaller. These effects are stronger at higher beta. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L-and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

  5. Thermal ion orbit loss and radial electric field in DIII-D

    DOE PAGES

    deGrassie, John S.; Boedo, Jose A.; Grierson, Brian A.

    2015-08-14

    A relatively simple model for the generation of the radial electric field, Er, near the outboard boundary in a tokamak is presented. The model posits that Er is established to supply the return current necessary to balance the thermal ion orbit loss current. Comparison with DIII-D data is promising. Features of the model that promote a more negative edge Er are higher ion temperature, lower density, lower impurity ion content, and a shorter pathlength for orbit loss. Lastly, these scalings are consistent with experimentally established access to the high-confinement mode edge transport barrier.

  6. Microturbulence in DIII-D tokamak pedestal. IV. Electrostatic turbulent transport

    NASA Astrophysics Data System (ADS)

    Liao, X.; Lin, Z.; Holod, I.; Li, B.; Sun, G. Y.

    2016-12-01

    Gyrokinetic simulations of electrostatic microturbulence in the edge plasmas of DIII-D shot 131997 find that the geodesic acoustic mode (GAM) is generated after nonlinear saturation both at the pedestal top and in the peak gradient region, and in turn, regulates the turbulence in both regions. Collisions significantly reduce the GAM amplitude and the associated GAM modulation of the turbulence in both regions. Collisions have little effects on the ion temperature gradient turbulent transport level at the pedestal top. On the other hand, collisional damping of the TEM significantly reduces the turbulent transport level in the peak gradient region.

  7. Asymmetries in the motional Stark effect emission 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.; Thorman, A.

    2016-11-01

    Spectrometer measurements and filter upgrades to a motional Stark effect polarimeter measuring the outer half-radius of the DIII-D tokamak helped to identify asymmetries in the polarization angle of Stark-split emission. The measured polarization angle of the π components differs and is not orthogonal to the σ component. These differences persist over a range of densities and with low levels of background light. It is suggested that the difference in the polarization angle between components is from a change in the ellipticity of the emitted light across the Stark components coupled with imperfect polarization preservation from an in-vessel mirror.

  8. Embedded calibration system for the DIII-D Langmuir probe analog fiber optic links

    SciTech Connect

    Watkins, J. G.; Rajpal, R.; Mandaliya, H.; Watkins, M.; Boivin, R. L.

    2012-10-15

    This paper describes a generally applicable technique for simultaneously measuring offset and gain of 64 analog fiber optic data links used for the DIII-D fixed Langmuir probes by embedding a reference voltage waveform in the optical transmitted signal before every tokamak shot. The calibrated data channels allow calibration of the power supply control fiber optic links as well. The array of fiber optic links and the embedded calibration system described here makes possible the use of superior modern data acquisition electronics in the control room.

  9. 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.

  10. 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.

  11. 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.

  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. Enhanced confinement regimes and control technology in the DIII-D tokamak

    SciTech Connect

    Lohr, J.; Burrell, K.H.; Coda, S.

    1993-07-01

    Advanced tokamak performance has been demonstrated in the DIII-D tokamak in a series of experiments which brought together developments in technology and improved understanding of the physical principles underlying tokamak operation. The achievement of greatly improved confinement coupled with development of new systems for real time plasma control have permitted investigation of the heretofore hidden or poorly controlled variables which together determine global confinement. These experiments, which included work in transport and control of the plasma boundary, point toward development of operationally and economically attractive reactors based on the tokamak. Some of these experiments are described.

  14. Microturbulence in DIII-D tokamak pedestal. III. Effects of collisions

    NASA Astrophysics Data System (ADS)

    Liao, X.; Lin, Z.; Holod, I.; Xiao, Y.; Li, B.; Snyder, P. B.

    2016-12-01

    Gyrokinetic simulations of the H-mode pedestal in DIII-D discharge 145701 find that the kinetic ballooning mode (KBM) is the most unstable mode for low toroidal numbers (n ≤ 25) and that the trapped electron mode (TEM) dominates over the KBM at higher toroidal mode numbers for realistic pressure gradients in the pedestal. Collisions reduce the TEM growth rate but have little effects on the KBM. KBM has the conventional ballooning mode structure peaking at the outer mid-plane, while TEM has an unconventional mode structure peaking at the top and bottom of the poloidal plane.

  15. 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.

  16. Asymmetries in the motional Stark effect emission on the DIII-D tokamak.

    PubMed

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

    2016-11-01

    Spectrometer measurements and filter upgrades to a motional Stark effect polarimeter measuring the outer half-radius of the DIII-D tokamak helped to identify asymmetries in the polarization angle of Stark-split emission. The measured polarization angle of the π components differs and is not orthogonal to the σ component. These differences persist over a range of densities and with low levels of background light. It is suggested that the difference in the polarization angle between components is from a change in the ellipticity of the emitted light across the Stark components coupled with imperfect polarization preservation from an in-vessel mirror.

  17. Embedded calibration system for the DIII-D Langmuir probe analog fiber optic links.

    PubMed

    Watkins, J G; Rajpal, R; Mandaliya, H; Watkins, M; Boivin, R L

    2012-10-01

    This paper describes a generally applicable technique for simultaneously measuring offset and gain of 64 analog fiber optic data links used for the DIII-D fixed Langmuir probes by embedding a reference voltage waveform in the optical transmitted signal before every tokamak shot. The calibrated data channels allow calibration of the power supply control fiber optic links as well. The array of fiber optic links and the embedded calibration system described here makes possible the use of superior modern data acquisition electronics in the control room.

  18. Commissioning of electron cyclotron emission imaging instrument on the DIII-D tokamak and first data

    SciTech Connect

    Tobias, B.; Domier, C. W.; Liang, T.; Kong, X.; Yu, L.; Luhmann, N. C. Jr.; Yun, G. S.; Park, H. K.; Classen, I. G. J; Boom, J. E.; Donne, A. J. H.; Munsat, T.; Nazikian, R.; Van Zeeland, M.; Boivin, R. L.

    2010-10-15

    A new electron cyclotron emission imaging diagnostic has been commissioned on the DIII-D tokamak. Dual detector arrays provide simultaneous two-dimensional images of T{sub e} fluctuations over radially distinct and reconfigurable regions, each with both vertical and radial zoom capability. A total of 320 (20 verticalx16 radial) channels are available. First data from this diagnostic demonstrate the acquisition of coherent electron temperature fluctuations as low as 0.1% with excellent clarity and spatial resolution. Details of the diagnostic features and capabilities are presented.

  19. Comparison of Moderate to High Ion Cyclotron Absorption on Energetic Ions in NSTX and DIII-D

    NASA Astrophysics Data System (ADS)

    Burby, J.; Pinsker, R. I.; Choi, M.

    2009-11-01

    Strong absorption of fast waves (FWs) on injected deuterons at ion cyclotron harmonic numbers in the 4-10 range is observed on both DIII-D and NSTX. The results from fast ion Dα spectroscopic measurements from the two devices differ significantly: deposition on fast ions peaks near the cyclotron harmonic layer closest to the magnetic axis in the conventional-aspect-ratio DIII-D, while results from the low-aspect-ratio NSTX show a broader deposition profile [1]. One root of the difference stems from the absorbing fast ions sampling more harmonic layers in NSTX than in DIII-D. We investigate cyclotron absorption in cases with multiple harmonic layers within a single ion gyroradius and related phenomena numerically and analytically by examining the response of individual charged particles to rf fields in various field configurations. 8pt [1] M. Podesta et al., RF Power in Plasmas (Proc.18th Top. Conf., Gent, Belgium, 2009), to be published.

  20. 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.

  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. Installation and Testing of a 110 GHz Gyrotron With Depressed Collector Potential on the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Lohr, John; Gorelov, I. A.; Grunloh, H. J.; Kajiwara, K.; Pawley, C.; Ponce, D.; Tooker, J. F.; Chu, T. S.; Blank, M.; Borchard, P.; Cahalan, P.; Cauffman, S.; Felch, K.

    2006-10-01

    A single stage depressed collector gyrotron has been installed for testing on the DIII-D tokamak. The tube has operated at the 1.25 MW level at 44% efficiency for short pulses and at 0.5 MW for 10 s pulses at CPI and now is being tested to full parameters at DIII-D. A two-stage mode conversion dummy load has been built to handle the higher ultimate power from this gyrotron for pulses up to 5 s in length. Modifications to the high voltage power supply system were required to provide 30 kV depression and sequencing of the application of the voltages.

  3. Fabrication development and usage of vanadium alloys in DIII-D

    SciTech Connect

    Smith, J.P.; Johnson, W.R.; Reis, E.E.

    1996-10-01

    GA is procuring material, designing components, and developing fabrication techniques for use of V alloy into the DIII-D divertor as elements of the Radiative Divertor Project modification. This program was developed to assist in the development of low activation alloys for fusion use by demonstrating the fabrication and installation of V alloy components in an operating tokamak. Along with fabrication development, the program includes multiple steps starting with small coupons installed in DIII-D to measure the environmental effects on V. This is being done in collaboration with DOE Fusion Materials Program (particularly at ANL and ORNL). Procurement of the material has been completed; the world`s largest heat of V alloy (1200 kg V-4Cr-4Ti) was produced and converted into various products. Manufacturing process is described and chemistry results presented. Research into potential fabrication methods is being performed. Joining of V alloys was identified as the most critical fabrication issue for its use in the Radiative Divertor program. Successful welding trials were done using resistance, friction, and electron beam methods; metallography and mechanical tests were done to evaluate the welds.

  4. Experimental Evidence of Edge Fluctuation Broadening of ECH Deposition at DIII-D

    NASA Astrophysics Data System (ADS)

    Brookman, M. W.; Austin, M. E.; Gentle, K. W.; Petty, C. C.; Peysson, Y.; Decker, J.; Barada, K.; Ernst, D. E.

    2016-10-01

    This work provides experimental evidence for broadening of the ECH and ECCD deposition by edge density fluctuations. Results on the DIII-D tokamak suggest a deposition FWHM 1.7-2.8 times wider than TORAY-GA. A 1D ECH deposition profile was measured through gyrotron power modulation. From 500 kHz, 48-channel ECE measurements and trial ECH deposition functions, a Fourier transformed heat flux is found and fit to transport drive terms. Radially broader ECH deposition best fit calculated fluxes in discharges with higher levels of edge density turbulence. Broadening of deposition does not arise from anomalous transport, which is minimal on DIII-D. Simulation and theory suggest edge (ρ.9) turbulent n _ e fluctuations refract RF waves that pass through them, broadening radial deposition of ECH and ECCD. On ITER, this effect could hinder NTM suppression by broadening ECCD deposition outside the 3/2 island. Work supported by the U.S. DOE under Award DE-FC02-04ER54698.

  5. Tokamak Equilibrium Reconstruction with MSE-LS Data in DIII-D

    NASA Astrophysics Data System (ADS)

    Lao, L.; Grierson, B.; Burrell, K. H.

    2016-10-01

    Equilibrium analysis of plasmas in DIII-D using EFIT was upgraded to include the internal magnetic field determined from spectroscopic measurements of motional-Stark-effect line-splitting (MSE-LS). MSE-LS provides measurements of the magnitude of the internal magnetic field, rather than the pitch angle as provided by MSE line-polarization (MSE-LP) used in most tokamaks to date. EFIT MSE-LS reconstruction algorithms and verifications are described. The capability of MSE-LS to provide significant constraints on the equilibrium analysis is evaluated. Reconstruction results with both synthetic and experimental MSE-LS data from 10 DIII-D discharges run over a range of conditions show that MSE-LS measurements can contribute to the equilibrium reconstruction of pressure and safety factor profiles. Adequate MSE-LS measurement accuracy and number of spatial locations are necessary. The 7 available experimental measurements provide useful additional constraints when used with other internal measurements. Using MSE-LS as the only internal measurement yields less current profile information. Work supported by the PPPL Subcontract S013769-F and US DOE under DE-FC02-04ER54698.

  6. Automatic Fault-Checking System on the DIII-D Tokamak

    SciTech Connect

    Scoville, J.T.; Walker, M.L.

    2005-04-15

    Modern tokamaks are highly sophisticated devices consisting of a large number of state-of-the-art systems that must function in unison to obtain a successful plasma discharge. An unsuccessful discharge can result if one or more systems fail, and diagnosis in an efficient and timely manner can be difficult. The resulting reduction in tokamak availability and productivity can be expensive, justifying a significant effort for automated fault diagnosis.For the DIII-D tokamak, a software system has been used for the past 5 years to automatically monitor and test the performance of hundreds of tokamak systems. The Fault Identification and Communication System (FICS) is automatically triggered to run immediately after each tokamak discharge and report its results via a simple color-coded graphical user interface. In addition to saving the operator time, the significant advantage of FICS is its ability to detect insipient faults that could lead to future machine failures. It has been estimated that FICS has saved an average of one to two shots per day, which equates to approximately 5% of all DIII-D pulses. The significant experience gained through the development and use of this post-discharge analysis tool also provides insight into future methods for on-line process monitoring of steady state devices.

  7. Systematic Characterization of Component Failures for the DIII-D Tokamak

    SciTech Connect

    Petersen, P.I.

    1999-08-01

    A fusion reactor will be a fairly complex system consisting of many components. All the components are required to work in order to produce a plasma and control it. Some of the components will be large, and for economic reasons there will not be spares for all components. It is therefore important to have a system whereby troubles are communicated, recorded and analyzed. Such a trouble report system has been in place at the DIII-D tokamak facility for many years. The purpose of the system is to easily facilitate communication between the people that discover problems and those that fix the problems. The trouble sheets are logged into a computer database that is used to characterize the kind of problems that the facility experiences, and determine which equipment, software, or human errors are causing significant downtime. The information is also used to evaluate whether sufficient maintenance is done to the equipment and to provide a basis for replacing it. The original system was based on paper forms. About a year ago the system was changed to a web-based system. In the new system a trouble report is filled out using a web browser, and the information is emailed to the repair personnel and managers as soon as the form is submitted through the web. The paper will discuss the problems experienced at the DIII-D facility, and how the information is used to adjust the preventive maintenance schedule.

