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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. Non-axisymmetric Field Effects on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

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

    2003-10-01

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

  3. Alcator C-Mod predictive modeling

    NASA Astrophysics Data System (ADS)

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

    2001-10-01

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

  4. First results from Alcator-C-MOD*

    NASA Astrophysics Data System (ADS)

    Hutchinson, I. H.; Boivin, R.; Bombarda, F.; Bonoli, P.; Fairfax, S.; Fiore, C.; Goetz, J.; Golovato, S.; Granetz, R.; Greenwald, M.; Horne, S.; Hubbard, A.; Irby, J.; LaBombard, B.; Lipschultz, B.; Marmar, E.; McCracken, G.; Porkolab, M.; Rice, J.; Snipes, J.; Takase, Y.; Terry, J.; Wolfe, S.; Christensen, C.; Garnier, D.; Graf, M.; Hsu, T.; Luke, T.; May, M.; Niemczewski, A.; Tinios, G.; Schachter, J.; Urbahn, J.

    1994-05-01

    Early operation of the Alcator-C-MOD tokamak [I.H. Hutchinson, Proceedings of IEEE 13th Symposium on Fusion Engineering, Knoxville, TN, edited by M. Lubell, M. Nestor, and S. Vaughan (Institute of Electrical and Electronic Engineers, New York, 1990), Vol. 1, p. 13] is surveyed. Reliable operation, with plasma current up to 1 MA, has been obtained, despite the massive conducting superstructure and the associated error fields. However, vertical disruptions are not slowed by the long vessel time constant. With pellet fueling, peak densities up to 9×1020 m-3 have been attained and ``snakes'' are often seen. Initial characterization of divertor and scrape-off layer is presented and indicates approximately Bohm diffusion. The edge plasma shows a wealth of marfe-like phenomena, including a transition to detachment from the divertor plates with accompanying radiative divertor regions. Energy confinement generally appears to exceed the expectations of neo-Alcator scaling. A transition to Ohmic H mode has been observed. Ion cyclotron heating experiments have demonstrated good power coupling, in agreement with theory.

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

    SciTech Connect

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

    2004-08-10

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

  6. Benchmarking Nonlinear Turbulence Simulations on Alcator C-Mod

    SciTech Connect

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

    2004-06-22

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

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

    SciTech Connect

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

    2002-07-10

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

  8. Modeling of Alcator C-Mod Divertor Baffling Experiments

    SciTech Connect

    D. P. Stotler; C. S. Pitcher; C. J. Boswell; T. K. Chung; B. LaBombard; B. Lipschultz; J. L. Terry; R. J. Kanzleiter

    2000-11-29

    A specific Alcator C-Mod discharge from the series of divertor baffling experiments is simulated with the DEGAS 2 Monte Carlo neutral transport code. A simple two-point plasma model is used to describe the plasma variation between Langmuir probe locations. A range of conductances for the bypass between the divertor plenum and the main chamber are considered. The experimentally observed insensitivity of the neutral current flowing through the bypass and of the D alpha emissions to the magnitude of the conductance is reproduced. The current of atoms in this regime is being limited by atomic physics processes and not the bypass conductance. The simulated trends in the divertor pressure, bypass current, and D alpha emission agree only qualitatively with the experimental measurements, however. Possible explanations for the quantitative differences are discussed.

  9. 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.

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

    SciTech Connect

    Niemczewski, A.P.

    1995-08-01

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

  11. Correlation ECE diagnostic in Alcator C-Mod

    DOE PAGES

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

    2015-03-12

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

  12. 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.

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

    SciTech Connect

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

    2013-03-15

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

  14. 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.

  15. Lower Hybrid Current Drive on Alcator C-MOD

    NASA Astrophysics Data System (ADS)

    Wilson, J. R.; Bernabei, S.; Hosea, J.; Phillips, C. K.; Parker, R.; Bonoli, P. T.; Hubbard, A. E.; Liptac, J.; Schmidt, A. E.; Wallace, G.

    2006-10-01

    A Lower Hybrid Current Drive (LHCD) system has been installed on the Alcator C-MOD tokomak at MIT. This system was designed for maximum flexibility in the launched parallel wave-number spectrum. This should allow tailoring of the lower hybrid deposition under a variety of plasma conditions. Initial results from this system will be presented. Power levels up to 800 kW have been injected into the tokomak. The parallel wave number has been varied over a wide range, n||˜ 2-4. Driven currents up to ˜280 kA have been inferred from magnetic measurements, in reasonable agreement with modeling. Sawtooth oscillations vanish and central q is inferred to rise above unity, indicating off-axis CD as expected. Measurements of non-thermal x-ray and electron cyclotron emission confirm the presence of a significant fast electrons population that varies with phase and plasma density. This variation will be compared to that predicted by detailed propagation and absorption codes.

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

    SciTech Connect

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

    2007-10-09

    A Lower Hybrid Current Drive (LHCD) system has been installed on the Alcator C-MOD tokamak at MIT. Twelve klystrons at 4.6 GHz feed a 4x22 waveguide array. This system was designed for maximum flexibility in the launched parallel wave-number spectrum. This flexibility allows tailoring of the lower hybrid deposition under a variety of plasma conditions. Power levels up to 900 kW have been injected into the tokomak. The parallel wave number has been varied over a wide range, n|| ~ 1.6–4. Driven currents have been inferred from magnetic measurements by extrapolating to zero loop voltage and by direct comparison to Fisch-Karney theory, yielding an efficiency of n20IR/P ~ 0.3. Modeling using the CQL3D code supports these efficiencies. Sawtooth oscillations vanish, accompanied with peaking of the electron temperature (Te0 rises from 2.8 to 3.8 keV). Central q is inferred to rise above unity from the collapse of the sawtooth inversion radius, indicating off-axis cd as expected. Measurements of non-thermal x-ray and electron cyclotron emission confirm the presence of a significant fast electron population that varies with phase and plasma density. The x-ray emission is observed to be radialy broader than that predicted by simple ray tracing codes. Possible explanations for this broader emission include fast electron diffusion or broader deposition than simple ray tracing predictions (perhaps due to diffractive effects).

  17. 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.

  18. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2007-09-01

    A Lower Hybrid Current Drive (LHCD) system has been installed on the Alcator C-MOD tokamak at MIT. Twelve klystrons at 4.6 GHz feed a 4×22 waveguide array. This system was designed for maximum flexibility in the launched parallel wave-number spectrum. This flexibility allows tailoring of the lower hybrid deposition under a variety of plasma conditions. Power levels up to 900 kW have been injected into the tokomak. The parallel wave number has been varied over a wide range, n∥˜1.6-4. Driven currents have been inferred from magnetic measurements by extrapolating to zero loop voltage and by direct comparison to Fisch-Karney theory, yielding an efficiency of n20IR/P˜0.3. Modeling using the CQL3D code supports these efficiencies. Sawtooth oscillations vanish, accompanied with peaking of the electron temperature (Te0 rises from 2.8 to 3.8 keV). Central q is inferred to rise above unity from the collapse of the sawtooth inversion radius, indicating off-axis cd as expected. Measurements of non-thermal x-ray and electron cyclotron emission confirm the presence of a significant fast electron population that varies with phase and plasma density. The x-ray emission is observed to be radialy broader than that predicted by simple ray tracing codes. Possible explanations for this broader emission include fast electron diffusion or broader deposition than simple ray tracing predictions (perhaps due to diffractive effects).

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

    SciTech Connect

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

    2007-09-28

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

  1. 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

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

    SciTech Connect

    D.P. Stotler; B. LaBombard

    2004-06-15

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

  3. 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.

  4. Overview of Alcator C-Mod Contribution to the 2015 JRT on Off Axis Current Drive

    NASA Astrophysics Data System (ADS)

    Wallace, G. M.; Bonoli, P. T.; Faust, I. C.; Mumgaard, R. T.; Parker, R. R.; Rice, J. E.; Shirawia, S.; Scott, S. D.; Bhattacharjee, A.; Ebrahimi, F.; Poli, F.; Gerhardt, S.; Podesta, M.; Solomon, W.; Wilson, J. R.; Holcomb, C.

    2015-11-01

    The goal of the US Department of Energy FES 2015 Joint Research Target (JRT) is to conduct experiments and analysis to quantify the impact of broadened current and pressure profiles on tokamak plasma confinement and stability. Broadened current profiles are achieved on Alcator C-Mod through the use of Lower Hybrid Current Drive (LHCD). C-Mod experiments from the recent run campaign have focused on several areas, such as MHD stability of discharges with broad current profiles, energy transport barriers and high-Z impurity transport in non-inductive discharges, momentum transport in discharges with altered safety factor profiles, and validation of LHCD actuator models. This work was conducted on the Alcator C-Mod tokamak, a DoE Office of Science user facility supported by US Department of Energy cooperative agreement DE-FC02-99ER54512.

  5. 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.

  6. Near infrared spectroscopy of the divertor region in the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Furno, I.; Wurden, G. A.

    2004-10-01

    A new system for near infrared spectroscopic measurements (1.2-2.6 μm) is designed for the Alcator C-Mod tokamak. These measurements will provide a new window into understanding the physics of detached and/or high radiating plasmas. The system, located on top of the Alcator C-Mod concrete igloo, will view a 30×30 cm2 region of the lower divertor by means of a re-entrant 5 m long ZnSe based periscope. A McPherson 218 spectrometer will disperse the light that will then be detected by a thermoelectrically cooled InGaAs, 256-element linear image sensor. The signals from each pixel will be digitized by a 16-bit, 333 kHz analog/digital board and then transmitted over optical link to a personal computer (PC) located in Alcator C-Mod control room. The PC will remotely control the system and store the data in the MDSplus database.

  7. 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

  8. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

    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∥, 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 ne˜0.5×1020m-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.

  10. 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

  11. 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

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

    SciTech Connect

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

    1994-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-08-01

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

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

    SciTech Connect

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

    2009-11-12

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

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

    PubMed

    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)] 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.

  16. 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.

  17. 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.

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

    SciTech Connect

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

    2008-10-15

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

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

    SciTech Connect

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

    2010-09-24

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

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

    SciTech Connect

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

    1999-11-03

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

  1. 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

  2. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2013-05-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-04-01

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

  5. Experimental and numerical characterization of ion-cyclotron heated protons on the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Tang, V.; Parker, R. R.; Bonoli, P. T.; Wright, J. C.; Granetz, R. S.; Harvey, R. W.; Jaeger, E. F.; Liptac, J.; Fiore, C. L.; Greenwald, M.; Irby, J. H.; Lin, Y.; Wukitch, S. J.

    2007-06-01

    Energetic minority protons with ~100 keV effective temperature are routinely created in Alcator C-Mod plasmas with the application of ion-cyclotron-range-of-frequency (ICRF) heating. A multichannel compact neutral particle analyzer (CNPA) is used to make measurements of these distributions in Alcator C-Mod's unique and reactor-relevant operating space via active and passive charge-exchange techniques. A radially injected 50 keV diagnostic hydrogen neutral beam is used for active analysis. Using a detailed model that accounts for beam, halo and impurity electron donors, core proton temperatures of ~30-120 keV are directly measured for the first time in lower density (ne0 ~ (0.8-1.5) × 1020 m-3) Alcator C-Mod plasmas with up to only ~0.5 MW of ICRF power. The model found that the minority proton temperatures are peaked spatially away from r/a = 0, even for an on-axis resonance. Additionally, noticeable phase-space anisotropy is seen as expected for ICRF heating. The measured effective temperatures also scale approximately with the Stix parameter. The CNPA temperature measurements are compared with several leading simulation packages such as the TORIC/FPPRF and AORSA/CQL3D full-wave/Fokker-Planck (FW/FP) solvers. Preliminary comparisons with the AORSA/CQL3D code which include results from a new synthetic diagnostic show good agreement and demonstrate that accurate tracking of the minority distribution during iterations of the FW and FP solvers is required to simulate Alcator C-Mod's energetic minority populations with accuracy. Physically, poor wave focusing and preferential heating of trapped energetic protons are found to move the fast proton temperature profiles off-axis. These FW/FP analyses represent the first comparison between predictions of these detailed codes and core minority tail experimental measurements on Alcator C-Mod.

  6. 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.

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

    PubMed

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

    2012-10-01

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

  8. 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.

  9. Measurement of Fast Electron Transport by Lower Hybrid Modulation Experiments in Alcator C-Mod

    SciTech Connect

    Schmidt, A. E.; Bonoli, P. T.; Parker, R.; Porkolab, M.; Wallace, G.; Wright, J. C.; Wilson, J. R.; Harvey, R. W.; Smirnov, A. P.

    2009-11-26

    The Lower Hybrid Current Drive (LHCD) system on Alcator C-Mod can produce spectra with a wide range of peak parallel refractive index (n{sub parallel}). An experiment in which LH power is square-wave modulated on a time scale much faster than the current relaxation time does not significantly alter the poloidal magnetic field inside the plasma and thus allows for realistic modeling and consistent plasma conditions for different ny spectra. Boxcar binning of hard x-rays during LH power modulation allows for time resolution sufficient to resolve the build-up, steady-state, and slowing-down of fast electrons. A transport model built in Matlab has been used to determine a fast electron pinch velocity for a high-n{sub parallel} case of 1-2 m/s.

  10. 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.

  11. 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).

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

    SciTech Connect

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

    2014-03-15

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

  13. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-03-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-06-01

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

  16. 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.

  17. Measurements of plasma sheath heat flux in the Alcator C-Mod divertor

    NASA Astrophysics Data System (ADS)

    Brunner, Dan; Labombard, Brian; Terry, Jim; Reinke, Matt

    2010-11-01

    Heat flux is one of the most important parameters controlling the lifetime of first-wall components in fusion experiments and reactors. The sheath heat flux coefficient (γ) is a parameter relating heat flux (from a plasma to a material surface) to the electron temperature and ion saturation current. Being such a simple expression for a kinetic process, it is of great interest to plasma edge fluid modelers. Under the assumptions of equal ion and electron temperatures, no secondary electron emission, and no net current to the surface the value of γ is approximately 7 [1]. Alcator C-Mod provides a unique opportunity among today's experiments to measure reactor-relevant heat fluxes (100's of MW/m^2 parallel to the magnetic field) in reactor-like divertor geometry. Motivated by the DoE 2010 joint milestone to measure heat flux footprints, the lower outer divertor of Alcator has been instrumented with a suite of Langmuir probes, novel surface thermocouples, and calorimeters in tiles purposefully ramped to eliminate shadowing; all within view of an IR camera. Initial results indicate that the experimentally inferred values of γ are found to agree with simple theory in the sheath limited regime and diverges to lower values as the density increases.

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

    NASA Astrophysics Data System (ADS)

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

    2009-11-01

    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.

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

    SciTech Connect

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

    2014-02-12

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

  1. 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.

  2. 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.

  3. 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.

  4. A CO{sub 2} laser polarimeter for measurement of plasma current profile in Alcator C-Mod

    SciTech Connect

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

    1994-06-01

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

  5. 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. 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.

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

    SciTech Connect

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

    2006-08-15

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

  8. 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.

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

    DOE PAGES

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

    2016-02-09

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

  10. Neutral gas compression in the Alcator C-Mod divertor, experimental observations

    SciTech Connect

    Niemczewski, A.; LaBombard, B.; Lipschultz, B.; McCracken, G.

    1994-11-01

    One of the high heat flux solutions envisioned for ITER is the gas target divertor. This scheme requires high neutral pressure to be sustained in the divertor chamber with a minimal effect on the pressure in the main tokamak chamber (i.e. high gas compression). The neutral gas compression has been studied in the Alcator C-Mod closed divertor under various central and edge plasma conditions. The neutral pressure measured by a fast, in-situ, ionization gauge, installed behind the divertor target plate was compared with the midplane pressure, measured by a shielded Bayard-Alpert gauge. Divertor pressures up to 30 mTorr with compression factors p{sub div}/p{sub mid} {le} 70 have been observed. It has been found that the neutral pressure in the divertor does not depend strongly on the fueling location but rather on the core plasma density and the resulting divertor plasma regime. Divertor detachment leads to a considerable drop in the compression ratio, suggesting a partial {open_quotes}unplugging{close_quotes} of the divertor volume. An examination of the local particle flux balance in the divertor indicates that the single most important factor determining divertor pressure and compression is the private-flux plasma channel opacity to neutrals.

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

    SciTech Connect

    C Kessel, et al

    2013-05-07

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

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

    PubMed

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

    2014-05-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

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

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

    SciTech Connect

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

    2014-05-15

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

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

    SciTech Connect

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

    2003-07-31

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

  16. 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.

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

    NASA Astrophysics Data System (ADS)

    Kurz, Christian

    1995-01-01

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

  18. 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.

  19. Scaling and transport analysis of divertor conditions on the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    LaBombard, B.; Goetz, J.; Kurz, C.; Jablonski, D.; Lipschultz, B.; McCracken, G.; Niemczewski, A.; Boivin, R. L.; Bombarda, F.; Christensen, C.; Fairfax, S.; Fiore, C.; Garnier, D.; Graf, M.; Golovato, S.; Granetz, R.; Greenwald, M.; Horne, S.; Hubbard, A.; Hutchinson, I.; Irby, J.; Kesner, J.; Luke, T.; Marmar, E.; May, M.; O'Shea, P.; Porkolab, M.; Reardon, J.; Rice, J.; Schachter, J.; Snipes, J.; Stek, P.; Takase, Y.; Terry, J.; Tinios, G.; Watterson, R.; Welch, B.; Wolfe, S.

    1995-06-01

    Detailed measurements and transport analysis of divertor conditions in Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are presented for a range of line-averaged densities, 0.7

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

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

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

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

    SciTech Connect

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

    2016-01-01

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

  3. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

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

    NASA Astrophysics Data System (ADS)

    Mumgaard, Robert T.

    2014-10-01

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

  6. 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.

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

    SciTech Connect

    Garnier, D.T.

    1996-06-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-01-01

    An extensive experimental survey of plasma potentials induced by ion cyclotron range-of frequency (ICRF) heating was carried out in the scrape-off layer (SOL) plasmas on the Alcator C-Mod tokamak. Enhanced plasma potentials >100 V are observed at locations where local magnetic fields map to active ICRF antennas. In these cases, the enhanced potential appears only when a local plasma density threshold is surpassed—a threshold that is quantitatively consistent with slow wave (SW) RF rectification theory. However, in many cases large potential enhancements are found in locations that do not map along magnetic field lines to active antennas without obstruction, i.e. locations that are inaccessible to SWs launched by the active antennas. Enhanced potentials in these ‘unmapped’ locations are correlated with local plasma parameters, ICRF electromagnetic fields associated with the fast wave (FW) and SW, launched wave spectra, and the boundary surface geometry. It is found that enhanced plasma potentials in unmapped locations correlate with the FW field strength. These observations are qualitatively consistent with a model that accounts for the conversion of FWs to SWs at conducting surfaces oriented at an oblique angle with respect to the magnetic field, with the SW leading to sheath rectification. In addition, enhanced plasma potentials are found far into the shadow of passive limiter structures. These are correlated with the magnitude of the local FW field strength, yet the effect does not follow any present model. Overall, ICRF-induced plasma potentials may appear in regions far removed from the active antennas, yet due to the complex response of the SOL potentials at a variety of boundary surfaces, it remains unclear what part of the plasma-facing wall should be targeted to mitigate ICRF-induced impurities. The results also suggest that operating active ICRF antennas in a high single pass absorption regime is crucial in minimizing the effects of the FW fields on

  9. 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.

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

    SciTech Connect

    Urbahn, J.A.

    1994-05-01

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

  11. 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.

  12. Scanning ion sensitive probe for plasma profile measurements in the boundary of the Alcator C-Mod tokamak

    NASA Astrophysics Data System (ADS)

    Brunner, D.; LaBombard, B.; Ochoukov, R.; Whyte, D.

    2013-05-01

    A new Ion Sensitive Probe head has been created for the outer-midplane scanning probe system on the Alcator C-Mod tokamak. The new probe head contains three elements: an ion sensitive probe to measure ion temperature and plasma potential, a Langmuir probe to measure electron temperature, density, and floating potential, and a second Langmuir probe to measure ion saturation current and the density fluctuations arising from ``blob'' events. The ion sensitive probe current is normalized to this measurement to reduced deleterious effects of the strong fluctuations. Design of the high heat flux probe (>100 MW/m2) and initial results are presented.

  13. Scanning ion sensitive probe for plasma profile measurements in the boundary of the Alcator C-Mod tokamak

    SciTech Connect

    Brunner, D.; LaBombard, B.; Ochoukov, R.; Whyte, D.

    2013-05-15

    A new Ion Sensitive Probe head has been created for the outer-midplane scanning probe system on the Alcator C-Mod tokamak. The new probe head contains three elements: an ion sensitive probe to measure ion temperature and plasma potential, a Langmuir probe to measure electron temperature, density, and floating potential, and a second Langmuir probe to measure ion saturation current and the density fluctuations arising from ''blob'' events. The ion sensitive probe current is normalized to this measurement to reduced deleterious effects of the strong fluctuations. Design of the high heat flux probe (>100 MW/m{sup 2}) and initial results are presented.

  14. Scanning ion sensitive probe for plasma profile measurements in the boundary of the Alcator C-Mod tokamak.

    PubMed

    Brunner, D; LaBombard, B; Ochoukov, R; Whyte, D

    2013-05-01

    A new Ion Sensitive Probe head has been created for the outer-midplane scanning probe system on the Alcator C-Mod tokamak. The new probe head contains three elements: an ion sensitive probe to measure ion temperature and plasma potential, a Langmuir probe to measure electron temperature, density, and floating potential, and a second Langmuir probe to measure ion saturation current and the density fluctuations arising from ''blob'' events. The ion sensitive probe current is normalized to this measurement to reduced deleterious effects of the strong fluctuations. Design of the high heat flux probe (>100 MW/m(2)) and initial results are presented.

  15. Current Profile Measurements from Moderate to Strong Lower Hybrid Single-Pass Damping on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Mumgaard, R. T.; Wallace, G. M.; Scott, S. D.; Shiraiwa, S.; Faust, I.; Parker, R. R.

    2015-11-01

    Lower Hybrid Current Drive (LHCD) is an effective tool to non-inductively drive up to 100% of the plasma current on Alcator C-Mod. Measurements with an upgraded MSE diagnostic show that the fast-electron current profile is broader than the Ohmic current profile but still located the plasma core in agreement with strongly centrally peaked fast electron bremsstrahlung (FEB) measurements. Scans in a regime of high current drive efficiency across a range of density, LHCD power, launched n||, and plasma current show the driven current profile, FEB profile shapes, and current drive efficiency are sensitive only to total plasma current. Simulations using ray-tracing Fokker Planck codes show that the rays make 1-3 bounces through the plasma edge to bridge the spectral gap. Although in agreement with the total current, the simulations qualitatively disagree with experiment regarding current and FEB profiles as well as sensitivity to power and density. Simulations at higher plasma temperature and current predict stronger single-pass damping and preliminary experiments show increased current drive efficiency. Experiments to determine if the profile discrepancies persist when the ray bounces play a reduced role are planned, including companion experiments in D and He resulting in different edge plasma conditions. This work was performed on the Alcator C-Mod tokamak, a DoE Office of Science user facility, and is supported by USDoE awards DE-FC02-99ER54512 and DE-AC02-09CH11466.

  16. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  18. 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.

  19. First Results From The New High Resolution Imaging X-ray Crystal Spectrometer On Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Ince-Cushman, Alexander; Rice, John; Bitter, Manfred; Reinke, Mathew; Hill, Kenneth; Scott, Steven

    2007-11-01

    In an effort to improve the diagnostic capabilities for measuring plasma rotation on Alcator C-Mod, an imaging x-ray spectrometer has been designed and installed. This instrument utilizes a spherically bent quartz crystal and a set of 2D x-ray detectors to image the entire plasma cross section with a spectral resolving power of approximately 10,000 with vertical spatial resolution of about 1cm. Line emission from highly ionized states of argon and molybdenum are measured at frame rates up to 200Hz. Using spectral tomographic techniques the line integrated spectra can be inverted to determine impurity density, velocity and temperature profiles. An overview of the instrument, analysis and example profiles are presented. Work supported by USDoE Coop. Agree. No. DE-FC02-99ER54512 & DE-AC02-76CH03073.

  20. Nonlinear transfer in heated L-modes approaching the L-H transition threshold in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Cziegler, I.; Tynan, G. R.; Diamond, P. H.; Hubbard, A. E.; Hughes, J. W.; Irby, J.; Terry, J. L.

    2015-08-01

    Nonlinear transfer processes between large-scale edge flows and the ambient broadband fluctuations have been shown to play a significant role in the dynamics of edge turbulence, including spreading power from coherent modes and suppressing turbulence at the formation of edge transport barriers. In order to predict thresholds of confinement regimes, both the transition dynamics and the parametric dependence of the nonlinear energy transfer must be studied. Since the expected flow damping terms depend on ion collision rates and local safety factors, recent experiments aimed also to explore the nonlinear drive at various values of the plasma current, density and amount of auxiliary heating. Nonlinear interactions between zonal flows and turbulence in L-mode are estimated using bispectral as well as time-resolved methods based on gas-puff-imaging in Alcator C-Mod [1].

  1. 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.

  2. Study on the Characteristics of Wire Grid Polarizer for Alcator C-Mod Motional Stark Effect Diagnostic

    NASA Astrophysics Data System (ADS)

    Ko, Jinseok; Scott, Steve

    2005-10-01

    A wire grid polarizer (WGP) has been installed on the in-vessel optics module of the Motional Stark Effect (MSE) diagnostic system in Alcator C-Mod tokamak. The polarized light produced by this WGP can be utilized to measure the change in polarization angle by Faraday rotation through the diagnostic system when the tokamak is operated in its normal condition. Intensive WGP tests on the off-site optical table have been done to investigate the effect of non-normal incidence on the WGP which is indeed the situation inside the tokamak. A model has been developed to understand and simulate various effects of the WGP structure such as changes in angle-of-incidence and reflection/transmission coefficients due to the anti-reflection coating layer and the glass substrate. These analytic predictions and test results are compared with the tests in the tokamak where the in-vessel WGP is used.

  3. 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)

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

    NASA Astrophysics Data System (ADS)

    Long, Andrew; Ernst, Darin

    2005-10-01

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

  5. Validation of MMM7.1 and TGLF anomalous transport models for predicting the evolution of Alcator C-Mod temperature profiles

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

    The Multi-Mode MMM7.1 [T. Rafiq, et al. Phys. Plasmas, 20, 032506, 2013] and the Trapped Gyro-Landau Fluid (TGLF)[G.M. Staebler, et al., Phys. Plasmas 14, 055909, 2007] anomalous transport models are validated employing experimental data for Alcator C-Mod discharges that represent a plasma density scan. The MMM7.1 and the TGLF models compute the anomalous transport driven by the ITG, TEM, ETG, KBM and collisional drift modes. The validation study is carried out with simulations that employ the new numerical solver PT-SOLVER in the PTRANSP code and that utilize Alcator C-Mod experimental boundary and initial conditions. The predicted evolving temperature profiles are compared with corresponding Alcator C-Mod experimental data. The comparison is quantified by calculating the RMS deviations and Offsets. Supported by USDoE FES grants DE-FG02-92ER54141, DE-SC0012174 and C-Mod award DE-FC02-99ER54512.

  6. Demonstrating the Physics Basis for the ITER 15 MA Inductive Discharge on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Kessel, C. E.; Wolfe, S. M.; Hutchinson, I. H.; Hughes, J. W.; Lin, Y.; Ma, Y.; Mikkelsen, D. R.; Poli, F.; Reinke, M. L.; Wukitch, S. J.

    2012-10-01

    Rampup discharges in C-Mod, matching ITE's current diffusion times show ICRF heating can save V-s but results in only weak effects on the current profile, despite strong modifications of the central electron temperature. Simulation of these discharges with TSC, and TORIC for ICRF, using multiple transport models, do not reproduce the temperature profile evolution, or the experimental internal self-inductance li, by sufficiently large amounts to be unacceptable for projections to ITER operation. For the flattop phase experiments EDA H-modes approach the ITER parameter targets of q95=3, H98=1, n/nGr=0.85, betaN=1.7, and k=1.8, and sustain them similar to a normalized ITER flattop time. The discharges show a degradation of energy confinement at higher densities, but increasing H98 with increasing net power to the plasma. For these discharges intrinsic impurities (B, Mo) provided radiated power fractions of 25-37%. Experiments show the plasma can remain in H-mode in rampdown with ICRF injection, the density will decrease with Ip while in the H-mode, and the back transition occurs when the net power reaches about half the L-H transition power. C-Mod indicates that faster rampdowns are preferable. Work supported by US Dept of Energy under DE-AC02-CH0911466 and DE-FC02-99ER54512.

  7. 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

  8. 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.

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

    SciTech Connect

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

    2006-05-15

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

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

    DOE PAGES

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

    2014-09-30

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

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

    SciTech Connect

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

    2014-11-15

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

  12. Helium-3 transport experiments in the scrape-off layer with the Alcator C-Mod omegatron ion mass spectrometer

    SciTech Connect

    Nachtrieb, R.; LaBombard, B.

    2000-11-01

    {sup 3}He gas was puffed from the wall into Ohmic low confinement-mode discharges of the Alcator C-Mod [I. H. Hutchinson , Phys. Plasmas 1, 1551 (1994)] tokamak and the charged states were measured near the wall with the omegatron ion mass spectrometer. Analysis of the data shows that the concentrations of singly- and doubly-ionized helium near the wall are approximately equal. The electron temperature and density at the omegatron are too low to account for ionization of helium in the local flux tube, therefore the helium is ionized in a hotter region of the edge plasma and is transported to the omegatron. A simple one-dimensional radial transport model reproduces the observed values of charge state flux and density, but only if rapid cross-field transport is included, increasing with distance from the separatrix. A constant cross-field diffusion coefficient of order 2m{sup 2}/s and an outward convection velocity profile increasing to of order 100 m/s in the far scrape-off layer is implied.

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

    NASA Astrophysics Data System (ADS)

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

    2013-08-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2012-06-01

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

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

    SciTech Connect

    Brunner, D.; LaBombard, B.

    2012-03-15

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

  16. Edge turbulence flows at two different poloidal angles in Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Zweben, Stewart; Terry, James; Agostini, Matteo; Davis, William; Grulke, Olaf; Hughes, Jerry; Labombard, Brian; Landreman, Matt; Ma, Yunxing; Pace, David; Scott, Bruce

    2012-10-01

    High resolution edge turbulence movies have been obtained simultaneously at both the outer midplane and near the lower X-point region of C-Mod, using gas puff imaging (GPI) with two high speed cameras at 400,000 frames/sec. The time-resolved turbulence flow speeds at the outer midplane GPI view was previously estimated using a time-resolved cross-correlation technique [1], but previous results also showed a significantly different turbulence structure in these two regions [2]. Preliminary results indicate that the poloidal turbulence flows are not necessarily the same at these two poloidal angles. For instance, in one shot there is a strong time-averaged poloidal flow near the X-region toward the outer midplane, but mainly poloidally-fluctuating flows at the outer midplane. Examples of these flow measurements will be shown for plasmas with and without ICRH and in L-mode and H-mode plasmas. Evidence for fluctuating zonal flows preceding the L-H transition will be assessed. This work is supported in part by DOE Contracts DE-AC02-09CH11466 and DE-FC02-99ER5412.[4pt] [1] S.J. Zweben, J.L. Terry et al, Plasma Phys. Cont. Fusion 54 (2012) 025008[0pt] [2] J.L. Terry, S.J. Zweben et al, J. Nucl. Mat. 390-291 (2009) 339

  17. 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.

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

    SciTech Connect

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

    2002-06-18

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

  19. 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.