  8. Technical overview of the millimeter-wave imaging reflectometer on the DIII-D tokamak (invited)

    NASA Astrophysics Data System (ADS)

    Muscatello, C. M.; Domier, C. W.; Hu, X.; Kramer, G. J.; Luhmann, N. C.; Ren, X.; Riemenschneider, P.; Spear, A.; Tobias, B. J.; Valeo, E.; Yu, L.

    2014-11-01

    The two-dimensional mm-wave imaging reflectometer (MIR) on DIII-D is a multi-faceted device for diagnosing electron density fluctuations in fusion plasmas. Its multi-channel, multi-frequency capabilities and high sensitivity permit visualization and quantitative diagnosis of density perturbations, including correlation length, wavenumber, mode propagation velocity, and dispersion. The two-dimensional capabilities of MIR are made possible with 12 vertically separated sightlines and four-frequency operation (corresponding to four radial channels). The 48-channel DIII-D MIR system has a tunable source that can be stepped in 500 μs increments over a range of 56 to 74 GHz. An innovative optical design keeps both on-axis and off-axis channels focused at the cutoff surface, permitting imaging over an extended poloidal region. The integrity of the MIR optical design is confirmed by comparing Gaussian beam calculations to laboratory measurements of the transmitter beam pattern and receiver antenna patterns. Measurements are presented during the density ramp of a plasma discharge to demonstrate unfocused and focused MIR signals.

  9. Characterizing the DIII-D divertor conditions during the tungsten ring experiment

    NASA Astrophysics Data System (ADS)

    Barton, J. L.; Watkins, J. G.; Wang, H. Q.; Nygren, R. E.; McLean, A.; Makowski, M.; Unterberg, E.; Thomas, D. M.; Guo, H. Y.; Guterl, J.; Buchenauer, B.

    2016-10-01

    Tungsten (W) is the leading divertor material in tokamaks, but the core W impurity fraction must be kept below 5 ×10-5 in a reactor. The DIII-D tokamak, having all graphite PFCs, has done a series of experiments with two W-coated molybdenum rings in the lower divertor to track W migration after plasma exposure. We characterize the divertor plasma conditions at the DIII-D target plate in L- and ELMing H-mode, and ELM suppressed plasmas. We will present data from an array of Langmuir probes in the divertor and divertor Thomson-scattering. We also compare the heat flux from fast thermocouples (7.5 mm below the surface of the metal tile inserts) and IRTV heat flux profiles from graphite tiles. The plasma conditions will be used to benchmark ERO modeling to aid in understanding the migration of sputtered W onto other plasma facing surfaces and will be compared to post exposure W distribution measured on the graphite tiles. Supported by US DOE under DE-AC04-94AL85000, DE-FC02-04ER54698, DE-AC05-000R22725, and DE-AC52-07NA27344.

  10. Energetic particle effects on n=1 MHD instabilities in a DIII-D hybrid discharge

    NASA Astrophysics Data System (ADS)

    Brennan, D. P.; Halfmoon, M. R.; Kim, C. C.; La Haye, R. J.

    2012-03-01

    The δf kinetic-MHD model in the 3-D extended MHD code NIMROD is used to perform a simulation study of energetic particle effects on the n=1 mode in a DIII-D hybrid discharge. The hybrid has low qmin>˜1 at high confinement, and is a candidate operational scenario for burning plasma experiments. However hybrid discharges are limited to moderate βN by the m/n=2/1 instability. Using realistic DIII-D equilibria, the stability of the n=1 mode is computed over a (qmin,βN) space. Unstable modes are driven by energetic particles far into the MHD stable region in this space. The drive is associated with the fishbone mode or BAE mode, depending on qmin. The stability boundary is found near the experimental (qmin,βN), where the unstable mode has a m/n=1/1 component localized near the axis. Experimentally, a m/n=1/1 structure is observed in agreement with the computed mode in key physical respects. At higher qmin and βN a mode with a broad m/n=2/1 structure is unstable. This suggests that the m/n=2/1 mode is triggered by energetic particles in these discharges, as βN is increased. A group of several similar discharges shows strong agreement with this computational explanation of onset.

  11. Modeling carbon production and transport during ELMs in DIII-D

    NASA Astrophysics Data System (ADS)

    Hogan, J.; Wade, M.; Coster, D.; Lasnier, C.

    2004-11-01

    Large-scale Type I ELM events could provide a significant C source in ITER, and C production rates depend on incident D flux density and surface temperature, quantities which can vary significantly during an ELM event. Recent progress on DIII-D has improved opportunities for code comparison. Fast time-scale measurements of divertor CIII evolution [1] and fast edge CER measurements of C profile evolution during low-density DIII-D LSN ELMy H-modes (type I) [2] have been modeled using the solps5.0/Eirene99 coupled edge code and time dependent thermal analysis codes. An ELM model based on characteristics of MHD peeling-ballooning modes reproduces the pedestal evolution. Qualitative agreement for the CIII evolution during an ELM event is found using the Roth et al annealing model for chemical sputtering and the sensitivity to other models is described. Significant ELM-to-ELM variations in observed maximum divertor target IR temperature during nominally identical ELMs are investigated with models for C emission from micron-scale dust particles. [1] M Groth, M Fenstermacher et al J Nucl Mater 2003, [2] M Wade, K Burrell et al PSI-16

  12. Investigation of He–W interactions using DiMES on DIII-D

    DOE PAGES

    Doerner, R. P.; Rudakov, D. L.; Chrobak, C. P.; ...

    2016-01-22

    Here, tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225–850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in themore » surface of W buttons after only a single DIII-D (3 s, ~150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04–0.1 MJ m–2.« less

  13. Exposures of tungsten nanostructures to divertor plasmas in DIII-D

    DOE PAGES

    Rudakov, D. L.; Wong, C. P. C.; Doerner, R. P.; ...

    2016-01-22

    Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with littlemore » obvious damage except in the areas where unipolar arcing occurred. In conclusion, arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.« less

  14. Simulation of RSAE/EAE and TAE activity in a reversed shear DIII-D case

    NASA Astrophysics Data System (ADS)

    Spong, Donald; van Zeeland, Michael

    2008-11-01

    The energetic particle gyrofuild model TAEFL has been applied to a DIII-D case where RSAE, TAE and EAE activity was present. This is a useful analysis tool because: (1) it retains the primary non-perturbative affect of fast ions on Alv'enic mode structures, i.e., the coupling of MHD cosine and sine parities by fast ion diamagnetic flows [which in the phi = 0 plane causes up-down asymmetries], (2) it is applicable to non-circular shaped tokamak equilibria, (3) it focuses on only the most unstable modes, and (4) an efficient implicit stepping algorithm has been developed, allowing rapid scans of linear growth rates and mode structures for cases with significant poloidal coupling. Analysis of the above DIII-D case has identified EAE, RSAE and TAE modes at sub-Alfv'enic beam velocities (down to ˜0.28 times the Alfv'en velocity). The dominant AE mode is a sensitive function of the q-profile and fast ion parameters. Also, the non-perturbative nature of this model yields mode structures that are consistent with the experimental coherence measurements.

  15. Fast Ion Effects During Test Blanket Module Simulation Experiments in DIII-D

    NASA Astrophysics Data System (ADS)

    Kramer, G. J.; Budny, R.; Nazikian, R.; Heidbrink, W. W.; Kurki-Suonio, T.; Salmi, A.; Schaffer, M. J.; van Zeeland, M. A.; Shinohara, K.; Snipes, J. A.; Spong, D.

    2010-11-01

    The fast beam-ion confinement in the presence of a scaled mock-up of two Test Blanket Modules (TBM) for ITER was studied in DIII-D. The TBM on DIII-D has four vertically arranged protective carbon tiles with thermocouples placed at the back of each tile. Temperature increases of up to 200^oC were measured for the two tiles closest to the midplane when the TBM fields were present. These measurements agree qualitatively with results from the full orbit-following beam-ion code, SPIRAL, that predict beam-ion losses to be localized on the central two carbon tiles when the TBM fields present. Within the experimental uncertainties no significant change in the fast-ion population was found in the core of these plasmas which is consistent with SPIRAL analysis. These experiments indicate that the TBM fields do not affect the fast-ion confinement in a harmful way which is good news for ITER.

  16. Time-dependent Radial Transport of Electron Distributions Due to ECCD in DIII-D

    NASA Astrophysics Data System (ADS)

    Harvey, R. W.; Smirnov, A. P.; Prater, R.; Petty, C. C.

    2007-11-01

    The radial transport modeling capability in the CQL3D bounce-averaged Fokker-Planck collisional-rf quasilinear code[1] has been greatly improved and the self-consistent time-dependent toroidal electric field added, making the code truly a ``Fokker-Planck-Transport'' code. The time-dependent, coupled 3D Fokker-Planck equation and the Ampere-Faraday Law equation are solved for the electron distribution, f( u,θu,ρ,t ), and the toroidal loop voltage, Vloop( ρ,t ). A fully 3D, time-implicit solution of the FP equation using sparse-matrix methods[2] is coupled to a new iterative toroidal electric field solve. The DIII-D ECH experiment is in an intermediate driven regime with τtransport τslowing[3] for the EC driven electrons. Results will be reported for time-evolution of radial profiles of current density, fast electrons, and toroidal loop voltage due to EC heating and current drive in DIII-D. [1] R.W. Harvey and M.G. McCoy, IAEA TCM on Advances in Simulation and Modeling of Thermonuclear Plasmas, Montreal, 1992; USDOC NTIS No. 93002962. [2] Y. Peysson et al., Radio Frequency Power in Plasmas, 15th Topical Conference, Moran, Wyoming (2003). [3] R.W. Harvey et al., Phys. Rev. Lett. 88, 205001 (2002).

  17. Characterization of the Radiation Environment During and Following Operation of the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Riso, Victoria; Pace, D. C.; Cooper, C. M.

    2015-11-01

    A survey of the gamma ray spectrum throughout the machine hall of the DIII-D tokamak provides a detailed mapping of its energy and temporal evolution. Engineering issues related to the structural effects of radiation produced by a fusion power plant will significantly affect the cost-effectiveness of the resulting energy. While existing magnetic confinement facilities produce considerably less neutron and gamma radiation than that expected from a power plant-scale facility, it remains useful to examine the latent gamma spectrum of the surrounding structures. The DIII-D tokamak produces ~1016 neutrons per run day (resulting primarily from beam-target DD fusion), with ~75 run days per year, leading to the activation of support structures with a short half-life. Measurements are made using bismuth germinate scintillator detectors operated in pulse height analysis mode. These detectors are placed throughout the machine hall and acquire gamma data both during experiments and for some time afterward. Results of these surveys from the 2015 experiments will be presented. Supported in part by US DOE under DE-FC02-04ER54698.

  18. Development of electron thermal transport model in DIII-D discharges

    NASA Astrophysics Data System (ADS)

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

    2008-11-01

    The electron thermal transport in tokamak plasmas is investigated with predictive integrated modeling simulations using a choice of different electron thermal transport models. Two models for transport driven by Electron Temperature Gradient (ETG) modes are considered: (1) the ETG part of the GLF23 transport model; and (2) the Horton model for the the electromagnetic part of the ETG anomalous transport [1]. These models are combined with the paleoclassical model [2] for electron thermal transport. ASTRA predictive simulation results obtained using these models are compared with one another and compared with experimental data from DIII-D H-mode discharges in an effort to discriminate among the models. It is found that the electromagnetic limit of the Horton model is important near the magnetic axis where the ETG mode in the GLF23 model is below threshold. The paleoclassical model is found to be needed to produce the observed edge pedestal in the DIII-D simulations. [1] W. Horton, B. G. Hong, and W. M. Tang, Phys. Fluids 31, 2971 (1988). [2] J. D. Callen, Nucl. Fusion 45, 1120 (2005).

  19. Development of GPU-Optimized EFIT for DIII-D Equilibrium Reconstructions

    NASA Astrophysics Data System (ADS)

    Huang, Y.; Lao, L. L.; Xiao, B. J.; Luo, Z. P.; Yue, X. N.

    2015-11-01

    The development of a parallel, Graphical Processing Unit (GPU)-optimized version of EFIT for DIII-D equilibrium reconstructions is presented. This GPU-optimized version (P-EFIT) is built with the CUDA (Compute Unified Device Architecture) platform to take advantage of massively parallel GPU cores to significantly accelerate the computation under the EFIT framework. The parallel processing is implemented with the Single-Instruction Multiple-Thread (SIMT) architecture. New parallel modules to trace plasma surfaces and compute plasma parameters have been constructed. DIII-D magnetic benchmark tests show that P-EFIT could accurately reproduce the EFIT reconstruction algorithms at a fraction of the computational cost. The acceleration factor continues to increase as the (R, Z) spatial grids are increased from 65 × 65 to 257 × 257 , suggesting there may be rooms for further optimization by further reducing the communication cost. Details of the P-EFIT optimization algorithms will be discussed. Work supported by US DOE DE-FC02-04ER54698, and by China MOST under 2014GB103000, China NNSF 11205191, China CAS GJHZ201303.

  20. Identifying the Location of the Separatrix at the OMP in DIII-D Using Power Accounting

    NASA Astrophysics Data System (ADS)

    Stangeby, P. C.

    2013-10-01

    Two methods are used which employ power accounting to improve the accuracy of identifying R-sep-omp, the location of the separatrix at the outside midplane (omp). The first method uses the measured deposited power profile at the outer target as the primary input, the P-SOL-exhaust method. The other uses the measured power input to the SOL, obtained from the total heating power less the power radiated from inside the separatrix, the P-SOL-input method. The methods were applied to experimental data for 21 H-mode DIII-D discharges. High spatial resolution Thomson scattering measured profiles of between-ELM ne and Te were used to calculate the electron parallel conducted heat flux profile which was then matched to the measured P-SOL-exhaust and P-SOL-input by adjusting R-sep-omp relative to that of the Thomson data. The values of R-sep-omp from the 2 methods agree to within ~1 mm of each other and to within ~1 mm of the values given by the ``standard DIII-D method''. This results in only modest changes to ne and Te at R-sep-omp relative to the ``standard'' values, increasing ne by ~10% and Te by ~20%. Work supported in part by the US DOE under DE-FC02-04ER54698.