  20. Impurity Screening in Ohmic and H-Mode Plasmas in the Alcator C-Mod Tokamak^*

    NASA Astrophysics Data System (ADS)

    McCracken, G. M.

    1996-11-01

    The impurity density in the confined plasma is determined not only by the impurity production rate but also by screening, i.e. the balance between the perpendicular and parallel transport in the SOL. The relative importance of screening has been studied by injecting gaseous recycling (Ne, Ar) and non-recycling impurities (N, C) into various poloidal positions of the SOL and divertor in C-Mod. The density of the non-recycling impurities in the core is a function of the poloidal position of injection, while the screening of recycling impurities is not. In both cases screening is significantly worse ( ~3x) during divertor detachment. For a given injection rate of N2 gas into H-mode discharges, the number of impurities in the core is typically a factor of 3 greater than for ohmic discharges. Optical imaging of low charge states of the injected non-recycling impurities using a camera and spectral filters shows a directed plume, indicating flow of impurities towards the divertor target at all positions studied, even at the inboard midplane. This implies that the friction force due to plasma flow dominates the parallel ion temperature gradient force. The spatial distribution of low charge states in the divertor has been studied using a multichord visible spectrometer, and the distribution of the nitrogen radiation has been been studied using a 20 chord bolometer array. The results show that for attached discharges the nitrogen radiation is predominantly in the SOL below the X-point. The position of the radiation is not strongly dependent on the position of nitrogen injection. The screening has been compared with calculations using the DIVIMP Monte Carlo code [1] and with an analytical model of the impurity transport. Work supported by U.S. DOE Contract No. DE-AC02-78ET51013. În collaboration with B Lipschultz, B LaBombard, J A Goetz, R Granetz, D Jablonski, H Ohkawa, J Terry, MIT, S Lisgo, P C Stangeby, University of Toronto, ^1P.C. Stangeby and D. Elder, J. Nucl. Mater

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

    SciTech Connect

    Jablonski, D.F.

    1996-05-01

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

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

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

    SciTech Connect

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

    2009-11-26

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

  8. 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.

  9. Beta Scaling of Transport on the DIII-D Tokamak - Is Transport Electrostatic or Electromagnetic?

    NASA Astrophysics Data System (ADS)

    Petty, C. C.

    2003-10-01

    The related methods of dimensional analysis, similarity, and scale invariance in physics provide a powerful technique for analyzing physical systems. Previous experiments on the JET, DIII-D, and Alcator C-Mod tokamaks have validated the principle of similarity for energy transport in high-temperature plasmas. Recently the dependence of transport on beta, the ratio of the plasma kinetic pressure to the magnetic field pressure, has been measured for H-mode plasmas on DIII-D. Experimentally determining the beta scaling helps to differentiate between various proposed mechanisms of turbulent transport since theories for which ExB transport is dominant show little change with increasing beta up to the ideal ballooning limit, while transport models that invoke electromagnetic effects like magnetic flutter transport generally have a strong, unfavorable beta scaling. These experiments on DIII-D varied the normalized beta from 1.1 to 3.0 in several steps, covering a range from 25% to 85% of the ideal no-wall beta limit, and showed that the measured thermal diffusivities and global energy confinement times (normalized to Bohm) have little dependence on beta. This weak, possibly non-existent, beta scaling of transport confirms previous observations from the DIII-D and JET tokamaks as well as the ATF torsatron. This experimental result is in marked contrast to empirical scaling relations derived from multi-machine H-mode confinement databases, such as the ITER-98(y,2) relation that contains a strong, unfavorable beta dependence. New semi-empirical scaling relations, derived from the confinement databases, that are gyroBohm-like and electrostatic predict that the fusion performance in ITER will optimize at high beta, yielding twice the fusion power as the nominal beta scenario at higher fusion gain.

  10. 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.

  11. Size Scaling of Intrinsic Rotation in DIII-D

    NASA Astrophysics Data System (ADS)

    Degrassie, J. S.; Solomon, W. M.

    2015-11-01

    Despite the richness in the variety of the profiles of intrinsic rotation in axisymmetric tokamaks, a common feature is a co-Ip directed toroidal velocity on the outboard midplane in the region of ρ ~ 0 . 8 in DIII-D. This feature showed a ``Rice scaling'' (RS) in DIII-D and led to similarity experiments with C-Mod. RS correlates toroidal velocity with W/Ip, where W is the total plasma kinetic energy and Ip the plasma current. Subsequent analysis from DIII-D shows a clear ρ * dimensionless scaling of this intrinsic velocity in DIII-D, where ρ * ~ √Ti / aB , multiplying the βq scaling indicative of RS. The DIII-D scaling is MA ~βN ρ * , where MA is the Alfvén ``Mach'' value and βN is normalized β. In machine parameters it is very similar to the theoretical ``Parra scaling,'' which emphasizes the correlation of toroidal velocity with ion temperature as seen experimentally, but in this DIII-D scaling having an additional critical dependence on √β . Published data from C-Mod and low power ICRF in JET also fit with this DIII-D scaling. The relation to the RS will be described. Work supported in part by US DOE under DE-FC02-04ER54698 & DE-AC02-09CH11466.

  12. Measurements of ion cyclotron parametric decay of lower hybrid waves at the high-field side of Alcator C-Mod

    NASA Astrophysics Data System (ADS)

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

    2013-05-01

    Ion cyclotron parametric decay instability (PDI) of lower hybrid (LH) waves is surveyed using edge Langmuir probes on the Alcator C-Mod tokamak. The measurement is carried out simultaneously at the high-field side (HFS) and low-field side (LFS) mid-plane of the tokamak, as well as in the outer divertor region. Different LH spectra are observed depending on the location of the probes and magnetic configuration in L-mode plasmas, with \\overrightarrow{B}\\times\\bigtriangledown B drift direction downward. In lower single null (LSN) plasmas, strong ion cyclotron PDI occurring at the HFS is observed for the first time. This instability is characterized by a frequency separation in sidebands corresponding to the ion cyclotron frequency (ωci) near the HFS scrape-off layer and develops with threshold-like behavior as density increases. In inner wall limited (IWL) plasmas, this HFS instability shows a higher density threshold compared with that in LSN plasmas. The pump width becomes broadened even in the absence of the sidebands. In upper single null plasmas with similar plasma parameters, ion cyclotron PDI sidebands have a frequency separation corresponding to ωci near the LFS and are weaker than those observed in the LSN and IWL plasmas. Correlation between the onset of ion cyclotron PDI and the observed loss of lower hybrid current drive efficiency (Wallace et al 2012 Phys. Plasmas 19 062505) is discussed.

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

    NASA Astrophysics Data System (ADS)

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

    2014-05-01

    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/cm2. 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.

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

    SciTech Connect

    S.J. Zweben, et. al.

    2013-03-29

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

  15. 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.

  16. 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.; LaBombard, B. A.; Lipschultz, B.; Terry, J. L.; Pitts, R. A.; Feng, Y.

    2015-04-28

    We study experiments in Alcator C-Mod to assess the level of toroidal asymmetry in divertor conditions resulting from poloidally and toroidally localized extrinsic impurity gas seeding show a weak toroidal peaking (~1.1) in divertor electron temperatures for high-power enhanced D-alpha H-modeplasmas. This is in contrast to similar experiments in Ohmically heated L-modeplasmas, which showed a clear toroidal modulation in the divertor electron temperature. Modeling of these experiments using the 3D edge transport code EMC3-EIRENE [Y. Feng et al., J. Nucl. Mater. 241, 930 (1997)] qualitatively reproduces these trends, and indicates that the different response in the simulations is due tomore » the ionization location of the injected nitrogen. Low electron temperatures in the private flux region (PFR) in L-mode result in a PFR plasma that is nearly transparent to neutral nitrogen, while in H-mode the impurities are ionized in close proximity to the injection location, with this latter case yielding a largely axisymmetric radiation pattern in the scrape-off-layer. In conclusion, the consequences for the ITER gas injection system are discussed. Quantitative agreement with the experiment is lacking in some areas, suggesting potential areas for improving the physics model in EMC3-EIRENE.« less

  17. 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.; LaBombard, B.; Hughes, J. W.; Terry, J. L.; Zweben, S. J.

    2016-06-10

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

  18. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-06-01

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

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

    SciTech Connect

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

    2011-12-23

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

  1. A study of hydrogenic retention in a tokamak with reactor-like plasma-facing surfaces; Alcator C-Mod

    NASA Astrophysics Data System (ADS)

    Lipschultz, Bruce

    2008-11-01

    Tritium retention is an important safety concern for ITER; Operation for 1000 discharges without a major stoppage will require the fraction of ion fluence to Plasma Facing Components (PFCs) that is retained, R, to be < 0.001%. One year operation of a reactor, where tungsten (W) PFCs are envisioned, requires R to be 100x smaller! Co-deposition of H with carbon projects to unacceptably high T retention in ITER. We present the results of the first in-depth study of fuel retention for high-Z PFCs with ITER divertor ne, Te, particle and heat fluxes. We utilize molybdenum (Mo, with a small fraction of W), which is very similar to tungsten in terms of hydrogenic retention. The retention observed in a series of disruption-free C-Mod discharges is high, R˜1%, 1000x than expected from inherent Mo properties. These retention characteristics are exhibited regardless if the Mo surfaces are bare or partially covered by B films; D co-deposition with B is not contributing significantly to retention. Retention appears linear in fluence up to the limit of the discharge sequence, ˜20s, approaching one ITER discharge. Comparison of He- and D-fueled discharges gives support to a model of retention site creation in the lattice (`traps') due to D neutral buildup and accompanying lattice distortion driven by recombination-limited release (D->D2) from the front surface. Disruptions can be used to rapidly heat surfaces, releasing the H/D for recovery, potentially applicable to ITER. Naturally-occurring disruptions appear to balance single-discharge retention reducing the campaign-integrated retention by at least 100. Comparisons to laboratory-based retention studies indicate that the tokamak environment leads to additional enhancements of retention. This work is supported by U.S. Dept. of Energy Coop. Agreement DE-FC02-99ER54512.

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  3. 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.

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

    SciTech Connect

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

    2009-01-15

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

  5. 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.

  6. 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.

  7. 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)].

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Power Deposition on the DIII-D Inner Wall Limiter

    NASA Astrophysics Data System (ADS)

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

    2014-10-01

    Power deposition on the inner wall limiter (IWL) of DIII-D was measured by infrared (IR) thermography and calculated from plasma profiles measured by an inner column Swing-Probe for 6 ohmic discharges. In some cases clear evidence was found for a narrow feature with λshort ~ ion poloidal gyro-radius ~ a few mm, and of strength q| | 0 _short /q| | 0 _long ~ 0 . 5 +/- a factor of 5, where q| | is the parallel power flux density. The objective of the experiment was to check the assumptions made in defining the shape of the ITER IWL, in particular to see if the radial gradient of q| | increases near the last closed flux surface on DIII-D in agreement with observations in other tokamaks [JET, COMPASS, TCV, C-Mod]. On the basis of the results from the IWL experiments done on the 5 tokamaks, ITER decided in April 2014 to re-design the limiter shape to accommodate a narrow power feature. Work supported in part by the US Department of Energy under DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-FG02-07ER54917.

  14. DIII-D PLASMA CONTROL SIMULATION ENVIRONMENT

    SciTech Connect

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

    2003-10-01

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

  15. Boronization in DIII-D

    SciTech Connect

    Jackson, G.L.; Burrell, K.H.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Hodapp, T.; Kellman, A.G.; Lee, R.; Lippman, S.I.; Phillips, J.; Taylor, T.S.; West, W.P. ); Winter, J. . Inst. fuer Plasmaphysik); Moyer, R. ); Watkins, J. (Sandia National Labs., Livermore,

    1992-05-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

  17. Infrared thermography system on DIII-D

    SciTech Connect

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

    1990-10-01

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

  18. 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.

  19. Dust Studies in DIII-D Tokamak

    SciTech Connect

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

    2008-04-15

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

  20. Dust Studies in DIII-D Tokamak

    SciTech Connect

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

    2008-09-07

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

  1. 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.

  2. Testing Gyrokinetics on C-Mod and NSTX

    SciTech Connect

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

    2005-06-20

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

  3. Carbonization of the DIII-D tokamak

    SciTech Connect

    Jackson, G.L.; Lippmann, S.; Petrie, T.W.; DeBoo, J.C.; Ferron, J.R.; Schissel, D.P.; Taylor, T.S. ); Winters, J. . Inst. fuer Plasmaphysik); Hill, D.N. )

    1990-12-01

    The DIII-D tokamak has been carbonized by the application of a carbon film covering all plasma facing surfaces. Carbonization was done in order to reduce the metal impurity influx and central metal accumulation especially during beam heated D{sup O} {yields} D{sup +} H-mode discharges. After carbonization, nickel impurity line radiation was reduced by a factor of 10 during the ohmic phase of the discharge and up to a factor of 30 during the H-mode phase. The reduction of metal impurities also produced a reduction of total radiated power and allowed high current operation. After carbonization, the highest plasma current in a double null divertor, and the highest stored energy ever achieved on DIII-D were observed, 3 MA and 3.6 MJ respectively. A toroidal beta, {beta}{sub T} = 5.1%, at full field, 2.1 T, was obtained. On the first day after carbonization, H-mode density profiles were more peaked than pre-carbonization discharges. 20 refs., 6 figs.

  4. 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.

  5. DIII-D Advanced Tokamak Research Overview

    SciTech Connect

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

    1999-12-01

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

  6. Disruption mitigation studies in DIII-D

    NASA Astrophysics Data System (ADS)

    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. 8, 2A 441 (1985)]. 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.

  7. Restoration of the DIII-D solenoid

    SciTech Connect

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

    1997-11-01

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

  8. 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.

  9. 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.}

  10. Helium transport studies on DIII-D

    SciTech Connect

    Finkenthal, D.F.; Hillis, D.L.; Wade, M.R.; Hogan, J.T.; Klepper, C.C.; Mioduszewski, P.K.; West, W.P.; Burrell, K.H.; Seraydarian, R.P.; Groebner, R.J.; Gohil, P.

    1992-05-01

    The measurement of Helium density profiles in tokamak plasmas is necessary for helium transport studies. These studies are important in predicting the helium ash transport properties for ITER and win have important implications for the design. Poor helium transport in reactors could lead to a buildup of fusion ash, causing fuel dilution and increased radiation that will result in degraded fusion power and possibly quench ignition altogether. Present estimates indicate that He concentrations in the core must be kept below 10% in order to maintain continuous reactor operation. Helium transport studies have begun on the DM-D tokamak using charge exchange recombination (CER) spectroscopy for helium density measurements. Helium transport behavior has been observed by injecting helium gas puffs into DM-D plasmas and measuring the He density profile evolution. The profiles are used to calculate the relevant helium transport properties. This paper covers the results obtained from puffing He gas into L-mode plasmas of various electron densities. The results obtained in DIII-D L-mode plasmas are similar to measurements made at TEXTOR and JT-60.

  11. Density limit studies on DIII-D

    SciTech Connect

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

    1998-08-01

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

  12. Energetic Ion Experiments in DIII-D

    SciTech Connect

    Heidbrink, W.W.

    2005-10-15

    A summary of fast ion experiments in the DIII-D tokamak is given. Most of the experiments involve {approx}80-keV deuterium beam ions. Deceleration of dilute fast-ion populations is accurately described by coulomb scattering theory. Fast waves with frequencies several times the deuterium cyclotron frequency interact with beam ions when the product of wave number and gyroradius k{sub perpendicular{rho}}{sub i} is {approx}>1.4. Global confinement of fast ions is often excellent although sawteeth, tearing modes, and beam-driven instabilities can cause additional transport. Intense beam-ion populations often drive instabilities. Toroidicity-induced Alfven eigenmodes (TAE) and somewhat lower frequency modes (originally called beta-induced Alfven eigenmodes) are often observed in a wide variety of plasma conditions. Over 50% of the beam power is lost during strong activity. Damping mechanisms such as mode coupling or radiative damping are needed to explain the observed TAE stability threshold. The most unstable toroidal mode number agrees well with theoretical expectations, but the radial and poloidal structure of the mode and the observed beam-ion transport have not been adequately explained. The modes with frequencies below the TAE are probably two types of energetic particle modes: the resonant TAE and the resonant kinetic ballooning mode.

  13. Initial deuterium confinement studies in DIII-D tokamak: DIII-D milestone report number 58

    NASA Astrophysics Data System (ADS)

    Deboo, J. C.; Schissel, D. P.; Burrell, K.; Stjohn, H.

    1989-08-01

    The operational space of DIII-D was extended to include high current, high power deuterium neutral beam heating of deuterium discharges in both the single null and double null configurations. Recent completion of the neutron shielding of the DIII-D facility has allowed operation in these new parameter regimes while maintaining a low neutron dose at the site boundary. The plasma current and neutral beam power ranges with D(0) yields D(+) operation were extended to P(sub T) is less than or equal to 12 MW and I(sub p) is less than or equal to 2.5 MA. Comparison of H-mode discharges in pure H with those in pure D indicates a continuing superior confinement in deuterium for values of q is greater than or equal to 3; however at lower values of q the confinement becomes similar. Analysis of the energy confinement in this regime has shown no major differences in current or power scaling compared to H(0) yields D(+) operation. Energy confinement during the ELMing phase of D(0) yields D(+) H-mode discharges scales linearly with the plasma current and deteriorates with neutral beam power. Recent, ongoing time dependent analysis of discharges with long ELM-free periods following the L to H transition indicate that the energy confinement time may be independent of power so that power degradation may be a manifestation of ELMs.

  14. Ech System on the Diii-D Tokamak

    NASA Astrophysics Data System (ADS)

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

    2009-04-01

    A significant upgrade to the 110 GHz DIII-D ECH system was completed last year. Two additional Communication and Power Industries (CPI) diode gyrotrons were installed and tested to half the designed pulse length of 10 s. For the 2008 experimental campaign, the DIII-D system comprised five long pulse CPI gyrotrons. One additional high voltage power supply is being tested to support operation of up to 6 gyrotrons simultaneously at full parameters. The five gyrotrons in operation have chemical-vapor-deposition (CVD) diamond windows that are monitored by infrared camera during full parameter testing and operation during plasma experiments. A sixth CPI gyrotron has been repaired after collector failure and is being conditioned for high power at DIII-D. New equipment for gyrotron collector heat load monitoring was tested and used to measure the collector power deposition profile. A new fast fault processing system based on FPGA technology is being commissioned.

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

    NASA Astrophysics Data System (ADS)

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

    2000-10-01

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

  16. FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT

    SciTech Connect

    O'NEIL, RC; STAMBAUGH, RD

    2002-06-01

    OAK A271 FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT. The Radiative Divertor Project originated in 1993 when the DIII-D Five Year Plan for the period 1994--1998 was prepared. The Project Information Sheet described the objective of the project as ''to demonstrate dispersal of divertor power by a factor of then with sufficient diagnostics and modeling to extend the results to ITER and TPX''. Key divertor components identified were: (1) Carbon-carbon and graphite armor tiles; (2) The divertor structure providing a gas baffle and cooling; and (3) The divertor cryopumps to pump fuel and impurities.

  17. 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.

  18. DIII-D Program Results and Future Plans

    NASA Astrophysics Data System (ADS)

    Stambaugh, R. D.

    1999-03-01

    This paper summarizes recent research results and future plans for the DIII-D National Fusion Program. The full set of transparencies for this presentation can be found at http://fusion.gat.com/pubs-ext/presentations/Stambaugh_FPA00.pdf

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

    SciTech Connect

    Nerem, A.

    1995-02-01

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

  20. Fast wave current drive system design for DIII-D

    SciTech Connect

    deGrassie, J.S.; Callis, R.; Lin-Liu, Y.R.; Moeller, C..; Petty, C.C.; Phelps, D.R.; Pinsker, R.I.; Remsen, D.; Baity, F.W.; Hoffman, D.J.; Taylor, D.J.; Arnold, W.; Martin, S.

    1992-09-01

    DIII-D has a major effort underway to develop the physics and technology of fast wave electron heating and current drive in conjunction with electron cyclotron heating. The present system consists of a four strap antenna driven by one 2 MW transmitter in the 32--60 MHz band. Experiments have been successful in demonstrating the physics of heating and current drive. In order to validate fast wave current drive for future machines a greater power capability is necessary to drive all of the plasma current. Advanced tokamak modeling for DIII-D has indicated that this goal can be met for plasma configurations of interest (i.e. high {beta} VH-mode discharges) with 8 MW of transmitter fast wave capability. It is proposed that four transmitters drive fast wave antennas at three locations in DIII-D to provide the power for current drive and current profile modification. As the next step in acquiring this capability, two modular four strap antennas are in design and the procurement of a high power transmitter in the 30--120 MHz range is in progress. Additionally, innovations in the technology are being investigated, such as the use of a coupled combine antenna to reduce the number of required feedthroughs and to provide for parallel phase velocity variation with a relatively small change in frequency, and the use of fast ferrite tuners to provide millisecond timescale impedance matching. A successful test of a low power fast ferrite prototype was conducted on DIII-D.

  1. Fast wave current drive system design for DIII-D

    SciTech Connect

    deGrassie, J.S.; Callis, R.; Lin-Liu, Y.R.; Moeller, C..; Petty, C.C.; Phelps, D.R.; Pinsker, R.I.; Remsen, D. ); Baity, F.W.; Hoffman, D.J.; Taylor, D.J. ); Arnold, W.; Martin, S. )

    1992-09-01

    DIII-D has a major effort underway to develop the physics and technology of fast wave electron heating and current drive in conjunction with electron cyclotron heating. The present system consists of a four strap antenna driven by one 2 MW transmitter in the 32--60 MHz band. Experiments have been successful in demonstrating the physics of heating and current drive. In order to validate fast wave current drive for future machines a greater power capability is necessary to drive all of the plasma current. Advanced tokamak modeling for DIII-D has indicated that this goal can be met for plasma configurations of interest (i.e. high [beta] VH-mode discharges) with 8 MW of transmitter fast wave capability. It is proposed that four transmitters drive fast wave antennas at three locations in DIII-D to provide the power for current drive and current profile modification. As the next step in acquiring this capability, two modular four strap antennas are in design and the procurement of a high power transmitter in the 30--120 MHz range is in progress. Additionally, innovations in the technology are being investigated, such as the use of a coupled combine antenna to reduce the number of required feedthroughs and to provide for parallel phase velocity variation with a relatively small change in frequency, and the use of fast ferrite tuners to provide millisecond timescale impedance matching. A successful test of a low power fast ferrite prototype was conducted on DIII-D.

  2. DIII-D experimental plan for FY-1989

    SciTech Connect

    Luxon, J.L.

    1988-11-01

    This document summarizes the Experimental Plan for the DIII-D tokamak facility for the fiscal year 1989. The long-range DIII-D 5 yr plan is directed ultimately at the goal of achieving good confinement at high beta in a plasma with non-inductively driven current. This is important to the design of a steady-state reactor. This program may be thought of as occurring in two phases. In the first phase of the program we axe separately investigating high beta plasma confinement in inductively-driven plasmas, and non-inductive current drive. In the second phase we will combine these two elements to investigate high beta plasma confinement with non-inductive current drive. The FY 89 plan continues the first phase of the DIII-D experimental effort that contains a strong focus on beta and confinement in non-circular plasma configurations and in the divertor configuration in particular. Important work also continues in the development of rf heating systems for heating, profile control, and current drive. This research is coupled to theoretical efforts at General Atomics. The FY 89 research program outlined herein is diverse and multifaceted. However, it is also characterized by a greater synthesis of techniques toward a common goal. An example is the application of ECH for sawtooth suppression that would improve the low q confinement and allow higher {beta} to be obtained. We believe this research program will provide a solid foundation for the continued development of the tokamak toward high beta steady-state reactor application. The DIII-D FY 89 research program will provide results that will help resolve many CIT and ITER Physics R&D issues. In addition, DIII-D confinement studies will be an important input to the newly formed National Transport Task Force.

  3. 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.

  4. Shape Optimization for DIII-D Advanced Tokamak Plasmas

    SciTech Connect

    C.E. Kesse; J.R. Ferron; C.M. Greenfield; J.E. Menard; T.S. Taylor

    2003-07-30

    The advanced tokamak program on DIII-D is targeting the full integration of high-beta and high-bootstrap/noninductive current fraction for long-pulse lengths and the high confinement consistent with these features. Central to achieving these simultaneously is access to the highest ideal beta limits possible to maximize the headroom for experimental operation with RWM control. A study of the ideal-MHD stability is done for plasmas modeled after DIII-D advanced tokamak plasmas, varying the plasma elongation, triangularity, and outboard squareness. The highest beta(sub)N limits reach 6-7 for the n=1 kink mode for all elongation, outer squareness values, and plasma triangularity equals 0.8.

  5. Stability of TAE modes in DIII-D

    SciTech Connect

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

    1992-09-01

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

  6. 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.

  7. Overview of DIII--D 1997 Experiment Campaign

    NASA Astrophysics Data System (ADS)

    Murakami, M.

    1997-11-01

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

  8. Fast wave current drive antenna performance on DIII-D

    SciTech Connect

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

    1991-10-01

    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 {beta} target plasmas. A four-element phased-array antenna is used to launch a directional wave spectrum with the peak n{sub {parallel}} value ({approx equal} 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{degree}) 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.

  9. Fast wave current drive antenna performance on DIII-D

    SciTech Connect

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

    1991-10-01

    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 {beta} target plasmas. A four-element phased-array antenna is used to launch a directional wave spectrum with the peak n{sub {parallel}} value ({approx_equal} 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{degree}) 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.

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

    SciTech Connect

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

    2006-07-15

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

  11. 3D DIII-D Equilibrium Calculations with Magnetic Islands

    NASA Astrophysics Data System (ADS)

    Reiman, Allan; Monticello, Don; Lazerson, Sam

    2012-10-01

    We discuss 3D equilibrium calculations for the DIII-D tokamak that have been performed using the PIES code, focusing particularly on a single shot that has been studied by a number of different codes in the context of the FY 2012 FES Theory Milestone. The shot was part of an experimental campaign to study the suppression of edge localized modes (ELMs) using an externally imposed nonaxisymmetric magnetic perturbation.

  12. 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.

  13. Development of Negative Triangularity Plasmas in DIII-D

    NASA Astrophysics Data System (ADS)

    Walker, M. L.

    2015-11-01

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

  14. Modeling 3-D Effects in the DIII-D Boundary

    NASA Astrophysics Data System (ADS)

    Evans, T. E.; Moyer, R. A.; Reiter, D.; Kasilov, S. V.; Runov, A. M.

    2002-11-01

    Resonant magnetic perturbations δ br from the DIII-D locked and resistive wall mode control coils (C-coil and I-coil, respectively) affect ne and Te profiles in both the pedestal and core. To understand why these δ br perturbations change the plasma profiles we first model the edge magnetic topology with a field line integration code, TRIP3D code. In general, the TRIP3D results indicate that the control coils create stochastic layers with as much as 25% edge magnetic flux connected to the divertors and walls. While heat and particle transport modeling in open stochastic layers is inherently very difficult, Monte Carlo methods appear to provide the most reasonable approach with which to address these issues. As such, we have assessed the possibility of coupling a recently developed Monte Carlo heat transport code, the E3D code, [A.M. Runov et al., Phys. Plasmas 8, 916 (2001)] to TRIP3D. We will discuss how this coupling can best be accomplished and what must be done to benchmark the TRIP3D/E3D ensemble using DIII-D experimental data. We will also discuss the analysis of proposed designs for a dedicated DIII-D stochastic boundary layer coil which produce minimal δ br core perturbations.

  15. Divertor and midplane materials evaluation system in DIII-D

    NASA Astrophysics Data System (ADS)

    Wong, C. P. C.; Rudakov, D. L.; Allain, J. P.; Bastasz, R. J.; Brooks, N. H.; Brooks, J. N.; Doerner, R. P.; Evans, T. E.; Hassanein, A.; Jacob, W.; Krieger, K.; Litnovsky, A.; McLean, A. G.; Philipps, V.; Pigarov, A. Yu.; Wampler, W. R.; Watkins, J. G.; West, W. P.; Whaley, J.; Wienhold, P.

    2007-06-01

    The Divertor Materials Evaluation System (DiMES) at General Atomics has successfully advanced the understanding of plasma surface interaction phenomena involving ITER-relevant materials and has been utilized for advanced diagnostic designs in the lower divertor of DIII-D. This paper describes a series of recent successful experiments. These include the study of carbon deposition in gaps and metallic mirrors as a function of temperature, study of dust migration from the divertor, study of methane injection in order to benchmark chemical sputtering diagnostics, and the measurement of charge exchange neutrals with a hydrogen sensor. In concert with the modification of the lower divertor of DIII-D, the DiMES sample vertical location was modified to match the raised divertor floor. The new Mid-plane Material Exposure Sample (MiMES) design will also be presented. MiMES will allow the study and measurement of erosion and redeposition of material at the outboard mid-plane of DIII-D, including effects from convective transport. We will continue to expose relevant materials and advanced diagnostics to different plasma configurations under various operational regimes, including material erosion and redeposition experiments, and gaps and mirror exposures at elevated temperature.

  16. 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.

  17. 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.

  18. 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

  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. Recent results from the DIII-D tokamak

    SciTech Connect

    Petersen, P.I.

    1998-02-01

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

  1. 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.

  2. 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.

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

    SciTech Connect

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

    1993-10-01

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

  4. 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.

  5. Plasma rotation and rf heating in DIII-D

    SciTech Connect

    DeGrassie, J. S.; Baker, D. R.; Burrell, K. H.; Greenfield, C. M.; Lin-Liu, Y. R.; Luce, T. C.; Petty, C. C.; Prater, R.; Heidbrink, W. W.; Rice, B. W.

    1999-09-20

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

  6. Plasma rotation and rf heating in DIII-D

    SciTech Connect

    Grassie, J. S. de; Baker, D. R.; Burrell, K. H.; Greenfield, C. M.; Lin-Liu, Y. R.; Luce, T. C.; Petty, C. C.; Prater, R.; Heidbrink, W. W.; Rice, B. W.

    1999-09-20

    In a variety of discharge conditions on DIII-D it is observed that rf electron heating reduces the toroidal rotation speed and core ion temperature. The rf heating can be with either fast wave or electron cyclotron heating and this effect is insensitive to the details of the launched toroidal wavenumber spectrum. To date all target discharges have rotation first established with co-directed neutral beam injection. A possible cause is enhanced ion momentum and thermal diffusivity due to electron heating effectively creating greater anomalous viscosity. Another is that a counter directed toroidal force is applied to the bulk plasma via rf driven radial current. (c) 1999 American Institute of Physics.

  7. A design retrospective of the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Luxon, J. L.

    2002-05-01

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

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

    SciTech Connect

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

    1988-02-24

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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Development of a radiative divertor for DIII-D

    SciTech Connect

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

    1994-07-01

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

  16. Future directions of the DIII-D tokomak program

    SciTech Connect

    Simonen, T.C. )

    1992-06-01

    The author discusses possible future direction for the DII-D tokamak in terms of developing the US and international tokamak research programs. DII-D is the second largest US magnetic fusion facility; an advanced tokamak and divertor research device for addressing plasma confinement and control issues, a testbed for demonstrating and developing fusion technology components, and a proving ground for applied physics theories and diagnostic instrumentation. DII-D is the largest non-circular tokamak in the US and is operated with extensive national and international collaboration, including a long-standing Japanese collaboration and major multi-disciplinary collaborations with LLNL, ORNL, and UCLA. The DIII-D program is now demonstrating advanced tokamak operating modes transiently and plans in future experiments to sustain these improved confinement, stability, and divertor modes with non-inductive rf current drive. The long range research program focuses on divertor and advanced tokamak issues. The goal is to provide an integrated demonstration of well-confined high-beta divertor plasma with non-inductive current drive. The author presents ways that DIII-D can contribute to technology issues relevant to ITER physics issues, with the requesite investment of resources.