  1. Measurements of non-axisymmetric effects in the DIII-D divertor

    SciTech Connect

    Evans, T.E,; Leonard, A.W.; Petrie, T.W.; Schaffer, M.J.; Lasnier, C.J.; Hill, D.N.; Fenstermacher, M.E.

    1994-07-01

    Non-stationary toroidal asymmetries are observed in the DIII-D divertor heat flux and scrape-off layer (SOL) currents. Using the present DIII-D diagnostics asymmetries are seen much less frequently in single-null H-modes (<5%) than in double-null H-modes (>50%). Divertor heat flux asymmetries are characterized by toroidal variations in the radial profile (i.e., multiple or bifurcated peaks at some toroidal locations and single peaks at others) while SOL currents sometimes have a strongly bipolar toroidal structure. SOL current asymmetries are particularly large during Edge Localized Modes (ELMs). In some cases heat flux variations of as much as a factor of two are seen. The measurements reported here indicate that these asymmetries are best described by a model in which non-axisymmetric radial magnetic perturbations create magnetic islands in the plasma boundary and scrape-off layer which then cause toroidal variation in the divertor heat flux and the scrape-off layer currents.

  2. Application of the Ideal Perturbed Equilibrium Code to DIII-D Magnetic Diagnostic Upgrade Designs

    NASA Astrophysics Data System (ADS)

    Logan, N. C.; Menard, J. E.; Park, J. K.; Strait, E. J.

    2012-10-01

    The Ideal Perturbed Equilibrium Code (IPEC) has been upgraded with advanced visualization tools and synthetic diagnostics to make its output directly comparable with DIII-D diagnostic measurements. Using the synthetic magnetic diagnostics, IPEC has been used to assist in the design of an advanced 3D magnetic field diagnostic currently being built as an upgrade to the DIII-D tokamak experiment. This poster outlines the application of IPEC modeling to the magnetic diagnostic design, highlighting the power and versatility of both the computational tools and proposed diagnostics. Of the many new measurements that will be possible with the magnetic diagnostic upgrade, special emphasis is given here to the ability to directly measure electromagnetic torques on the plasma. The magnetic diagnostic design will be able to simultaneously measure electromagnetic torque from non-axisymmetric fields with toroidal mode numbers 1, 2 and 3. This will open the door to many new possibilities in studying rotational braking effects that will be further supported by IPEC.

  3. 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.

  4. Improved Spectral Fitting Models for the B-Stark Diagnostic at DIII-D

    NASA Astrophysics Data System (ADS)

    Pablant, N. A.; Grierson, B. A.; Burrell, K. H.; Groebner, R. J.; Kaplan, D. H.; Holcomb, C. T.

    2010-11-01

    Recent results are presented from the B-Stark diagnostic installed on the DIII-D tokamak. This diagnostic provides measurements of the magnitude and direction of the internal magnetic field. The B-Stark system is a version of a motional Stark effect (MSE) diagnostic based on the relative line intensities and spacing of the Stark split Dα emission from injected neutral beams. Improvements to the spectral fitting model are presented, including the addition of an analytical model for Dα emission from the fast-ion distribution. We discuss the accuracy of using in-situ beam-into-gas calibrations to find the beam emission line profiles, the viewing direction and the transmission properties of the collection optics. We also present results of efforts to improve the determination of the beam emission line profiles. Finally, the magnetic field measured with the B-Stark system is compared to values found from plasma equilibrium reconstructions (EFIT) and the MSE polarimetry system on DIII-D.

  5. Inferring Core Tungsten Behavior Using SPRED During the DIII-D Metal Rings Campaign

    NASA Astrophysics Data System (ADS)

    Thomas, D. M.; Kaplan, D.; Groebner, R.; Grierson, B.; Unterberg, Z.; Victor, B.

    2016-10-01

    The GA SPRED EUV spectrometer was used to study core emission of highly charged tungsten ions (W40+-W45+) in the 120-135Å region during the recent Metal Rings Campaign. These experiments used two 5-cm wide toroidal rings of W-coated metal inserts exposed to a variety of DIII-D discharges to study effects of high-Z divertor erosion, migration, core uptake, and effects on advanced tokamak performance. For the proper core temperature range (2-4 keV), the measured multistate W emission forms a well defined spectral pattern that can be used to study the relative importance of strike point location, flux expansion, injected power, ELM characteristics and magnetic drift direction for high-Z core contamination in DIII-D. The spectra are fit using simple Gaussians to estimate concentrations using the historical SPRED intensity calibration. Calibration shots using known core dosages of pellet injected W are used to help infer the relative response of the instrument. Supported by US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466, DE-AC05-00OR22725, DE-AC52-07NA27344.

  6. Electron cyclotron heating and core intrinsic rotation reversal in DIII-D

    SciTech Connect

    Grassie, J. S. de; Boedo, J. A.; Grierson, B. A.

    2015-12-10

    The effect of electron cyclotron heating (ECH) on the intrinsic rotation profile in DIII-D is shown experimentally. Former DIII-D experiments have shown that ECH tends to cause an interior reduction in the normally co-Ip directed intrinsic rotation profile, and this core rotation can be fully reversed to the opposite direction. This effect is due to a turbulent rearrangement of the interior rotation profile. Here, we show results that there is more than one mechanism causing this. We compare two low density L-mode discharges where the only operational difference is the location of the ECH deposition. At low ECH power, comparable to the Ohmic power, the primary change is in the q-profile accompanied by a reversal of the core intrinsic rotation direction for the more off-axis deposition. The change in the shear of the q-profile fits well with a recent theoretical prediction for this rotation reversal. At higher ECH power, the primary change is in the core electron temperature, Te, accompanied by a hollowing of the rotation profile near the magnetic axis. This effect appears to be due to the change in electron collisionality, consistent with another theoretical, gyrokinetic prediction. The variety of phenomena that could allow ECH to modify the intrinsic rotation profile give some expectation that regions of large velocity shear in the interior could be generated, with the possibility of triggering internal transport barriers.

  7. Understanding distortions in the DIII-D tokamak magnetic field trim coils

    NASA Astrophysics Data System (ADS)

    Paulsen, R. T.; Paz-Soldan, C.; Strait, E. J.

    2016-10-01

    Trim coils were originally incorporated onto the DIII-D tokamak to reduce the error fields that arise from distortions in the preexisting coils that confine the plasma. However, as a result of numerous obstacles crowding DIII-D prior to their installation, the effective geometries of the trim coils were forced to stray from a nominal rectangular shape, causing the trim coils to induce error fields of their own apart from the ones they serve to combat. Since these secondary error fields have the potential to hamper plasma operation, it is imperative to discover to what extent the trim coil geometry is understood. This work seeks to investigate the observed magnetic field measurements for the energized trim coils in the absence of plasma, assess the distortions, and compare these calculations to the same quantities estimated for schematics of the as-built trim coil configurations. Furthermore, measurements taken with plasma with and without trim coils are analyzed by tools for extracting the optimal current for error field correction and compared to the predictions for the as-built system. Supported by US DOE DE-FC02-04ER54698 and SULI program.

  8. Operational performance and plans for the ECH system on DIII-D

    NASA Astrophysics Data System (ADS)

    Lohr, J.; Cengher, M.; Gorelov, Y. A.; Ponce, D.; Torrezan, A.; Ives, L.; Read, M.; Leviness, A.

    2016-10-01

    The ECH system on the DIII-D tokamak currently comprises 6 gyrotrons operating at 110 GHz and injecting 3.5 MW for administratively limited pulse lengths up to 5 sec. A 7th gyrotron generating 1.5 MW at 117.5 GHz is planned for installation late in 2016. Production of this tube was delayed due to issues related to reflected electrons resulting in internal arcs during initial testing. Performance reliability of the individual gyrotrons in the DIII-D complex has exceeded 90% for a wide variety of operational modes, including fast modulation and rapid poloidal sweeping of the rf beams using high speed dc motors and magnetic position encoders. Measures have been taken to reduce the risk of damage to launcher hardware and diagnostics from inadvertent operation of the tokamak at densities such that the right hand cutoff frequency is present in the plasma. The system has occasionally been used in non-fusion applications. The most recent of these has been testing a CVD diamond waveguide blocking window at the Brewster angle. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC02-09CH11466.

  9. 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.

  10. 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.

  11. Effect of resonant magnetic perturbations on microturbulence in DIII-D pedestal

    DOE PAGES

    Holod, I.; Lin, Z.; Taimourzadeh, S.; ...

    2016-10-03

    Vacuum resonant magnetic perturbations (RMP) applied to otherwise axisymmetric tokamak plasmas produce in general a combination of non-resonant effects that preserve closed flux surfaces (kink response) and resonant effects that introduce magnetic islands and/or stochasticity (tearing response). The effect of the plasma kink response on the linear stability and nonlinear transport of edge turbulence is studied using the gyrokinetic toroidal code GTC for a DIII-D plasma with applied n = 2 vacuum RMP. GTC simulations use the 3D equilibrium of DIII-D discharge 158103 (Nazikian et al 2015 Phys. Rev. Lett. 114 105002), which is provided by nonlinear ideal MHD VMECmore » equilibrium solver in order to include the effect of the plasma kink response to the external field but to exclude island formation at rational surfaces. Analysis using the GTC simulation results reveal no increase of growth rates for the electrostatic drift wave instability and for the electromagnetic kinetic-ballooning mode in the presence of the plasma kink response to the RMP. Moreover, nonlinear electrostatic simulations show that the effect of the 3D equilibrium on zonal flow damping is very weak and found to be insufficient to modify turbulent transport in the electrostatic turbulence.« less

  12. Effect of resonant magnetic perturbations on microturbulence in DIII-D pedestal

    SciTech Connect

    Holod, I.; Lin, Z.; Taimourzadeh, S.; Nazikian, R.; Spong, D.; Wingen, A.

    2016-10-03

    Vacuum resonant magnetic perturbations (RMP) applied to otherwise axisymmetric tokamak plasmas produce in general a combination of non-resonant effects that preserve closed flux surfaces (kink response) and resonant effects that introduce magnetic islands and/or stochasticity (tearing response). The effect of the plasma kink response on the linear stability and nonlinear transport of edge turbulence is studied using the gyrokinetic toroidal code GTC for a DIII-D plasma with applied n = 2 vacuum RMP. GTC simulations use the 3D equilibrium of DIII-D discharge 158103 (Nazikian et al 2015 Phys. Rev. Lett. 114 105002), which is provided by nonlinear ideal MHD VMEC equilibrium solver in order to include the effect of the plasma kink response to the external field but to exclude island formation at rational surfaces. Analysis using the GTC simulation results reveal no increase of growth rates for the electrostatic drift wave instability and for the electromagnetic kinetic-ballooning mode in the presence of the plasma kink response to the RMP. Moreover, nonlinear electrostatic simulations show that the effect of the 3D equilibrium on zonal flow damping is very weak and found to be insufficient to modify turbulent transport in the electrostatic turbulence.

  13. 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.

  14. First Measurements of W Erosion from Ultraviolet Emission in DIII-D

    NASA Astrophysics Data System (ADS)

    Johnson, C. A.; Ennis, D. A.; Loch, S. D.; Allen, S.; Victor, B.; Samuell, C. M.; Ballance, C. P.; Abrams, T.; Unterberg, E.

    2016-10-01

    Erosion of Plasma Facing Components (PFCs) will play a critical role in establishing the performance of future reactor-relevant fusion devices. Erosion can be diagnosed from spectral line emission together with atomic coefficients called 'ionizations per photon' (S/XB). New ultraviolet survey spectrometers (200 to 400 nm) have been commissioned on the DIII-D experiment for enhanced diagnosis of W erosion in the divertor region. Previous atomic calculations predict neutral W will radiate most strongly at ultraviolet wavelengths. Ultraviolet measurements from DIII-D experiments with W PFCs in the divertor have identified new candidate spectral lines for more accurately diagnosing neutral W erosion rates. For example, W emission lines at 265.65 and 363.19 nm were observed to be more intense than the widely used 400.89 nm. Complete UV spectra will be presented and compared to synthetic spectra generated by new atomic calculations of neutral W using ADAS for varying plasma conditions. Work supported by USDOE Grant E-FC02-04ER54698 and DE-SC0015877.

  15. Measurements of the Ech Power and of the Transmission Line Losses on Diii-D

    NASA Astrophysics Data System (ADS)

    Cengher, M.; Lohr, J.; Gorelov, I. A.; Grosnickle, W. H.; Ponce, D.; Johnson, P.

    2009-04-01

    The measurement of the power injected by the electron cyclotron heating (ECH) system in the DIII-D tokamak is a critical requirement for analysis of experiments, for tuning the gyrotrons for maximum power and efficiency, for tracking long-term operational trends and for providing a warning of problems with the system. The ECH system at General Atomics consists of six 110 GHz, 1 MW class gyrotrons. The rf power generated by each gyrotron is determined from calorimetry, using the relevant temperature and flow measurements from the cooling circuits of cavity, matching optics unit and dummy loads. The rf pulse length and time dependence are measured using an rf monitor at the first miter bend in the transmission line. The direct measurement of the efficiencies of four of the transmission lines was performed using a high power, small dummy load (SDL) placed alternately in 2 positions of each DIII-D waveguide line, at accessible points close the beginning and the end of each line. Total losses in the transmission lines range from 21.2% to 30.7%. Experimental results are compared to theoretical predictions of the performance of the components and waveguide lines.

  16. Measurement and Analysis of ECH Power Injected Into DIII-D

    NASA Astrophysics Data System (ADS)

    Johnson, P. S.; Lohr, J.