  17. A TALIF Diagnostic for the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Magee, R. M.; Galante, M. E.; McCarren, D. W.; Scime, E. E.; Boivin, R. L.; Brooks, N. H.

    2011-10-01

    The density profile of hydrogenic neutrals in the edge of DIII-D plays an important role in the problems of momentum transport, pedestal formation, and plasma-wall interaction, but an accurate measurement has proven difficult. A two-photon absorption laser induced fluorescence (TALIF) diagnostic is under construction and is intended to provide temporally and spatially resolved neutral density measurements in the pedestal region. This three-level TALIF scheme offers the advantages of direct excitation of ground state atoms, emission in the visible portion of the spectrum, a high degree of spatial localization, and the potential for a Doppler-free measurement. The large background of Dα emission, the principal challenge of the measurement, can be overcome by the focusing of a high power (1 MW) UV laser. Calculations of the SNR show that densities of 1015 m-3 or lower can be measured with a spatial resolution of 0.3 mm. We present design details of the proposed laser system, calculations of the expected performance in DIII-D and in a helicon source plasma, and measurements of the HI profile in the helicon plasma. Work supported in part by US DOE under DE-FC02-04ER54698.

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

    SciTech Connect

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

    1992-10-01

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

  19. Tritium in the DIII-D carbon tiles

    SciTech Connect

    Taylor, P.L.; Kellman, A.G.; Lee, R.L.

    1993-06-01

    The amount of tritium in the carbon tiles used as a first wall in the DIII-D tokamak was measured recently when the tiles were removed and cleaned. The measurements were made as part of the task of developing the appropriate safety procedures for processing of the tiles. The surface tritium concentration on the carbon tiles was surveyed and the total tritium released from tile samples was measured in test bakes. The total tritium in all the carbon tiles at the time the tiles were removed for cleaning is estimated to be 15 mCi and the fraction of tritium retained in the tiles from DIII-D operations has a lower bound of 10%. The tritium was found to be concentrated in a narrow surface layer on the plasma facing side of the tile, was fully released when baked to 1,000{degree}C, and was released in the form of tritiated gas (DT) as opposed to tritiated water (DTO) when baked.

  20. 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.

  1. 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.

  2. High Performance Stationary Discharges in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Luce, T. C.

    2003-10-01

    Qualification of scenarios for burning plasma experiments requires demonstration of performance on the resistive equilibration timescale. Progress has been made in DIII-D in demonstrating performance on this timescale ( ˜2 s) at or well above the level needed in next-step devices such as ITER. By raising the energy content before the onset of sawteeth, two types of stationary discharges have been obtained with high fusion performance. The first occurs at edge safety factor q_95 > 4. In this case, sawteeth never appear and the plasma is held by feedback at the estimated no-wall beta limit for >2 s. In this case, βN = 3.2 and confinement quality (H_89 = 2.5) is high. For comparison to burning plasma projections, the fusion performance figure-of-merit G = βN H_89/q_95^2 is used. For the baseline Q = 10 goal in ITER, G = 0.42 at q_95 = 3. Higher q_95 discharges in DIII-D achieve G = 0.41. This has two substantial implications for ITER. First, the baseline goal of Q = 10 could be met at 75% of the rated current with corresponding reduction of risk from disruption. Second, the inductive capability of ITER could sustain this level of performance for >1 h, opening the possibility for significant nuclear testing (hybrid scenarios). The second class of discharges exists for q_95 < 4. Sawteeth appear, but the effect of the early heating seems to soften their impact. Discharges have been operated for >2 s with βN = 2.8 and H_89 = 2.3, corresponding to G = 0.63 at q_95 = 3.2 or 50% higher than the ITER baseline scenario. We will compare these discharges to the conventional scenarios achieved in DIII-D and promising advanced tokamak discharges with q_min > 1.5 which approach full noninductive operation. Outstanding physics issues in the projection to next-step devices will be addressed.

  3. DIII-D Research in Support of ITER

    SciTech Connect

    Strait, E

    2008-10-14

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

  4. 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.

  5. 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.

  6. 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.

  7. Calorimetric Power Measurements of the DIII-D Gyrotron System

    NASA Astrophysics Data System (ADS)

    Gorelov, I.; Lohr, J.; Callis, R. W.; Cary, W. P.; Ponce, D.; Pinsker, R. I.; Chiu, H.; Baity, F. W.

    2001-10-01

    Gyrotron power measurements are an integral part of rf experiments on DIII-D. The ECH complex at General Atomics is built up from four 110 GHz, 1 MW gyrotrons, one from Communication and Power Industry (CPI) and three from Russia's Gyrotron Company (Gycom). Power measurements are made calorimetrically using the temperature and flow measurements of the gyrotron cooling circuits. Three such circuits are measured on the Gycom gyrotrons: window, MOU and dummy load. Interior cooling circuits are additionally measured on the CPI gyrotron that are very useful when tuning for maximum power and efficiency. Calorimetric signals from each cooling circuit are acquired by computer, where the dissipated energy is calculated with a LabView program. From these calculations, total rf power and efficiency were determined. Thus, calorimetry measurements were effectuated during gyrotron operations to provide the average power of each pulse.

  8. Experimental studies of fast wave propagation in DIII-D

    SciTech Connect

    Ikezi, H.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.

    1995-06-01

    Fast Alfven waves radiated from the phased array antenna in the DIII-D tokamak and used for heating and current drive are studied by employing a B-loop array mounted on the vacuum vessel wall. The wave propagation direction controlled by the antenna phasing is clearly observed. A small divergence of the rays arising from the anisotropic nature of the fast wave is found. Comparison with a ray tracing code confirms that the ray position calculated by the code is accurate up to at least one toroidal turn of the rays. Conservation of Rk{sub t} which is a basic assumption in computer codes is tested. Although the upshift of toroidal wavenumber k{sub t} at small major radius R is confirmed, Rk{sub t} is not well conserved. A mass density interferometer is demonstrated by employing the extraordinary fast wave.

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

    DOE PAGES

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

    2016-08-02

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

  10. 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.

  11. Recent DIII-D neutral beam calibration results

    SciTech Connect

    Wight, J.; Hong, R.M.; Phillips, J.

    1991-10-01

    Injected DIII-D neutral beam power is estimated based on three principle quantities: the fraction of ion beam that is neutralized in the neutralizer gas cell, the beamline transmission efficiency, and the fraction of beam reionized in the drift duct. System changes in the past few years have included a new gradient grid voltage operating point, ion source arc regulation, routine deuterium operations and new neutralizer gas flow controllers. Additionally, beam diagnostics have been improved and better calibrated. To properly characterize the beams the principle quantities have been re-measured. Two diagnostics are primarily used to measure the quantities. The beamline waterflow calorimetry system measures the neutralization efficiency and the beamline transmission efficiency, and the target tile thermocouples measure the reionization loss. An additional diagnostic, the target tile pyrometer, confirmed the reionization loss measurement. Descriptions and results of these measurements will be presented. 4 refs., 5 figs., 2 tabs.

  12. Thomson scattering diagnostic upgrade on DIII-D

    SciTech Connect

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

    2010-10-15

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

  13. Detached divertor operation in DIII-D helium plasmas

    SciTech Connect

    Hill, D. N., LLNL

    1998-05-01

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

  14. Measurements of the neutron source strength at DIII-D

    SciTech Connect

    Heidbrink, W.W.; Taylor, P.L.; Phillips, J.A.

    1997-01-01

    A set of neutron counters and a pair of scintillators measure the 2.5 MeV neutron emission produced by the DIII-D tokamak. The neutron counter set provides a large dynamic range ({approximately}7 orders of magnitude) while the scintillators provide the very fast resolution needed for studying transient events. The counters are absolutely calibrated {ital in situ} with a {sup 252}Cf source and the scintillators are cross calibrated to the counters. The historic variations in the emission measured by the various detectors have been compared and are consistent within the estimated accuracy of the absolute calibration (15{percent}). In the discharges with the highest emission levels (2.4{times}10{sup 16} n/s), the signals from the neutron counters and the scintillators agree well. Comparisons with other diagnostics also corroborate the neutron measurements.{copyright} {ital 1997 American Institute of Physics.}

  15. Closed Loop Feedback of MHD Instabilities on DIII-D

    SciTech Connect

    Fredrickson, E.D.; Bialek, J.; Garofalo, A.M.; Johnson, L.C.; La Haye, R.J.; Lazarus, E.A.

    2001-01-16

    A system of coils, sensors and amplifiers has been installed on the DIII-D tokamak to study the physics of feedback stabilization of low-frequency MHD [magnetohydrodynamic] modes such as the Resistive Wall Mode (RWM). Experiments are being performed to assess the effectiveness of this minimal system and benchmark the predictions of theoretical models and codes. In the last campaign, the experiments have been extended to a regime where the RWM threshold is lowered by a fast ramp of the plasma current. In these experiments, the onset time of the RWM is very reproducible. With this system, the onset of the RWM has been delayed by up to 100 msec without degrading plasma performance. The growth rate of the mode increases proportional to the length of delay, suggesting that the plasma is evolving towards a more unstable configuration. The present results have suggested directions for improving the feedback system including better sensors and improved feedback algorithms.

  16. Personal computer applications in DIII-D neutral beam operation

    SciTech Connect

    Glad, A.S.

    1986-08-01

    An IBM PC AT has been implemented to improve operation of the DIII-D neutral beams. The PC system provides centralization of all beam data with reasonable access for on-line shot-to-shot control and analysis. The PC hardware was configured to interface all four neutral beam host minicomputers, support multitasking, and provide storage for approximately one month's accumulation of beam data. The PC software is composed of commercial packages used for performance and statistical analysis (i.e., LOTUS 123, PC PLOT, etc.), host communications software (i.e., PCLink, KERMIT, etc.), and applications developed software utilizing f-smcapso-smcapsr-smcapst-smcapsr-smcapsa-smcapsn-smcaps and b-smcapsa-smcapss-smcapsIc-smcaps. The objectives of this paper are to describe the implementation of the PC system, the methods of integrating the various software packages, and the scenario for on-line control and analysis.

  17. Recent sheath physics studies on DIII-D

    NASA Astrophysics Data System (ADS)

    Watkins, J. G.; Labombard, B.; Stangeby, P. C.; Lasnier, C. J.; McLean, A. G.; Nygren, R. E.; Boedo, J. A.; Leonard, A. W.; Rudakov, D. L.

    2015-08-01

    A study to examine some current issues in the physics of the plasma sheath has been recently carried out in DIII-D low power Ohmic plasmas using both flush and domed Langmuir probes, divertor Thomson scattering (DTS), an infrared camera (IRTV), and a new calorimeter triple probe assembly mounted on the Divertor Materials Evaluation System (DIMES). The sheath power transmission factor was found to be consistent with the theoretically predicted value of 7 (±2) for low power plasmas. Using this factor, the three heat flux profiles derived from the LP, DTS, and calorimeter diagnostic measurements agree. Comparison of flush and domed Langmuir probes and divertor Thomson scattering indicates that proper interpretation of flush probe data to get target plate density and temperature is feasible and could potentially yield accurate measurements of target plate conditions where the probes are located.

  18. 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.

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

    SciTech Connect

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

    1999-08-01

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

  20. 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.

  1. 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.

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

    SciTech Connect

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

    1999-06-01

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

  3. 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.

  4. The DIII-D 3 MW, 110 GHz ECH system

    NASA Astrophysics Data System (ADS)

    Callis, R. W.; Lohr, J.; Ponce, D.; O'Neill, R. C.; Prater, R.; Luce, T. C.

    1999-09-01

    Three 110 GHz gyrotrons with nominal output power of 1 MW each have been installed and are operational on the DIII-D tokamak. One gyrotron is built by Gycom and has a nominal rating of 1 MW and a 2 s pulse length, with the pulse length being determined by the maximum temperature allowed on the edge cooled Boron Nitride window. The second and third gyrotrons were built by Communications and Power Industries (CPI). The first CPI gyrotron uses a double disc FC-75 cooled sapphire window which has a pulse length rating of 0.8 s at 1 MW, 2 s at 0.5 MW and 10 s at 0.35 MW. The second CPI gyrotron, utilizes a single disc chemical-vapor-deposition diamond window, that employs water cooling around the edge of the disc. Calculation predict that the diamond window should be capable of full 1 MW cw operation. All gyrotrons are connected to the tokamak by a low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HE11 mode. Each waveguide system incorporates a two mirror launcher which can steer the rf beam poloidally from the center to the outer edge of the plasma. Central current drive experiments with the two gyrotrons with 1.5 MW of injected power drove about 0.17 MA. Results from using the three gyrotron systems will be reported as well as the plans to upgrade the system to 6 MW.

  5. Upgrades for the ECH System on DIII-D

    NASA Astrophysics Data System (ADS)

    Lohr, J.; Cengher, M.; Doane, J. L.; Gorelov, Y. A.; Moeller, C. P.; Ponce, D.; Noraky, S.; Penaflor, B. G.; Kolemen, E.

    2012-10-01

    The gyrotron system for electron cyclotron heating on the DIII-D tokamak is being upgraded with the addition of higher efficiency gyrotrons having collector potential depression. Two new gyrotrons, operating at the present frequency of 110 GHz and generating 1.2 MW per unit, have been manufactured and are being installed and tested. The subsequent group of gyrotrons have been designed to generate 1.5 MW for 10 s pulses at 117.5 GHz. The first of these tubes is presently being manufactured at Communications and Power Industries. By the end of 2013, the system will comprise eight high power gyrotrons and, pending the successful performance of the 1.5 MW tube, an upgrade to a 15 MW system will begin. High voltage power supplies, transmission lines, launchers and associated control and data acquisition systems are included in the upgrades as is enhanced ability to steer the rf beams under a variety of pre-programed and reactive scenarios.

  6. 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.

  7. Ion Temperature Measurements in the DIII--D Divertor

    NASA Astrophysics Data System (ADS)

    Brooks, N. H.; Isler, R. C.; McKee, G. R.; Tugarinov, S.

    1996-11-01

    Doppler profile measurements of the D_α, He II, C II and C III line emission in the DIII--D divertor have been performed with two high resolution spectrometers: an instrument of Russian design with high optical throughput and 7 ms readout, and a conventional Czerny Turner spectrometer with slower response time, but greater dynamic range in its detector system. In continuous ELMing H--mode plasmas the Doppler profiles are usually single-gaussian, but during operation at low density or during large, discrete ELM events the profiles of the hydrogen-like species are often multi-gaussian. Comparison of ion temperatures inferred from the single-gaussian profiles and electron temperatures measured by Thomson scattering and by spectroscopic line ratio methods yields good agreement for the higher charge states, where equilibration of ion and electron temperatures is expected. When strong D2 puffing triggers the MARFE-like conditions of Partially Detached Divertor plasmas, the D_α line profile, usually a complex asymmetric profile with multiple components, evolves into a single-gaussian profile fitted by a very low temperature (<2 eV) similar to that measured for the electrons by Thomson scattering.

  8. Central Thomson scattering upgrade on DIII-D

    SciTech Connect

    Nilson, D.G.; Stallard, B.W.; Carlstrom, T.N.; Hsieh, C.L.; Stockdale, R.E.

    1997-11-01

    The existing 36 channel Thomson scattering system on the DIII-D tokamak measures the plasma temperature and density in the core region. However, with the recent interest in core peaked density profiles, coverage needs to be extended into the magnetic axis. This paper addresses the technical issues involved with extending the viewing region from a major radius of 194 cm to 165 cm. At least one of the existing seven core laser beams will be rerouted to probe the plasma horizontally instead of vertically. To do this, a rigid extension of the existing laser/collection optics tower will be built to route the laser to a nearby tangential port. A fiber bundle array from one of the two existing core plasma collection optics sets will be rotated to allow up to 10 of the 36 core channels to view along this new beam path. A new in-vessel absorbing glass laser dump must be developed since there are no appropriate laser beam exit ports. The close proximity of this laser dump to the viewing region presents stray light issues that must be resolved to allow for an accurate density calibration using Rayleigh scattering in argon gas.

  9. Experimental studies of fast wave propagation in DIII-D

    SciTech Connect

    Ikezi, H.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.

    1996-02-01

    Fast Alfv{acute e}n waves radiated from the phased array antenna in the DIII-D tokamak and used for heating and current drive are studied by employing a {dot {ital B}}-loop array mounted on the vacuum vessel wall. The wave propagation direction controlled by the antenna phasing is clearly observed. A small divergence of the rays arising from the anisotropic nature of the fast wave is found. Comparison with a ray tracing code confirms that the ray position calculated by the code is accurate up to at least one toroidal turn of the rays. Conservation of {ital Rk}{sub t} which is a basic assumption in computer codes is tested. Although the upshift of toroidal wavenumber {ital k}{sub t} at small major radius {ital R} is confirmed, {ital Rk}{sub {ital t}} is not well conserved. A mass density interferometer is demonstrated by employing the extraordinary fast wave. {copyright} {ital 1996 American Institute of Physics.}

  10. 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.

  11. 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.

  12. Simulation of DIII-D Flat q Discharges

    SciTech Connect

    C.E. Kessel; A. Garofalo; T. Terpstra

    2004-06-25

    The Advanced Tokamak plasma configuration has significant potential for the economical production of fusion power. Research on various tokamak experiments are pursuing these plasmas to establish high {beta}, high bootstrap current fraction, 100% noninductive current, and good energy confinement, in a quasi-stationary state. One candidate is the flat q discharge produced in DIII-D, where the safety factor varies from 2.0 on axis, to slightly below 2.0 at the minimum, and then rises to about 3.5 at the 95% surface. This plasma is prototypical of those studied for power plants in the ARIES tokamak studies. The plasma is produced by ramping up the plasma current and ramping down the toroidal field throughout the discharge. The plasma current reaches 1.65 MA, and the toroidal field goes from 2.25 to 1.6 T. The q{sub min} remains high and at large radius, {rho} {approx} 0.6. The plasma establishes an internal transport barrier in the ion channel, and transitions to H-mode. The free-boundary Tokamak Simulation Code (TSC) is being used to model the discharge and project the impact of changes in the plasma current, toroidal field, and injected power programming.

  13. 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.

  14. 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.

  15. Gamma ray imager on the DIII-D tokamak.

    PubMed

    Pace, D C; Cooper, C M; Taussig, D; Eidietis, N W; Hollmann, E M; Riso, V; Van Zeeland, M A; Watkins, M

    2016-04-01

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1-60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons. PMID:27131674

  16. 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).

  17. 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.

  18. 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.

  19. Array of neutral particle analyzers at DIII-D

    SciTech Connect

    Carolipio, E.M.; Heidbrink, W.W.

    1997-01-01

    Local measurements of the fast-ion distribution in auxiliary-heated plasmas are key to understanding the behavior of energetic particles under a variety of conditions, such as beam{endash}ion transport during Alfv{acute e}n instabilities and the acceleration of beam ions by fast waves. For the first time at DIII-D, line-averaged and local measurements of the energetic-particle density (for E=5{endash}75 keV) are possible using an array of four compact charge{endash}exchange analyzers [P. Beiersdorfer {ital et al.}, Rev. Sci. Instrum. {bold 58}, 2092 (1987)]. The installation consists of three vertically viewing analyzers with fixed sightlines, measuring particles with {chi}=90{degree} (where {chi} is the angle between the particle{close_quote}s velocity and the toroidal direction), and one horizontally viewing analyzer with a variable sightline, measuring particles with 2{degree}{approx_lt}{chi}{approx_lt}60{degree}. All the analyzers can make passive measurements while three detectors, with sightlines that intersect deuterium heating beams, can make active charge{endash}exchange measurements. {copyright} {ital 1997 American Institute of Physics.}

  20. Array of neutral particle analyzers at DIII-D

    SciTech Connect

    Carolipio, E.M.; Heidbrink, W.W.

    1996-12-01

    Local measurements of the fast-ion distribution in auxiliary-heated plasmas are key to understanding the behavior of energetic particles under a variety of conditions, such as beam-ion transport during Alfven instabilities and the acceleration of beam ions by fast waves. For the first time at DIII-D, line-averaged and local measurements of the energetic-particle density (for E = 5--75 keV) are possible using an array of four compact charge-exchange analyzers. The installation consists of three vertically-viewing analyzers with fixed sightlines, measuring particles with {chi} = 90{degree} (where {chi} is the angle between the particle`s velocity and the toroidal direction) and one horizontally-viewing analyzer with a variable sightline, measuring particles with 2{degree}{grave U} {chi} {acute U} 60{degree}. All the analyzers can make passive measurements while three detectors, with sightlines that intersect deuterium heating beams, can make active charge-exchange measurements.

  1. Kinetic Alfven eigenmodes in JET and DIII-D

    SciTech Connect

    Jaun, A.; Hellsten, T.; Heidbrink, W.W.; Carolipio, E.

    1996-12-31

    Kinetic effects are studied for global Alfven eigenmodes in realistic tokamak equilibria with finite aspect ratio and plasmas, comparing calculations from the full wave code PENN with experimental measurements. The kinetic plasma model is based on a Larmor radius expansion in toroidal geometry and takes into account the gradients in the equilibrium density and temperatures. It allows for a consistent description of the mode conversion to the kinetic Alfven wave (KAW) and the effect of diamagnetic drifts on electromagnetic waves. Comparisons axe first carried out for a JET discharge, showing that multiple peeks measured in the low frequency Alfven spectrum are the signature of kinetic Alfven eigenmodes (KAE) induced through coupling between a global ellipticity Alfven eigenmode (EAE) and the KAW. In general, series of modes appear in the proximity of global fluid modes, some with a regular spacing in frequency and a very weak Landau damping of {vert_bar}{gamma}/{omega}{vert_bar} {approx_equal} 0.0007. A kinetic analysis of a DIII-D discharge shows that TAE mode wavefields reach the plasma core through electromagnetic drift waves which propagate because of finite temperature gradients in the regions of small k{sub {parallel}}. They can lead to particle diffusion and may explain the large losses of beam ions observed during the TAE instabilities. Comparisons of frequency and eigenmode structure axe carried out for resistive and kinetic models, between the theoretical calculations using the PENN code and the experimental measurements from magnetic probes.

  2. Multipoint Thomson scattering diagnostic for DIII-D

    NASA Astrophysics Data System (ADS)

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

    1988-02-01

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

  3. Multipoint Thomson scattering diagnostic for DIII-D

    NASA Astrophysics Data System (ADS)

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

    1988-08-01

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

  4. 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.

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

    NASA Astrophysics Data System (ADS)

    Lopez, N. A.; Crowley, B.

    2014-10-01

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

  6. Turbulent Ion Temperature Fluctuation Measurements on DIII-D

    NASA Astrophysics Data System (ADS)

    Uzun-Kaymak, I. U.; Fonck, R. J.; McKee, G. R.

    2012-10-01

    A novel dual-channel, high throughput, high efficiency, charge exchange spectrometer (UF-CHERS) has been developed to measure impurity ion temperature and toroidal velocity fluctuations with 1 μs time resolution. These measurements are primarily needed for identifying underlying instabilities and validation of transport simulation codes. Based on the measured photon flux levels for the entire spectral line, a photon noise floor ˜1% is expected. Statistical averaging over long data records should provide turbulence measurements to 1/5 to 1/10 of the estimated photon noise floor. Correlation measurements in DIII-D plasmas demonstrate broadband ion-temperature fluctuations from 0-150 kHz, while cross-correlation with BES measurements of density fluctuations exhibits cross-power between density and ion temperature fluctuations to 250 kHz in ECH-heated low-collisionality L-mode discharges. The fluctuation level is Ti/Ti˜2%, with correlation lengths in the range of 3-5 cm. A new detector system will provide improved signal-to-noise level and higher frequency measurements.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Divertor heat and particle control experiments on the DIII-D tokamak

    SciTech Connect

    Mahdavi, M.A; Baker, D.R.; Allen, S.L.

    1994-05-01

    In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D{sub 2} gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models.

  14. 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.

  15. GYRO Simulations of Core Momentum Transport in DIII-D and JET Plasmas

    SciTech Connect

    R.V. Budny; J. Candy; R.E. Waltz; and contributors to the DIII-D and JET-EFDA work programs

    2005-06-27

    Momentum, energy, and particle transport in DIII-D and JET ELMy H-mode plasmas is simulated with GYRO and compared with measurements analyzed using TRANSP. The simulated transport depends sensitively on the nabla(T(sub)i) turbulence drive and the nabla(E(sub)r) turbulence suppression inputs. With their nominal values indicated by measurements, the simulations over-predict the momentum and energy transport in the DIII-D plasmas, and under-predict in the JET plasmas. Reducing |nabla(T(sub)i)| and increasing |nabla(E(sub)r)| by up to 15% leads to approximate agreement (within a factor of two) for the DIII-D cases. For the JET cases, increasing |nabla(T(sub)i)| or reducing |nabla(E(sub)r)| results in approximate agreement for the energy flow, but the ratio of the simulated energy and momentum flows remains higher than measurements by a factor of 2-4.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Transport in high performance weak and negative central shear discharges in DIII-D

    SciTech Connect

    Greenfield, C.M.; Schissel, D.P.; Stallard, B.W.

    1996-07-01

    In recent experiments in the DIII-D tokamak, the previously reported enhanced performance regime with negative central magnetic shear (NCS) has been extended to further improve fusion performance. This was done by using controlled L-H transitions to further broaden the pressure profile, thereby delaying the onset of MHD activity which would lead to the termination of the high performance phase. Such discharges have achieved record parameters for DIII-D, including D-D fusion power up to 28 kW and stored energy in excess of 4 MJ.

  2. 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.

  3. 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.

    1998-05-18

    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 alloy structure in the DIII-D radiative diverter upgrade. Data from the test series indicate that the performance of the V-4Cr4Ti 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.

  4. 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.

  5. Wide-view charge exchange recombination spectroscopy diagnostic for Alcator C-Mod

    SciTech Connect

    Rowan, W. L.; Bespamyatnov, I. O.; Granetz, R. S.

    2008-10-15

    This diagnostic measures temperature, density, and rotation for the fully stripped boron ion between the pedestal top and the plasma core with resolution consistent with the profile gradients. The diagnostic neutral beam used for the measurements generates a 50 keV, 6 A hydrogen beam. The optical systems provide views in both poloidal and toroidal directions. The imaging spectrometer is optimized to simultaneously accept 45 views as input with minimum cross-talk. In situ calibration techniques are applied for spatial location, spectral intensity, and wavelength. In the analysis, measured spectra are fitted to a model constructed from a detailed description of the emission physics. Methods for removal of interfering spectra are included. Applications include impurity and thermal transport.

  6. Comparison of impurity transport in Alcator C-Mod with fluid models of drift wave turbulence

    NASA Astrophysics Data System (ADS)

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

    2010-11-01

    Using a new theory, we investigate the influence of the impurity density and impurity density gradient on turbulent particle transport. Heavy impurities (argon is the example here) appear to have the strongest influence while light impurities (boron, for example) may simply transport as passive tracers when in the presence of a significant quantity of heavy impurity. The theory describes how collisional and trapped electron drift wave dynamics are modified by impurities and how the turbulence transports the impurities in a background hydrogenic plasma. The collisional regime is described using a Hasegawa-Wakatani system of equations. The trapped electron mode is modeled with a generalized form of the Terry-Horton system of equations. Measured positive and negative impurity gradients can be predicted, but new experiments will be required to verify the implications of the theory.

  7. 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.

  8. 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.

  9. Investigation of the heat handling capabilities of DIII-D neutral beamline internal components

    SciTech Connect

    Phillips, J.C.; Baxi, C.B.; Hong, R.

    1993-10-01

    The current DIII-D neutral beam system is a nominal five second pulse length upgrade of a previous design rated for only 500 msec operation. While the ion sources are rated for 60 sec operation, in practice pulse lengths are limited both by the beamline internal components ability to handle the fraction of the power which is scraped off, and by the power supplies ability to provide pulse lengths of greater than 5 sec. This paper examines the capability of the existing DIII-D neutral beamline heat removing components both in terms of present and desired operating parameters. To date, at 2.5 MW per ion source, pulses are limited to 3.5 sec by beamline internal components, while at lower ratings of 2.0 MW per ion source, up to 5 sec pulses have been achieved. Recent advances and demonstration of the extraction of 3.5 MW per DIII-D ion source give an even wider window of operating conditions. A full series of beamline thermocouple data has been collected to determine the heat loading and implied surface temperatures for the various DIII-D neutral beamline internal components. These data will be presented along with an analysis of the needs for specific component upgrades, given a desire for 10 sec operation.

  10. Alpha-channeling simulation experiment in the DIII-D tokamak.

    PubMed

    Wong, K L; Budny, R; Nazikian, R; Petty, C C; Greenfield, C M; Heidbrink, W W; Ruskov, E

    2004-08-20

    Alfvén instabilities can reduce the central magnetic shear via redistribution of energetic ions. They can sustain a steady state internal transport barrier as demonstrated in this DIII-D tokamak experiment. Improvement in burning plasma performance based on this mechanism is discussed.

  11. RECENT DEVELOPMENTS IN ALTERNATIVES TO CAMAC FOR DATA ACQUISITION AT DIII-D

    SciTech Connect

    KELLMAN,D.H; CAMPBELL,G.L; FERRON,J.R; PIGLOWSKI,D.A; AUSTIN,M.E; MCKEE,G.R

    2003-10-01

    OAK-B135 For over twenty years, data acquisition hardware at DIII-D has been based on the CAMAC platform. These rugged and reliable systems, however, are gradually becoming obsolete due to end-of-life issues, ever-decreasing industry support of older hardware, and the availability of modern alternative hardware with superior performance. Efforts are underway at DIII-D to adopt new data acquisition solutions which exploit modern technologies and surpass the limitations of the CAMAC standard. These efforts have involved the procurement and development of data acquisition systems based on the PCI and Compact-PCI platform standards. These systems are comprised of rack-mount computers containing data acquisition boards (digitizers), Ethernet connectivity, and the drivers and software necessary for control. Each digitizer contains analog-to-digital converters, control circuitry, firmware and memory to collect, store, and transfer waveform data acquired using internal or external triggers and clocks. Software has been developed which allows DIII-D computers to program the operational parameters of the digitizers, as well as to upload acquired data into the DIII-D acquisition database. All communication between host computers and the new acquisition systems occurs via standard Ethernet connections, a vast improvement over the slower, serial loop highways used for control and data transfer with CAMAC systems. In addition, the capabilities available in modern integrated and printed circuit manufacture result in digitizers with high channel count and memory density. Cost savings are also realized by utilizing a platform based on standards of the personal computer industry. Details of the new systems at DIII-D are presented, along with initial experience with their use, and plans for future expansion and improvement.

  12. 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.

  13. 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.

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

    NASA Astrophysics Data System (ADS)

    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.

    2016-02-01

    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. 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.

  15. 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.

  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. Model-based Adaptive Control of Resistive Wall Modes in DIII-D

    NASA Astrophysics Data System (ADS)

    Xie, F.; Schuster, E.; Humphreys, D. A.; Walker, M. L.

    2009-11-01

    One of the major non-axisymmetric instabilities under study in the DIII-D tokamak is the resistive wall mode (RWM), a form of plasma kink instability whose growth rate is moderated by the influence of a resistive wall. The General Atomics/FARTECH DIII-D/RWM dynamic model represents the plasma surface as a toroidal current sheet and the wall using an eigenmode approach. We report first on the experimental validation and reconciliation of the proposed dynamic model, which is a required step previous to the potential implementation in the Plasma Control System (PCS) of any model-based controller. The dynamic model is then used to synthesize an adaptive control law for the stabilization of the RWM under time-varying β conditions. Simulation results are presented comparing the performance of the model-based adaptive controller and present non-model-based PD controllers.

  18. Plasma radiometry with 30 chord resolution for fast transients in the DIII-D tokamak

    SciTech Connect

    Gray, D.S.; Hollmann, E.M.; Luckhardt, S.C.; Chalfant, J.; Chousal, L.; Hernandez, R.; Jones, E.; Kellman, A.G.

    2004-10-01

    A diagnostic capable of providing time resolved measurements of plasma radiated power during disruptions and other fast transients, e.g., edge localized modes has been employed in the DIII-D tokamak. The radiation is detected with absolute extreme ultraviolet (AXUV) photodiode arrays. Thirty chords from a single port provide measurements from a full slice of the plasma at one toroidal location. The analog bandwidth is up to 1 MHz for the brightest events, i.e., disruptions. Active cooling of the diode arrays prevents damage during high temperature vessel baking. Effective responsivity values of 0.12-0.18 A/W are taken from previous work on the application of AXUV diodes in DIII-D. The total radiated energy in disruptions typically agrees with bolometer measurements within about 12%.

  19. 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.

  20. Scintillator-based diagnostic for fast ion loss measurements on DIII-D.

    PubMed

    Fisher, R K; Pace, D C; García-Muñoz, M; Heidbrink, W W; Muscatello, C M; Van Zeeland, M A; Zhu, Y B

    2010-10-01

    A new scintillator-based fast ion loss detector has been installed on DIII-D with the time response (>100 kHz) needed to study energetic ion losses induced by Alfvén eigenmodes and other MHD instabilities. Based on the design used on ASDEX Upgrade, the diagnostic measures the pitch angle and gyroradius of ion losses based on the position of the ions striking the two-dimensional scintillator. For fast time response measurements, a beam splitter and fiberoptics couple a portion of the scintillator light to a photomultiplier. Reverse orbit following techniques trace the lost ions to their possible origin within the plasma. Initial DIII-D results showing prompt losses and energetic ion loss due to MHD instabilities are discussed.

  1. Measurements of fast-ion transport by mode-particle resonances on DIII-D

    NASA Astrophysics Data System (ADS)

    Muscatello, C. M.; Grierson, B. A.; Harvey, R. W.; Heidbrink, W. W.; Pace, D. C.; Van Zeeland, M. A.

    2012-10-01

    Magnetohydrodynamic (MHD) instabilities in tokamak plasmas manifest in a variety of ways, characterized by different scale lengths and mode frequencies. MHD activity can cause significant degradation of plasma performance due to transport of particles, energy and current. Among the many different types of MHD, arguably fishbones, sawteeth and Alfvén eigenmodes (AEs) are observed to cause the largest fluxes of superthermal ions. DIII-D's expansive suite of diagnostics makes it possible to rigorously characterize these instabilities and study their interaction with fast ions. This review paper first presents an overview of the recent additions to DIII-D's collection of fast-ion diagnostics. The extended diagnostic capabilities are employed in a series of experiments to investigate fast-ion dynamics in the presence of fishbones, sawteeth and AEs. Results from these seemingly unrelated studies are highlighted, and they reveal that mode-particle resonances play the central role in the observed deterioration of fast-ion confinement.

  2. Scintillator-based diagnostic for fast ion loss measurements on DIII-D

    SciTech Connect

    Fisher, R. K.; Van Zeeland, M. A.; Pace, D. C.; Heidbrink, W. W.; Muscatello, C. M.; Zhu, Y. B.; Garcia-Munoz, M.

    2010-10-15

    A new scintillator-based fast ion loss detector has been installed on DIII-D with the time response (>100 kHz) needed to study energetic ion losses induced by Alfven eigenmodes and other MHD instabilities. Based on the design used on ASDEX Upgrade, the diagnostic measures the pitch angle and gyroradius of ion losses based on the position of the ions striking the two-dimensional scintillator. For fast time response measurements, a beam splitter and fiberoptics couple a portion of the scintillator light to a photomultiplier. Reverse orbit following techniques trace the lost ions to their possible origin within the plasma. Initial DIII-D results showing prompt losses and energetic ion loss due to MHD instabilities are discussed.

  3. 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.

  4. Novel current drive experiments on the CDX-U, HIT, and DIII-D Tokamaks

    SciTech Connect

    Ono, M.; Forest, C.B.; Hwang, Y.S.; Armstrong, R.J.; Choe, W.; Darrow, D.S.; Greene, G.; Jones, T.; Jarboe, T.R.; Martin, A.; Nelson, B.A.; Orvis, D.; Painter, C.; Zhou, L.; Rogers, J.A.; Schaffer, M.J.; Hyatt, A.W.; Pinsker, R.I.; Staebler, G.M.; Stambaugh, R.D.; Strait, E.J.; Greene, K.L.; Leuer, J.A.; Lohr, J.M.

    1992-10-01

    Two types of novel, non-inductive current drive concepts for starting-up and maintaining tokamak discharges have been developed on the CDX-U, HIT, and DIII-D Tokamaks. On CDX-U, a new, non-inductive current drive technique utilizing fully internally generated pressure driven currents has been demonstrated. The measured current density profile shows a non-hollow profile which agrees with a modeling calculation including helicity conserving non-classical current transport providing the ``seed current``. Another current drive concept, dc-helicity injection, has been investigated on, CDX-U, HIT and DIII-D. This method utilizes injection of magnetic helicity via low energy electron currents, maintaining the plasma current through helicity conserving relaxiation. In these experiments, non-ohmic tokamak plasmas were formed and maintained in the tens of kA range.

  5. Novel current drive experiments on the CDX-U, HIT, and DIII-D Tokamaks

    SciTech Connect

    Ono, M.; Forest, C.B.; Hwang, Y.S.; Armstrong, R.J.; Choe, W.; Darrow, D.S.; Greene, G.; Jones, T. . Plasma Physics Lab.); Jarboe, T.R.; Martin, A.; Nelson, B.A.; Orvis, D.; Painter, C.; Zhou, L.; Rogers, J.A. ); Schaffer, M.J.; Hyatt, A.W.; Pinsker, R.I.; Staebler, G.M.; Stambaugh, R.D.; Strait, E.J.; Greene, K.L.; Leuer, J.A.; Lohr, J.

    1992-01-01

    Two types of novel, non-inductive current drive concepts for starting-up and maintaining tokamak discharges have been developed on the CDX-U, HIT, and DIII-D Tokamaks. On CDX-U, a new, non-inductive current drive technique utilizing fully internally generated pressure driven currents has been demonstrated. The measured current density profile shows a non-hollow profile which agrees with a modeling calculation including helicity conserving non-classical current transport providing the seed current''. Another current drive concept, dc-helicity injection, has been investigated on, CDX-U, HIT and DIII-D. This method utilizes injection of magnetic helicity via low energy electron currents, maintaining the plasma current through helicity conserving relaxiation. In these experiments, non-ohmic tokamak plasmas were formed and maintained in the tens of kA range.

  6. 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.

  7. 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.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; et al

    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

  8. 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.

  9. Nonlinear MHD simulations of Quiescent H-mode plasmas in DIII-D

    NASA Astrophysics Data System (ADS)

    Liu, F.; Huijsmans, G. T. A.; Loarte, A.; Garofalo, A. M.; Solomon, W. M.; Snyder, P. B.; Hoelzl, M.; Zeng, L.

    2015-09-01

    In the Quiescent H-mode (QH-mode) regime, the edge harmonic oscillation (EHO), thought to be a saturated kink-peeling mode (KPM) driven unstable by current and rotation, is found in experiment to provide sufficient stationary edge particle transport to avoid the periodic expulsion of particles and energy by edge localized modes (ELMs). In this paper, both linear and nonlinear MHD modelling of QH-mode plasmas from the DIII-D tokamak have been investigated to understand the mechanism leading to the appearance of the EHO in QH-mode plasmas. For the first time nonlinear MHD simulations with low-n modes both with ideal wall and resistive wall boundary conditions have been carried out with the 3D non-linear MHD code JOREK. The results show, in agreement with the original conjectures, that in the non-linear phase, kink peeling modes are the main unstable modes in QH-mode plasmas of DIII-D and that the kink-peeling modes saturate non-linearly leading to a 3D stationary state. The characteristics of the kink-peeling modes, in terms of mode structure and associated decrease of the edge plasma density associated with them, are in good agreement with experimental measurements of the EHO in DIII-D. The effect of plasma resistivity, the role of plasma parallel rotation as well as the effect of the conductivity of the vacuum vessel wall on the destabilization and saturation of kink-peeling modes have been evaluated for experimental QH-mode plasma conditions in DIII-D.

  10. 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.

  11. Suppression of type-I ELMs with reduced RMP coil set on DIII-D

    DOE PAGES

    Orlov, D. M.; Moyer, R. A.; Evans, T. E.; Paz-Soldan, Carlos; Ferraro, N. M.; Nazikian, R.; deGrassie, John S.; Grierson, Brian A.; Eldon, D.; Fenstermacher, Max E.; et al

    2016-02-19

    Recent experiments on DIII-D have demonstrated that having a toroidally-monochromatic spectral content of edge-resonant magnetic perturbations (RMPs) is not a necessary condition for suppression of edge localized modes (ELMs). Robust ELM suppression has been reproducibly obtained on DIII-D during experiments in which various non-axisymmetric coil loops were turned off pseudo-randomly producing a variety of n=1 , n=2, and n=3 spectral contributions. It was demonstrated that RMP ELM suppression could be achieved with as few as 5 out of 12 internal coil loops (I-coils) on DIII-D at similar coil currents and with good plasma confinement. Linear MHD plasma response (M3DC1, IPEC,more » MARS) and vacuum (SURFMN, TRIP3D) modelling have been performed in order to understand the effects of the perturbation spectrum on the plasma response and ELM suppression. The results suggest that reduction of the dominant n=3 perturbation field is compensated by increased n=2 field in the plasma that may lead to RMP ELM suppression at lower levels of n=3 perturbative magnetic flux from the I-coils. In conclusion, these results provide additional confidence that ITER may be capable of RMP ELM suppression in the event of multiple internal coil failures.« less

  12. 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.

  13. The 110 GHz Gyrotron Installation on DIII--D: Status and Experimental Results

    NASA Astrophysics Data System (ADS)

    Lohr, John; Ponce, Dan; Callis, R. W.; Popov, L.; Zerbini, M.; Cahalan, P.

    1997-11-01

    The 110 GHz installation on DIII--D consists of two gyrotrons each of which operates at generated power levels between 0.5 and 1.0 MW for pulse lengths up to 2.0 s. The gyrotrons are connected to DIII--D by windowless evacuated transmission lines. The greatest experience to date has been accumulated with the Gycom Centaur gyrotron, a diode tube which has been operated reliably at generated rf power levels in excess of 0.80 MW for pulse durations of 2.0 s. This tube has been modulated at 100% depth at frequencies up to 1 kHz. The second gyrotron is a Communications and Power Industries model VGT-8011A, a triode geometry, which is in initial testing. For this gyrotron, collector power loading has been measured, the beam steering has been set and pulse/power extension is in progress. DIII--D tests of the system performance are ongoing and initial tokamak experiments on transport, H--mode physics and scaling have begun.

  14. 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.

  15. 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.

  16. 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.

  17. Suppression of type-I ELMs with reduced RMP coil set on DIII-D

    NASA Astrophysics Data System (ADS)

    Orlov, D. M.; Moyer, R. A.; Evans, T. E.; Paz-Soldan, C.; Ferraro, N. M.; Nazikian, R.; deGrassie, J. S.; Grierson, B. A.; Eldon, D.; Fenstermacher, M. E.; King, J. D.; Logan, N. C.; Lanctot, M. J.; Maingi, R.; Snyder, P. B.; Strait, E. J.; Wingen, A.

    2016-03-01

    Recent experiments on DIII-D have demonstrated that having a toroidally-monochromatic spectral content of edge-resonant magnetic perturbations (RMPs) is not a necessary condition for suppression of edge localized modes (ELMs). Robust ELM suppression has been reproducibly obtained on DIII-D during experiments in which various non-axisymmetric coil loops were turned off pseudo-randomly producing a variety of n=1 , n= 2 , and n= 3 spectral contributions. It was shown that RMP ELM suppression could be achieved with as few as 5 out of 12 internal coil loops (I-coils) on DIII-D at similar coil currents and with good plasma confinement. Linear MHD plasma response (m3dc1, ipec, mars) and vacuum (surfmn, trip3d) modelling have been performed in order to understand the effects of the perturbation spectrum on the plasma response and ELM suppression. The results suggest that reduction of the dominant n= 3 perturbation field is compensated by increased n= 2 field in the plasma that may lead to RMP ELM suppression at lower levels of n= 3 perturbative magnetic flux from the I-coils. These results provide additional confidence that ITER may be capable of RMP ELM suppression in the event of multiple internal coil failures.

  18. An algorithm to provide real time neutral beam substitution in the DIII-D tokamak

    SciTech Connect

    Phillips, J.C.; Greene, K.L.; Hyatt, A.W.; McHarg, B.B. Jr.; Penaflor, B.G.

    1999-06-01

    A key component of the DIII-D tokamak fusion experiment is a flexible and easy to expand digital control system which actively controls a large number of parameters in real-time. These include plasma shape, position, density, and total stored energy. This system, known as the PCS (plasma control system), also has the ability to directly control auxiliary plasma heating systems, such as the 20 MW of neutral beams routinely used on DIII-D. This paper describes the implementation of a real-time algorithm allowing substitution of power from one neutral beam for another, given a fault in the originally scheduled beam. Previously, in the event of a fault in one of the neutral beams, the actual power profile for the shot might be deficient, resulting in a less useful or wasted shot. Using this new real-time algorithm, a stand by neutral beam may substitute within milliseconds for one which has faulted. Since single shots can have substantial value, this is an important advance to DIII-D`s capabilities and utilization. Detailed results are presented, along with a description not only of the algorithm but of the simulation setup required to prove the algorithm without the costs normally associated with using physics operations time.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

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

    NASA Astrophysics Data System (ADS)

    Sung, C.; White, A. E.; Mikkelsen, D. R.; Greenwald, M.; Holland, C.; Howard, N. T.; Churchill, R.; Theiler, C.

    2016-04-01

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

  4. 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.

  5. 4 MW upgrade to the DIII-D fast wave current drive system

    SciTech Connect

    deGrassie, J.S.; Pinsker, R.I.; Cary, W.P.

    1993-10-01

    The DIII-D fast wave current drive (FWCD) system is being upgraded by an additional 4 MW in the 30 to 120 MHz frequency range. This capability adds to the existing 2 MW 30 to 60 MHz system. Two new ABB transmitters of the type that are in use on the ASDEX-Upgrade tokamak in Garching will be used to drive two new water-cooled four-strap antennas to be installed in DIII-D in early 1994. The transmission and tuning system for each antenna will be similar to that now in use for the first 2 MW system on DIII-D, but with some significant improvements. One improvement consists of adding a decoupler element to counter the mutual coupling between the antenna straps which results in large imbalances in the power to a strap for the usual current drive intrastrap phasing of 90{degrees}. Another improvement is to utilize pressurized, ceramic-insulated transmission lines. The intrastrap phasing will again be controlled in pairs, with a pair of straps coupled in a resonant loop configuration, locking their phase difference at either 0 or 180{degrees}, depending upon the length of line installed. These resonant loops will incorporate a phase shifter so that they will be able to be tuned to resonance at several frequencies in the operating band of the transmitter. With the frequency change capability of the ABB generators, the FWCD frequency will thus be selectable on a shot-to-shot basis, from this preselected set of frequencies. The schedule is for experiments to begin with this added 4 MW capability in mid-1994. The details of the system are described.

  6. 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.

  7. Implementation of a quasi-realtime display of DIII-D neutral beam heating waveforms

    SciTech Connect

    Phillips, J.C.

    1993-10-01

    The DIII-D neutral beam system employs eight 80 keV ion sources mounted on four beamlines to provide plasma heating to the DIII-D tokamak. The neutral beam system is capable of injecting over 20 MW of deuterium power with flexibility in terms of timing and modulation of the individual neutral beams. To maintain DIII-D`s efficient tokamak shot cycle and make informed control decisions, it is important to be able to determine which beams fired, and exactly when, by the time the tokamak shot is over. Previously this information was available in centralized form only after a several minute wait. A cost-effective alternative to the traditional eight-channel storage oscilloscope has been implemented using off the shelf PC hardware and software. The system provides a real time display of injected neutral beam accelerator voltages and tokamak plasma current, as well an a summation waveform indicative of the total injected power as a function of time. The hardware consists of a Macintosh Centris 650 PC with a Motorola 68040 microprocessor. Data acquisition is accomplished using a National Instrument`s 16-channel analog to digital conversion board for the Macintosh. The color displays and functionality were developed using National Instruments` LabView environment. Because the price of PCs has been decreasing rapidly and their capabilities increasing, this system is far less expensive than an eight-channel storage oscilloscope. As a flexible combination of PC and software, the system also provides much more capability than a dedicated oscilloscope, acting as the neutral beam coordinator`s logbook, recording comments and availability statistics. Data such as shot number and neutral beam parameters are obtained over the local network from other computers and added to the display. Waveforms are easily archived to disk for future recall. Details of the implementation will be discussed along with samples of the displays and a description of the system`s function and capabilities.

  8. 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.

  9. Toroidal Rotation and Core Ion Confinement with RF Heating in DIII-D

    SciTech Connect

    deGrassie, J.S.; Greenfield, C.M.; Baker, D.R.; Burrell, K.H.; Lin-Liu, Y.R.; Lohr, J.; Luce, T.C.; Petty, C.C.; Prater, R.; Staebler, G.M.; Heidbrink, W.W.; Rice, B.W. Rice, Mau, T.K.; Porkolab, M.

    1999-07-01

    Shear in the E x B flow velocity can stabilize turbulent transport [1], and so it is of interest to understand the physics behind electric field generation and modification in the tokamak. In DIII-D the core radial electric field in many regimes is generated by flow velocities driven by momentum input from neutral beam injection (NBI). In a variety of conditions it is observed that direct electron heating is accompanied by a reduction in the NBI driven toroidal rotation velocity, U{sub {phi}}, and the ion temperature, T{sub i}, primarily in the core, {rho} <0.5 (where {rho} is a radial coordinate of the normalized toroidal flux). This electron heating can be done with either electron cyclotron heating (ECH) or fast wave electron heating (FWEH). Both can be accompanied by the reduction in U{sub {phi}} and T{sub i} [2-4]. Details of the parallel wavenumber (k//) spectrum of the launched rf do not seem to be important in either case for the effect to exist. Reductions are observed for EC waves launched with nonzero k// for current drive or launched radially with k//=0; and for FWEH with waves directed either co or counter, using the DIII-D four strap antennas [5], This universality indicates that increased electron temperature, T{sub e}, is increasing ion momentum and thermal transport, at least in the parameter regimes of these experiments. It is also possible that nonambipolar transport of resonantly heated particles is playing a role. To date, the great majority of the DIII-D experiments have been conducted with the rf target discharges driven by co-injected NBI.

  10. Off-axis neutral beam current drive for advanced scenario development in DIII-D

    SciTech Connect

    Murakami, Masanori; Park, Jin Myung; Petty, C C.; Luce, T.C.; Heidbrink, W. W.; Osborne, T.H.; Prater, R.; Wade, M R; Unterberg, E. A.

    2009-01-01

    Modification of the two existing DIII-D neutral beamlines is planned to allow vertical steering to provide off-axis neutral beam current drive (NBCD) peaked as far off-axis as half the plasma minor radius. New calculations for a downward-steered beam indicate strong current drive with good localization off-axis so long as the toroidal magnetic field, B-T, and the plasma current, I-p, point in the same direction. This is due to good alignment of neutral beam injection (NBI) with the local pitch of the magnetic field lines. This model has been tested experimentally on DIII-D by injecting equatorially mounted NBs into reduced size plasmas that are vertically displaced with respect to the vessel midplane. The existence of off-axis NBCD is evident in the changes seen in sawtooth behaviour in the internal inductance. By shifting the plasma upwards or downwards, or by changing the sign of the toroidal field, off-axis NBCD profiles measured with motional Stark effect data and internal loop voltage show a difference in amplitude (40-45%) consistent with differences predicted by the changed NBI alignment with respect to the helicity of the magnetic field lines. The effects of NBI direction relative to field line helicity can be large even in ITER: off-axis NBCD can be increased by more than 30% if the B-T direction is reversed. Modification of the DIII-D NB system will strongly support scenario development for ITER and future tokamaks as well as provide flexible scientific tools for understanding transport, energetic particles and heating and current drive.

  11. 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).

  12. The Bootstrap Current and Neutral Beam Current Drive in DIII-D

    SciTech Connect

    Politzer, P.A.

    2005-10-15

    Noninductive current drive is an essential part of the implementation of the DIII-D Advanced Tokamak program. For an efficient steady-state tokamak reactor, the plasma must provide close to 100% bootstrap fraction (f{sub bs}). For noninductive operation of DIII-D, current drive by injection of energetic neutral beams [neutral beam current drive (NBCD)] is also important. DIII-D experiments have reached {approx}80% bootstrap current in stationary discharges without inductive current drive. The remaining current is {approx}20% NBCD. This is achieved at {beta}{sub N} [approximately equal to] {beta}{sub p} > 3, but at relatively high q{sub 95} ({approx}10). In lower q{sub 95} Advanced Tokamak plasmas, f{sub bs} {approx} 0.6 has been reached in essentially noninductive plasmas. The phenomenology of high {beta}{sub p} and {beta}{sub N} plasmas without current control is being studied. These plasmas display a relaxation oscillation involving repetitive formation and collapse of an internal transport barrier. The frequency and severity of these events increase with increasing {beta}, limiting the achievable average {beta} and causing modulation of the total current as well as the pressure. Modeling of both bootstrap and NBCD currents is based on neoclassical theory. Measurements of the total bootstrap and NBCD current agree with calculations. A recent experiment based on the evolution of the transient voltage profile after an L-H transition shows that the more recent bootstrap current models accurately describe the plasma behavior. The profiles and the parametric dependences of the local neutral beam-driven current density have not yet been compared with theory.

  13. Multi-megawatt 110 GHz ECH system for the DIII-D tokamak

    SciTech Connect

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

    1997-11-01

    Two 110 GHz gyrotrons with nominal output power of 1 MW each have been installed on the DIII-D tokamak. The first 110 GHz gyrotron built by Gycom has a nominal rating of 1 MW and a 2s pulse length, with the pulse length being determined by the maximum temperature allowed on the edge cooled boron nitride window. This gyrotron was first operated into the DIII-D tokamak in late 1996. The second gyrotron was built by Communications and Power Industries (CPI) was commissioned during the spring of 1997. The CPI gyrotron uses a double disc FC-75 cooled sapphire window which has a pulse length rating of 0.8s at 1 MW, 2s at 0.5 MW and 10s at 0.2 MW. Both gyrotrons are connected to the tokamak by a low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HE(11) mode. Using short pulse lengths to avoid breakdown inside the air filled waveguide, the microwave beam has been measured inside the DIII-D vacuum vessel using a paper target and an IR camera. The resultant microwave beam was found to be well focused with a spot size of approximately 8 cm. The beam can be steered poloidially from the center to the outer edge of the plasma. The initial operation of the Gycom gyrotron with about 0.5 MW delivered to a low density plasma for 0.5 s showed good central electron heating, with peak temperature in excess of 10 keV. A third gyrotron, being built by CPI, will be installed later this year. Progress with the first CPI tube will also be discussed and future plans for the ECH installation and physics experiments will be presented.

  14. 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.

  15. Spatially resolved measurements of two-dimensional turbulent structures in DIII-D plasmas

    NASA Astrophysics Data System (ADS)

    Zemedkun, S. E.; Che, S.; Chen, Y.; Domier, C. W.; Luhmann, N. C.; Munsat, T.; Parker, S. E.; Tobias, B.; Wan, W.; Yu, L.

    2015-12-01

    Two-dimensional observations of spatially coherent electron temperature fluctuations at drift-wave scales (k ˜ 1 cm-1) have been made using the electron cyclotron emission imaging diagnostic on the DIII-D tokamak. These measurements enable the extraction of spectral properties, including poloidal dispersion relations. Temperature fluctuation levels are found to be Te ˜/⟨Te⟩=1.2 % , and the phase velocity of the fluctuations is found to be constant across frequencies, consistent with modes having real frequencies low compared to the rotation-induced Doppler shifts. Comparisons with radially global linear gyrokinetic simulations suggest that the observed modes may be trapped electron modes.

  16. 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.

  17. 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.

  18. Current driven due to localized electron power deposition in DIII-D

    SciTech Connect

    Harvey, R.W.; Lin-Liu, Y.R.; Luce, T.C.; Prater, R.; Sauter, O.; Smirnov, A.P.

    1999-05-01

    Due to spatial localization of electron cyclotron wave injection in DIII-D, electrons heated in an off-axis region must toroidally transit the tokamak 25--50 times before re-entering the heating region. This distance is of the order of the mean free path. The effect of such RF localization is simulated with a time-dependent Fokker-Planck code which is 2D-in-velocity, 1D-in-space-along-B, and periodic in space. An effective parallel electric field arises to maintain continuity of the driven current. Somewhat surprisingly, the localized current drive efficiency remains equal to that for a uniform medium.

  19. Rapid Inward Impurity Transport during Impurity Pellet Injection on the DIII-D Tokamak

    SciTech Connect

    Evans, T.E.; Hyatt, A.W.; Lee, R.L.; Kellman, A.G.; Parks, P.B.; Stockdale, R.; Taylor, P.L.; Whyte, D.G.; Jernigan, T.C.

    1998-11-01

    Neon killer pellets are injected into the DIII-D tokamak plasma in order to radiatively quench the plasma{close_quote}s stored energy and mitigate disruption effects. Inward radial transport on the time scale of the pellet ablation ({le}1 ms) results in central deposition of the neon inside the ablation penetration radius of the pellet, causing effective radiative energy dissipation. This result is in contrast to the radially outward deposition measured for fueling (hydrogenic) pellets. The observed magnitudes of magnetic fluctuations ({delta}B/B{approximately}0.2{percent}) are shown to be capable of causing the radial transport. {copyright} {ital 1998} {ital The American Physical Society }

  20. Design of long-pulse fast wave current drive antennas for DIII-D

    NASA Astrophysics Data System (ADS)

    Baity, F. W.; Batchelor, D. B.; Bills, K. C.; Fogelman, C. H.; Jaeger, E. F.; Ping, J. L.; Riemer, B. W.; Ryan, P. M.; Stallings, D. C.; Taylor, D. J.; Yugo, J. J.

    1994-10-01

    Two new long-pulse fast wave current drive (FWCD) antennas will be installed on DIII-D in early 1994. These antennas will increase the available FWCD power from 2 MW to 6 MW for pulse lengths of up to 2 s, and to 4 MW for up to 10 s. Power for the new antennas is from two ASDEX-type 30- to 120-MHz transmitters. When operated at 90° phasing into a low-density plasma (˜4×1019m-3) with hot electrons (˜10 keV), these two new antennas are predicted to drive approximately 1 MA of plasma current.

  1. Commissioning of electron cyclotron emission imaging instrument on the DIII-D tokamak and first dataa)

    NASA Astrophysics Data System (ADS)

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

    2010-10-01

    A new electron cyclotron emission imaging diagnostic has been commissioned on the DIII-D tokamak. Dual detector arrays provide simultaneous two-dimensional images of Te fluctuations over radially distinct and reconfigurable regions, each with both vertical and radial zoom capability. A total of 320 (20 vertical×16 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.

  2. Embedded calibration system for the DIII-D Langmuir probe analog fiber optic linksa)

    NASA Astrophysics Data System (ADS)

    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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Chaotic density fluctuations in L-mode plasmas of the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Maggs, J. E.; Rhodes, T. L.; Morales, G. J.

    2015-04-01

    Analysis of the time series obtained with the Doppler backscattering system (Hillsheim et al 2009 Rev. Sci. Instrum. 80 0835070) in the DIII-D tokamak (Luxon 2005 Fusion Sci. Technol. 48 828) shows that intermediate wave number plasma density fluctuations in low confinement (L-mode) tokamak plasmas are chaotic. The supporting evidence is based on the shape of the power spectrum; the location of the signal in the complexity-entropy plane (C-H plane) (Rosso et al 2007 Phys. Rev. Lett. 99 154102); and the population of the corresponding Bandt-Pompe (Bandt and Pompe 2002 Phys. Rev. Lett. 88 174102) probability distributions.

  9. HIGH POWER LONG PULSE PERFORMANCE OF THE DIII-D GYROTRON INSTALLATION

    SciTech Connect

    J. LOHR; Y.A. GORELOV; R.W. CALLIS; H.J. GRUNLOH; J.J. PEAVY; R.I. PINSKER; D. PONCE; R. PRATER; R.A. ELLIS,III

    2002-05-01

    At DIII-D, five 110 GHz gyrotrons are operating routinely for 2.0 s pulses at generated power levels {ge}750 kW per gyrotron. A sixth gyrotron is being installed, which should bring the generated power level to >4 MW and the injected power to about 3.0 MW. The output power now can be modulated by the plasma control system to fix T{sub e} at a desired value. The system is being used as a tool for control of current diffusion, for current profile control and other experiments leading to advanced tokamak operation.

  10. 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.

  11. Alfvén Eigenmode Induced Fast Ion Transport in DIII-D Resulting From Finite Larmor Radius Effects

    NASA Astrophysics Data System (ADS)

    Kramer, G. J.; Fu, G. Y.; Nazikian, R.; van Zeeland, M. A.; Fisher, R. K.; Pace, D. C.; Chen, L.; Chen, X.; Heidbrink, W. W.

    2013-10-01

    Alfvén Eigenmode (AE) induced fast-ion redistribution and loss are commonly observed in DIII-D. In those experiments the perpendicular wave vector times the fast-ion Larmor radius is of order unity (k⊥ρi ~ 1), which allows fast ion orbits to traverse the AE mode structures asymmetrically causing a significant change in magnetic moment. Full-orbit simulations of the expected transport in DIII-D plasmas show that this effect can lead to fast-ion radial diffusion constants of 5 m2/s at the measured mode amplitudes. This level of diffusion is not captured with a guiding center approximation and is in agreement with that deduced in DIII-D experiments when AE activity was present. Work supported by the US Department of Energy under DE-AC02-09CH11466, DE-FC02-04ER54698, and SC-G903402.

  12. Aspect ratio effects on neoclassical tearing modes from comparison between DIII-D and National Spherical Torus Experiment

    SciTech Connect

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

    2012-06-15

    Neoclassical tearing mode islands are sustained by helically perturbed bootstrap currents arising at finite beta from toroidal effects that trap a fraction of the particles in non-circulating orbits. DIII-D and NSTX are here operated with similar shape and cross-sectional area but almost a factor of two difference in inverse aspect ratio a/R. In these experiments, destabilized n=1 tearing modes were self-stabilized (reached the 'marginal point') by reducing neutral-beam power and thus beta. The measure of the marginal island gives information on the small-island stabilizing physics that in part (with seeding) governs onset. The marginal island width on NSTX is found to be about three times the ion banana width and agrees with that measured in DIII-D, except for DIII-D modes closer to the magnetic axis, which are about two times the ion banana width. There is a balance of the helically perturbed bootstrap term with small island effects with the sum of the classical and curvature terms in the modified Rutherford equation for tearing-mode stability at the experimental marginal point. Empirical evaluation of this sum indicates that while the stabilizing effect of the curvature term is negligible in DIII-D, it is important in NSTX. The mode temporal behavior from the start of neutral-beam injection reduction also suggests that NSTX operates closer to marginal classical tearing stability; this explains why there is little hysteresis in beta between mode onset, saturation, and self-stabilization (while DIII-D has large hysteresis in beta). NIMROD code module component calculations based on DIII-D and NSTX reconstructed experimental equilibria are used to diagnose and confirm the relative importance of the stabilizing curvature effect, an advantage for low aspect ratio; the relatively greater curvature effect makes for less susceptibility to NTM onset even if the classical tearing stability index is near marginal.