    2007-11-01

    The 6 ECH waveguides at DIII-D are on the order of 100 meters in length with up to 16 miter bends. Accurate measurement of the ratio of generated-to-transmitted power gives the transmission line efficiency directly and is essential for analysis of experiments. The power generated by the gyrotrons is measured calorimetrically for each pulse, but direct measurements of the injected power have relied on analysis of modulated plasma heating, which can overlook significant power where plasma volumes are large and ECH driven temperature fluctuations are small. High power tests of efficiencies of individual components have been difficult due to mutual interaction of components, sensitivity of power monitors to polarization, and the generally high efficiency of the components. We report a direct measurement of the efficiencies of complete transmission lines, using a high power dummy load placed at the end of each DIII-D waveguide. Experimental results will be compared to previous measurements and to theoretical calculations of the performance of the components and waveguide lines.

  17. Power Calibration for the Electron Cyclotron Heating System on DIII-D

    NASA Astrophysics Data System (ADS)

    Cengher, M.; Lohr, J.; Gorelov, I. A.; Ponce, D.; Kajiwara, K.

    2007-11-01

    The generated rf power for each of the five gyrotrons in the DIII-D system is calculated based on calorimetry, using temperature and flow measurements from the gyrotron and waveguide system cooling circuits for the cavity, window, collector, matching optics unit (MOU) and dummy loads. Analysis of the data involves fitting the dissipated energy versus time curves and integration of the energy for each of the circuits. The cavity signal is used to calculate the total generated energy, using a previously determined relationship between cavity loading and rf production, with other cooling circuits as a check. The time dependence of the rf power is determined using a diode pickoff at the first miter bend in the transmission line normalized to the integrated calorimetry measurement. The MOU calorimetry response provides a direct measurement of the percentage of rf in the Gaussian mode and the efficiency of coupling the rf into the waveguide. The losses in each transmission line are taken into account to calculate the power transmitted to DIII-D.

  18. Investigation of He-W interactions using DiMES on DIII-D

    NASA Astrophysics Data System (ADS)

    Doerner, R. P.; Rudakov, D. L.; Chrobak, C. P.; Briesemeister, A. R.; Corr, C.; De Temmerman, G.; Kluth, P.; Lasnier, C. J.; McLean, A. G.; Pace, D. C.; Pitts, R. A.; Schmitz, O.; Thompson, M.; Winters, V.

    2016-02-01

    Tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225-850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in the surface of W buttons after only a single DIII-D (3 s, ˜150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04-0.1 MJ m-2.

  19. Effect of resonant magnetic perturbations on microturbulence in DIII-D pedestal

    NASA Astrophysics Data System (ADS)

    Holod, I.; Lin, Z.; Taimourzadeh, S.; Nazikian, R.; Spong, D.; Wingen, A.

    2017-01-01

    Vacuum resonant magnetic perturbations (RMP) applied to otherwise axisymmetric tokamak plasmas produce in general a combination of non-resonant effects that preserve closed flux surfaces (kink response) and resonant effects that introduce magnetic islands and/or stochasticity (tearing response). The effect of the plasma kink response on the linear stability and nonlinear transport of edge turbulence is studied using the gyrokinetic toroidal code GTC for a DIII-D plasma with applied n  =  2 vacuum RMP. GTC simulations use the 3D equilibrium of DIII-D discharge 158103 (Nazikian et al 2015 Phys. Rev. Lett. 114 105002), which is provided by nonlinear ideal MHD VMEC equilibrium solver in order to include the effect of the plasma kink response to the external field but to exclude island formation at rational surfaces. Analysis using the GTC simulation results reveal no increase of growth rates for the electrostatic drift wave instability and for the electromagnetic kinetic-ballooning mode in the presence of the plasma kink response to the RMP. Furthermore, nonlinear electrostatic simulations show that the effect of the 3D equilibrium on zonal flow damping is very weak and found to be insufficient to modify turbulent transport in the electrostatic turbulence.

  20. Testing of the ITER-ECE prototype receiver and related components on DIII-D

    NASA Astrophysics Data System (ADS)

    Austin, M. E.; Brookman, M. W.; Phillips, P. E.; Rowan, W. L.; Danani, S.

    2015-11-01

    Real-world testing of advanced plasma diagnostic instruments and techniques intended for use on ITER is crucial to ensure their success. A prototype millimeter-wave receiver developed by Virginia Diodes, Inc. was brought to DIII-D to check its performance by measuring third harmonic ECE in high temperature plasmas. The receiver is state-of-the art, employing a waveguide based triplexer and a DRO-based local oscillator with an integrated tripler, subharmonic mixer and amplifier to detect emission in the 200-300 GHz range. Comparisons of ECE measurements with those from the DIII-D Michelson interferometer will evaluate linearity, sensitivity, and noise temperature. Also, transmission measurements of a double wedged quartz window, very similar to that proposed for the ITER vacuum interface, are given, showing no interference effects and good broadband performance. Additionally, results of the testing of a new high intensity LED light source for alignment of transmission line components are shown. Supported by US DOE DE-FG02-97ER54415, DE-FC02-04ER54698.

  1. A tangentially viewing visible TV system for the DIII-D divertor

    SciTech Connect

    Fenstermacher, M.E.; Meyer, W.H.; Wood, R.D.

    1996-02-01

    A video camera system has been installed on the DIII-D tokamak for 2-D spatial studies of line emission in the lower divertor region. The system views the divertor tangentially from an outer port at approximately the height of the X-point. At the tangency plane the entire divertor from inner wall to outside the DIII-D bias ring is viewed with spatial resolution of approximately 1 cm. The image contains information from approximately 90 degrees of toroidal angle. In a recent upgrade, remotely controllable filter changers were added which have produced images from nominally identical shots using a series of spectral lines. Software was developed to calculate the response function matrix using distributed computing techniques and assuming toroidal symmetry. Standard sparse matrix algorithms are then used to invert the 3-D images onto a poloidal plane. Spatial resolution of the inverted images is 2 cm; higher resolution simply increases the size of the response function matrix. Initial results from a series of experiments with multiple identical shots show that the emission from CII and CIII, which appears along the inner scrape-off layer above and below the X-point during ELMing H-mode, moves outward and becomes localized near the X-point in Partially Detached Divertor (PDD) operation.

  2. RESISTIVE WALL STABILIZATION OF HIGH BETA PLASMAS IN DIII-D

    SciTech Connect

    STRAIT,EJ; BIALEK,J; BOGATU,N; CHANCE,M; CHU,MS; EDGELL,D; GAROFALO,AM; JACKSON,GL; JENSEN,TH; JOHNSON,LC; KIM,JS; LAHAYE,RJ; NAVRATIL,G; OKABAYASHI,M; REIMERDES,H; SCOVILLE,JT; TURNBULL,AD; WALKER,ML

    2002-09-01

    OAK A271 RESISTIVE WALL STABILIZATION OF HIGH BETA PLASMAS IN DIII-D. Recent DIII-D experiments show that ideal kink modes can be stabilized at high beta by a resistive wall, with sufficient plasma rotation. However, the resonant response by a marginally stable resistive wall mode to static magnetic field asymmetries can lead to strong damping of the rotation. Careful reduction of such asymmetries has allowed plasmas with beta well above the ideal MHD no-wall limit, and approaching the ideal-wall limit, to be sustained for durations exceeding one second. Feedback control can improve plasma stability by direct stabilization of the resistive wall mode or by reducing magnetic field asymmetry. Assisted by plasma rotation, direct feedback control of resistive wall modes with growth rates more than 5 times faster than the characteristic wall time has been observed. These results open a new regime of tokamak operation above the free-boundary stability limit, accessible by a combination of plasma rotation and feedback control.

  3. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D

    SciTech Connect

    Johnson, W.R.; Smith, J.P.

    1997-08-01

    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 Program (RDP), has been completed by Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). CVN impact tests on sheet material indicate that the material has properties comparable to other previously-processed V-4Cr-4Ti and V-5Cr-5Ti alloys. Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RDP, and research into several joining methods for fabrication of the RDP components, including resistance seam, friction, and electron beam welding, and explosive bonding is being pursued. Preliminary trials have been successful in the joining of V-alloy to itself by resistance, friction, and electron beam welding processes, and to Inconel 625 by friction welding. In addition, an effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625, in both tube-to-bar and sheet-to-sheet configurations, has been initiated, and results have been encouraging.

  4. Performance and data analysis aspects of the new DIII-D monostatic profile reflectometer system

    SciTech Connect

    Zeng, L. Peebles, W. A.; Doyle, E. J.; Rhodes, T. L.; Crocker, N.; Nguyen, X.; Wannberg, C. W.; Wang, G.

    2014-11-15

    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 (n{sub e}) 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 n{sub e} 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.

  5. Access to DIII-D data located in multiple files and multiple locations

    SciTech Connect

    McHarg, B.B. Jr.

    1993-10-01

    The General Atomics DIII-D tokamak fusion experiment is now collecting over 80 MB of data per discharge once every 10 min, and that quantity is expected to double within the next year. The size of the data files, even in compressed format, is becoming increasingly difficult to handle. Data is also being acquired now on a variety of UNIX systems as well as MicroVAX and MODCOMP computer systems. The existing computers collect all the data into a single shot file, and this data collection is taking an ever increasing amount of time as the total quantity of data increases. Data is not available to experimenters until it has been collected into the shot file, which is in conflict with the substantial need for data examination on a timely basis between shots. The experimenters are also spread over many different types of computer systems (possibly located at other sites). To improve data availability and handling, software has been developed to allow individual computer systems to create their own shot files locally. The data interface routine PTDATA that is used to access DIII-D data has been modified so that a user`s code on any computer can access data from any computer where that data might be located. This data access is transparent to the user. Breaking up the shot file into separate files in multiple locations also impacts software used for data archiving, data management, and data restoration.

  6. Performance Of The DIII-D ECH High Power Transmission Lines And Launching Systems

    NASA Astrophysics Data System (ADS)

    Cengher, Mirela; Lohr, J.; Gorelov, Y.; Torrezan, A.; Ponce, D.; Moeller, C.; Ellis, R. A.; Kolemen, E.

    2015-11-01

    The Electron Cyclotron Heating (ECH) transmission system on the DIII D tokamak consists of corrugated coaxial 31.75 mm waveguide transmission lines and steerable launching mirrors. Total power injected into plasma can reach up to 3.5 MW, with pulse length up to 5 seconds. The ECH power injected to the tokamak from each gyrotron is measured on a shot-to-shot basis and shows individual average injected powers from a gyrotron into the plasma between 540 and 700 kW. The transmission coefficient including the waveguide line and the MOU is between -1.04 dB and -1.43 dB. The maximum ECH energy injected into DIII-D is 16.6 MJ. The HE11 mode content is over 85 % for all the lines. The four dual waveguide launchers have increased poloidal scanning speed, and can steer the RF beams 40 degrees poloidally in 200 ms, with real-time poloidal motion control by the plasma control system. A new method of in-situ calibration of the mirror angle was used in conjunction with the upgrading of the encoders and motors for the launchers. Work supported by US DOE under DE-FC02-04ER54698 and DE-AC02-09CH11466.

  7. Real-Time Variation of the Injected Neutral Beam Energy on the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Scoville, J. T.; Crowley, B. J.; Pace, D. C.; Rauch, J. M.

    2016-10-01

    A powerful new technique for smoothly controlling the time evolution of injected energy from neutral beams has recently been implemented on the DIII-D tokamak. Upgrades to the high voltage circuitry of the neutral beams and to the tokamak's Plasma Control System have enabled the first-ever continuous variation of beam voltage during plasma shots. This avoids the perturbative effects of pulse modulation, which was the previously employed method for changing the injected beam power. The new technique allows much finer control of the injected energy, with beam voltage able to be varied smoothly over a 20 kV range (within the 45-85 kV beam operating space) in 0.5 sec. This capability enables fundamentally new experiments that require precise control of beam ion phase space, including the minimization of undesirable energetic ion instabilities and scans across low torque regimes at fixed power. We present a description of the beam system modifications and initial results from plasma experiments using the new variable beam energy capability on the DIII-D tokamak. Work supported by U.S. DOE under DE-FC02-04ER54698.

  8. Investigation of He–W interactions using DiMES on DIII-D

    SciTech Connect

    Doerner, R. P.; Rudakov, D. L.; Chrobak, C. P.; Briesemeister, A. R.; Corr, C.; De Temmerman, G.; Kluth, P.; Lasnier, C. J.; McLean, A. G.; Pace, D. C.; Pitts, R. A.; Schmitz, O.; Thompson, M.; Winters, V.

    2016-01-22

    Here, tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225–850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in the surface of W buttons after only a single DIII-D (3 s, ~150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04–0.1 MJ m–2.

  9. Technical overview of the millimeter-wave imaging reflectometer on the DIII-D tokamak (invited).

    PubMed

    Muscatello, C M; Domier, C W; Hu, X; Kramer, G J; Luhmann, N C; Ren, X; Riemenschneider, P; Spear, A; Tobias, B J; Valeo, E; Yu, L

    2014-11-01

    The two-dimensional mm-wave imaging reflectometer (MIR) on DIII-D is a multi-faceted device for diagnosing electron density fluctuations in fusion plasmas. Its multi-channel, multi-frequency capabilities and high sensitivity permit visualization and quantitative diagnosis of density perturbations, including correlation length, wavenumber, mode propagation velocity, and dispersion. The two-dimensional capabilities of MIR are made possible with 12 vertically separated sightlines and four-frequency operation (corresponding to four radial channels). The 48-channel DIII-D MIR system has a tunable source that can be stepped in 500 μs increments over a range of 56 to 74 GHz. An innovative optical design keeps both on-axis and off-axis channels focused at the cutoff surface, permitting imaging over an extended poloidal region. The integrity of the MIR optical design is confirmed by comparing Gaussian beam calculations to laboratory measurements of the transmitter beam pattern and receiver antenna patterns. Measurements are presented during the density ramp of a plasma discharge to demonstrate unfocused and focused MIR signals.

  10. Technical overview of the millimeter-wave imaging reflectometer on the DIII-D tokamak (invited)

    SciTech Connect

    Muscatello, C. M. Domier, C. W.; Hu, X.; Luhmann, N. C.; Ren, X.; Riemenschneider, P.; Spear, A.; Valeo, E.; Yu, L.; Kramer, G. J.; Tobias, B. J.