  13. 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.

  14. 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.

  15. 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.

  16. Thermal analysis and design of polarizer mirrors for waveguides in the DIII-D tokamak

    SciTech Connect

    Baxi, C.B.; Doane, J.L.; Sevier, D.L.

    1992-07-01

    A new high power electron cyclotron heating (ECH) system designed to operate at 110 GHz with a power output of 2 MW has been introduced on DIII-D. All components of the system are capable of handling a 10 second pulse at an interval of 10 minutes. Transmission of ECH power from the the source (a millimeter-wave gyrotron) to the plasma through waveguide miter bends may change the polarization and rotate the polarization major axis. Polarizing elements are therefore required to correct for the effect of transmission lines and also to generate proper polarization for coupling into the plasma. Rotating mirrors with different rectangular grooved gratings in two successive miter bends can generate the required wide range of elliptical polarizations. Peak heat fluxes due to ohmic losses in these mirrors are several MW/m{sub 2} for a 0.5 MW gyrotron power. The complex distribution of losses in the grooves requires a detailed thermal stress analysis to ensure that temperature and stress limits are not exceeded. The desired pulse length is 10 sec, with a cooling time of 10 min between pulses. The temperature rise in the polarizing mirrors must be limited to less than 300{degree}C to prevent thermal fatigue and outgassing in the vacuum lines. This paper presents an analysis for the polarizing mirrors for the DIII-D ECH system.

  17. 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.

  18. Progress in GYRO Validation Studies of DIII-D H-mode Plasmas

    NASA Astrophysics Data System (ADS)

    Holland, C.; Burrell, K. H.; Petty, C. C.; Candy, J.; Holcomb, C. T.

    2011-10-01

    The need for a validated predictive capability of turbulent transport in ITER is now widely recognized. However, to date most validation studies of nonlinear codes such as GYRO have focused upon low power L-mode discharges, which have significant differences in key dimensionless parameters such as β and ρ* =ρs / a from more ITER-relevant H-mode discharges. In order to address this gap, the results of nonlinear GYRO simulations of a range of DIII-D H-mode discharges (including quiescent H-mode, hybrid, and steady state discharges) are presented. These H-mode studies focus upon three physics effects, motivated by key differences between L- and H-mode plasmas. The first is the linear and nonlinear effects of finite β in plasmas where βN = β /(I / aB) varies from 1.5-2.5, 3-5 times higher than typical L-mode values. Second is the impact of nonlocality at typical DIII-D H-mode ρ* values, which in contrast to typical L-modes can be non-negligible. Third is the stabilizing effect of a significant density of energetic particles. Work supported by US DOE under DE-FG02-07ER54917, DE-FC02-04ER54698, DE-FG02-95ER54309, and DE-AC52-07NA27344.

  19. 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.

  20. Vanadium alloys for the radiative divertor program of DIII-D

    SciTech Connect

    Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

    1995-10-01

    Vanadium alloys provide an attractive solution for fusion power plants as they exhibit a potential for low environmental impact due to low level of activation from neutron fluence and a relatively short half-life. They also have attractive material properties for use in a reactor. General Atomics along with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan to utilize vanadium alloys as part of the Radiative Divertor Project (RDP) modification for the DIII-D tokamak. The goal for using vanadium alloys is to provide a meaningful step towards developing advanced materials for fusion power applications by demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak in conjunction with developing essential fabrication technology for the manufacture of full-scale vanadium alloy components. A phased approach towards utilizing vanadium in DIII-D is being used starting with small coupons and samples, advancing to a small component, and finally a portion of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. A major portion of the program is research and development to support fabrication and resolve key issues related to environmental effects.

  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. Regimes of improved confinement and stability in DIII-D obtained through current profile modifications

    SciTech Connect

    Lao, L.L.; Ferron, J.R.; Taylor, T.S.; Chan, V.S.; Osborne, T.H.; Burrell, K.H.; Chu, M.S.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Jackson, G.L.; St. John, H.; Strait, E.J.; Thompson, S.J.; Turnbull, A.D.; Doyle, E.J.; Rettig, C.; James, R.; Wroblewski, D.; Lazarus, E.A.; Zohm, H.

    1992-09-01

    Several regimes of improved confinement and stability have been obtained in recent experiments in the DIII-D tokamak by dynamically varying the toroidal current density profile to transiently produce a poloidal magnetic field profile with more favorable confinement and stability properties. A very peaked current density profile with high plasma internal inductance, {ell}{sub i}, is produced either by a rapid change in the plasma poloidal cross section or by a rapid change in the total plasma current. Values of thermal energy confinement times nearly 1.8 times the JET/DIII-D ELM-free H-mode thermal confinement scaling are obtained. The confinement enhancement factor over the ITER89-P L-mode confinement scaling, H, is as high as 3. Normalized toroidal beta, {beta}{sub N}, greater than 6%-m-T/MA and values of the product {beta}{sub N}H greater than 15 have also been obtained. Both the confinement and the maximum achievable {beta} vary with {ell}{sub i} and decrease as the current profile relaxes. For strongly shaped H-mode discharges, in addition to the current density profile peakedness, as measured by {ell}{sub i} other current profile parameters, such as its distribution near the edge region, may also affect the confinement enhancement.

  3. Regimes of improved confinement and stability in DIII-D obtained through current profile modifications

    SciTech Connect

    Lao, L.L.; Ferron, J.R.; Taylor, T.S.; Chan, V.S.; Osborne, T.H.; Burrell, K.H.; Chu, M.S.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Jackson, G.L.; St. John, H.; Strait, E.J.; Thompson, S.J.; Turnbull, A.D. ); Doyle, E.J.; Rettig, C. ); James, R.; Wroblewski, D. (Lawrence Livermore National Lab., CA (United Sta

    1992-09-01

    Several regimes of improved confinement and stability have been obtained in recent experiments in the DIII-D tokamak by dynamically varying the toroidal current density profile to transiently produce a poloidal magnetic field profile with more favorable confinement and stability properties. A very peaked current density profile with high plasma internal inductance, [ell][sub i], is produced either by a rapid change in the plasma poloidal cross section or by a rapid change in the total plasma current. Values of thermal energy confinement times nearly 1.8 times the JET/DIII-D ELM-free H-mode thermal confinement scaling are obtained. The confinement enhancement factor over the ITER89-P L-mode confinement scaling, H, is as high as 3. Normalized toroidal beta, [beta][sub N], greater than 6%-m-T/MA and values of the product [beta][sub N]H greater than 15 have also been obtained. Both the confinement and the maximum achievable [beta] vary with [ell][sub i] and decrease as the current profile relaxes. For strongly shaped H-mode discharges, in addition to the current density profile peakedness, as measured by [ell][sub i] other current profile parameters, such as its distribution near the edge region, may also affect the confinement enhancement.

  4. 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.

  5. 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.

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

    DOE PAGES

    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.; et al

    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

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

    DOE PAGES

    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 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

  8. 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.

  9. 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.

  10. Boundary and PMI Diagnostics for the DIII-D National Fusion Facility

    NASA Astrophysics Data System (ADS)

    Thomas, D. M.; Bray, B. D.; Chrobak, C.; Leonard, A. W.; Allen, S. L.; Lasnier, C. J.; McLean, A. G.; Briesemeister, A. R.; Boedo, J. A.; Elder, D.; Watkins, J. G.

    2014-10-01

    The Boundary and Plasma Materials Interaction Center is planning an improved set of boundary and divertor diagnostics for DIII-D in order to develop and validate robust heat flux solutions for future fusion devices on a timescale relevant to the design of FNSF. We intend to develop and test advanced divertor configurations on DIII-D using high performance plasma scenarios that are compatible with advanced tokamak operations in FNSF as well as providing a comprehensive testbed for modeling. Simultaneously, candidate PFC material solutions can be easily tested in these scenarios. Additional diagnostic capability is vital to help understand and validate these solutions. We will describe a number of desired measurements and our plans for deployment. These include better accounting of divertor radiation, including species identification and spatial distribution, divertor/SOL main ion temperature and neutral pressure, fuller 2D Te /ne imaging, and toroidally separated 3D heat flux measurements. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC05-00OR22725, DE-FG02-07EAR54917, and DE-AC04-94AL85000.

  11. Ion Bernstein wave antenna loading measurements on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Mayberry, M. J.; Pinsker, R. I.; Petty, C. C.; Porkolab, M.; Chiu, S. C.; Cary, W. P.; Prater, R.

    1993-04-01

    Antenna loading measurements carried out during high power ion Bernstein wave (IBW) heating experiments on the DIII-D tokamak indicate that efficient, direct coupling to the IBW at ω lesssim 2ωci as predicted by linear coupling theory did not occur. Whereas strong peaking in the loading resistance near cyclotron harmonics is predicted for high edge densities (ω < ωLH|edge), the observed loading resistance is nearly independent of the toroidal magnetic field. The loading anomaly can be explained by invoking the ponderomotive force which decreases the edge density immediately in front of the antenna, thus allowing coupling to the cold plasma lower hybrid wave (LHW). A linear LHW coupling code including the modified density profile due to the ponderomotive force reproduces the measured dependence of antenna loading on toroidal magnetic field, edge density, antenna frequency and antenna phasing. Further evidence for the ponderomotive force is obtained from reactive loading measurements which indicate that the plasma is pushed away from the antenna as the radiofrequency power level is increased. The results indicate that the lack of central ion heating observed during DIII-D IBW experiments may be due to a lack of efficient mode transformation from the coupled LHW to a centrally propagating IBW, possibly as a result of nonlinear mechanism(s)

  12. 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.

  13. BOUT Simulations of Edge Turbulence in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Cohen, Bruce; Umansky, Maxim; Nevins, William; Makowski, Mike; Boedo, Jose; Rudakov, Dmitry; Holland, Chris; McKee, George; Yan, Zheng

    2012-03-01

    Progress is reported on simulations of electromagnetic drift-resistive ballooning turbulence in realistic single-null tokamak geometry using the BOUT three-dimensional fluid code [1] that solves Braginskii-based fluid equations [2]. The simulation domain models the actual magnetic geometry of the DIII-D tokamak. The simulations follow unstable drift-resistive ballooning turbulence in the edge region to saturation. Fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes are compared to experimental probe and beam-emission-spectroscopy data for a well-characterized L-mode discharge in DIII-D. Post-processing of the simulation data using synthetic diagnostics facilitates the comparisons. The simulations are comprised of a suite of runs in which the physics model evolves to include more fluid fields and physics terms. The relative agreement of the simulation results with the experimental data improves as more physics is included. [4pt] [1] X. Q. Xu, and R. H. Cohen, Contrib. Plasma Phys. 36 (1998) 158. [0pt] [2] S. Braginskii, ``Transport Processes in a Plasma,'' in Rev. Plas. Phys., Vol 1, ed. M. A. Leontovich (Consltnts. Bureau, New York, 1965), p. 205.

  14. PROGRESS TOWARD FULLY NONINDUCTIVE, HIGH BETA DISCHARGES IN DIII-D

    SciTech Connect

    GREENFIELD,CM; FERRON,JR; MURAKAMI,M; WADE,MR; BUDNY,RV; BURRELL,KH; CASPER,TA; DeBOO,JC; DOYLE,EJ; GAROFALO,AM; JAYAKUMAR,RJ; KESSEL,C; LAO,LL; LOHR,J; LUCE,TC; MAKOWSKI,MA; MENARD,JE; PETRIE,TW; PETTY,CC; PINSKER,RI; PRATER,R; POLITZER,PA; St JOHN,HE; TAYLOR,TS; WEST,WP; DIII-D NATIONAL TEAM

    2003-08-01

    OAK-B135 Advanced Tokamak (AT) research in DIII-D focuses on developing a scientific basis for steady-state, high performance operation. For optimal performance, these experiments routinely operate with {beta} above the n = 1 no-wall limit, enabled by active feed-back control. The ideal wall {beta} limit is optimized by modifying the plasma shape, current and pressure profile. Present DIII-D AT experiments operate with f{sub BS} {approx} 50%-60%, with a long-term goal of {approx} 90%. Additional current is provided by neutral beam and electron cyclotron current drive, the latter being localized well away from the magnetic axis ({rho} {approx} 0.4-0.5). Guided by integrated modeling, recent experiments have produced discharges with {beta} {approx} 3%, {beta}{sub N} {approx} 3, f{sub BS} {approx} 55% and noninductive fraction f{sub NI} {approx} 90%. Additional control is anticipated using fast wave current drive to control the central current density.

  15. 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.

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

    PubMed

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

    2014-11-01

    A new frequency-modulated profile reflectometer system, featuring a monostatic antenna geometry (using one microwave antenna for both launch and receive), has been installed on the DIII-D tokamak, providing a first experimental test of this measurement approach for profile reflectometry. Significant features of the new system are briefly described in this paper, including the new monostatic arrangement, use of overmoded, broadband transmission waveguide, and dual-polarization combination/demultiplexing. Updated data processing and analysis, and in-service performance aspects of the new monostatic profile reflectometer system are also presented. By using a raytracing code (GENRAY) to determine the approximate trajectory of the probe beam, the electron density (ne) profile can be successfully reconstructed with L-mode plasmas vertically shifted by more than 10 cm off the vessel midplane. Specifically, it is demonstrated that the new system has a capability to measure ne profiles with plasma vertical offsets of up to ±17 cm. Examples are also presented of accurate, high time and spatial resolution density profile measurements made over a wide range of DIII-D conditions, e.g., the measured temporal evolution of the density profile across a L-H transition.

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

    PubMed

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

    2014-11-01

    A new frequency-modulated profile reflectometer system, featuring a monostatic antenna geometry (using one microwave antenna for both launch and receive), has been installed on the DIII-D tokamak, providing a first experimental test of this measurement approach for profile reflectometry. Significant features of the new system are briefly described in this paper, including the new monostatic arrangement, use of overmoded, broadband transmission waveguide, and dual-polarization combination/demultiplexing. Updated data processing and analysis, and in-service performance aspects of the new monostatic profile reflectometer system are also presented. By using a raytracing code (GENRAY) to determine the approximate trajectory of the probe beam, the electron density (ne) profile can be successfully reconstructed with L-mode plasmas vertically shifted by more than 10 cm off the vessel midplane. Specifically, it is demonstrated that the new system has a capability to measure ne profiles with plasma vertical offsets of up to ±17 cm. Examples are also presented of accurate, high time and spatial resolution density profile measurements made over a wide range of DIII-D conditions, e.g., the measured temporal evolution of the density profile across a L-H transition. PMID:25430256

  18. Performance and data analysis aspects of the new DIII-D monostatic profile reflectometer systema)

    NASA Astrophysics Data System (ADS)

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

    2014-11-01

    A new frequency-modulated profile reflectometer system, featuring a monostatic antenna geometry (using one microwave antenna for both launch and receive), has been installed on the DIII-D tokamak, providing a first experimental test of this measurement approach for profile reflectometry. Significant features of the new system are briefly described in this paper, including the new monostatic arrangement, use of overmoded, broadband transmission waveguide, and dual-polarization combination/demultiplexing. Updated data processing and analysis, and in-service performance aspects of the new monostatic profile reflectometer system are also presented. By using a raytracing code (GENRAY) to determine the approximate trajectory of the probe beam, the electron density (ne) profile can be successfully reconstructed with L-mode plasmas vertically shifted by more than 10 cm off the vessel midplane. Specifically, it is demonstrated that the new system has a capability to measure ne profiles with plasma vertical offsets of up to ±17 cm. Examples are also presented of accurate, high time and spatial resolution density profile measurements made over a wide range of DIII-D conditions, e.g., the measured temporal evolution of the density profile across a L-H transition.

  19. 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.

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

    DOE PAGES

    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.; et al

    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

  1. 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. PMID:25430212

  2. Effects of plasma rotation in reconstructed 3-D equilibria for DIII-D

    NASA Astrophysics Data System (ADS)

    Wingen, A.; Shafer, M. W.; Unterberg, E. A.; Wilcox, R. S.; Cianciosa, M. R.; Hirshman, S. P.; Hillis, D. L.; Lao, L.; Paz-Soldan, C.

    2015-11-01

    A technique for tokamak equilibrium reconstructions when weakly 3-D fields (δB/B ~ 10-3) are applied is used for inner-wall-limited DIII-D discharges. The technique couples diagnostics to the non-linear, ideal MHD equilibrium solver VMEC, using the V3FIT code, to find the most likely 3-D equilibrium based on a suite of measurements. Observations at DIII-D show that plasma rotation larger than 20 krad/s changes the relative phase between the applied 3-D fields and the measured plasma response. Numerical simulations of linear, resistive, 2-fluid MHD show, that large plasma rotation increases flux surface corrugations1. Discharges with low averaged (~10 krad/s) and peaked rotation profiles (~40 krad/s) are reconstructed. Similarities and differences to forward modeled VMEC equilibria, which do not include rotational effects, are shown. The resulting significance of including rotational effects in VMEC is discussed. Supported by US DOE DE-AC05-00OR22725, DE-FC02-04ER54698.

  3. Testing Pedestal Models for Joint Research Target on DIII-D

    NASA Astrophysics Data System (ADS)

    Groebner, R. J.; Snyder, P. B.; Osborne, T. H.; Smith, S. P.; Leonard, A. W.; Bray, B. D.; Deterly, T. M.; Liu, C.; Boivin, R. L.; Degrassie, J. S.; Prater, R.; Eldon, D.; Rhodes, T. L.; Zeng, L.; Hillesheim, J. C.; Yan, Z.; McKee, G. R.

    2011-10-01

    As part of the FY2011 DOE Joint Research Target on Pedestal Physics, tests are being performed on DIII-D for gyrokinetic modes that have been proposed as physics mechanisms for controlling the H-mode pedestal structure. These modes include kinetic ballooning modes (KBM), candidates for limiting the total pedestal pressure gradient, ion temperature gradient modes (ITGM), candidates for limiting the Ti gradient at the pedestal top and electron temperature gradient modes (ETGM), candidates for limiting the Te gradient both in the pedestal and on top of the pedestal. The theoretical control parameters for these modes are αMHD for KBM, ηe for ETGM and ηi for ITGM. Experiments are being performed in DIII-D to determine if measured values of these parameters are close to the theoretical threshold values for the linear onset of these modes. Work supported by US DOE under DE-FC02-04ER54698, DE-FG02-07ER54917, DE-FG02-08ER54984, DE-FG02-89ER53296, and DE-FG02-08ER54999.

  4. TRANSPORT BY INTERMITTENCY IN THE BOUNDARY OF THE DIII-D TOKAMAK

    SciTech Connect

    BOEDO,JA; RUDAKOV,DL; MOYER,RA; MCKEE,GR; COLCHIN,RJ; SCHAFFER,MJ; STANGEBY,PG; WEST,WP; ALLEN,SL; EVANS,TE; FONCK,RJ; HOLLMANN,EM; KRASHENINNIKOV,S; LEONARD,AW; NEVINS,W; MAHDAVI,MA; PORTER,GD; TYNAN,GR; WHYTE,DG; XU,X

    2002-11-01

    A271 TRANSPORT BY INTERMITTENCY IN THE BOUNDARY OF THE DIII-D TOKAMAK. Intermittent plasma objectives (IPOs) featuring higher pressure than the surrounding plasma, are responsible for {approx} 50% of the E x B{sub T} radial transport in the scrape off layer (SOL) of the DIII-D tokamak in L- and H-mode discharges. Conditional averaging reveals that the IPOs are positively charged and feature internal poloidal electric fields of up to 4000 V/m. The IPOs move radially with E x B{sub T}/B{sup 2} velocities of {approx} 2600 m/s near the last closed flux surface (LCFS), and {approx} 330 m/s near the wall. The IPOs slow down as they shrink in radial size from 4 cm at the LCFS to 0.5 cm near the wall. The skewness (i.e. asymmetry of fluctuations from the average) of probe and beam emission spectroscopy (BES) data indicate IPO formation at or near the LCFS and the existence of positive and negative IPOs which move in opposite directions. The particle content of the IPOs at the LCFS is linearly dependent on the local density and decays over {approx} 3 cm into the SOL while their temperature decays much faster ({approx} 1 cm).

  5. Extended fast-ion D-alpha diagnostic on DIII-D

    SciTech Connect

    Muscatello, C. M.; Heidbrink, W. W.; Taussig, D.; Burrell, K. H.

    2010-10-15

    A fast-ion deuterium-alpha (FIDA) diagnostic, first commissioned on DIII-D in 2005, relies on Doppler-shifted light from charge-exchange between beam neutrals and energetic ions. The second generation (2G) system was installed on DIII-D in 2009. Its most obvious improvement is the spatial coverage with 11 active in-beam and three passive off-beam views; the latter allows for simultaneous monitoring of the background signal. Providing extended coverage in fast-ion velocity space, the new views possess a more tangential component with respect to the toroidal field compared to their first generation counterparts. Each viewing chord consists of a bundle of three 1.5 mm core fibers to maximize light gathering. For greater throughput, fast f/1.8 optical components are used throughout. The signal is transmitted via fiber optics to a patch panel, so the user is able to choose the detector. FIDA was originally installed with a spectrometer and charge-coupled device (CCD) camera to monitor the full D{sub {alpha}} spectrum for two spatial views. 2G adds another spectrometer and CCD that monitor the blue-shifted wing for six spatial views at 1 kHz. In addition, a photomultiplier tube and fast digitizer provide wavelength-integrated signals at 1 MHz for eight spatial views.

  6. Extended fast-ion D-alpha diagnostic on DIII-D.

    PubMed

    Muscatello, C M; Heidbrink, W W; Taussig, D; Burrell, K H

    2010-10-01

    A fast-ion deuterium-alpha (FIDA) diagnostic, first commissioned on DIII-D in 2005, relies on Doppler-shifted light from charge-exchange between beam neutrals and energetic ions. The second generation (2G) system was installed on DIII-D in 2009. Its most obvious improvement is the spatial coverage with 11 active in-beam and three passive off-beam views; the latter allows for simultaneous monitoring of the background signal. Providing extended coverage in fast-ion velocity space, the new views possess a more tangential component with respect to the toroidal field compared to their first generation counterparts. Each viewing chord consists of a bundle of three 1.5 mm core fibers to maximize light gathering. For greater throughput, fast f/1.8 optical components are used throughout. The signal is transmitted via fiber optics to a patch panel, so the user is able to choose the detector. FIDA was originally installed with a spectrometer and charge-coupled device (CCD) camera to monitor the full D(α) spectrum for two spatial views. 2G adds another spectrometer and CCD that monitor the blue-shifted wing for six spatial views at 1 kHz. In addition, a photomultiplier tube and fast digitizer provide wavelength-integrated signals at 1 MHz for eight spatial views.

  7. High heat flux Langmuir probe array for the DIII-D divertor plates

    SciTech Connect

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

    2008-10-15

    Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m{sup 2} 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 deg. 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 J{sub sat}, T{sub e}, and V{sub f} with 4 mm spatial resolution are shown.

  8. High spatial resolution upgrade of the electron cyclotron emission radiometer for the DIII-D tokamak.

    PubMed

    Truong, D D; Austin, M E

    2014-11-01

    The 40-channel DIII-D electron cyclotron emission (ECE) radiometer provides measurements of Te(r,t) at the tokamak midplane from optically thick, second harmonic X-mode emission over a frequency range of 83-130 GHz. The frequency spacing of the radiometer's channels results in a spatial resolution of ∼1-3 cm, depending on local magnetic field and electron temperature. A new high resolution subsystem has been added to the DIII-D ECE radiometer to make sub-centimeter (0.6-0.8 cm) resolution Te measurements. The high resolution subsystem branches off from the regular channels' IF bands and consists of a microwave switch to toggle between IF bands, a switched filter bank for frequency selectivity, an adjustable local oscillator and mixer for further frequency down-conversion, and a set of eight microwave filters in the 2-4 GHz range. Higher spatial resolution is achieved through the use of a narrower (200 MHz) filter bandwidth and closer spacing between the filters' center frequencies (250 MHz). This configuration allows for full coverage of the 83-130 GHz frequency range in 2 GHz bands. Depending on the local magnetic field, this translates into a "zoomed-in" analysis of a ∼2-4 cm radial region. Expected uses of these channels include mapping the spatial dependence of Alfven eigenmodes, geodesic acoustic modes, and externally applied magnetic perturbations. Initial Te measurements, which demonstrate that the desired resolution is achieved, are presented.

  9. Absorption of Fast Waves at Moderate to High Ion Cyclotron Harmonics on DIII-D

    NASA Astrophysics Data System (ADS)

    Pinsker, R. I.; Porkolab, M.; Heidbrink, W. W.; Luo, Y.; Petty, C. C.; Prater, R.; Choi, M.; Baity, F. W.; Fredd, E.; Hosea, J. C.; Harvey, R. W.; Smirnov, A. P.; Murakami, M.; Van Zeeland, M. A.

    2005-09-01

    The absorption of fast Alfvén waves (FW) by ion cyclotron harmonic damping in the range of harmonics from fourth to eighth is studied theoretically and with experiments in the DIII-D tokamak. A formula for linear ion cyclotron absorption on Maxwellian ion species is used to estimate the single-pass damping for various cases of experimental interest. It is found that damping on fast ions from neutral beam injection can be significant even at the eighth harmonic if the fast ion beta and the background plasma density are both high enough. The predictions are tested in several L-mode experiments in DIII-D with FW power at 60 MHz and at 116 MHz. It is found that 4th and 5th harmonic absorption of the 60 MHz power on the beam ions can be quite strong, but 8th harmonic absorption of the 116 MHz power appears to be weaker than expected. Possible explanations of the discrepancy are discussed.

  10. Absorption of fast waves at moderate to high ion cyclotron harmonics on DIII-D

    NASA Astrophysics Data System (ADS)

    Pinsker, R. I.; Porkolab, M.; Heidbrink, W. W.; Luo, Y.; Petty, C. C.; Prater, R.; Choi, M.; Schaffner, D. A.; Baity, F. W.; Fredd, E.; Hosea, J. C.; Harvey, R. W.; Smirnov, A. P.; Murakami, M.; Van Zeeland, M. A.

    2006-07-01

    The absorption of fast Alfvén waves (FW) by ion cyclotron harmonic damping in the range of harmonics from 4th to 8th is studied theoretically and with experiments in the DIII-D tokamak. A formula for linear ion cyclotron absorption on ions with an arbitrary distribution function which is symmetric about the magnetic field is used to estimate the single-pass damping for various cases of experimental interest. It is found that damping on fast ions from neutral beam injection can be significant even at the 8th harmonic if the fast ion beta, the beam injection energy and the background plasma density are high enough and the beam injection geometry is appropriate. The predictions are tested in several L-mode experiments in DIII-D with FW power at 60 MHz and at 116 MHz. It is found that 4th and 5th harmonic absorption of the 60 MHz power on the beam ions can be quite strong, but 8th harmonic absorption of the 116 MHz power appears to be weaker than expected. The linear modelling predicts a strong dependence of the 8th harmonic absorption on the initial pitch-angle of the injected beam, which is not observed in the experiment. Possible explanations of the discrepancy are discussed.

  11. Absorption of Fast Waves at Moderate to High Ion Cyclotron Harmonics on DIII-D

    SciTech Connect

    Pinsker, R.I.; Petty, C.C.; Prater, R.; Choi, M.; Porkolab, M.; Heidbrink, W.W.; Luo, Y.; Baity, F.W.; Murakami, M.; Fredd, E.; Hosea, J.C.; Harvey, R.W.; Smirnov, A.P.; Van Zeeland, M.A.

    2005-09-26

    The absorption of fast Alfven waves (FW) by ion cyclotron harmonic damping in the range of harmonics from fourth to eighth is studied theoretically and with experiments in the DIII-D tokamak. A formula for linear ion cyclotron absorption on Maxwellian ion species is used to estimate the single-pass damping for various cases of experimental interest. It is found that damping on fast ions from neutral beam injection can be significant even at the eighth harmonic if the fast ion beta and the background plasma density are both high enough. The predictions are tested in several L-mode experiments in DIII-D with FW power at 60 MHz and at 116 MHz. It is found that 4th and 5th harmonic absorption of the 60 MHz power on the beam ions can be quite strong, but 8th harmonic absorption of the 116 MHz power appears to be weaker than expected. Possible explanations of the discrepancy are discussed.

  12. 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).

  13. Quantification of Chemical Erosion in the DIII-D Divertor and Implications for ITER

    SciTech Connect

    McLean, A. G.; Stangeby, P. C.; Bray, B. D.; Brezinsek, S.; Brooks, N. H.; Davis, J. W.; Isler, R. C.; Kirschner, A.; Laengner, R.; Lasnier, C. J.; Mu, Y.; Munoz-Burgos, J. M.; Rudakov, D. L.; Schmitz, O.; Unterberg, Ezekial A; Watkins, J. G.; Whyte, D. G.; Wong, C. P. C.

    2011-01-01

    The Porous Plug Injector (PPI) has proven to be an invaluable diagnostic for in situ characterization and quantification of erosion phenomena in DIII-D. Previous work has led to derivation of three primary figures of merit for chemical erosion (CE) in attached and cold divertor conditions: relative intensity of C+ chemical and physical sources, the CE yield (Y-chem) and effective photon efficiencies for chemically eroded products. Application of these figures for accounting of observed absolutely calibrated CI and CII emission intensities is demonstrated to produce a self-consistent solution at the DIII-D targets. Reinterpretation of the CI (C degrees) spectral lineshape profile supports the relative roles of local chemical versus physical sputtering as previously determined for CII (C+). Comparison of calculated in situ Y-chem to that measured ex situ suggests a tokamak-specific lower energy threshold for CE and has potentially major implications for prediction of tritium co-deposition near the divertor targets in ITER.

  14. GPU-Based Optimal Control Techniques for Resistive Wall Mode Control on DIII-D

    NASA Astrophysics Data System (ADS)

    Clement, M.; Navratil, G. A.; Hanson, J. M.; Strait, E. J.

    2014-10-01

    The DIII-D tokamak can excite strong, locked or nearly locked kink modes whose rotation frequencies do not evolve quickly and are slow compared to their growth rates. To control such modes, DIII-D plans to implement a Graphical Processing Unit (GPU) based feedback control system in a low-latency architecture based on system developed on the HBT-EP tokamak. Up to 128 local magnetic sensors will be used to extrapolate the state of the rotating kink mode, which will be used by the feedback algorithm to calculate the required currents for the internal and/or external control coils. Offline techniques for resolving the mode structure of the resistive wall mode (RWM) will be presented and compared along with the proposed GPU implementation scheme and potential real-time estimation algorithms for RWM feedback. Work supported by the US Department of Energy under DE-FG02-07ER54917, DE-FG02-04ER54761, and DE-FC02-04ER54698.

  15. 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.

  16. 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.

  17. High-performance finite-difference time-domain simulations of C-Mod and ITER RF antennas

    NASA Astrophysics Data System (ADS)

    Jenkins, Thomas G.; Smithe, David N.