    2014-11-15

    The two-dimensional mm-wave imaging reflectometer (MIR) on DIII-D is a multi-faceted device for diagnosing electron density fluctuations in fusion plasmas. Its multi-channel, multi-frequency capabilities and high sensitivity permit visualization and quantitative diagnosis of density perturbations, including correlation length, wavenumber, mode propagation velocity, and dispersion. The two-dimensional capabilities of MIR are made possible with 12 vertically separated sightlines and four-frequency operation (corresponding to four radial channels). The 48-channel DIII-D MIR system has a tunable source that can be stepped in 500 μs increments over a range of 56 to 74 GHz. An innovative optical design keeps both on-axis and off-axis channels focused at the cutoff surface, permitting imaging over an extended poloidal region. The integrity of the MIR optical design is confirmed by comparing Gaussian beam calculations to laboratory measurements of the transmitter beam pattern and receiver antenna patterns. Measurements are presented during the density ramp of a plasma discharge to demonstrate unfocused and focused MIR signals.

  11. Nonlinear hybrid simulation of internal kink with beam ion effects in DIII-D

    NASA Astrophysics Data System (ADS)

    Shen, Wei; Fu, G. Y.; Tobias, Benjamin; Van Zeeland, Michael; Wang, Feng; Sheng, Zheng-Mao

    2015-04-01

    In DIII-D sawteething plasmas, long-lived (1,1) kink modes are often observed between sawtooth crashes. The saturated kink modes have two distinct frequencies. The mode with higher frequency transits to a fishbone-like mode with sufficient on-axis neutral beam power. In this work, hybrid simulations with the global kinetic-magnetohydrodynamic (MHD) hybrid code M3D-K have been carried out to investigate the linear stability and nonlinear dynamics of the n = 1 mode with effects of energetic beam ions for a typical DIII-D discharge where both saturated kink mode and fishbone were observed. Linear simulation results show that the n = 1 internal kink mode is unstable in MHD limit. However, with kinetic effects of beam ions, a fishbone-like mode is excited with mode frequency about a few kHz depending on beam pressure profile. The mode frequency is higher at higher beam power and/or narrower radial profile consistent with the experimental observation. Nonlinear simulations have been performed to investigate mode saturation as well as energetic particle transport. The nonlinear MHD simulations show that the unstable kink mode becomes a saturated kink mode after a sawtooth crash. With beam ion effects, the fishbone-like mode can also transit to a saturated kink mode with a small but finite mode frequency. These results are consistent with the experimental observation of saturated kink mode between sawtooth crashes.

  12. High heat flux Langmuir probe array for the DIII-D divertor platesa)

    NASA Astrophysics Data System (ADS)

    Watkins, J. G.; Taussig, D.; Boivin, R. L.; Mahdavi, M. A.; Nygren, R. E.

    2008-10-01

    Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m2 for 10 s exposures have been installed in the lower divertor target plates of the DIII-D tokamak. The 20 pyrolytic graphite probe tips have more than three times higher thermal conductivity and 16 times larger mass than the original DIII-D isotropic graphite probes. The probe tips have a fixed 12.5° surface angle to distribute the heat flux more uniformly than the previous 6 mm diameter domed collectors and a symmetric "rooftop" design to allow operation with reversed toroidal magnetic field. A large spring-loaded contact area improves heat conduction from each probe tip through a ceramic insulator into a cooled graphite divertor floor tile. The probe tips, brazed to molybdenum foil to ensure good electrical contact, are mounted in a ceramic tray for electrical isolation and reliable cable connections. The new probes are located 1.5 cm radially apart in a staggered arrangement near the entrance to the lower divertor pumping baffle and are linearly spaced 3 cm apart on the shelf above the in-vessel cryopump. Typical target plate profiles of Jsat, Te, and Vf with 4 mm spatial resolution are shown.

  13. Nonlinear hybrid simulation of internal kink with beam ion effects in DIII-D

    SciTech Connect

    Shen, Wei; Sheng, Zheng-Mao; Fu, G. Y.; Tobias, Benjamin; Zeeland, Michael Van; Wang, Feng

    2015-04-15

    In DIII-D sawteething plasmas, long-lived (1,1) kink modes are often observed between sawtooth crashes. The saturated kink modes have two distinct frequencies. The mode with higher frequency transits to a fishbone-like mode with sufficient on-axis neutral beam power. In this work, hybrid simulations with the global kinetic-magnetohydrodynamic (MHD) hybrid code M3D-K have been carried out to investigate the linear stability and nonlinear dynamics of the n = 1 mode with effects of energetic beam ions for a typical DIII-D discharge where both saturated kink mode and fishbone were observed. Linear simulation results show that the n = 1 internal kink mode is unstable in MHD limit. However, with kinetic effects of beam ions, a fishbone-like mode is excited with mode frequency about a few kHz depending on beam pressure profile. The mode frequency is higher at higher beam power and/or narrower radial profile consistent with the experimental observation. Nonlinear simulations have been performed to investigate mode saturation as well as energetic particle transport. The nonlinear MHD simulations show that the unstable kink mode becomes a saturated kink mode after a sawtooth crash. With beam ion effects, the fishbone-like mode can also transit to a saturated kink mode with a small but finite mode frequency. These results are consistent with the experimental observation of saturated kink mode between sawtooth crashes.

  14. On the formation and stability of long-lived impurity-ion snakes in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Delgado-Aparicio, L.; Sugiyama, L.; Granetz, R.; Gates, D.; Rice, J.; Reinke, M. L.; Bergerson, W.; Bitter, M.; Brower, D. L.; Fredrickson, E.; Gao, C.; Greenwald, M.; Hill, K.; Hubbard, A.; Irby, J.; Hughes, J. W.; Marmar, E.; Pablant, N.; Scott, S.; Wilson, R.; Wolfe, S.; Wukitch, S.

    2013-04-01

    Long-lived (1, 1) ‘snake’ modes were discovered nearly three decades ago, but basic questions regarding their formation, stability, and superb particle confinement—shown by surviving tens to hundreds of sawtooth cycles—have remained unanswered. High-resolution spectroscopic imaging diagnostics permit studies of heavy-impurity-ion snakes with unprecedented temporal and spatial resolution, making it possible to positively identify the SXR signals with specific ion charge states and to infer, for the first time, the perturbed impurity density, Zeff, and resistivity at the centre of these long-lived helical modes. The results show a new scenario for the formation of heavy-impurity-ion snakes, which can begin as a broad 1/1 kink asymmetry of the central impurity-ion density, that grows and undergoes a seamless transition to a large crescent-shaped helical island-like structure inside q < 1, with a regularly sawtoothing core. This type of formation departs strongly from the nonlinear island model based on a modified Rutherford equation proposed originally to describe the pellet-induced snakes and expanded further to account for the impurity effects (e.g. \\tilde{P}_rad and \\tilde{Z}_eff ). These new high-resolution observations show details of their evolution and the accompanying sawtooth oscillations that suggest important differences between the density and temperature dynamics, ruling out a purely pressure-driven process. Instead, many features arise naturally from nonlinear interactions in a 3D MHD model that separately evolves the plasma density and temperature.

  15. I-mode Plasmas with Combined ICRF Mode Conversion Flow Drive and Minority Heating on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lin, Y.; Reinke, M. L.; Rice, J. E.; Wukitch, S. J.; Granetz, R.; Greenwald, M.; Hubbard, A. E.; Marmar, E. S.; Podpaly, Y. A.; Porkolab, M.; Tsujii, N.; Wolfe, S.; Alcator C-Mod Team

    2011-10-01

    High performance I-mode plasmas have been obtained via ICRF mode conversion heating and flow drive, plus ICRF minority heating. The plasmas have reversed magnetic field (Bt0 ~ 5.1 T) and lower-single-null shape. External 3He is puffed to have nHe3/ne ~ 0.1 and the residual H level nH/ne is about 0.05. We use 50 MHz RF power (<=2.5 MW) for D(3He) mode conversion flow drive and heating, and 80 MHz RF power (<=2.5 MW) for D(H) minority heating. The obtained I-mode plasmas have high Te (<=8 keV) and large rotation (<=100 km/s). The rotation profile has a large shear at r/a ~ 0.9, which may help further enhance the plasma confinement. Because of the long slowdown time at high Te and relatively low density, a high energy H tail is generated. The tail also affects the sawtooth oscillation and produces large sawtooth crashes, which also trigger neoclassical tearing modes. The appearance of the NTMs often coincides with a slowdown of plasma rotation. The onset condition of these NTMs is at the low collisionality regime within the criterion established by studies on other tokamaks. Supported by USDoE award DE-FC02-99ER54512.

  16. Comparison of edge fluctuation measurements and energy and particle transport for I-mode plasmas on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Dominguez, A.; Marmar, E. S.; Terry, J. L.; Hughes, J. W.; Cziegler, I.; Ennever, P.; Golfinopoulos, T.; Hubbard, A. E.; Labombard, B.; Lipschultz, B.; Porkolab, M.; White, A. E.; Whyte, D. G.

    2011-10-01

    The I-mode regime features high energy confinement and low particle confinement, making it of interest both as a potential reactor regime as well as a physics case study for particle and energy channel separation. It has been observed that the I-mode is accompanied by two significant changes in edge fluctuations: The presence of a weakly coherent mode (WCM) at ~250 kHz and the suppression of broadband fluctuations at ~50-150 kHz during L-mode to I-mode transitions. Recently, experiments have been conducted which make use of D α emission measurements close to the last closed flux surface (LCFS) in order to estimate the ionization source profile and, consequently, the particle transport across the LCFS. In addition, estimates of the energy transport across the LCFS are acquired using the TRANSP code. Results and analysis will be presented studying the features of the edge fluctuations and their connection to the energy and particle transport across the LCFS. Supported by USDoE award DE-FC02-99ER54512.

  17. Characterization and performance of a field aligned ion cyclotron range of frequency antenna in Alcator C-Mod

    SciTech Connect

    Wukitch, S. J.; Garrett, M. L.; Ochoukov, R.; Terry, J. L.; Hubbard, A.; Labombard, B.; Lau, C.; Lin, Y.; Lipschultz, B.; Miller, D.; Reinke, M. L.; Whyte, D.; Collaboration: Alcator C-Mod Team

    2013-05-15

    Ion cyclotron range of frequency (ICRF) heating is expected to provide auxiliary heating for ITER and future fusion reactors where high Z metallic plasma facing components (PFCs) are being considered. Impurity contamination linked to ICRF antenna operation remains a major challenge particularly for devices with high Z metallic PFCs. Here, we report on an experimental investigation to test whether a field aligned (FA) antenna can reduce impurity contamination and impurity sources. We compare the modification of the scrape of layer (SOL) plasma potential of the FA antenna to a conventional, toroidally aligned (TA) antenna, in order to explore the underlying physics governing impurity contamination linked to ICRF heating. The FA antenna is a 4-strap ICRF antenna where the current straps and antenna enclosure sides are perpendicular to the total magnetic field while the Faraday screen rods are parallel to the total magnetic field. In principle, alignment with respect to the total magnetic field minimizes integrated E|| (electric field along a magnetic field line) via symmetry. A finite element method RF antenna model coupled to a cold plasma model verifies that the integrated E|| should be reduced for all antenna phases. Monopole phasing in particular is expected to have the lowest integrated E||. Consistent with expectations, we observed that the impurity contamination and impurity source at the FA antenna are reduced compared to the TA antenna. In both L and H-mode discharges, the radiated power is 20%–30% lower for a FA-antenna heated discharge than a discharge heated with the TA-antennas. However, inconsistent with expectations, we observe RF induced plasma potentials (via gas-puff imaging and emissive probes to be nearly identical for FA and TA antennas when operated in dipole phasing). Moreover, the highest levels of RF-induced plasma potentials are observed using monopole phasing with the FA antenna. Thus, while impurity contamination and sources are indeed reduced with the FA antenna configuration, the mechanism determining the SOL plasma potential in the presence of ICRF and its impact on impurity contamination and sources remains to be understood.

  18. Taming the ICRF Antenna - Plasma Edge Interaction via Novel Field-Aligned ICRF Antenna on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Wukitch, S. J.; Lin, Y.; Terry, J.; Hubbard, A.; Mumgaard, R. T.; Reinke, M. L.; the Alcator C-Mod Team

    2016-10-01

    Although ICRF is attractive for bulk plasma heating due to favorable wave propagation, ICRF antenna - edge plasma interaction remains a challenge. Recent experiments reveal that RF-induced potentials in the scrape-off layer and antenna impurity source are dependent on the power ratio between the inner and outer current staps, Pcent/Pout. Using a modified field aligned antenna, the transmission line network connected the center two straps at [0,pi] to one transmitter and the outer two straps another transmitter. This experiment was motivated by positive three strap antenna results from ASDEX-U. With -30 dB decoupling, we scanned Pcent/Pout from zero to greater than 1000. A minimum in the RF enhanced potential and local impurity source is observed for Pcent/Pout greater than 1 and less than 4 with a gradual rise in impurity source for Pcent/Pout greater than 4. This minimum correlates where the image currents in the antenna limiters are expected to be smallest. We also tested antenna operation in [0,0,pi,pi] antenna phasing and found excessive local impurity production despite the antenna being field aligned. This antenna phasing excites low k and potentially have higher coupling. Latest results and analysis will be presented Supported by US DOE Award DE-FC02-99ER54512.

  19. Measurements of electron temperature profiles on Alcator C-Mod using a novel energy-resolving x-ray camera

    NASA Astrophysics Data System (ADS)

    Maddox, J.; Delgado, L.; Pablant, N.; Hill, K. W.; Bitter, M.; Efthimion, P.; Rice, J.

    2015-11-01

    The most common electron temperature diagnostics, Thomson Scattering (TS) and Electron Cyclotron Emission (ECE), both require large diagnostic footprints and expensive optics. Another electron temperature diagnostic is the Pulse-Height-Analysis (PHA) system, which derives the electron temperature from the x-ray bremsstrahlung continuum. However, the main disadvantage of the PHA method is poor temporal resolution of the Si(Li) diode detectors. This paper presents a novel x-ray pinhole camera, which uses a pixilated Pilatus detector that allows single photon counting at a rate 2MHz per pixel and the setting of energy thresholds. The detector configuration is optimized by Shannon-sampling theory, such that spatial profiles of the x-ray continuum intensity can be obtained simultaneously for different energies, in the range from 4 to 16 keV. The exponential-like dependence of the x-ray intensity with photon energies is compared with a model describing the Be filter, attenuation in air, and detector efficiency, as well as different sets of energy thresholds. Electron temperature measurements are compared with TS and ECE measurements. This work was supported by the US DOE Contract No.DE-AC02-09CH11466 and the DoE Summer Undergraduate Laboratory Internship (SULI) program.

  20. Measurement and simulation of ICRF wave intensity with a recalibrated phase contrast imaging diagnostic on 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.

    2015-12-10

    Waves in the ion cyclotron range of frequencies (ICRF) are one of the major tools to heat fusion plasmas. Full-wave simulations are essential to predict the wave propagation and absorption quantitatively, and it is important that these codes be validated against actual experimental measurements. In this work, the absolute intensity of the ICRF waves previously measured with a phase contrast imaging diagnostic was recalibrated and compared once more with full-wave predictions. In the earlier work, significant discrepancies were found between the measured and the simulated mode converted wave intensity [N. Tsujii et al., Phys. Plasmas 19, 082508]. With the new calibration of the detector array, the measured mode converted wave intensity is now in much better agreement with the full-wave predictions. The agreement is especially good for comparisons performed close to the antenna.

  1. DIII-D research operations annual report to the U.S. Department of Energy, October 1, 1995--September 30, 1996

    SciTech Connect

    1997-07-01

    The mission of the DIII-D research program is to advance fusion energy science understanding and predictive capability and to improve and optimize the tokamak concept. A long term goal remains to integrate these products into a demonstration of high confinement, high plasma pressure (plasma {beta}), sustained long pulse operation with fusion power plant relevant heat and particle handling capability. The DIII-D program is a world recognized leader in tokamak concept improvement and a major contributor to the physics R and D needs of the International Thermonuclear Experimental Reactor (ITER). The scientific objectives of the DIII-D program are given in Table 1-2. The FY96 DIII-D research program was highly successful, as described in this report. A moderate sized tokamak, DIII-D is a world leader in tokamak innovation with exceptional performance, measured in normalized parameters.

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

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

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

  3. Metallurgical bonding development of V 4Cr 4Ti alloy for the DIII-D radiative divertor program

    NASA Astrophysics Data System (ADS)

    Smith, J. P.; Johnson, W. R.; Trester, P. W.

    1998-10-01

    General Atomics (GA), in conjunction with the Department of Energy's (DOE) DIII-D Program, is carrying out a plan to utilize a vanadium alloy in the DIII-D tokamak as part of the DIII-D Radiative Divertor (RD) upgrade. The V-4Cr-4Ti alloy has been selected in the U.S. as the leading candidate vanadium alloy for fusion applications. This alloy will be used for the divertor fabrication. Manufacturing development with the V-4Cr-4Ti alloy is a focus of the DIII-D RD Program. The RD structure, part of which will be fabricated from V-4Cr-4Ti alloy, will require many product forms and types of metal/metal bonded joints. Metallurgical bonding methods development on this vanadium alloy is therefore a key area of study by GA. Several solid-state (non-fusion weld) and fusion weld joining methods are being investigated. To date, GA has been successful in producing ductile, high-strength, vacuum leak-tight joints by all of the methods under investigation. The solid-state joining was accomplished in air, i.e., without the need for a vacuum or inert gas environment to prevent interstitial impurity contamination of the V-4Cr-4Ti alloy.

  4. Metallurgical Bonding Development of V-4Cr-4Ti Alloy for the DIII-D Radiative Divertor Program

    SciTech Connect

    Smith, J.P.; Johnson, W.R.; Trester, P.W.

    1998-06-01

    General Atomics (GA), in conjunction with the Department of Energy`s (DOE) DIII-D Program, is carrying out a plan to utilize a vanadium alloy in the DIII-D tokamak as part of the DIII-D Radiative Divertor (RD) upgrade. The V-4Cr-4Ti alloy has been selected in the U.S. as the leading candidate vanadium alloy for fusion applications. This alloy will be used for the divertor fabrication. Manufacturing development with the V-4Cr-4Ti alloy is a focus of the DIII-D RD Program. The RD structure, part of which will be fabricated from V-4Cr-4Ti alloy, will require many product forms and types of metal/metal bonded joints. Metallurgical bonding methods development on this vanadium alloy is therefore a key area of study by GA. Several solid state (non-fusion weld) and fusion weld joining methods are being investigated. To date, GA has been successful in producing ductile, high strength, vacuum leak tight joints by all of the methods under investigation. The solid state joining was accomplished in air, i.e., without the need for a vacuum or inert gas environment to prevent interstitial impurity contamination of the V-4Cr-4Ti alloy.

  5. COMPLETE SUPPRESSION OF THE M=2/N-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D

    SciTech Connect

    PETTY,CC; LAHAYE,LA; LUCE,TC; HUMPHREYS,DA; HYATT,AW; PRATER,R; STRAIT,EJ; WADE,MR

    2003-03-01

    A271 COMPLETE SUPPRESSION OF THE M=2/N-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D. The first suppression of the important and deleterious m=2/n=1 neoclassical tearing mode (NTM) is reported using electron cyclotron current drive (ECCD) to replace the ''missing'' bootstrap current in the island O-point. Experiments on the DIII-D tokamak verify the maximum shrinkage of the m=2/n=1 island occurs when the ECCD location coincides with the q = 2 surface. The DIII-D plasma control system is put into search and suppress mode to make small changes in the toroidal field to find and lock onto the optimum position, based on real time measurements of dB{sub {theta}}/dt, for complete m=2/n=1 NTM suppression by ECCD. The requirements on the ECCD for complete island suppression are well modeled by the modified Rutherford equation for the DIII-D plasma conditions.

  6. Advances in understanding quiescent H-mode plasmas in DIII-D

    SciTech Connect

    Burrell, K.H.; West, W.P.; Gohil, P.; Greenfield, C.M.; Groebner, R.J.; Hyatt, A.W.; Kaplan, D.H.; Lao, L.L.; Leonard, A.W.; Osborne, T.H.; Snyder, P.B.; Thomas, D.M.; Strait, E.J.; Doyle, E.J.; Rhodes, T.L.; Wang, G.; Zeng, L.; Austin, M.E.; Casper, T.A.; Jayakumar, R.J.

    2005-05-15

    Recent QH-mode research on DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] has used the peeling-ballooning modes model of edge magnetohydrodynamic stability as a working hypothesis to organize the data; several predictions of this theory are consistent with the experimental results. Current ramping results indicate that QH modes operate near the edge current limit set by peeling modes. This operating point explains why QH mode is easier to get at lower plasma currents. Power scans have shown a saturation of edge pressure with increasing power input. This allows QH-mode plasmas to remain stable to edge localized modes (ELMs) to the highest powers used in DIII-D. At present, the mechanism for this saturation is unknown; if the edge harmonic oscillation (EHO) is playing a role here, the physics is not a simple amplitude dependence. The increase in edge stability with plasma triangularity predicted by the peeling-ballooning theory is consistent with the substantial improvement in pedestal pressure achieved by changing the plasma shape from a single null divertor to a high triangularity double null. Detailed ELITE calculations for the high triangularity plasmas have demonstrated that the plasma operating point is marginally stable to peeling-ballooning modes. Comparison of ELMing, coinjected and quiescent, counterinjected discharges with the same shape, current, toroidal field, electron density, and electron temperature indicates that the edge radial electric field or the edge toroidal rotation are also playing a role in edge stability. The EHO produces electron, main ion, and impurity particle transport at the plasma edge which is more rapid than that produced by ELMs under similar conditions. The EHO also decreases the edge rotation while producing little change in the edge electron and ion temperatures. Other edge electromagnetic modes also produce particle

  7. Calculation of stochastic broadening in real space due to noise and field errors in the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Brodsky, Lisa; Punjabi, Alkesh; Ali, Halima

    2008-11-01

    The equilibrium EFIT data for the DIII-D shot 115467 at 3000 ms is used to construct the equilibrium generating function for magnetic field line trajectories in the DIII-D tokamak in natural canonical coordinates. A canonical transformation is used to construct an area-preserving map for field line trajectories in the natural canonical coordinates in the DIII-D. Maps in natural canonical coordinates have the advantage that natural canonical coordinates can be inverted to calculate real space coordinates (R,Z,φ), and there is no problem in crossing the separatrix. This is not possible for magnetic coordinates. This map is applied to calculate stochastic broadening due to magnetic noise and field errors in the DIII-D. Mode numbers for noise + field errors are (m,n)=(3,1), (4,1), (6,2), (7,2), (8,2), (9,3), (10,3), (11,3), (12,3). The common amplitude δ is varied from 0.8X10-5 to 2.0X10-5. Preliminary results suggest that the width of stochastic layer from noise and field errors in the DIII-D varies from about 7 to 16 cm near X-point, and about 0.6 to 3% of poloidal flux is lost from inside ideal separatrix. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

  8. Real-time protection of the ohmic heating coil force limits in DIII-D

    SciTech Connect

    Broesch, J.D.; Scoville, J.T.; Hyatt, A.W.; Coon, R.M.

    1997-11-01

    The maximum safe operating limits of the DIII-D tokamak are determined by the force produced in the ohmic heating coil and the toroidal field coil during a plasma pulse. This force is directly proportional to the product of the current in the coils. Historically, the current limits for each coil were set statically before each pulse without regard for the time varying nature of the currents. In order to allow the full time-dependent capability of the ohmic coil to be used, a system was developed for monitoring the product of the currents dynamically and making appropriate adjustments in real time. This paper discusses the purpose, implementation, and results of this work.

  9. Progress in snowflake divertor research in DIII-D, NSTX and NSTX-U

    NASA Astrophysics Data System (ADS)

    Soukhanovskii, V. A.; Allen, S.; Fenstermacher, M.; Izacard, O.; Lasnier, C.; Makowski, M.; McLean, A.; Myer, W.; Ryutov, D.; Scotti, F.; Eldon, D.; Kolemen, E.; Vail, P.; Canal, G.; Groebner, R.; Hyatt, A.; Leonard, A.; Osborne, T.; Bell, R.; Diallo, A.; Gerhardt, S.; Kaye, S.; Leblanc, B.; Menard, J.; Podesta, M.

    2016-10-01

    Recent snowflake (SF) divertor DIII-D experiments focused on divertor heat transport under attached and radiative divertor conditions, incl 1-understanding of increased scrape-off layer width in SF-plus configuration at lower densities; 2-particle, heat and radiation distribution in the SF divertor with CD4 seeding. NSTX data was analyzed to understand the link between SF divertor and ELM (de)stabilization with and without CD4 seeding and lithium conditioning. Prep for SF divertor experiments in NSTX-U include 1-equilibria modeling with ISOLVER code using various sets of divertor coils and L- and H-mode plasma scenarios; 2-transport and impurity radiation modeling with UEDGE code; 3-new diagnostics (ie-a 100-200 kHz camera for null-region mode observations). Supported by DOE under DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FC02-04ER54698.

  10. Identifying the location of the OMP separatrix in DIII-D using power accounting

    DOE PAGES

    Stangeby, Peter C.; Canik, John M.; Elder, J. D.; ...

    2015-08-07

    In order to identify reliable scalings for the scrape-off layer (SOL) power width it is necessary to know the location of the separatrix in divertor tokamaks as accurately as possible, specifically its location at the outside midplane (OMP) the standard reference location. Two methods are described which use power accounting to improve the accuracy of identifying the location of the OMP separatrix. The first uses the infrared-measured deposited power profile at the outer target as the primary input, the 'more » $$P_{{\\rm SOL}}^{{\\rm exhaust}}$$ method'. The second uses the measured power input to the SOL, obtained by subtracting the power radiated from inside the separatrix from the total heating power, the ' $$P_{{\\rm SOL}}^{{\\rm input}}$$ method'. Furthermore, these two power accounting methods are illustrated with the examples of 21 H-mode DIII-D discharges.« less

  11. Identifying the location of the OMP separatrix in DIII-D using power accounting

    SciTech Connect

    Stangeby, Peter C.; Canik, John M.; Elder, J. D.; Lasnier, C. J.; Leonard, A. W.; Eldon, D.; Makowski, M. A.; Osborne, T. H.; Grierson, Brian A.

    2015-08-07

    In order to identify reliable scalings for the scrape-off layer (SOL) power width it is necessary to know the location of the separatrix in divertor tokamaks as accurately as possible, specifically its location at the outside midplane (OMP) the standard reference location. Two methods are described which use power accounting to improve the accuracy of identifying the location of the OMP separatrix. The first uses the infrared-measured deposited power profile at the outer target as the primary input, the ' $P_{{\\rm SOL}}^{{\\rm exhaust}}$ method'. The second uses the measured power input to the SOL, obtained by subtracting the power radiated from inside the separatrix from the total heating power, the ' $P_{{\\rm SOL}}^{{\\rm input}}$ method'. Furthermore, these two power accounting methods are illustrated with the examples of 21 H-mode DIII-D discharges.

  12. Properties of toroidal Alfvén eigenmode in DIII-D plasma

    SciTech Connect

    Wang, Zhixuan; Lin, Zhihong Holod, Ihor; Heidbrink, W. W.; Deng, Wenjun; Xiao, Y.; Zhang, H.; Zhang, W.; Van Zeeland, M.