    2015-12-01

    Finite-difference time-domain methods have, in recent years, developed powerful capabilities for modeling realistic ICRF behavior in fusion plasmas [1, 2, 3, 4]. When coupled with the power of modern high-performance computing platforms, such techniques allow the behavior of antenna near and far fields, and the flow of RF power, to be studied in realistic experimental scenarios at previously inaccessible levels of resolution. In this talk, we present results and 3D animations from high-performance FDTD simulations on the Titan Cray XK7 supercomputer, modeling both Alcator C-Mod's field-aligned ICRF antenna and the ITER antenna module. Much of this work focuses on scans over edge density, and tailored edge density profiles, to study dispersion and the physics of slow wave excitation in the immediate vicinity of the antenna hardware and SOL. An understanding of the role of the lower-hybrid resonance in low-density scenarios is emerging, and possible implications of this for the NSTX launcher and power balance are also discussed. In addition, we discuss ongoing work centered on using these simulations to estimate sputtering and impurity production, as driven by the self-consistent sheath potentials at antenna surfaces.

  18. Multi-species particle transport in GYRO simulations of low-collisionality, peaked-density H-mode plasmas in C-Mod

    NASA Astrophysics Data System (ADS)

    Mikkelsen, D. R.; Greenwald, M.; Waltz, R.; Candy, J.

    2009-11-01

    Experimental results from Alcator C-Mod have confirmed earlier AUG and JET findings that spontaneous peaking of the density profile in H-mode plasmas depends on collisionality. Previously reported nonlinear, 'full-radius' GYRO simulations [1] of low-collisionality, peaked-density H-mode plasmas in C-Mod generated a particle pinch that was produced exclusively by higher-k modes. Nonlinear simulations of AUG have a similar character [2], and recent detailed linear analyses [2,3] suggest that density peaking may be common in low collisionality plasmas. Here we increase the number of ion species in the simulations to determine whether impurity pinches are also expected, and whether the degree of density peaking is predicted to differ for the three hydrogen isotopes. These simulations include experimentally relevant levels of several impurities, and a range of H/D and D/T mixes.[4pt] [1] D.R. Mikkelsen, et al., Bull. Am. Phys. Soc. 52, (2007) No. 16, 221, NP8.71 [0pt] [2] C. Angioni, et al., Phys. Plasmas 16 (2009) 060702 [0pt] [3] M. Maslov, et al., Nucl. Fusion 49 (2009) 075037

  19. 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.

  20. 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.

  1. 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.

  2. Experimental Tests of Linear and Nonlinear 3D Equilibrium Models in DIII-D

    NASA Astrophysics Data System (ADS)

    King, J. D.

    2014-10-01

    A major upgrade to the DIII-D magnetic diagnostics has been used to show that linear ideal MHD captures the key features of plasma response to 3D fields over a wide range of conditions, while a nonlinear model disagrees with observations. New measurements on the low and high field sides of the torus allow detailed comparisons with ``synthetic diagnostic'' predictions of several MHD models, in plasmas with edge safety factor (q95) and plasma pressure (β) spanning the range expected for ITER's baseline scenario. Model comparisons with DIII-D data confirm that the linear ideal MHD code MARS-F provides a quantitative description of the plasma response to applied fields with toroidal mode numbers n = 1 and n = 3. Similarly, good experimental agreement is seen for the linearized two-fluid M3DC-1 code and ideal MHD IPEC code. In contrast, the n = 1 plasma response predicted by the nonlinear 3D equilibrium code VMEC is found to disagree with measurement in both amplitude (by a factor of 3 or more) and qualitative structure - perhaps related to VMEC's strong sensitivity to the current density near the plasma edge. The measured plasma response to n = 3 perturbations is characterized by strong variations in amplitude at the high field side as the edge safety factor is varied, in good qualitative agreement with ideal MHD predictions of resonant behavior as rational surfaces reach the edge (q95 = m / 3). However, the resonance at q95 = 11 / 3 appears significantly weaker than the others, which suggests the presence of non-ideal effects - possibly loss of screening currents and formation of islands. This observation may shed new light on the physics of the suppression of edge localized modes (ELMs) by n = 3 magnetic perturbations, which in DIII-D occurs most easily with q95 near 11/3. The success of linear, ideal MHD models, as well as the understanding of the limits of their validity (e.g. edge resonances, and the well-known role of kinetic effects in high β plasmas), will be

  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. Collisionality scaling of turbulence and transport in advanced inductive plasmas in DIII-D

    NASA Astrophysics Data System (ADS)

    Yan, Z.; McKee, G. R.; Petty, C.; Luce, T.; Chen, X.; Holland, C.; Rhodes, T.; Schmitz, L.; Wang, G.; Zeng, L.; Marinoni, A.; Solomon, W.; DIII-D Team

    2015-11-01

    The collisionality scaling of multiscale turbulence properties and thermal transport characteristics in high-beta, high confinement Advanced Inductive (AI) plasmas was determined via systematic dimensionless scaling experiments on DIII-D. Preliminary estimate indicates a weak collisionality dependence of energy confinement as v* varied by a factor of ~2. Electron density and scaled (~Bt2) temperature profiles are well matched in the scan. Interestingly, low-k density fluctuation amplitudes are observed to decrease at lower v* near ρ ~ 0 . 75 . Ion and electron thermal transport values, computed with ONETWO using experimentally measured profiles and sources, will be presented, along with multi-scale turbulence measurements obtained with various fluctuation diagnostics. Altering collisionality should change the relative contribution of different modes to transport.

  5. 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.

  6. 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.

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

    DOE PAGES

    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 '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

  8. Near midplane scintillator-based fast ion loss detector on DIII-D

    SciTech Connect

    Chen, X.; Heidbrink, W. W.; Fisher, R. K.; Pace, D. C.; Chavez, J. A.; Van Zeeland, M. A.; Garcia-Munoz, M.

    2012-10-15

    A new scintillator-based fast-ion loss detector (FILD) installed near the outer midplane of the plasma has been commissioned on DIII-D. This detector successfully measures coherent fast ion losses produced by fast-ion driven instabilities ({<=}500 kHz). Combined with the first FILD at {approx}45 Degree-Sign below the outer midplane [R. K. Fisher, et al., Rev. Sci. Instrum. 81, 10D307 (2010)], the two-detector system measures poloidal variation of losses. The phase space sensitivity of the new detector (gyroradius r{sub L}{approx}[1.5-8] cm and pitch angle {alpha}{approx}[35 Degree-Sign -85 Degree-Sign ]) is calibrated using neutral beam first orbit loss measurements. Since fast ion losses are localized poloidally, having two FILDs at different poloidal locations allows for the study of losses over a wider range of plasma shapes and types of loss orbits.

  9. Infrared Monitoring of 110 GHz Gyrotron Windows at Diii-D

    NASA Astrophysics Data System (ADS)

    Gorelov, Y.; Lohr, J.; Callis, R. W.; Ponce, D.

    2003-02-01

    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.

  10. Comparison of Experimental and Theoretical Thermal Diffusivities in the DIII-D Edge Plasma

    NASA Astrophysics Data System (ADS)

    Roveto, J. J.; Stacey, W. M.; Wilks, T. M.

    2015-11-01

    The capability of the Georgia Tech GTEDGE edge transport interpretation code has been upgraded to include improved ion-orbit-loss models for neutral beam and thermalized ions in the edge plasma. We are undertaking a new comparison of various theoretical thermal diffusivity models with the improved interpretation of experimental edge transport now possible. The initial effort is examining two DIII-D shots, #123302, a reference ELMing H-mode shot, and #123301, a matched RMP shot. The improved interpretation leads to quite different experimental thermal diffusivity profiles in the edge than previously reported when ion-orbit-loss effects are included. The experimental values are being compared with various theoretical models, including paleoclassical, neoclassical, ITG, drift ballooning mode, TEM, and ETG. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

  11. 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.

  12. 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.

  13. 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).

  14. 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.

  15. Particle exhaust schemes in the DIII-D advanced divertor configuration

    SciTech Connect

    Menon, M.M.; Mioduszewski, P.K.

    1989-01-01

    For density control in long-pulse operation, the open divertor on the DIII-D tokamak will be equipped with a baffled chamber and a pumping system. The throat of the baffle chamber is sized to provide optimal pumping for the typical plasma equilibrium configuration. Severe limitations on the toroidal conductance of this baffle chamber require the use of in-vessel pumping to achieve the desired particle exhaust of about 25 Torr{center dot}l/s. Two separate pumping schemes are considered: an array of titanium getter modules based on the design developed by the Tore Supra team and a cryocondensation pump. The merits and demerits of each scheme are analyzed, and the design considerations introduced by the tokamak environment are brought out. 3 refs., 5 figs.

  16. Electron Cyclotron Current Drive at High Electron Temperature on DIII-D

    NASA Astrophysics Data System (ADS)

    Petty, C. C.; Austin, M. E.; Harvey, R. W.; Lohr, J.; Luce, T. C.; Makowski, M. A.; Prater, R.

    2007-09-01

    Experiments on DIII-D have measured the electron cyclotron current drive (ECCD) efficiency for co- and counter-injection in low density plasmas with radiation temperatures from electron cyclotron emission (ECE) above 20 keV. The radiation temperature is generally higher than the Thomson scattering temperature, indicating that there is a significant population of non-thermal electrons. The experimental ECCD profile measured with motional Stark effect (MSE) polarimetry is found to agree with quasi-linear theory except for the highest power density cases (QEC/ne2≫1). Radial transport of the energetic electrons with diffusion coefficients of ˜0.4 m2/s is needed to model the broadened ECCD profile at high power density.

  17. ACTIVE FILTER HARDWARE DESIGN & PERFORMANCE FOR THE DIII-D PLASMA CONTROL SYSTEM

    SciTech Connect

    SELLERS,D; FERRON,J.R; WALKER,M.L; BROESCH,J.D

    2003-10-01

    OAK-B135 The digital plasma control system (PCS), currently in operation on the DIII-D tokamak, requires inputs from a large number of sensors. Due to the nature of the digitizers and the relative noisy environment from which these signals are derived, each of the 32 signals must be conditioned via an active filter. Two different types of filters, Chebyshev and Bessel with fixed frequencies: 100 Hz Bessel was used for filtering the motional Stark effect diagnostic data. 800 Hz Bessel was designed to filter plasma control data and 1200 Hz Chebyshev is used with closed loop control of choppers. The performance of the plasma control system is greatly influenced by how well the actual filter responses match the software model used in the control system algorithms. This paper addresses the various issues facing the designer in matching the electrical design with the theoretical.

  18. 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.

  19. DIII-D in-vessel port cover and shutter assembly for the phase contrast interferometer

    SciTech Connect

    Phelps, R.D.

    1994-01-01

    The entire outer wall of the DIII-D vacuum vessel interion is covered with a regular array of graphite tiles. Certain of the diagnostic ports through the outer vessel wall contain equipment which is shielded from the plasma by installing port covers designed to withstand energy deposition. If the diagnostic contained in the port must communicate with the vessel volume, a shutter assembly is usually provided. In the ports at 285 degrees, R+1 and R-1, interferometer mirrors have been installed to provide a means for transmitting a large diameter CO-2 laser beam through the edge of the plasma. To protect the mirrors and other hardware contained in these ports, a special protective plate and shutter arrangement has been designed. This report describes the details of design, fabrication, and installation of these protective covers and shutters.

  20. Demonstration of ELM pacing by Pellet Injection on DIII-D and Extrapolation to ITER

    SciTech Connect

    Baylor, Larry R; Commaux, Nicolas JC; Jernigan, Thomas C; Parks, P. B.; Evans, T.E.; Osborne, T. H.; Strait, E. J.; Fenstermacher, M. E.; Lasnier, C. J.; Moyer, R.A.; Yu, J.H.

    2010-01-01

    Demonstration of ELM pacing by pellet injection on DIII-D and extrapolation to ITER<#_ftn1>* L.R. Baylor1, N. Commaux1, T.C. Jernigan1, P.B. Parks2, T.E. Evans2, T.H. Osborne2, E.J. Strait2, M.E. Fenstermacher3, C.J. Lasnier3, R.A. Moyer4, J.H. Yu4 1Oak Ridge National Laboratory, Oak Ridge, TN, USA 2General Atomics, San Diego, CA, USA 3 Lawrence Livermore National Laboratory, Livermore, CA, USA 4University of California San Diego, La Jolla, CA, USA Deuterium pellet injection has been used in experiments on the DIII-D tokamak to investigate the possibility of triggering small rapid edge localized modes (ELMs) in reactor relevant plasma regimes. ELMs have been observed to be triggered from small 1.8 mm pellets injected from all available locations and under all H-mode operating scenarios in DIII-D. Experimental details have shown that the ELMs are triggered before the pellets reach the top of the H-mode pedestal, implying that very small shallow penetrating pellets are sufficient to trigger ELMs. Fast camera images of the pellet entering the plasma from the low field side show a single plasma filament becoming visible near the pellet cloud and striking the outer vessel wall within 200 ms. Additional ejected filaments are then observed to subsequently reach the wall. The plasma stored energy loss from the pellet triggered ELMs is a function of the elapsed time after a previous ELM. Pellet ELM pacing has been proposed as a method to prevent large ELMs that can damage the ITER plasma facing components [1]. A demonstration of pacing of ELMs on DIII-D was made by injecting slow 14 Hz pellets on the low field side in an ITER shape plasma with low natural ELM frequency and a normalized b of 1.8. The non-pellet discharge natural ELM frequency was ~5 Hz with ELM energy losses up to 85 kJ (>10% of total stored energy) while the case with pellets was able to demonstrate >20 Hz ELMs with an average ELM energy loss less than 22 kJ (<3% of the total). The resulting ELM frequency

  1. Attainment of quasi steady-state H-mode plasmas in the DIII-D tokomak

    SciTech Connect

    Mahdavi, M.A.; Kellman, A.G.; Gohil, P.; Brooks, N.; Burrell, K.H.; Groebner, R.J.; Haas, G.; Hill, D.; Jackson, G.; Janeschitz, G.

    1989-01-01

    We report the results of experiments on DIII-D where quasi steady-state H-mode plasma with low amplitude Edge Localized Modes (ELMs) were maintained for 5 seconds. Within one second from the onset of the H-mode, most plasma parameters reach their asymptotic values (/bar n//sub e/ approx. 7 /times/ 10/sup 13/ cm/sup /minus/3/, /tau//sub E/ approx. = 100 ms, Z/sub eff/ = 1.7) and the electron density and temperature profiles are no longer evolving. By this time, frequent ELMs and effective helium wall conditioning limit the rate of increase of /bar n//sub e/ (less than or equal to10% of the beam fueling rate) and maintain the total radiated power form impurities at a constant level. 5 refs., 2 figs.

  2. Detailed OEDGE modeling of core-pedestal fueling in DIII-D

    SciTech Connect

    Elder, J. D.; Leonard, A. W.; Stangeby, P. C.; Boedo, J.A.; Bray, B. D.; Brooks, N. H.; Fenstermacher, M. E.; Reiter, D.; Unterberg, Ezekial A; Watkins, J. G.; Lisgo, S.

    2013-01-01

    The OEDGE code is used to model core fueling for attached L-mode plasmas and between edge localized modes (ELMs) for attached H-mode plasmas in DIII-D. Empirical plasma reconstruction has been used to determine the plasma conditions in these discharges. EIRENE is used to model the hydrogen recycling. Divertor recycling accounts for 65 100% of the core fueling. The fraction of the total divertor target flux ionized inside the separatrix ranges from 5% to 20%. The fraction of total wall flux ionized inside the separatrix ranges from 20% to 50%. Neutrals originating from wall regions closer to the separatrix are more likely to ionize in the confined plasma. Ionization in the confined plasma is concentrated below the midplane with peaks in the poloidal profiles just above the X-point. Radial core ionization in high density H-mode is peaked strongly near the separatrix.

  3. Statistical Analysis of Locked Modes and their Disruptivity at DIII-D

    NASA Astrophysics Data System (ADS)

    Sweeney, R.; Choi, W.; Olofsson, K. E. J.; Volpe, F. A.

    2014-10-01

    A database has been developed to study locking and disruptivity of neoclassical tearing modes with poloidal and toroidal mode numbers m = 2 and n = 1. Approximately 30,000 DIII-D discharges are studied providing statistics on the fraction of disruptions containing locked modes (LMs) and the ratio of disruptive LMs to all LMs. Other quantities analyzed include the time-scales between mode-formation and locking, and between locking and disruption, the amplitude of the mode upon locking and disruption, the existence or lack of a rotating precursor, and the toroidal phase of locking. Correlations are examined between locking and disruptivity and parameters such as plasma beta and neutral beam torque. Simple interpretations are provided in terms of island size and torques acting on the island, and implications for an automatic locked mode controller are discussed. This work was supported by the US Department of Energy under DE-SC0008520.

  4. Comparisons of Runaway Electron Production in DIII--D with the CQL3D Model

    NASA Astrophysics Data System (ADS)

    Harvey, R. W.; Chan, V. S.; Chiu, S. C.; Evans, T. E.; Whyte, D. G.; Rosenbluth, M. N.

    1997-11-01

    The time-dependent CQL3D relativistic, bounce-averaged Fokker-Planck code is well suited to modeling the electron runaway generation by ``tail slideaway'' and ``knockon'' processes associated with rapid plasma temperature drop as occurs during plasma disruption and pellet injection. ``Rapid temperure drop'' means in a time short compared to the resistive time thus leading to large inductive electric field; and also short compared to tail electron slowing down time thereby providing a significant number of high velocity electrons. The tail electrons find themselves beyond the critical velocity for runaway. During DIII--D pellet injection experiments, in the plasma interior we calculate nearly 100 percent of the plasma current is transferred to runaway electrons by tail slideaway. This effect diminishes towards the plasma edge, where the knockon process can become dominant on a longer time scale. These computational results will be benchmarked against the experimental results.

  5. Striation pattern of target particle and heat fluxes in three dimensional simulations for DIII-D

    SciTech Connect

    Frerichs, H.; Schmitz, O.; Reiter, D.; Evans, T. E.; Feng, Y.

    2014-02-15

    The application of resonant magnetic perturbations results in a non-axisymmetric striation pattern of magnetic field lines from the plasma interior which intersect the divertor targets. The impact on related particle and heat fluxes is investigated by three dimensional computer simulations for two different recycling conditions (controlled via neutral gas pumping). It is demonstrated that a mismatch between the particle and heat flux striation pattern (splitting vs. no splitting), as is repeatedly observed in ITER similar shape H-mode plasmas at DIII-D, can be reproduced by the simulations for high recycling conditions at the onset of partial detachment. These results indicate that a detailed knowledge of the particle and energy balance is at least as important for realistic simulations as the consideration of a change in the magnetic field structure by plasma response effects.

  6. 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.

  7. 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.

  8. Near midplane scintillator-based fast ion loss detector on DIII-D.

    PubMed

    Chen, X; Fisher, R K; Pace, D C; García-Muñoz, M; Chavez, J A; Heidbrink, W W; Van Zeeland, M A

    2012-10-01

    A new scintillator-based fast-ion loss detector (FILD) installed near the outer midplane of the plasma has been commissioned on DIII-D. This detector successfully measures coherent fast ion losses produced by fast-ion driven instabilities (≤500 kHz). Combined with the first FILD at ∼45° below the outer midplane [R. K. Fisher, et al., Rev. Sci. Instrum. 81, 10D307 (2010)], the two-detector system measures poloidal variation of losses. The phase space sensitivity of the new detector (gyroradius r(L) ∼ [1.5-8] cm and pitch angle α ∼ [35°-85°]) is calibrated using neutral beam first orbit loss measurements. Since fast ion losses are localized poloidally, having two FILDs at different poloidal locations allows for the study of losses over a wider range of plasma shapes and types of loss orbits.

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

    NASA Astrophysics Data System (ADS)

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

    2015-02-01

    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.

  10. Experiments on ion cyclotron damping at the deuterium fourth harmonic in DIII-D

    SciTech Connect

    Pinsker, R. I.; Baity, F. W.; Bernabei, S.; Greenough, N.; Heidbrink, W. W.; Mau, T. K.; Petty, C. C.; Porkolab, M.

    1999-09-20

    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 {approx}1 MW. (c) 1999 American Institute of Physics.

  11. Coupling of global toroidal Alfven eigenmodes and reversed shear Alfven eigenmodes in DIII-D

    SciTech Connect

    Van Zeeland, M. A.; Turnbull, A. D.; Austin, M. E.; Gorelenkov, N. N.; Kramer, G. J.; Nazikian, R.; Heidbrink, W. W.; Ruskov, E.; Makowski, M. A.; McKee, G. R.

    2007-05-15

    Reversed shear Alfven eigenmodes (RSAEs) are typically thought of as being localized near the minima in the magnetic safety factor profile, however, their spatial coupling to global toroidal Alfven eigenmodes (TAEs) has been observed in DIII-D discharges. For a decreasing minimum magnetic safety factor, the RSAE frequency chirps up through that of stable and unstable TAEs. Coupling creates a small gap at the frequency degeneracy point forming two distinct global modes. The core-localized RSAE mode structure changes and becomes temporarily global. Similarly, near the mode frequency crossing point, the global TAE extends deeper into the plasma core. The frequency splitting and spatial structure of the two modes throughout the various coupling stages, as measured by an array of internal fluctuation diagnostics, are in close agreement with linear ideal MHD calculations using the NOVA code. The implications of this coupling for eigenmode stability is also investigated and marked changes are noted throughout the coupling process.

  12. Modeling of fast wave absorption by beam ions in DIII-D discharges

    SciTech Connect

    Mau, T. K.; Petty, C. C.; Porkolab, M.; Heidbrink, W. W.

    1999-09-20

    In recent discharges on DIII-D, neutron measurements indicated absorption of the fast wave by energetic deuterium beam ions when the fourth harmonic resonance is on axis, but little or no interaction for the fifth harmonic. In this work, a geometric optics code is used to quantify the beam ion absorption of fast waves as the frequency (or on-axis harmonic resonance) is varied. Isotropic and anisotropic Maxwellians are used to model the beam ion distribution. Wave power flow in this harmonic range has been found to exhibit a strong poloidal and toroidal behavior in its initial transits across the plasma. Absorption along the rays is calculated using the fully thermal and magnetized treatment. Competing with the beam ions for absorption are the minority hydrogen and background electrons. The modeling results are only in partial agreement with experimental observations, indicating that more detailed physics may need to be included. (c) 1999 American Institute of Physics.

  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. 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.

  15. 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.

  16. 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.

  17. The 110 GHz Gyrotron System on DIII-D: Gyrotron Tests and Physics Results

    SciTech Connect

    J. Lohr; P. Calahan; R.W. Callis; T.S. Chu; J.S. deGrassie; I. Gorelov; H. Ikezi; R.A. Legg; T.C. Luce; C.C. Petty; D. Ponce; R. Prater; D.J. Schuster; S.E. Tsimring

    1999-12-01

    The DIII-D tokamak has installed a system with three gyrotrons at the 1 MW level operating at 110 GHz. Physics experiments on electron cyclotron current drive, heating, and transport have been performed. Good efficiency has been achieved both for on-axis and off-axis current drive with relevance for control of the current density profile leading to advanced regimes of tokamak operation, although there is a difference between off-axis ECCD efficiency inside and outside the magnetic axis. Heating efficiency is excellent and electron temperatures up to 10 keV have been achieved. The gyrotron system is versatile, with poloidal scan and control of the polarization of the injected rf beam. Phase correcting mirrors form a Gaussian beam and focus it into the waveguide. Both perpendicular and oblique launch into the tokamak have been used. Three different gyrotron designs are installed and therefore unique problems specific to each have been encountered, including parasitic oscillations, mode hops during modulation and polarization control problems. Two of the gyrotrons suffered damage during operations, one due to filament failure and one due to a vacuum leak. The repairs and subsequent testing will be described. The transmission system uses evacuated, windowless waveguide and the three gyrotrons have output windows of three different materials. One gyrotron uses a diamond window and generates a Gaussian beam directly. The development of the system and specific tests and results from each of the gyrotrons will be presented. The DIII-D project has committed to an upgrade of the system, which will add three gyrotrons in the 1 MW class, all using diamond output windows, to permit operation at up to ten seconds per pulse at one megawatt output for each gyrotron.

  18. DISCHARGE IMPROVEMENT THROUGH CONTROL OF NEOCLASSICAL TEARING MODES BY LOCALIZED ECCD IN DIII-D

    SciTech Connect

    PRATER,R; LAHAYE,RJ; LOHR,J; LUCE,TC; PETTY,CC; FERRON,JR; HUMPHREYS,DA; STRAIT,EJ; PERKINS,FW; HARVEY,RW

    2002-10-01

    A271 DISCHARGE IMPROVEMENT THROUGH CONTROL OF NEOCLASSICAL TEARING MODES BY LOCALIZED ECCD IN DIII-D. Neoclassical tearing modes (NTMs) are MHD modes which can limit the performance of high beta discharges in tokamaks, in some cases leading to a major disruption. The destabilizing effect which results in NTM growth is a helical decrease in the bootstrap current caused by a local reduction of the plasma pressure gradient by seed magnetic islands. The NTM is particularly well suited to control since the mode is linearly stable although nonlinearly unstable, so if the island amplitude can be decreased below a threshold size the mode will decay and vanish. One means of shrinking the island is the replacement of the missing bootstrap current by a localized current generated by electron cyclotron current drive (ECCD). This method has been applied to the m=3/n=2 neoclassical tearing mode in DIII-D, in H-mode plasmas with ongoing ELMs and sawteeth, both of which generate seed islands periodically. In the case of the 3/2 mode, full suppression was obtained robustly by applying about 1.5 MW of ECCD very near the rational surface of the mode. When the mode first appears in the plasma the stored energy decreases by 20%, but after the mode is stabilized by the ECCD the beta may be raised above the initial threshold pressure by 20% by additional neutral beam heating, thereby generating an improvement in the limiting beta of nearly a factor 2. An innovative automated search algorithm was implemented to find and retain the optimum location for the ECCD in the presence of the mode.

  19. MULTI-MODE ERROR FIELD CORRECTION ON THE DIII-D TOKAMAK

    SciTech Connect

    SCOVILLE, JT; LAHAYE, RJ

    2002-10-01

    OAK A271 MULTI-MODE ERROR FIELD CORRECTION ON THE DIII-D TOKAMAK. Error field optimization on DIII-D tokamak plasma discharges has routinely been done for the last ten years with the use of the external ''n = 1 coil'' or the ''C-coil''. The optimum level of correction coil current is determined by the ability to avoid the locked mode instability and access previously unstable parameter space at low densities. The locked mode typically has toroidal and poloidal mode numbers n = 1 and m = 2, respectively, and it is this component that initially determined the correction coil current and phase. Realization of the importance of nearby n = 1 mode components m = 1 and m = 3 has led to a revision of the error field correction algorithm. Viscous and toroidal mode coupling effects suggested the need for additional terms in the expression for the radial ''penetration'' field B{sub pen} that can induce a locked mode. To incorporate these effects, the low density locked mode threshold database was expanded. A database of discharges at various toroidal fields, plasma currents, and safety factors was supplement4ed with data from an experiment in which the fields of the n = 1 coil and C-coil were combined, allowing the poloidal mode spectrum of the error field to be varied. A multivariate regression analysis of this new low density locked mode database was done to determine the low density locked mode threshold scaling relationship n{sub e} {proportional_to} B{sub T}{sup -0.01} q{sub 95}{sup -0.79} B{sub pen} and the coefficients of the poloidal mode components in the expression for B{sub pen}. Improved plasma performance is achieved by optimizing B{sub pen} by varying the applied correction coil currents.

  20. A tangentially viewing VUV TV system for the DIII-D divertor

    SciTech Connect

    Nilson, D.G.; Ellis, R.; Fenstermacher, M.E.; Brewis, G.; Jalufka, N.

    1998-07-01

    A video camera system capable of imaging VUV emission in the 120--160 nm wavelength range, from the entire divertor region in the DIII-D tokamak, was designed. The new system has a tangential view of the divertor similar to an existing tangential camera system which has produced two dimensional maps of visible line emission (400--800 nm) from deuterium and carbon in the divertor region. However, the overwhelming fraction of the power radiated by these elements is emitted by resonance transitions in the ultraviolet, namely the C IV line at 155.0 nm and Ly-{alpha} line at 121.6 nm. To image the ultraviolet light with an angular view including the inner wall and outer bias ring in DIII-D, a 6-element optical system (f/8.9) was designed using a combination of reflective and refractive optics. This system will provide a spatial resolution of 1.2 cm in the object plane. An intermediate UV image formed in a secondary vacuum is converted to the visible by means of a phosphor plate and detected with a conventional CID camera (30 ms framing rate). A single MgF{sub 2} lens serves as the vacuum interface between the primary and secondary vacuums; a second lens must be inserted in the secondary vacuum to correct the focus at 155 nm. Using the same tomographic inversion method employed for the visible TV, they reconstruct the poloidal distribution of the UV divertor light. The grain size of the phosphor plate and the optical system aberrations limit the best focus spot size to 60 {micro}m at the CID plane. The optical system is designed to withstand 350 C vessel bakeout, 2 T magnetic fields, and disruption-induced accelerations of the vessel.

  1. Software upgrade for the DIII-D neutral beam control systems

    SciTech Connect

    Cummings, J.W.; Thurgood, P.A.

    1991-11-01

    The neutral beams are used to heat the plasma in the DIII-D tokamak, a fusion energy research experiment operated by General Atomics (GA) and funded by the Department of Energy (DOE). The experiment is dedicated to demonstrating noninductive current drive of high beta high temperature divertor plasma with good confinement. The neutral beam heating system for the DIII-D tokamak uses four MODCOMP Classic computers for data acquisition and control of the four beamlines. The Neutral Beam Software Upgrade project was launched in early 1990. The major goals were to upgrade the MAX IV operating system to the latest revision (K.1), use standard MODCOMP software (as much as possible), and to develop a very user friendly,'' versatile system. Accomplishing these goals required new software to be developed and modifications to existing applications software to make it compatible with the latest operating system. The custom operating system modules to handle the message service and interrupt handling were replaced by the standard MODCOMP Inter Task Communication (ITC) and interrupt routines that are part of the MAX IV operating system. The message service provides the mechanism for doing shot task sequencing (task scheduling). The interrupt routines are used to connect external interrupts to the system. The new software developed consists of a task dispatcher, screen manager, and interrupt tasks. The existing applications software had to be modified to be compatible with the MODCOMP ITC services and consists of the Modcomp Infinity Data Base Manager, a multi-user system, and menu-driven operating system interface routines using the Infinity Data Base Manager.

  2. Recent results from the DIII-D tokamak and implications for future devices

    SciTech Connect

    Luxon, J.L.

    1995-02-01

    Improvements to the DIII-D tokamak have led to significant new research results and enhanced performance. These results provide important inputs to the design of next generation divertor systems including the upgrade of the DIII-D divertor. The use of graphite for the plasma facing components and careful wall preparation has enabled the routine achievement of regimes of enhanced energy confinement. In elongated discharges, triangularity has been found to be important in attaining good discharge performance as measured by the product of the normalized plasma pressure and the energy confinement time, {beta}{tau}{sub E} This constrains the design of the divertor configuration (X-point location). Active pumping of the divertor region using an in-situ toroidal cryogenic pump has demonstrated control of the plasma density in H-mode discharges and allowed the dependence of confinement on plasma density and current to be separately determined. Helium removal from the plasma edge sufficient to achieve effective ash removal in reactor discharges has also been demonstrated using this pumping configuration. The reduction of the heat flux to the divertor plates has been demonstrated using two different techniques to increase the radiation in the boundary regions of the plasma and thus reduce the heat flux to the divertor plates; deuterium gas injection has been used to create a strongly radiating localized zone near the X-point, and impurity (neon) injection to enhance the radiation from the plasma mantle. Precise shaping of the plasma current profile has been found to be important in achieving enhanced tokamak performance. Transiently shaped current profiles have been used to demonstrate regimes of plasmas with high beta and good confinement. Control of the current profile also is important to sustaining the plasma in the Very High (VH)-mode of energy confinement.