    2015-02-15

    Linear properties of the toroidal Alfvén eigenmode (TAE) excited by energetic particles (EP) in a DIII-D tokamak experiment have been studied in global gyrokinetic particle simulations treating self-consistently kinetic effects of EP, thermal ions, and electrons. Simulation results of the TAE frequency and mode structure agree very well with the experimental measurements. The non-perturbative EP contribution induces a radial localization of the TAE mode structure, a break-down of mode radial symmetry, as well as a frequency dependence on the toroidal mode number. The simulations further demonstrate the dependence of the growth rate and mode structure on EP pressure gradients. The in-out asymmetry of the mode structure and the experimental identification of the poloidal harmonics have also been clarified.

  13. Design Concepts For A Long Pulse Upgrade For The DIII-D Fast Wave Antenna Array

    SciTech Connect

    Ryan, Philip Michael; Baity Jr, F Wallace; Caughman, John B; Goulding, Richard Howell; Hosea, J.; Greenough, Nevell; Nagy, Alex; Pinsker, R.; Rasmussen, David A

    2009-01-01

    A goal in the 5-year plan for the fast wave program on DIII-D is to couple a total of 3.6 MW of RF power into a long pulse, H-mode plasma for central electron heating. The present short-pulse 285/300 antenna array would need to be replaced with one capable of at least 1.2 MW, 10 s operation at 60 MHz into an H-mode (low resistive loading) plasma condition. The primary design under consideration uses a poloidally-segmented strap (3 sections) for reduced strap voltage near the plasma/Faraday screen region. Internal capacitance makes the antenna structure self-resonant at 60 MHz, strongly reducing peak E-fields in the vacuum coax and feed throughs.

  14. IMPROVED CCD DETECTORS FOR HIGH SPEED, CHARGE EXCHANGE SPECTROSCOPY STUDIES ON THE DIII-D TOKAMAK

    SciTech Connect

    K.H. BURRELL; P. GOHIL; R.J. GROEBNER; D.H. KAPLAN; D.G. NILSON; J.I. ROBINSON; D.M. THOMAS

    2002-06-01

    The tokamak plasmas utilized in magnetic fusion research provide access to some of the longest duration, highest temperature plasmas on earth. For example, on the DIII-D tokamak, plasmas lasting up to 10 seconds have been produced with ion temperatures up to 27 keV and electron temperatures up to 15 keV in toroidal plasmas with 1.7 m major radius, 0.65 m half width and 2.5 m vertical height. Typical particle densities in these plasmas are in the range of 2.0 x 10{sup 19} m{sup -3} through 2.0 x 10{sup 20} m{sup -3}. Spectroscopic measurements of line radiation from highly ionized atoms provide one of the key techniques for determining the plasma parameters in tokamak discharges.

  15. Impurity mixing and radiation asymmetry in massive gas injection simulations of DIII-D

    SciTech Connect

    Izzo, V. A.

    2013-05-15

    Simulations of neon massive gas injection into DIII-D are performed with the 3D MHD code NIMROD. The poloidal and toroidal distribution of the impurity source is varied. This report will focus on the effects of the source variation on impurity mixing and radiated power asymmetry. Even toroidally symmetric impurity injection is found to produce asymmetric radiated power due to asymmetric convective heat flux produced by the 1/1 mode. When the gas source is toroidally localized, the phase relationship between the mode and the source location is important, affecting both radiation peaking and impurity mixing. Under certain circumstances, a single, localized gas jet could produce better radiation symmetry during the disruption thermal quench than evenly distributed impurities.

  16. Vertical-viewing electron cyclotron emission diagnostic for the DIII-D tokamak

    SciTech Connect

    James, R.; Janz, S.; Ellis, R.; Boyd, D.; Lohr, J.

    1988-08-01

    The vertical-viewing electron cyclotron emission diagnostic on DIII-D will be used to assess the nonthermal electron distributions resulting from electron cyclotron heating and electron cyclotron current drive experiments. Electron cyclotron emission along a vertical chord is collected using an ellipsoidal focusing mirror and retroreflector (the latter to minimize wall reflections). The emission is then transported approx.20 m using a quasioptical transport system composed of eight lenses and three mirrors, and detected between the 2nd and the 10th harmonics by a fast-scanning (40-Hz) Michelson interferometer. The entire system has been aligned using a Gaussian beam simulator and absolutely calibrated in situ using a cold liquid-nitrogen bath. Details of the design, installation, and calibration will be discussed.

  17. Internal Mode Structure of Resonant Field Amplification in DIII-D

    NASA Astrophysics Data System (ADS)

    Lanctot, M. J.; Navratil, G.; Reimerdes, H.; Bogatu, I. N.; in, Y.; Chu, M. S.; Garofalo, A. M.; Jackson, G. L.; La Haye, R. J.; Strait, E. J.; Turnbull, A. D.; Liu, Y. Q.; Okabayashi, M.; Solomon, W. M.

    2008-11-01

    The sensitivity of high-β plasmas to error fields is caused by a paramagnetic plasma response to error fields with a topology that is resonant with the structure of weakly-damped resistive wall modes (RWM), a phenomenon referred to as resonant field amplification (RFA) [1]. The RFA has been driven in DIII-D H-mode plasmas by applying slowly-rotating, low-n magnetic fields with a set of 12 coils located inside the vacuum vessel. Measurements of the RFA mode structure have been obtained using a pair of soft x-ray photodiode cameras. A virtual diagnostic has been developed to compare the measurements to the eigenfunctions for the free boundary external kink and the RWM, which were calculated using the stability codes GATO and MARS-F. Details of the analysis will be presented. 6pt [1] A.H. Boozer, Phys. Rev. Lett. 86, 5059 (2001).

  18. Modeling and Simulation of Resistive Wall Mode Control In DIII-D

    NASA Astrophysics Data System (ADS)

    Walker, M. L.; Humphreys, D. A.; Jensen, T. H.; Leuer, J. A.; Nerem, A.; Strait, E. J.; Garofalo, A. M.

    2001-10-01

    Detailed dynamic response models have been developed for all relevant subsystems comprising the DIII-D resistive wall mode (RWM) closed loop control system. These include the switching power amplifiers (SPA), digital plasma control system (PCS), acquisition and control circuitry, and a fully toroidal model of plasma/vessel dynamics based on specification of the marginal wall position from stability codes such as GATO and DCON. These models have been validated with experimental data, including open-loop excitation of the SPA, PCS, and vacuum vessel dynamic responses, and measurement of the growth rate and mode structure of the unstable plasma. These models are incorporated into a closed-loop control simulation to investigate the control limitations which are due to realistic power supply responses. Consequences of and approaches to the intrinsically multivariable RWM control problem are also investigated.

  19. Optimization of Feedback Control Coils for Resistive Wall Mode Stabilization in DIII-D

    NASA Astrophysics Data System (ADS)

    Bialek, J.; Boozer, A. H.; Garofalo, A. M.; Mauel, M. E.; Navratil, G. A.; Turnbull, A. D.

    1999-11-01

    Recent experiments in DIII--D on Resistive Wall Mode (RWM) stabilization with active feedback have been very promising. We investigated extensions to the sensor and control coil set that would further improve RWM stabilization. The VALEN computer code models the RWM as an equivalent current distribution on the unperturbed plasma boundary which duplicates the plasma external magnetic field of the mode, as calculated by GATO. This surface current determines the plasma interaction with all conducting structures. In three dimensions the VALEN code models the unstable plasma, passive structure, proposed sensors, and proposed control coils together with the control logic. The problem may be examined as a transient simulation, or for a linear power supply model, as an eigenvalue calculation. A summary of the configurations examined and their predicted effectiveness will be presented.

  20. Simulating the ITER Plasma Startup Scenario in the DIII-D Tokamak

    SciTech Connect

    Jackson, G; Casper, T; Luce, T; Humphreys, D; Ferron, J; Hyatt, A; Petrie, T; West, W

    2008-10-13

    DIII-D experiments have investigated ITER startup scenarios, including an initial phase where the plasma was limited on low field side (LFS) poloidal bumper limiters. Both the original ITER 'small-bore' (constant q{sub 95}) startup and a 'large-bore' lower internal inductance (l{sub i}) startup have been simulated. In addition, l{sub i} feedback control has been tested with the goal of producing discharges at the ITER design value, l{sub i}(3) = 0.85. These discharges have been simulated using the Corsica free boundary equilibrium code. High performance hybrid scenario discharges ({beta}{sub N} = 2.8, H{sub 98,y2} = 1.4) and ITER H-mode baseline discharges ({beta}{sub N} > 1.6, H{sub 98,y2} = 1-1.2) have been obtained experimentally in an ITER similar shape after the ITER-relevant startup.

  1. Modeling Tokamak Discharges during Startup in DIII-D and Predictions for ITER

    NASA Astrophysics Data System (ADS)

    Budny, R. V.; Andre, R.; Kessel, C. E.; McCune, D.; Jackson, G. L.; Humphreys, D. A.; Luce, T. C.; Politzer, P. A.

    2009-11-01

    The PTRANSP code is being developed for improving predictions of ITER and future Tokamaks [1]. Important goals are predicting plasma performance as well as safe and efficient startup and termination. PTRANSP is being tested on experiments in DIII-D to explore and optimize plasma startup and termination, and to benchmark of transport. An important aspect is to model the boundary accurately since parameters such as the flux consumption and induction li can vary sensitively on the boundary shape. PTRANSP is being modified to allow for direct input of the flux in R, Z space (from EFIT) instead of using Fourier moments of the boundary. PTRANSP will write quantities such as the inductance and surface voltage at the separatrix surface as well as the flux boundary usually used in TRANSP. The predictive capabilities are being improved also. These include models for predicting temperatures, densities, and toroidal momentum. Predictions for ITER will be discussed. [1] R.V. Budny, Nuclear Fusion 49 (2009) 085008.

  2. Simulations of DIII-D Sawtooth Oscillations Using Theory-Based Transport and Sawtooth Models

    NASA Astrophysics Data System (ADS)

    Jeon, Y. M.; Li, G.; Ren, Q.; Guo, W.; Lao, L. L.; St. John, H. E.; Chu, M. S.; Prater, R.; Park, J. M.

    2007-11-01

    Development and validation of a predictive sawtooth model is an important research topic for present-day tokamaks and ITER. Analysis using ONETWO to model the DIII-D sawtooth behavior due to the interactions between FW and NBI fast ions, predicts the observed reduction in the axis safety factor q0 due to current profile evolution. Preliminary analysis indicates that the predicted drops in q0 within a sawtooth cycle follow closely the experimental values from the EFIT code using MSE data. In ONETWO simulations, the evolution of q0 within a sawtooth period is modeled with neoclassical resistivity and the experimental density and temperature profiles in two neighboring giant sawtooth cycles. Initial analysis using the Kadomtsev sawtooth model indicates that the sawtooth crash can be qualitatively reproduced with an appropriately chosen triggering parameter. The Porcelli sawtooth model is being implemented into ONETWO to more comprehensively predict the sawtooth crash.

  3. Error Field Correction in DIII-D Ohmic Plasmas With Either Handedness

    SciTech Connect

    Park, Jong-Kyu; Schaffer, Micahel J.; La Haye, Robert J.; Scoville, Timothy J.; Menard, Jonathon E.

    2011-05-16

    Error field correction results in DIII-D plasmas are presented in various configurations. In both left-handed and right-handed plasma configurations, where the intrinsic error fields become different due to the opposite helical twist (handedness) of the magnetic field, the optimal error correction currents and the toroidal phases of internal(I)-coils are empirically established. Applications of the Ideal Perturbed Equilibrium Code to these results demonstrate that the field component to be minimized is not the resonant component of the external field, but the total field including ideal plasma responses. Consistency between experiment and theory has been greatly improved along with the understanding of ideal plasma responses, but non-ideal plasma responses still need to be understood to achieve the reliable predictability in tokamak error field correction.

  4. Progress in the Development of an Integrated Modeling Tool to Support DIII-D and EAST

    NASA Astrophysics Data System (ADS)

    Ren, Q.; Lao, L. L.; Chu, M. S.; St. John, H. E.; Abla, G.; Collier, A.; Prater, R.; Park, J. M.; Li, G.; Guo, W.; Pan, C.; Srinivasan, R.; Worrall, M.

    2009-11-01

    Recent progress in the development of the IMFIT Integrated modeling tool is presented. The goal of IMFIT is to develop a modern and efficient integrated modeling platform to support DIII-D and EAST research, including the capability to simulate the behavior of tokamak discharges. Recent progress includes user-friendly and Python-based GUIs with multi-links to equilibrium, transport, and stability codes to facilitate modeling and analysis, and EFIT F90 upgrade with dynamic memory allocation and MPI option to support multiple devices and grid sizes. Through the GUI, straightforward analysis for kinetic EFIT reconstruction is made available. Ongoing developments include design of efficient algorithms to support interactions among physics modules such as EFIT/ONETWO/TGLF coupling for scenario development and transport flux analysis and EFIT/PEST3/TORAY coupling for modeling of tearing mode stability. Details will be presented.

  5. Far Scrape-Off Layer and Near Wall Plasma Studies in DIII-D

    SciTech Connect

    Rudakov, D; Boedo, J; Moyer, R; Brooks, N; Doerner, R; Evans, T; Fenstermacher, M; Groth, M; Hollmann, E; Krashennijov, S; Lasnier, C; Mahdavi, M; McKee, G; McLean, A; Stangeby, P; Wampler, W; Watkins, J; West, W; Whyte, D; Wong, C

    2004-12-03

    Far scrape-off layer (SOL) plasma parameters in DIII-D depend strongly on the discharge density and confinement regime. In L-mode, cross-field transport increases with the average discharge density and elevates the far SOL density, thus increasing plasma-wall contact. Far SOL density near the low field side (LFS) of the main chamber wall also increases with decreasing plasma current and with decreasing outer wall gap. In H-mode, between edge localized modes (ELMs), plasma-wall contact is weaker than in L-mode. During ELMs plasma fluxes to the LFS wall increase to, or above the L-mode levels. A large fraction of the net cross-field fluxes is convected through the SOL by large amplitude intermittent transport events. In high density L-mode and during ELMs in H-mode, intermittent events propagate all the way to the LFS wall and may cause sputtering.