  3. Core barrier formation near integer q surfaces in DIII-D

    SciTech Connect

    Austin, M. E.; Gentle, K. W.; Burrell, K. H.; Waltz, R. E.; Gohil, P.; Greenfield, C. M.; Groebner, R. J.; Petty, C. C.; Prater, R.; Heidbrink, W. W.; Luo, Y.; Kinsey, J. E.; Makowski, M. A.; McKee, G. R.; Shafer, M. W.; Nazikian, R.; Rhodes, T. L.; Van Zeeland, M. A.

    2006-08-15

    Recent DIII-D experiments have significantly improved the understanding of internal transport barriers (ITBs) that are triggered close to the time when an integer value of the minimum in q is crossed. While this phenomenon has been observed on many tokamaks, the extensive transport and fluctuation diagnostics on DIII-D have permitted a detailed study of the generation mechanisms of q-triggered ITBs as pertaining to turbulence suppression dynamics, shear flows, and energetic particle modes. In these discharges, the evolution of the q profile is measured using motional Stark effect polarimetry and the integer q{sub min} crossings are further pinpointed in time by the observation of Alfven cascades. High time resolution measurements of the ion and electron temperatures and the toroidal rotation show that the start of improved confinement is simultaneous in all three channels, and that this event precedes the traversal of integer q{sub min} by 5-20 ms. There is no significant low-frequency magnetohydrodynamic activity prior to or just after the crossing of the integer q{sub min} and hence magnetic reconnection is determined not to be the precipitant of the confinement change. Instead, results from the GYRO code point to the effects of zonal flows near low order rational q values as playing a role in ITB triggering. A reduction in local turbulent fluctuations is observed at the start of the temperature rise and, concurrently, an increase in turbulence poloidal flow velocity and flow shear is measured with the beam emission spectroscopy diagnostic. For the case of a transition to an enduring internal barrier the fluctuation level remains at a reduced amplitude. The timing and nature of the temperature, rotation, and fluctuation changes leading to internal barriers suggests transport improvement due to increased shear flow arising from the zonal flow structures.

  4. ELM PARTICLE AND ENERGY TRANSPORT IN THE SOL AND DIVERTOR OF DIII-D

    SciTech Connect

    FENSTERMACHER,ME; LEONARD,AW; SNYDER,PB; BOEDO,JA; COLCHIN,RJ; GROEBNER,RJ; GRAY,DS; GROTH,M; HOLLMANN,E; LASNIER,CJ; OSBORNE,TH; PETRIE,TW; RUDAKOV,DL; TAKAHASHI,H; WATKINS,JG; ZENG,L

    2003-04-01

    A271 ELM PARTICLE AND ENERGY TRANSPORT IN THE SOL AND DIVERTOR OF DIII-D. Results from a series of dedicated experiments measuring the effect of particle and energy pulses from Type-I Edge Localized Modes (ELMs) in the DIII-D scrape-off layer (SOL) and divertor are compared with a simple model of ELM propagation in the boundary plasma. The simple model asserts that the propagation of ELM particle and energy perturbations is dominated by ion parallel convection along SOL fields lines and the recovery from the ELM perturbation is determined by recycling physics. Time scales associated with the initial changes of boundary plasma parameters are expected to be on the order of the ion transit time from the outer midplane, where the ELM instability is initiated, to the divertor targets. To test the model, the ion convection velocity is changed in the experiment by varying the plasma density. At moderate to high density, n{sub e}/n{sub Gr} = 0.5-0.8, the delays in the response of the boundary plasma to the midplane ELM pulses, the density dependence of those delays and other observations are consistent with the model. However, at the lowest densities, n{sub e}/n{sub Gr} {approx} 0.35, small delays between the response sin the two divertors, and changes in the response of the pedestal thermal energy to ELM events, indicate that additional factors including electron conduction in the SOL, the pre-ELM condition of the divertor plasma, and the ratio of ELM instability duration to SOL transit time, may be playing a role. The results show that understanding the response of the SOL and divertor plasmas to ELMs, for various pre-ELM conditions, is just as important to predicting the effect of ELM pulses on the target surfaces of future devices as is predicting the characteristics of the ELM perturbation of the core plasma.

  5. Progress in GYRO validation studies of DIII-D H-mode plasmas

    NASA Astrophysics Data System (ADS)

    Holland, C.; Petty, C. C.; Schmitz, L.; Burrell, K. H.; McKee, G. R.; Rhodes, T. L.; Candy, J.

    2012-11-01

    The need for a validated predictive capability of turbulent transport in ITER is now widely recognized. However, to date most validation studies of nonlinear codes such as GYRO (Candy and Waltz 2003 J. Comput. Phys. 186 545) have focused upon low power L-mode discharges, which have significant differences in key dimensionless parameters such as ρ* = ρs/a from more ITER-relevant H-mode discharges. In order to begin addressing this gap, comparisons of the turbulent transport and fluctuations predicted by nonlinear GYRO simulations for a number of DIII-D (Luxon 2002 Nucl. Fusion 42 614) H-mode discharges to power balance analyses and experimental measurements are presented. The results of two H-mode studies are presented in this paper, this first of which investigates the importance of nonlocality at typical DIII-D H-mode ρ* values. Electrostatic global GYRO simulations of H-mode discharges at low and high rotation are shown to predict turbulence and transport levels lower than corresponding local simulations, and which are consistent with or slightly above experimental measurements and power balance analyses, even at ‘near-edge’ radii where gyrofluid and gyrokinetic models systematically underpredict turbulence and transport levels. The second study addresses the stabilizing effect of a significant density of energetic particles on turbulent transport. The results of local GYRO simulations of low-density QH-mode plasmas are presented, which model the fast beam ion population as a separate, dynamic ion species. The simulations show a significant reduction of transport with this fast ion treatment, which helps to understand previously reported results (Holland et al 2011 Phys. Plasmas 18 056113) in which GYRO simulations without this treatment significantly overpredicted (by a factor of 10 or more) power balance calculations. These results are contrasted with simulations of a high-density, low fast ion fraction QH-mode discharge, which predict transport levels

  6. High spatial resolution upgrade of the electron cyclotron emission radiometer for the DIII-D tokamak

    SciTech Connect

    Truong, D. D.; Austin, M. E.

    2014-11-15

    The 40-channel DIII-D electron cyclotron emission (ECE) radiometer provides measurements of T{sub e}(r,t) at the tokamak midplane from optically thick, second harmonic X-mode emission over a frequency range of 83–130 GHz. The frequency spacing of the radiometer's channels results in a spatial resolution of ∼1–3 cm, depending on local magnetic field and electron temperature. A new high resolution subsystem has been added to the DIII-D ECE radiometer to make sub-centimeter (0.6–0.8 cm) resolution T{sub e} measurements. The high resolution subsystem branches off from the regular channels’ IF bands and consists of a microwave switch to toggle between IF bands, a switched filter bank for frequency selectivity, an adjustable local oscillator and mixer for further frequency down-conversion, and a set of eight microwave filters in the 2–4 GHz range. Higher spatial resolution is achieved through the use of a narrower (200 MHz) filter bandwidth and closer spacing between the filters’ center frequencies (250 MHz). This configuration allows for full coverage of the 83–130 GHz frequency range in 2 GHz bands. Depending on the local magnetic field, this translates into a “zoomed-in” analysis of a ∼2–4 cm radial region. Expected uses of these channels include mapping the spatial dependence of Alfven eigenmodes, geodesic acoustic modes, and externally applied magnetic perturbations. Initial T{sub e} measurements, which demonstrate that the desired resolution is achieved, are presented.

  7. 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.

  8. Measurement and simulation of ICRF wave intensity with a recalibrated phase contrast imaging diagnostic on Alcator C-Mod

    NASA Astrophysics Data System (ADS)

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

    2015-12-01

    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.

  9. 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.

  10. 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.

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

    NASA Astrophysics Data System (ADS)

    Lin, Yijun

    2014-10-01

    For ICRF antenna utilization in future fusion reactors, taming the antenna-plasma edge interaction while robustly coupling RF power is a critical challenge. Using a novel field-aligned (FA) ICRF antenna where the antenna straps are perpendicular to the total magnetic field, we have shown dramatically improved ICRF antenna performance. The FA antenna has significantly reduced antenna impurity sources, core impurity contamination and radiated power compared to conventional toroidally aligned antennas. The FA antenna also has load tolerance to plasma transients and significantly reduced RF-enhanced heat flux. The emerging physics picture is that the FA antenna minimizes generation of slow wave fields (E//B polarization). This reduction in slow wave lowers the local RF sheath around the ICRF antenna, and thus lowers the impurity source at local antenna structure. Simplified antenna simulations show a strong reduction in slow wave fields. The reduction of the slow wave field also impacts the antenna load tolerance. With the slow wave present, the antenna impedance is strongly modified by the slow wave coupling between antenna straps and this coupling is dependent upon the local density. With reduced slow wave coupling, the antenna reactive impedance is defined by the strap geometry and independent of the plasma whereas the real impedance is determined by the fast wave coupling. Experimentally we have found that the FA antenna loading has similar trends versus plasma current and densities to TA antennas, but the FA antenna reflection coefficient has significantly reduced variation, particularly during L-H and H-L transitions, and ELMs. Further comparisons of the FA and TA antennas are underway with an extensive array of diagnostics to characterize the RF plasma edge interaction and the latest results will be presented. Supported by US DoE awards DE-FC02-99ER54512 at MIT.

  12. Control of Plasma-Stored Energy for Burn Control using DIII-D In-Vessel Coils

    SciTech Connect

    Hawryluk, R. J.; Eidietis, N. W.; Grierson, B. A.; Hyatt, A. W.; Koleman, E.; Logan, N. C.; Nazikian, R.; Paz-Soldan, C.; Wolf, S.

    2014-09-01

    A new approach has been experimentally demonstrated to control the stored energy by applying a non-axisymmetric magnetic field using the DIII-D in-vessel coils to modify the energy confinement time. In future burning plasma experiments as well as magnetic fusion energy power plants, various concepts have been proposed to control the fusion power. The fusion power in a power plant operating at high gain can be related to the plasma-stored energy and hence, is a strong function of the energy confinement time. Thus, an actuator, that modifies the confinement time, can be used to adjust the fusion power. In relatively low collisionality DIII-D discharges, the application of non-axisymmetric magnetic fields results in a decrease in confinement time and density pumpout. Gas puffing was used to compensate the density pumpout in the pedestal while control of the stored energy was demonstrated by the application of non-axisymmetric fields.

  13. High resolution main-ion charge exchange spectroscopy in the DIII-D H-mode pedestal

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

    A new high spatial resolution main-ion (deuterium) charge-exchange spectroscopy system covering the tokamak boundary region has been installed on the DIII-D tokamak. Sixteen new edge main-ion charge-exchange recombination sightlines have been combined with nineteen impurity sightlines in a tangentially viewing geometry on the DIII-D midplane with an interleaving design that achieves 8 mm inter-channel radial resolution for detailed profiles of main-ion temperature, velocity, charge-exchange emission, and neutral beam emission. At the plasma boundary, we find a strong enhancement of the main-ion toroidal velocity that exceeds the impurity velocity by a factor of two. The unique combination of experimentally measured main-ion and impurity profiles provides a powerful quasi-neutrality constraint for reconstruction of tokamak H-mode pedestals.

  14. Projected profile similarity in gyrokinetic simulations of Bohm and gyro-Bohm scaled DIII-D L and H modes

    SciTech Connect

    Waltz, R. E.; Candy, J.; Petty, C. C.

    2006-07-15

    Global gyrokinetic simulations of DIII-D [M. A. Mahdavi and J. L. Luxon, in 'DIII-D Tokamak Special Issue', Fusion Sci. Technol. 48, 2 (2005)] L- and H-mode dimensionally similar discharge pairs are treated in detail. The simulations confirm the Bohm scaling of the well-matched L-mode pair. The paradoxical but experimentally apparent gyro-Bohm scaling of the H-mode pair at larger relative gyroradius (rho-star) and lower transport levels is due to poor profile similarity. Simulations of projected experimental plasma profiles with perfect similarity show both the L- and H-mode pairs to have Bohm scaling. A {rho}{sub *} stabilization rule for predicting the breakdown of gyro-Bohm scaling from simulations of a single discharge is presented.

  15. A frequency tunable, eight-channel correlation ECE system for electron temperature turbulence measurements on the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Sung, C.; Peebles, W. A.; Wannberg, C.; Rhodes, T. L.; Nguyen, X.; Lantsov, R.; Bardóczi, L.

    2016-11-01

    A new eight-channel correlation electron cyclotron emission diagnostic has recently been installed on the DIII-D tokamak to study both turbulent and coherent electron temperature fluctuations under various plasma conditions and locations. This unique system is designed to cover a broad range of operation space on DIII-D (1.6-2.1 T, detection frequency: 72-108 GHz) via four remotely selected local oscillators (80, 88, 96, and 104 GHz). Eight radial locations are measured simultaneously in a single discharge covering as much as half the minor radius. In this paper, we present design details of the quasi-optical system, the receiver, as well as representative data illustrating operation of the system.

  16. Validation of the thermal transport model used for ITER startup scenario predictions with DIII-D experimental data

    DOE PAGES

    Casper, T. A.; Meyer, W. H.; Jackson, G. L.; Luce, T. C.; Hyatt, A. W.; Humphreys, D. A.; Turco, F.

    2010-12-08

    We are exploring characteristics of ITER startup scenarios in similarity experiments conducted on the DIII-D Tokamak. In these experiments, we have validated scenarios for the ITER current ramp up to full current and developed methods to control the plasma parameters to achieve stability. Predictive simulations of ITER startup using 2D free-boundary equilibrium and 1D transport codes rely on accurate estimates of the electron and ion temperature profiles that determine the electrical conductivity and pressure profiles during the current rise. Here we present results of validation studies that apply the transport model used by the ITER team to DIII-D discharge evolutionmore » and comparisons with data from our similarity experiments.« less

  17. Rotation in the Plasma Flow Frame of Marginal Neoclassical Tearing Mode Islands in DIII-D and NSTX

    NASA Astrophysics Data System (ADS)

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

    2011-10-01

    Small island effects inhibit the pervasive occurrence of neoclassical tearing modes. m / n = 2 / 1 or 3/1 islands are reduced (by decreasing beta and thus the destabilizing bootstrap current density) to the self-stabilization size (``marginal point'') in DIII-D and NSTX. Non-zero island propagation in the plasma flow frame can produce a stabilizing polarization current provided it occurs in the direction of (but does not exceed) the ion diamagnetic drift. This ``polarization threshold'' naturally scales as a characteristic island size of several times the ion banana width, which is consistent with experiment. Comparison of measurements in DIII-D and NSTX discharges of similar cross-section (but different aspect ratio) with polarization current theory will be presented. Work supported in part by US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466, and DE-FG02-04ER54761.

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

    NASA Astrophysics Data System (ADS)

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

    2012-05-01

    New results suggest that changes observed in the intrinsic toroidal rotation influence the internal transport barrier (ITB) formation in the Alcator C-Mod tokamak [E. S. Marmar and Alcator C-Mod group, Fusion Sci. Technol. 51, 261 (2007)]. These arise when the resonance for ion cyclotron range of frequencies (ICRF) minority heating is positioned off-axis at or outside of the plasma half-radius. These ITBs form in a reactor relevant regime, without particle or momentum injection, with Ti ≈ Te, and with monotonic q profiles (qmin < 1). C-Mod H-mode plasmas exhibit strong intrinsic co-current rotation that increases with increasing stored energy without external drive. When the resonance position is moved off-axis, the rotation decreases in the center of the plasma resulting in a radial toroidal rotation profile with a central well which deepens and moves farther off-axis when the ICRF resonance location reaches the plasma half-radius. This profile results in strong E × B shear (>1.5 × 105 rad/s) in the region where the ITB foot is observed. Gyrokinetic analyses indicate that this spontaneous shearing rate is comparable to the linear ion temperature gradient (ITG) growth rate at the ITB location and is sufficient to reduce the turbulent particle and energy transport. New and detailed measurement of the ion temperature demonstrates that the radial profile flattens as the ICRF resonance position moves off axis, decreasing the drive for the ITG the instability as well. These results are the first evidence that intrinsic rotation can affect confinement in ITB plasmas.

  19. OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM

    SciTech Connect

    BURRELL,KH

    2002-11-01

    OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, the authors have made significant progress in developing the building blocks needed for AT operation: (1) the authors have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {le} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. They have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiation power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet

  20. Optimization of the Safety Factor Profile for High Noninductive Current Fraction Discharges in DIII-D

    SciTech Connect

    Ferron, J.R.; Holcomb, C T; Luce, T.C.; Politzer, P. A.; Turco, F.; White, A. E.; DeBoo, J. C.; Doyle, E. J.; Hyatt, A. W.; La Haye, R.; Murakami, Masanori; Petrie, T W; Petty, C C.; Rhodes, T. L.; Zeng, L.

    2011-01-01

    In order to assess the optimum q profile for discharges in DIII-D with 100% of the current driven noninductively (f(NI) = 1), the self-consistent response of the plasma profiles to changes in the q profile was studied in high f(NI), high beta(N) discharges through a scan of q(min) and q(95) at two values of beta(N). As expected, both the bootstrap current fraction, f(BS), and f(NI) increased with q(95). The temperature and density profiles were found to broaden as either q(min) or beta(N) is increased. A consequence is that f(BS) does not continue to increase at the highest values of q(min). A scaling function that depends on q(min), q(95), and the peaking factor for the thermal pressure was found to represent well the f(BS)/beta(N) inferred from the experimental profiles. The changes in the shapes of the density and temperature profiles as beta(N) is increased modify the bootstrap current density (J(BS)) profile from peaked close to the axis to relatively flat in the region between the axis and the H-mode pedestal. Therefore, significant externally driven current density in the region inside the H-mode pedestal is required in addition to J(BS) in order to match the profiles of the noninductive current density (J(NI)) to the desired total current density (J). In this experiment, the additional current density was provided mostly by neutral beam current drive with the neutral-beam-driven current fraction 40-90% of f(BS). The profiles of J(NI) and J were most similar at q(min) approximate to 1.35-1.65, q(95) approximate to 6.8, where f(BS) is also maximum, establishing this q profile as the optimal choice for f(NI) = 1 operation in DIII-D with the existing set of external current drive sources.

  1. Radiation asymmetries during disruptions on DIII-D caused by massive gas injection

    SciTech Connect

    Commaux, N.; Baylor, L. R.; Jernigan, T. C.; Foust, C. R.; Combs, S.; Meitner, S. J.; Hollmann, E. M.; Izzo, V. A.; Moyer, R. A.; Humphreys, D. A.; Wesley, J. C.; Eidietis, N. W.; Parks, P. B.; Lasnier, C. J.

    2014-10-15

    One of the major challenges that the ITER tokamak will have to face during its operations are disruptions. During the last few years, it has been proven that the global consequences of a disruption can be mitigated by the injection of large quantities of impurities. But one aspect that has been difficult to study was the possibility of local effects inside the torus during such injection that could damage a portion of the device despite the global heat losses and generated currents remaining below design parameter. 3D MHD simulations show that there is a potential for large toroidal asymmetries of the radiated power during impurity injection due to the interaction between the particle injection plume and a large n = 1 mode. Another aspect of 3D effects is the potential occurrence of Vertical Displacement Events (VDE), which could induce large poloidal heat load asymmetries. This potential deleterious effect of 3D phenomena has been studied on the DIII-D tokamak, thanks to the implementation of a multi-location massive gas injection (MGI) system as well as new diagnostic capabilities. This study showed the existence of a correlation between the location of the n = 1 mode and the local heat load on the plasma facing components but shows also that this effect is much smaller than anticipated (peaking factor of ∼1.1 vs 3-4 according to the simulations). There seems to be no observable heat load on the first wall of DIII-D at the location of the impurity injection port as well as no significant radiation asymmetries whether one or 2 valves are fired. This study enabled the first attempt of mitigation of a VDE using impurity injection at different poloidal locations. The results showed a more favorable heat deposition when the VDE is mitigated early (right at the onset) by impurity injection. No significant improvement of the heat load mitigation efficiency has been observed for late particle injection whether the injection is done “in the way” of the VDE

  2. Effects of ExB Velocity Shear and Magnetic Shear in the Formation of Core Transport Barriers in the DIII-D Tokamak

    SciTech Connect

    Burrell, K.H.; Greenfield, C.M.; Lao, L.L.; Staebler, G.M.; Austin, M.E.; Rice, B.W.; Stallard, B.W.

    1997-12-31

    Core transport barriers can be reliably formed in DIII-D by tailoring the evolution of the current density profile. This paper reports studies of the relative role of magnetic and ExB shear in creating core transport barriers in the DIII-D tokamak and considers the detailed dynamics of the barrier formation. The core barriers seen in DIII-D negative shear discharges form in a stepwise fashion during the initial current ramp. The reasons for the stepwise formation is not known; these steps do not correlate with integer values of q(O) or minimum q. The data from DIII-D is consistent with previous results that negative magnetic shear facilitates the formation of core transport barriers in the ion channel but is not necessary. However, strongly negative magnetic shear does allow formation of transport barriers in particle, electron thermal, ion thermal and angular momentum transport channels. Shots with strong negative magnetic shear have produced the steepest ion temperature and toroidal rotation profiles seen yet in DIII-D. In addition, the ExB shearing rates seen in these shots exceed the previous DIII-D record value by a factor of four.

  3. Effect of Energetic Trapped Particles Produced by ICRF Wave Heating on Sawtooth Instability in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Choi, M.; Chan, V. S.; Chu, M. S.; Jeon, Y. M.; Lao, L. L.; Li, G.; Pinsker, R. I.; Ren, Q.; Turnbull, A. D.

    2007-09-01

    We evaluate the accuracy of the Porcelli sawtooth model using more realistic numerical models from the ORBIT-RF and GATO codes in DIII-D fast wave heating experiments. Simulation results confirm that the fast wave-induced energetic trapped particles may stabilize the sawtooth instability. The crucial kinetic stabilizing contribution strongly depends on both the experimentally reconstructed magnetic shear at the q = 1 surface and the calculated poloidal beta of energetic trapped particles inside the q = 1 surface.

  4. Energetic Particle Effects Can Explain the Low Frequency of Alfvin Modes in the DIII-D Tokamak

    SciTech Connect

    Gorelenkov, N.N.; Heidbrink, W.W.

    2001-01-31

    During beam injection in the DIII-D tokamak, modes with lower frequencies than expected for toroidicity-induced Alfvin eigenmodes (TAE) are often observed. We present the analysis of one of these ''beta-induced Alfvin eigenmodes'' (BAE) with a high-n stability code HINST that includes the effect of the energetic ions on the mode frequency. It shows that the ''BAE'' could be the theoretically predicted resonant-TAE (RTAE), which is also called an energetic-particle mode (EPM).

  5. An Overview Of The Motional Stark Effect Diagnostic On DIII-D And Design Work For An ITER MSE

    SciTech Connect

    Holcomb, C T; Allen, S L; Makowski, M A; Jayakumar, R J; Gu, M F; Lerner, S; Morris, K L; Latkowski, J; Moller, J M; Meyer, W; Ellis, R; Geer, R; Behne, D; Chipman, R; Smith, P; McClain, S

    2007-09-20

    The advanced tokamak research program at DIII-D relies critically on the measurement of the current density profile. This was made possible by the development of a Motional Stark Effect (MSE) polarimeter that was first installed in 1992. Three major upgrades have since occurred, and improvements in our understanding of critical performance issues and calibration techniques are ongoing. In parallel with these improvements, we have drawn on our DIII-D experience to begin studies and design work for MSE on burning plasmas and ITER. This paper first reviews how Motional Stark Effect polarimetry (MSE) is used to determine the tokamak current profile. It uses the DIII-D MSE system as an example, and shows results from the latest upgrade that incorporates an array of channels from a new counter-Ip injected neutral beam. The various calibration techniques presently used are reviewed. High-leverage or unresolved issues affecting MSE performance and reliability in ITER are discussed. Next, we show a four-mirror collection optics design for the two ITER MSE views. Finally, we discuss measurements of the polarization properties of a few candidate mirrors for the ITER MSE.

  6. DIII-D research operations. Annual report to the US Department of Energy, October 1, 1994--September 30, 1995

    SciTech Connect

    1996-09-01

    The DIII-D research program funded by the U.S. Department of Energy (DOE) is aimed at developing the knowledge base for an economically and environmentally attractive energy source for the nation and the world. The DIII-D program mission is to advance fusion energy science understanding and predictive capability and improve the tokamak concept. The DIII-D scientific objectives are: (1) Advance understanding of fusion plasma physics and contribute to the physics base of ITER through extensive experiment and theory iteration in the following areas of fusion science - Magnetohydrodynamic (MHD) stability - Plasma turbulence and transport - Wave-particle interactions - Boundary physics plasma neutral interaction (2) Utilize scientific understanding in an integrated manner to show the tokamak potential to be - More compact by increasing plasma stability and confinement to increase the fusion power density ({Beta}{tau}) - Steady-state through disruption control, handling of divertor heat and particle loads and current drive (3) Acquire understanding and experience with environmentally attractive low activation material in an operating tokamak. This report contains the research conducted over the past year in search of these scientific objectives.

  7. DIII-D YPGRADE PROJECT FINAL REPORT FOR THE PERIOD OCTOBER 1, 1993 THROUGH MAY 31, 2003

    SciTech Connect

    STAMBAUGH, RD

    2003-06-01

    OAK-B135 Under DOE Contracts DE-AC03-89ER51114 and DE-AC03-99ER54463 to General Atomics (GA), three ''capital project'' upgrade projects were accomplished on DIII-D from FY93 to FY03 at a total GA cost of $27.2M. These projects included the Fast Wave Current Drive (FWCD) Upgrade ($8.2M), the Radiative Divertor Upgrade ($7.2M) and the Electron Cyclotron Heating (ECH) Upgrade ($11.8M). The ECH and FWCD upgrades provided DIII-D rf and microwave power for electron heating, driving plasma current, controlling the plasma current profile, controlling tearing mode instabilities, and modulated transport studies.The divertor provided adequate density and impurity control for high triangularity single null plasmas in the Advanced Tokamak (AT) Program and information for International Thermonuclear Experimental Reactor (ITER) divertor design. These upgrades provide the power and density control required to initiate the active control of advanced tokamak discharges, which is the key element in the DIII-D program.

  8. Analysis of shot-to-shot variability in post-disruption runaway electron currents for diverted DIII-D discharges

    NASA Astrophysics Data System (ADS)

    Izzo, V. A.; Humphreys, D. A.; Kornbluth, M.

    2012-09-01

    In DIII-D experiments, rapid termination by Ar-pellet injection sometimes produces a post-termination runaway electron (RE) current plateau, but this effect is highly non-reproducible on a shot-to-shot basis, particularly for diverted target plasmas. A set of DIII-D discharges is analyzed with two MHD codes to understand the relationship between the current profile of the target plasma and the amplitude of the RE current plateau. Using the linear stability code GATO, a correlation between the radial profile of the unstable n = 1 mode just after Ar-pellet injection and the observed appearance of an RE plateau is identified. Nonlinear NIMROD simulations with RE test-particle calculations directly predict RE confinement times during the disruption. With one exception, NIMROD predicts better RE confinement for shots in which higher RE currents were observed in DIII-D. But, the variation in confinement is primarily connected to the saturated n = 1 mode amplitude and not its radial profile. Still, both sets of analyses support the hypothesis that RE deconfinement by MHD fluctuations is a major factor in the shot-to-shot variability of RE plateaus, though additional factors such as seed current amplitude cannot be ruled out.

  9. Experiments on Helicon Excitation and Off-Axis Current Drive on DIII-D: Status and Plans

    NASA Astrophysics Data System (ADS)

    Pinsker, R. I.; Prater, R.; Moeller, C. P.; Degrassie, J. S.; Tooker, J. F.; Anderson, J. P.; Torreblanca, H.; Hansink, M.; Nagy, A.; Porkolab, M.

    2015-11-01

    Fast waves in the LHRF, also called ``whistlers'' or ``helicons,'' will be studied in experiments on the DIII-D tokamak beginning in autumn 2015. In the first stage, a 12-element traveling wave antenna (``comb-line'') is installed in the DIII-D vessel for operation at very low power (~ 0.1 kW) at 476 MHz, with a well-defined launched n| | spectrum peaked at 3.0. The goals of the low-power experiment include: (1) determining the efficiency with which the desired fast waves can be excited under a variety of plasma conditions in discharges relevant to the subsequent high-power current drive experiments and (2) proving that the radial and poloidal location at which the antenna will be mounted does not cause deleterious effects in the DIII-D discharges with high neutral beam power, and that the antenna is not damaged by fast ion losses, etc. Plans for 1 MW-level experiments with a single klystron beginning in FY17 are discussed. In addition to demonstrating off-axis current drive at an efficiency of ~ 60 kA/MW in high-performance plasmas, these experiments will explore non-linear aspects of wave excitation, propagation and absorption such as ponderomotive effects and parametric decay instabilities. Supported by US DOE DE-FC02-04ER54698, DE-AC02-09CH11466 and DE-FG02-94ER54084.

  10. Excitation of Alfven eigenmodes by low energy beam ions in the DIII-D and JET tokamaks

    SciTech Connect

    Nazikian, R.; Gorelenkov, N. N.; Budny, R. V.; Fu, G. Y.; Kramer, G. J.; Solomon, W. M.; White, R. B.; Alper, B.; Pinches, S. D.; Sharapov, S. E.; Borba, D.; Makowski, M. A.; Strait, E. J.; Van Zeeland, M. A.

    2008-05-15

    Core localized Alfven eigenmodes in DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] and Joint European Torus (JET) [P. H. Rebut and B. E. Keen, Fusion Technol. 11, 13 (1987)] plasmas are driven by deuterium neutral beam ions traveling well below the Alfven speed. Modes are observed in reverse magnetic shear discharges with deuterium ion velocities as low as 0.23 and 0.16 of the Alfven speed parallel to the magnetic field in DIII-D and JET plasmas, respectively. Ellipticity-induced Alfven eigenmodes in DIII-D and toroidicity-induced Alfven eigenmodes in JET are excited by deuterium ions traveling well below the fundamental passing ion resonance condition, indicating the role of high-order resonances in driving these modes. NOVA-K analysis reveals many high-order resonances as contributing to the mode drive at high central safety factor due to the correspondingly large poloidal orbit width and the decrease in the perpendicular scale length of the modes.

  11. Integrated magnetic and kinetic control of advanced tokamak plasmas on DIII-D based on data-driven models

    NASA Astrophysics Data System (ADS)

    Moreau, D.; Walker, M. L.; Ferron, J. R.; Liu, F.; Schuster, E.; Barton, J. E.; Boyer, M. D.; Burrell, K. H.; Flanagan, S. M.; Gohil, P.; Groebner, R. J.; Holcomb, C. T.; Humphreys, D. A.; Hyatt, A. W.; Johnson, R. D.; La Haye, R. J.; Lohr, J.; Luce, T. C.; Park, J. M.; Penaflor, B. G.; Shi, W.; Turco, F.; Wehner, W.; the ITPA-IOS Group members; experts

    2013-06-01

    The first real-time profile control experiments integrating magnetic and kinetic variables were performed on DIII-D in view of regulating and extrapolating advanced tokamak scenarios to steady-state devices and burning plasma experiments. Device-specific, control-oriented models were obtained from experimental data using a generic two-time-scale method that was validated on JET, JT-60U and DIII-D under the framework of the International Tokamak Physics Activity for Integrated Operation Scenarios (Moreau et al 2011 Nucl. Fusion 51 063009). On DIII-D, these data-driven models were used to synthesize integrated magnetic and kinetic profile controllers. The neutral beam injection (NBI), electron cyclotron current drive (ECCD) systems and ohmic coil provided the heating and current drive (H&CD) sources. The first control actuator was the plasma surface loop voltage (i.e. the ohmic coil), and the available beamlines and gyrotrons were grouped to form five additional H&CD actuators: co-current on-axis NBI, co-current off-axis NBI, counter-current NBI, balanced NBI and total ECCD power from all gyrotrons (with off-axis current deposition). Successful closed-loop experiments showing the control of (a) the poloidal flux profile, Ψ(x), (b) the poloidal flux profile together with the normalized pressure parameter, βN, and (c) the inverse of the safety factor profile, \\bar{\\iota}(x)=1/q(x) , are described.