  6. Benchmarking and Optimizing Techniques for Inverting Images of DIII-D Soft X-Ray Emissions

    NASA Astrophysics Data System (ADS)

    Chandler, E.; Unterberg, E. A.; Shafer, M. W.; Wingen, A.

    2012-10-01

    A tangential 2-D soft x-ray (SXR) imaging system is installed on DIII-D to directly measure the 3-D magnetic topology at the plasma edge. This diagnostic allows the study of the plasma SXR emissivity at time resolutions >=,0 ms and spatial resolutions ˜1 cm. Extracting 3-D structure from the 2-D image requires the inversion of large ill-posed matrices - a ubiquitous problem in mathematics. The goal of this work is to reduce the memory usage and computational time of the inversion to a point where image inversions can be processed between shots. We implement the Phillips-Tikohnov and Maximum Entropy regularization techniques on a parallel GPU processor. To optimize the memory demands of computing these matrixes, effects of reducing the inversion grid size and binning images are analyzed and benchmarked. Further benchmarking includes a characterization of the final image quality (with respect to numerical and instrumentation noise).

  7. DIII-D accomplishments and plans in support of fusion next steps

    SciTech Connect

    Buttery, R. J; Eidietis, N.; Holcomb, C.; Haye, R. J. La; Leonard, A.; Nazikian, R.; Solomon, W. M.; Baylor, L.; Burrell, K; Garofalo, A.; Jackson, G.

    2013-06-01

    DIII-D is using its flexibility and diagnostics to address the critical science required to enable next step fusion devices. We have adapted operating scenarios for ITER to low torque and are now being optimized for transport. Three ELM mitigation scenarios have been developed to near-ITER parameters. New control techniques are managing the most challenging plasma instabilities. Disruption mitigation tools show promising dissipation strategies for runaway electrons and heat load. An off axis neutral beam upgrade has enabled sustainment of high βN capable steady state regimes. Divertor research is identifying the challenge, physics and candidate solutions for handling the hot plasma exhaust with notable progress in heat flux reduction using the snowflake configuration. Our work is helping optimize design choices and prepare the scientific tools for operation in ITER, and resolve key elements of the plasma configuration and divertor solution for an FNSF.

  8. DIII-D accomplishments and plans in support of fusion next steps

    DOE PAGES

    Buttery, R. J; Eidietis, N.; Holcomb, C.; ...

    2013-06-01

    DIII-D is using its flexibility and diagnostics to address the critical science required to enable next step fusion devices. We have adapted operating scenarios for ITER to low torque and are now being optimized for transport. Three ELM mitigation scenarios have been developed to near-ITER parameters. New control techniques are managing the most challenging plasma instabilities. Disruption mitigation tools show promising dissipation strategies for runaway electrons and heat load. An off axis neutral beam upgrade has enabled sustainment of high βN capable steady state regimes. Divertor research is identifying the challenge, physics and candidate solutions for handling the hot plasmamore » exhaust with notable progress in heat flux reduction using the snowflake configuration. Our work is helping optimize design choices and prepare the scientific tools for operation in ITER, and resolve key elements of the plasma configuration and divertor solution for an FNSF.« less

  9. Carbon Ion Flow Measurements in DIII-D Divertors Using Coherence Imaging Spectroscopy

    NASA Astrophysics Data System (ADS)

    Allen, S. L.; Meyer, W. H.; Porter, G. D.; Howard, J.

    2013-10-01

    New, single-crystal imaging interferometers along with improved relay optics have been installed in the upper and lower DIII-D divertors. These provide improved images of the Doppler shift and thereby flow of CIII (465 nm) ions. An improved in-situ calibration technique has been implemented, providing zero velocity reference images and measured spectrometer phase vs wavelength. The temperature resolution of the system has been greatly improved, resulting in a stable wavelength calibration. Image intensified cameras have made possible measurements of flow during ELMS and in the non-active divertor. Streamlined data analysis has been used to look for flow trends. In general, we see flow in opposite directions on the inner and outer scrape-off layers in the divertor. Supported in part by the US Department of Energy under DE-AC52-07NA27344 and DE-FC02-04ER54698.

  10. Investigation of Millimeter Wave Bursts at ECE Frequencies in DIII-D Plasmas

    NASA Astrophysics Data System (ADS)

    Wu, S. M.; Austin, M. E.

    2005-10-01

    Intense bursts of narrowband millimeter wave radiation have been observed in low density H-mode and QH-mode plasmas in the DIII-D tokamak. These bursts occur in the range of second harmonic electron cyclotron frequencies and have bandwidths of 1 to 4 GHz. The narrow frequency width suggests that the bursting is not classic runaway electron emission which typically has a bandwidth of 10 GHz or more. ECE bursting has been observed in three different conditions. The burst occurs simultaneously with edge-localized mode (ELM) precursors, edge harmonic oscillations, or plasma disruption precursors. In the precursor cases the bursts precede the collapse phase and are coincident with similar types of MHD events. The data will be compared to models of stimulated and scattered emission for electrons in the high energy tail of the distribution.

  11. ECE-Imaging of the H-mode Pedestal on DIII-D

    NASA Astrophysics Data System (ADS)

    Tobias, B. J.; Domier, C. W.; Luhmann, N. C., Jr.; Austin, M. E.

    2012-10-01

    Forward modeling of ECE originating near the edge of DIII-D plasmas has improved our understanding of radiation properties in this region and enabled interpretation of ECE-Imaging and radiation temperature profiles of the H-mode pedestal. A variety of coherent edge modes have been imaged, revealing the nature of the edge harmonic oscillation (EHO) present in QH-mode plasmas, as well as directly diagnosing the plasma response to RMP fields applied for ELM suppression. Attempts to image the most fleeting aspects of ELMs in low density H-mode discharges have revealed intense bursts of millimeter wave radiation. Initiating during ELM precursor oscillations and prevalent at ITER relevant collisionality, these bursts appear to be coherent, stimulated emission from thermal electrons interacting with a non-axisymmetric perturbation of the plasma boundary.

  12. OEDGE modeling of DIII-D density scan discharges leading to detachment

    NASA Astrophysics Data System (ADS)

    Elder, J. D.; Stangeby, P. C.; Bray, B. D.; Brooks, N.; Leonard, A. W.; McLean, A. G.; Unterberg, E. A.; Watkins, J. G.

    2015-08-01

    The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for Dα, Dβ and Dγ for values of Te and ne characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to Te.

  13. GYROTRON POWER BALANCE BASED ON CALORIMETRIC MEASUREMENTS IN THE DIII-D ECH SYSTEM

    SciTech Connect

    GORELOV,I.A; LOHR,J.M; BAITY,JR.,F.W; CAHALAN,P; CALLIS,R.W; PONCE,D; CHIU,H.K

    2003-10-01

    OAK-B135 A powerful microwave system operating at the second harmonic of the electron cyclotron frequency on the DIII-D tokamak was upgraded up to six assemblies of 110 GHz gyrotrons in 2003. three Gycom gyrotrons nominally generate 750 kW for 2 s pulses, with the pulse length limit resulting from the peak temperature allowed on the boron nitride rf output window. Three Communications and Power Industries (CPI) gyrotrons with diamond windows have been recently installed and have been tested to 0.9-1.0 MW for 5 s pulses. Heat loading on internal parts of the gyrotrons, the matching optics unit and the dummy loads is measured calorimetrically. This paper discusses the calorimetry system and calorimetric measurements of gyrotron performance.

  14. Improved Measurements of Injected Electron Cyclotron Power in DIII-D

    NASA Astrophysics Data System (ADS)

    Gorelov, I. A.; Lohr, J.; Ponce, D.; Cengher, M.; Johnson, P. S.

    2007-11-01

    Direct measurements of the rf power injected into the DIII-D tokamak from the ECH gyrotrons are being made using a high power dummy load at the tokamak. The measurements will permitt power monitors, which measure the rf leakage from well-aligned gaps in the vacuum waveguides near the tokamak, to be calibrated for various elliptical polarizations of the rf propagating in the HE11 waveguide mode. Using these measurements, correlations with calorimetric measurements of the gyrotron cooling circuits, the usual basis for rf power measurements in the system, will be made. Low power rf measurements and theoretical and experimental estimates of the transmission efficiencies of the individual components in the transmission lines will be compared with the direct measurements.

  15. Investigation of contamination of correlation ECE signals by electron cyclotron heating at DIII-D

    NASA Astrophysics Data System (ADS)

    Baltzer, M. M.; Rhodes, T. L.; Wang, G.; Sung, C.; Carter, T. A.

    2016-10-01

    Correlation electron cyclotron emission (CECE) is commonly used at DIII-D to extract localized electron temperature fluctuations from passive emission. Very long coherence length signals (10's of cm in length) have been found to contaminate CECE measurements on some plasma discharges using electron cyclotron heating (ECH). Temperature fluctuations levels are observed to increase by as much as 50% due to this effect. Unabsorbed radiation from ECH is under investigation as a possible cause. A simple 1D code modeling the CECE data is used to investigate and test the effects of long coherence length contamination on the temperature fluctuation calculation. Data analysis methods to mitigate this effect have been designed and show significant promise in correcting the final results. Supported by US DOE DE-FG02-08ER54984 and DE-FC02-04ER54698.

  16. Method of neutral density determination near the X-point in DIII-D

    SciTech Connect

    Colchin, R.J.; Maingi, R.; Isler, R.C.; Owen, L.W.; Fenstermacher, M.E.; Carlstrom, T.N.

    1998-07-01

    A new method for determining the density of neutrals near the X-point of a diverted plasma is described. Code calculations have predicted that the neutral density peaks poloidally near the X-piont and there is evidence that neutrals play a role in the L-H transition and in damping the plasma rotation. The new method uses D{sub {alpha}} data from a tangentially-viewing video camera calibrated by a vertically-viewing photomultiplier. These data, combined with electron temperature and density measurements from the divertor Thomson scattering (DTS) system, provide sufficient informatino to determine the neutral density in the X-point region of DIII-D. Preliminary results show neutral densities above the X-point are order 10{sup 10}--10{sup 11} atomscm{sup 3}. The diagnostics used to determine these neutral densities, the data analysis method, and preliminary results are described in this paper.

  17. A New Neutron Calibration Technique with Fast Scintillators on DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Zhu, Y. B.; Heidbrink, W. W.; Taylor, P. L.; Carrig, W.

    2015-11-01

    Absolute calibrations are necessary for conventional neutron measurements based on proportional counters and fission chambers, at regular intervals. For the DIII-D tokamak, the wide span of fusion rates, approximately between 1.e9 - 1.e17 neutrons per second, from pure Ohmic to high power auxiliary heating plasmas requires careful cross-calibrations of a variety of neutron detectors with stepwise and overlapped sensitivities, with an intense isotope neutron source, e.g. californium-252 and real plasmas. Scintillators have been successfully utilized for fast time resolved neutron detection for decades. A new calibration approach with the help of scintillators is shown to be straightforward, simpler and trustworthy while the conventional approach is complicated, time consuming and costly. Details on the calibration setup and results will be presented. Supported by US DOE SC-G903402 and DE-FC02-04ER54698.

  18. A 250 GHz microwave interferometer for divertor experiments on DIII-D

    SciTech Connect

    James, R.A.; Nilson, D.G.; Stever, R.D.; Hill, D.N.; Casper, T.A.

    1994-01-31

    A new 250 GHz, two-frequency microwave interferometer system has been developed to diagnose divertor plasmas on DIII-D. This diagnostic will measure the line-averaged density across both the inner and outer, lower divertor legs. With a cut-off density of over 7 {times} 10{sup 14} cm{sup {minus}3}, temporal measurements of ELMs, MARFs and plasma detachment are expected. The outer leg system will use a double pass method while the inner leg system will be single pass. Two special 3D carbon composite tiles are used, one to protect the microwave antennas mounted directly under the strike point and the other as the outer leg reflecting surface. Performance, design constraints, and the thermalmechanical design of the 3D carbon composite tiles are discussed.

  19. INFRARED MONITORING OF 110GHz GYROTRON WINDOWS AT DIII-D

    SciTech Connect

    Y. GORELOV; J. LOHR; R.W. CALLIS; D. PONCE

    2002-05-01

    The combination of low millimeter wave losses and excellent thermal conductivity with good mechanical properties make artificial chemical vapor deposition (CVD) diamonds a compelling choice for 1 MW 110 GHz gyrotron windows. Five gyrotrons are currently operating at the DIII-D tokamak. Three Gycom gyrotrons have boron nitride (BN) ceramic windows. Due to temperature increases of the windows up to about 930 C, the pulse duration of these tubes is limited to 2 s for output power near 800 kW. Two Communications and Power Industries (CPI) gyrotrons with diamond windows are also installed and operating. The diamond disks of these windows and the construction of their water-cooling assemblies are different. This paper reviews the infrared (IR) measurements of both types of gyrotron windows, with emphasis on the two diamond designs.

  20. Fluctuation measurements in the DIII-D and TEXT tokamaks via collective scattering and reflectometry

    SciTech Connect

    Peebles, W.A.; Baang, S.; Brower, D.L.; Burrell, K.; Doyle, E.J.; Groebner, R.J.; Lehecka, T.; Luhmann, N.C. Jr.; Matsumoto, H.; Philipona, R.; Rettig, C.; Rhodes, T.L.; Yu, C.X. )

    1990-11-01

    Anomalous transport in fusion plasmas remains an enigma requiring explanation. A predictive capability is highly desirable if confinement enhancement regimes such as {ital H} mode or super shots are to be extrapolated to the next phase in development of the International Fusion program, epitomized, for example in ITER. Therefore, identification of the role that electrostatic turbulence plays in confinement is a critical issue requiring detailed experimental data capable of testing and challenging existing theoretical models. This article presents microturbulence measurements obtained on the DIII-D and TEXT tokamaks utilizing heterodyne, far-infrared collective scattering, and reflectometry techniques. The experimental systems are described on both machines and emphasis placed on results obtained during the {ital L}-{ital H} transition, ELM activity, and saturated ohmic operation where ion temperature gradient driven (ITGD) turbulence is theoretically predicted to exist.