  12. Sawtooth control using beam ions accelerated by fast waves in the DIII-D tokamak

    SciTech Connect

    Choi, M.; Turnbull, A. D.; Chan, V. S.; Chu, M. S.; Lao, L. L.; Jeon, Y. M.; Li, G.; Ren, Q.; Pinsker, R. I.

    2007-11-15

    The accuracy of the Porcelli sawtooth model is evaluated using realistic numerical calculations for a DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] experiment with neutral beam injection and fast wave heating. Simulation results confirm that beam ions accelerated by the fast waves play a crucial role in delaying the normal sawtooth crash and inducing giant sawteeth with large amplitude and long period. A single giant sawtooth period was analyzed in detail in an effort to evaluate the efficacy of the Porcelli model in quantitatively predicting a particular sawtooth crash by evaluating the model through the sawtooth period using equilibria reconstructed from the discharge data. The kinetic stabilizing contribution of fast trapped ions is found to depend strongly on both the experimentally reconstructed magnetic shear at the q=1 surface (s{sub 1}) and the calculated poloidal beta of trapped beam ions inside the q=1 surface. To within estimates of the error from the equilibrium reconstructions and the simulation fast ion particle statistics, the results are consistent with the observed sawtooth crash. The calculations indicate that the sawtooth crash is ultimately triggered by the resistive kink in the ion-kinetic regime after the stabilizing contribution from the fast ions is reduced due to an increase in s{sub 1} as the discharge evolves.

  13. DIII-D L-mode edge instabilities simulated in BOUT + +

    NASA Astrophysics Data System (ADS)

    Bass, Eric; Holland, Christopher; Xia, Tianyang; Xu, Xueqiao

    2015-11-01

    The well-known shortfall in predicted DIII-D tokamak L-mode edge transport is investigated using various one and two-fluid models implemented in the BOUT + + code. Five- and six-field models employed here contain the essential physics of peeling-ballooning modes, common in the H-mode edge, as well as drift-wave instabilities, predicted by various gyrokinetic and gyrofluid codes to dominate in the benchmark L-mode shot under investigation. We examine the extent to which instabilities driven at the separatrix can account for the systematic under-prediction of transport by these previous efforts. The focus is on the unstable linear spectrum for two domains: including and not including the separatrix. We compare one-fluid results (resistive ballooning unstable), two-fluid results (varying stability characteristics), and predictions from the gyrokinetic code GYRO (driftwave dominant), to establish which models make meaningful contact with the experiment in which regimes and locations. We comment on numerical pitfalls within BOUT + + revealed in the present study. Prepared by UCSD under Contract DE-FG02-06ER54871 and by LLNL under Contract DE-AC52-07NA27344.

  14. Investigation of physical processes limiting plasma density in H-mode on DIII-D

    SciTech Connect

    Maingi, R.; Mahdavi, M.A.; Jernigan, T.C.

    1996-12-01

    A series of experiments was conducted on the DIII-D tokamak to investigate the physical processes which limit density in high confinement mode (H-mode) discharges. The typical H-mode to low confinement mode (L-mode) transition limit at high density near the empirical Greenwald density limit was avoided by divertor pumping, which reduced divertor neutral pressure and prevented formation of a high density, intense radiation zone (MARFE) near the X-point. It was determined that the density decay time after pellet injection was independent of density relative to the Greenwald limit and increased non-linearly with the plasma current. Magnetohydrodynamic (MHD) activity in pellet-fueled plasmas was observed at all power levels, and often caused unacceptable confinement degradation, except when the neutral beam injected (NBI) power was {le} 3 MW. Formation of MARFEs on closed field lines was avoided with low safety factor (q) operation but was observed at high q, qualitatively consistent with theory. By using pellet fueling and optimizing discharge parameters to avoid each of these limits, an operational space was accessed in which density {approximately} 1.5 {times} Greenwald limit was achieved for 600 ms, and good H-mode confinement was maintained for 300 ms of the density flattop. More significantly, the density was successfully increased to the limit where a central radiative collapse was observed, the most fundamental density limit in tokamaks.

  15. Edge fluctuation measurements by phase contrast imaging on DIII-D

    NASA Astrophysics Data System (ADS)

    Coda, S.; Porkolab, M.

    1994-05-01

    A novel CO2 laser phase contrast imaging diagnostic has been developed for the DIII-D tokamak, where it is being employed to investigate density fluctuations at the outer edge of the plasma. This system generates 16-point, 1-D images of a 7.6 cm wide region in the radial direction, and is characterized by long wavelength (7.6 cm) and high frequency (100 MHz) capability, as well as excellent sensitivity (bar-n is approximately greater than 10(exp 9) cm(sup -3)). The effects of vertical line integration have been studied in detail, both analytically and numerically with actual flux surface geometries generated by the EFITD magnetic equilibrium code. It is shown that in the present configuration the measurement is mostly sensitive to radial wave vectors. Experimental results on fluctuation suppression at the L- to H-mode transition and on the L-mode wave number spectrum are discussed briefly. Finally, future plans for extending the measurement to the core of the plasma and for investigating externally launched fast waves are presented.

  16. Effect of ECH on Turbulent Fluctuations During ITER Baseline Discharges on DIII D

    NASA Astrophysics Data System (ADS)

    Marinoni, A.; Rost, J. C.; Porkolab, M.; Davis, E. M.; Pinsker, R. I.; Burrell, K. H.; DIII-D Team

    2015-11-01

    Recent experiments on the DIII-D tokamak simulating the ITER Baseline Scenario show that torque-free and spatially localized Electron Cyclotron Heating (ECH), compared to neutral beam heating, modifies flow shear and density fluctuations, resulting in a slightly worse confinement degradation than that given by the IPB98(y,2) scaling. After turning off ECH, the Phase Contrast Imaging diagnostic measures, on confinement time scales, a decrease in the intensity of fluctuations at frequencies lower than 200 kHz, consistent with the mean flow shear exceeding the maximum linear growth rate of ITG modes. In contrast, at higher frequencies the intensity of fluctuations increases promptly, due to ETG modes enhanced by the prompt increase of the electron temperature inverse scale length in the outer third of the minor radius. The latter effect is seen in preliminary non-linear gyro-kinetic simulations to generate a larger transient heat flux and an inward particle pinch. Work supported by the US DOE under DE-FG02-94ER54235 and DE-FC02-04ER54698.

  17. Effect of separatrix magnetic geometry on divertor behavior in DIII-D

    NASA Astrophysics Data System (ADS)

    Petrie, T. W.; Canik, J. M.; Lasnier, C. J.; Leonard, A. W.; Mahdavi, M. A.; Watkins, J. G.; Fenstermacher, M. E.; Ferron, J. R.; Groebner, R. J.; Hill, D. N.; Hyatt, A. W.; Holcomb, C. T.; Luce, T. C.; Moyer, R. A.; Stangeby, P. C.

    2013-07-01

    We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR ∝ [RTAR]2[L||]6/7 and TTAR ∝ [RTAR]-2[L||]-4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.

  18. 3D Magnetic Measurements of Kink and Locked Modes in DIII-D

    NASA Astrophysics Data System (ADS)

    King, J. D.; Strait, E. J.; Hanson, J. M.; Paz-Soldan, C.; Logan, N. C.; Lanctot, M. J.; Shiraki, D.

    2013-10-01

    The DIII-D magnetics diagnostic has been greatly expanded to fully characterize non-axisymmetric ``3D'' fields. Five poloidal locations now recover n <= 3 , while new HFS arrays provide poloidal spectral resolution of 7.4 cm. Initial measurements suggest externally driven kink structures deviate from MARS-F and IPEC models. These variations extend to the ideal regime, where toroidal agreement is observed. The plasma response to an n = 3 RMP increases monotonically as beta increases and q95 decreases, contrary to predictions of a screening to kink valley. Finally, the temporal evolution of the 3D eigenstructure of a slowly rotating (5 Hz) quasi-static, born locked, tearing mode provides the first evidence of an appreciable n = 2 error field, and an estimate of the phase for future correction. This new 3D capability will be used to understand and optimize control of RWMs, NTV torque, ELMs, and error field correction to extend stable tokamak operation. Work supported in part by the US Department of Energy under DE-AC05-00OR22725, DE-FC02-04ER54698, DE-FG02-95ER54309, DE-AC02-09CH11466, DE-FG02-04ER54761 and DE-AC52-07NA27344.

  19. OPTICAL DESIGN FOR Li BEAM ZEEMAN POLARIMETRY MEASUREMENTS ON DIII-D

    SciTech Connect

    T.N. CARLSTROM; D.M. THOMAS; G. BREWIS

    2002-08-01

    Measurements of the magnetic field pitch angle are obtained from the polarization characteristics of the {sigma} component of the Zeeman triplet of Li emission at 670.8 nm. A four element optical system images a horizontal Li beam on an array of 3 x 32, 1 mm diam. optical fibers, providing 32 spatial channels with 5 mm radial resolution in the plasma. Low Verdet constant glass is used for all optical elements near the DIII-D vessel to minimize the effects of Faraday rotation caused by stray magnetic fields. Before entering the fiber optics, the light passes through two crossed (45 degrees) photoelastic modulators (PEM) and a linear polarizer, which convert the various polarization states into an intensity modulated signal at the fundamental and 2nd harmonic of the PEM frequencies. For each spatial channel, light from a three fiber bundle is collimated and passes through a temperature tuned etalon (free spectral range, FSR = 0.3 nm; finesse, F = 5.7) in order to select only one {sigma} line of the triplet. The FSR is large enough to adequately cover the expected Zeeman triplet and small enough to achieve a low bandwidth (0.06 nm) at reasonably low F. A 1.0 nm wide interference filter is used to block all but 4-5 peaks of the etalon. The light is then focused on GaAs photo multiplier detectors. Details of the design and performance are presented.

  20. Regularization of soft-X-ray imaging in the DIII-D tokamak

    NASA Astrophysics Data System (ADS)

    Wingen, A.; Shafer, M. W.; Unterberg, E. A.; Hill, J. C.; Hillis, D. L.

    2015-05-01

    An image inversion scheme for the soft-X-ray imaging system (SXRIS) diagnostic at the DIII-D tokamak is developed to obtain the local soft-X-ray emission at a poloidal cross-section from the spatially line-integrated image taken by the SXRIS camera. The scheme uses the Tikhonov regularization method since the inversion problem is generally ill-posed. The regularization technique uses the generalized singular value decomposition (GSVD) to determine a solution that depends on a free regularization parameter. The latter has to be chosen carefully, and the so-called L-curve method to find the optimum regularization parameter is outlined. A representative test image is used to study the properties of the inversion scheme with respect to inversion accuracy, amount/strength of regularization, image noise and image resolution. The optimum inversion parameters are identified, while the L-curve method successfully computes the optimum regularization parameter. Noise is found to be the most limiting issue, but sufficient regularization is still possible at noise to signal ratios up to 10%-15%. Finally, the inversion scheme is applied to measured SXRIS data and the line-integrated SXRIS image is successfully inverted.

  1. Trapped gyro-Landau-fluid transport modeling of DIII-D hybrid discharges

    SciTech Connect

    Kinsey, J. E.; Staebler, G. M.; Petty, C. C.

    2010-12-15

    Previous work has summarized the physics and first results of benchmarking the trapped gyro-Landau-fluid (TGLF) model for turbulent transport driven by trapped ion and electron modes, ion and electron temperature gradient (ETG) modes, and electromagnetic kinetic ballooning modes including the effects of shaped geometry. Recently, an improved collision model was implemented which provides a more accurate fit to a transport database of nonlinear collisional GYRO[J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] simulations of long wavelength driftwave turbulence. The impact of the new collision model on TGLF modeling results was unknown. Using the improved TGLF model we obtain excellent agreement with the ion and electron temperature profiles from 30 DIII-D [A. Mahdavi and J. L. Luxon, Fusion Sci. Technol. 48, 2 (2005)] hybrid discharges. The transport results show that the electron energy transport tends to be dominated by short wavelength ETG modes in cases where the ion energy transport approaches neoclassical levels. The hybrid regime has significant energy confinement improvement from ExB velocity shear which is well predicted by TGLF. Weak magnetic shear and low safety factor are also shown to enhance the hybrid regime energy confinement. In high normalized {beta} hybrids, we find that finite {beta} effects noticably reduce the predicted electron energy transport and improve agreement with the measured electron temperature profiles.

  2. Lithium granule ablation and penetration during ELM pacing experiments at DIII-D

    DOE PAGES

    Lunsford, R.; Bortolon, A.; Roquemore, A. L.; Mansfield, D. K.; Nagy, A.; Maingi, R.; Parks, P. B.; Jackson, G.; Gilson, E.; Chrobak, C. P.

    2016-05-25

    At DIII-D, lithium granules were radially injected into the plasma at the outer midplane to trigger and pace edge localized modes (ELMs). Granules ranging in size from 300 to 1000 microns were horizontally launched into H-mode discharges with velocities near 100 m/s, and granule to granule injection frequencies less than 500 Hz. While the smaller granules were only successful in triggering ELMs approximately 20% of the time, the larger granules regularly demonstrated ELM triggering efficiencies of greater than 80%. A fast visible camera looking along the axis of injection observed the ablation of the lithium granules. We used the durationmore » of ablation as a benchmark for a neutral gas shielding calculation, and approximated the ablation rate and mass deposition location for the various size granules, using measured edge plasma profiles as inputs. In conclusion, this calculation suggests that the low triggering efficiency of the smaller granules is due to the inability of these granules to traverse the steep edge pressure gradient region and reach the top of the pedestal prior to full ablation.« less

  3. Protective Interlocks and Procedures on the DIII-D ECH System

    NASA Astrophysics Data System (ADS)

    Lohr, J.; Cengher, M.; Chen, X.; Gorelov, Y.; Ponce, D.; Prater, R.; Torrezan, A.

    2015-11-01

    Several new protective interlocks are being installed on the DIII-D ECH system to increase the safety margins for plasma operations at densities approaching cutoff. Inadvertent overdense operation has resulted in reflection of an rf beam back into one of the launchers causing extensive arcing and melt damage on one waveguide line. Therefore, protective steps have been taken to reduce the risk of such damage in the future. These include a density interlock generated by the plasma control system, enhanced video monitoring of the launchers, an ambient light monitor on each of the waveguide systems and versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, which are installed as the last miter bends in each waveguide line. Calculations of the rf beam trajectories in the plasmas are being performed using the TORAY ray tracing code with input from kinetic profile diagnostics, and strike points for refracted beams on the vacuum vessel are being calculated, which allows evaluating the risk of damage to sensitive diagnostics and hardware. Work supported by US DOE under DE-FC02-04ER54698.

  4. OEDGE Modeling of the DIII-D H-Mode 13CH4 Puffing Experiment

    SciTech Connect

    Elder, J D; McLean, A G; Stangeby, P C; Allen, S L; Boedo, J C; Bray, B D; Brooks, N H; Fenstermacher, M E; Groth, M; Leonard, A W; Reiter, D; Rudakov, D L; Wampler, W R; Watkins, J G; West, W P; Whyte, D G

    2006-06-01

    Use of carbon in tokamaks leads to a serious tritium retention issue due to co-deposition. To further investigate the processes involved, a detached ELMy H-mode (6.5 MW NBI) experiment was performed on DIII-D in which {sup 13}CH{sub 4} was puffed into the main vessel through the toroidally symmetric pumping plenum at the top of lower single-null discharges. Subsequently, the {sup 13}C content of tiles taken from the vessel wall was measured. The interpretive OEDGE code was used to model the results. The {sup 13}C deposition pattern could be reproduced, in general shape and magnitude, by assuming in the code the existence of a parallel flow and a radial pinch in the scrape-off layer. Parallel flows of Mach {approx} 0.3 toward the inner divertor and a radial pinch {approx}10 to 20 m/s (+ R-direction) were found to yield {sup 13}C deposition comparable to the experiment.

  5. Transport and performance in DIII-D discharges with weak or negative central magnetic shear

    SciTech Connect

    Greenfield, C.M.; Schissel, D.P.; Stallard, B.W.

    1996-12-01

    Discharges exhibiting the highest plasma energy and fusion reactivity yet realized in the DIII-D tokamak have been produced by combining the benefits of a hollow or weakly sheared central current profile with a high confinement (H-mode) edge. In these discharges, low power neutral beam injection heats the electrons during the initial current ramp, and {open_quotes}freezes in{close_quotes} a hollow or flat central current profile. When the neutral beam power is increased, formation of a region of reduced transport and highly peaked profiles in the core often results. Shortly before these plasmas would otherwise disrupt, a transition is triggered from the low (L-mode) to high (H-mode) confinement regimes, thereby broadening the pressure profile and avoiding the disruption. These plasmas continue to evolve until the high performance phase is terminated nondisruptively at much higher {beta}{sub T} (ratio of plasma pressure to toroidal magnetic field pressure) than would be attainable with peaked profiles and an L-mode edge. Transport analysis indicates that in this phase, the ion diffusivity is equivalent to that predicted by Chang-Hinton neoclassical theory over the entire plasma volume. This result is consistent with suppression of turbulence by locally enhanced E x B flow shear, and is supported by observations of reduced fluctuations in the plasma. Calculations of performance in these discharges extrapolated to a deuterium-tritium fuel mixture indicates that such plasmas could produce a DT fusion gain Q{sub DT} = 0.32.

  6. Investigation of torque generated by Test Blanket Module mock-up in DIII-D

    NASA Astrophysics Data System (ADS)

    Salmi, A.; Tala, T.; Lanctot, M.; Degrassie, J. S.; Paz-Soldan, C.; Logan, N.; Solomon, W. M.; Grierson, B. A.

    2015-11-01

    Experiments at DIII-D have investigated the scaling of Test Blanket Module (TBM) torque with plasma pressure and collisionality by performing dimensionless parameter scans. In each configuration, neutral beam torque modulation and TBM torque modulation were sequentially applied to allow experimental characterization of the TBM generated torque and the underlying transport. Calculations of the neoclassical toroidal viscosity (NTV) torque with PENT code of these plasmas find that TBM torque is strongly edge localized while the tentative experimental analysis indicates a more radially broad TBM torque profile. Both the experimental and PENT results will be elaborated and experimental TBM torque scaling with pressure and collisionality presented. Experimental validation of existing plasma response and NTV torque models is an important step toward understanding the impact of magnetic field ripple on plasma rotation, and for predicting the required compensation fields. Work supported by the US Department of Energy under DE-AC52-07NA27344, DE-FC02-04ER54698 and DE-AC02-09CH11466.

  7. DIII-D Equilibrium Reconstructions with New 3D Magnetic Probes

    NASA Astrophysics Data System (ADS)

    Lao, Lang; Strait, E. J.; Ferraro, N. M.; Ferron, J. R.; King, J. D.; Lee, X.; Meneghini, O.; Turnbull, A. D.; Huang, Y.; Qian, J. G.; Wingen, A.

    2015-11-01

    DIII-D equilibrium reconstructions with the recently installed new 3D magnetic diagnostic are presented. In addition to providing information to allow more accurate 2D reconstructions, the new 3D probes also provide useful information to guide computation of 3D perturbed equilibria. A new more comprehensive magnetic compensation has been implemented. Algorithms are being developed to allow EFIT to reconstruct 3D perturbed equilibria making use of the new 3D probes and plasma responses from 3D MHD codes such as GATO and M3D-C1. To improve the computation efficiency, all inactive probes in one of the toroidal planes in EFIT have been replaced with new probes from other planes. Other 3D efforts include testing of 3D reconstructions using V3FIT and a new 3D variational moment equilibrium code VMOM3D. Other EFIT developments include a GPU EFIT version and new safety factor and MSE-LS constraints. The accuracy and limitation of the new probes for 3D reconstructions will be discussed. Supported by US DOE under DE-FC02-04ER54698 and DE-FG02-95ER54309.

  8. ELM Behavior in High- βp EAST-Demonstration Plasmas on DIII-D

    NASA Astrophysics Data System (ADS)

    Li, G. Q.; Gong, X. Z.; Garofalo, A. M.; Lao, L. L.; Meneghini, O.; Snyder, P. B.; Ren, Q. L.; Ding, S. Y.; Guo, W. F.; Qian, J. P.; Wan, B. N.; Xu, G. S.; Holcomb, C. T.; Solomon, W. M.

    2015-11-01

    In the DIII-D high- βp EAST-demonstration experiment, for several similar discharges when the experimental parameters such as the toroidal magnetic field or ECH power are varied slightly, the changes in ELM frequency response are observed to be much larger. Kinetic EFIT equilibrium reconstructions for these discharges have been performed, which suggest that the ELM frequency changes are likely due to the variations of pedestal width, height, and edge current density. Kinetic profile analyses further indicate that the strong ITB that are located at large minor radii (rho=0.6 ~0.7) in these discharges are affecting the pedestal structure. The ITB could broaden the pedestal width and decrease the pedestal height, thus changing the ELM frequency and size. With the GATO and ELITE MHD codes, the linear growth rates and mode structures of these ELMs are analyzed. The impact of ITB on the ELMs behavior will be discussed. Work supported by China MOST under 2014GB106001 and 2015GB102001 and US DOE under DE-FC02-04ER54698 and DE-FG03-95ER54309.

  9. Equilibrium and Stability Characteristics of DIII-D Discharges with Low Edge Safety Factor

    NASA Astrophysics Data System (ADS)

    Jayakumar, R. J.; Garofalo, A. M.; Navratil, G. A.; Reimerdes, H.; Jensen, T. H.; La Haye, R. J.; Scoville, J. T.; Strait, E. J.; Turnbull, A. D.; Okabayashi, M.

    2002-11-01

    A low safety factor plasma has a good potential for reactor applications, since the fusion power and gain increase with reduction in safety factor. Experiments are being carried out in the DIII-D to investigate the equilibrium profiles and stability limits of such plasmas. Plasmas with flat q profiles q_min>1 and q_95<2.5 have been obtained with βN (>2.0) above no wall limits during current ramp and using resistive wall mode feedback. The discharge followed a current profile evolution predicted by the code CORSICA. The ideal wall and no wall beta limits for such discharges are being investigated using the codes GATO and DCON, for different current and pressure profiles. Simultaneously, the current profile and its evolution are being modeled for a variety of plasma cross sectional shapes and discharge formation scenarios to select an optimal scenario for the 2003 experimental run. The experimental results to date and modeling results will be presented.

  10. Robust Control of the Toroidal Rotation and Safety Factor Profiles in the DIII-D Tokamak

    NASA Astrophysics Data System (ADS)

    Shi, W.; Wehner, W.; Schuster, E.; Walker, M. L.; Humphreys, D. A.

    2011-10-01

    Because of the coupling between the different magnetic and kinetic plasma profiles, multi-input-multi-output (MIMO) model-based controllers are introduced to regulate the plasma rotation and safety factor profiles around particular target profiles. The approach is based on linear two-time-scale models identified from experimental data. The inputs are separated into slow and fast components by a low-pass filter that is incorporated into the overall plant. Then a singular value decomposition (SVD) of the plasma model is carried out to decouple the system and identify the most relevant control channels. Finally, the H∞ technique is used to determine a stabilizing feedback controller that minimizes the reference tracking errors and rejects disturbances with minimal control energy. Computer simulation results illustrate the performance of the robust profile controller, showing potential for improved performance. Experimental results in DIII-D are also reported. Supported by the NSF CAREER award program ECCS-0645086 and the US DOE under DE-FG02-09ER55064 and DE-FC02-04ER54698.

  11. Status and characterization of the lithium beam diagnostic on DIII-D

    SciTech Connect

    Stoschus, H.; Hudson, B.; Thomas, D. M.; Watkins, M.; Osborne, T. H.; Finkenthal, D. F.; Moyer, R. A.

    2013-08-15

    The 30 keV lithium beam diagnostic on DIII-D is suitable to measure both the radial electron density and poloidal magnetic field profiles in the pedestal. The refurbished system features a new setup to measure the Doppler shift allowing accurate alignment of the spectral filters. The injector has been optimized to generate a stable lithium neutral beam with a current of I= 15−20 mA and a diameter of 1.9 ± 0.1 cm measured by beam imaging. The typical temporal resolution is Δt= 1−10 ms and the radial resolution of ΔR= 5 mm is given by the optical setup. A new analysis technique based on fast Fourier transform avoids systematic error contributions from the digital lock-in analysis and accounts intrinsically for background light correction. Latest upgrades and a detailed characterization of the system are presented. Proof-of-principle measurements of the poloidal magnetic field with a statistical error of typically 2% show a fair agreement with the predictions modeled with the Grad-Shafranov equilibrium solver EFIT within 4%.

  12. Summary of recent studies of beam-driven BAE and chirping modes in DIII-D

    SciTech Connect

    Heidbrink, W.W.; Carolipio, E.M.; James, R.A.; Strait, E.J.

    1995-05-01

    In high beta DIII-D plasmas with intense neutral-beam injection, beta-induced Alfven eigenmodes (BAE modes) are observed. These instabilities can cause concentrated losses of >50% of the fast ions and thus are of concern for future devices. We have now observed BAE modes and resultant fast-ion loss in full-field (2.0 T) discharges where the ratio of parallel velocity to the Alfven speed is {upsilon}{sub parallel}/{upsilon}{sub A} {approx_equal}0.3. In a few discharges, we have also observed a new instability, a ``chirping`` mode. These modes have frequencies between 50 and 200 kHz that ``whistle`` down a factor of two in a single 2 ms burst. They occur in plasmas with relatively large values of fast-ion beta ({l_angle}{beta}{sub {integral}}{r_angle} {approx_gt}1%), Alfven speed ({upsilon}{sub {parallel}}/{upsilon}{sub A}{approx_gt}0.5), and plasma rotation ({integral}rot >20 kHz). In contrast to the usual Alfven modes, which are fluid modes of the background plasma, the chirping instabilities seem to be beam modes that are nearly stationary in the plasma frame.

  13. DIII-D Upgrade to Prepare the Basis for Steady-State Burning Plasmas

    NASA Astrophysics Data System (ADS)

    Buttery, R. J.; Guo, H. Y.; Taylor, T. S.; Wade, M. R.; Hill, D. N.

    2014-10-01

    Future steady-state burning plasma facilities will access new physics regimes and modes of plasma behavior. It is vital to prepare for this both experimentally using existing facilities, and theoretically in order to develop the tools to project to and optimize these devices. An upgrade to DIII-D is proposed to address the three critical aspects where research must go beyond what we can do now: (i) torque free electron heating to address the energy, particle and momentum transport mechanisms of burning plasmas using electron cyclotron (EC) heating and full power balanced neutral beams; (ii) off-axis heating and current drive to develop the path to true fusion steady state by reorienting neutral beams and deploying EC and helicon current drive; (iii) a new divertor with hot walls and reactor relevant materials to develop the basis for benign detached divertor operation compatible with wall materials and a high performance fusion core. These elements with modest incremental cost and enacted as a user facility for the whole US program will enable the US to lead on ITER and take a decision to proceed with a Fusion Nuclear Science Facility. Work supported by the US Department of Energy under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  14. PROGRESS IN QUANTIFYING THE EDGE PHYSICS OF H-MODE REGIME IN DIII-D

    SciTech Connect

    R.J. GROEBNER; D.R. BAKER; J.A. BOEDO; K.H. BURRELL; T.N. CARLSTROM; R.D. DERANIAN; E.J. DOYLE; J.R. FERRON; P. GOHIL; G.R. MOYER; C.L. RETTIG; T.L. RHODES; D.M. THOMAS; T.H. OSBORNE; W.P. WEST

    2000-10-01

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

  15. Electron heat transport in improved confinement discharges in DIII-D

    SciTech Connect

    Stallard, B.W.; Greenfield, C.M.; Staebler, G.M.

    1999-01-01

    In DIII-D tokamak plasmas with an internal transport barrier (ITB), the comparison of gyrokinetic linear stability (GKS) predictions with experiments in both low and strong negative magnetic shear plasmas provide improved understanding for electron thermal transport within the plasma. Within a limited region just inside the ITB, the electron temperature gradient (ETG) modes appear to control the electron temperature gradient and, consequently, the electron thermal transport. The increase in the electron temperature gradient with more strongly negative magnetic shear is consistent with the increase in the ETG mode marginal gradient. Closer to the magnetic axis the T{sub e} profile flattens and the ETG modes are predicted to be stable. With additional core electron heating, FIR scattering measurements near the axis show the presence of high k fluctuations (12 cm{sup {minus}1}), rotating in the electron diamagnetic drift direction. This turbulence could impact electron transport and possibly also ion transport. Thermal diffusivities for electrons, and to a lesser degree ions, increase. The ETG mode can exist at this wavenumber, but it is computed to be robustly stable near the axis. Consequently, in the plasmas the authors have examined, calculations of drift wave linear stability do not explain the observed transport near the axis in plasmas with or without additional electron heating, and there are probably other processes controlling transport in this region.

  16. Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D

    SciTech Connect

    Lasnier, C. J. Allen, S. L.; Ellis, R. E.; Fenstermacher, M. E.; McLean, A. G.; Meyer, W. H.; Morris, K.; Seppala, L. G.; Crabtree, K.; Van Zeeland, M. A.

    2014-11-15

    An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

  17. Current/Pressure Profile Effects on Tearing Mode Stability in DIII-D Hybrid Discharges

    NASA Astrophysics Data System (ADS)

    Kim, K.; Park, J. M.; Murakami, M.; La Haye, R. J.; Na, Yong-Su

    2015-11-01

    It is important to understand the onset threshold and the evolution of tearing modes (TMs) for developing a high-performance steady state fusion reactor. As initial and basic comparisons to determine TM onset, the measured plasma profiles (such as temperature, density, rotation) were compared with the calculated current profiles between a pair of discharges with/without n=1 mode based on the database for DIII-D hybrid plasmas. The profiles were not much different, but the details were analyzed to determine their characteristics, especially near the rational surface. The tearing stability index calculated from PEST3, Δ' tends to increase rapidly just before the n=1 mode onset for these cases. The modeled equilibrium with varying pressure or current profiles parametrically based on the reference discharge is reconstructed for checking the onset dependency on Δ' or neoclassical effects such as bootstrap current. Simulations of TMs with the modeled equilibrium using resistive MHD codes will also be presented and compared with experiments to determine the sensibility for predicting TM onset. Work supported by US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344.

  18. Modeling and Analysis for Tearing Mode Stability in DIII-D Hybrid Discharges

    NASA Astrophysics Data System (ADS)

    Kim, Kyungjin; Park, J. M.; Murakami, M.; La Haye, R. J.; Na, Yong-Su; DIII-D Team

    2014-10-01

    Plasma rotation in DIII-D hybrid scenario plasmas is found to change the stability of tearing modes (TMs) in a profound manner. It is important to understand the onset threshold and the evolution of TMs for developing a high-performance steady-state fusion reactor. The modified Rutherford equation (MRE) estimates the growth rate of an island and is used to analyze the TM stability. The change in TM stability was investigated in hybrid plasmas with various conditions including rotation, normalized beta, q profile, and so on. The measured island width is larger in low q95 cases and increased as plasma rotation was reduced. The island width calculated by MRE with TM stability index Δ' assumed from its poloidal mode number, -m/r, showed a good agreement during high rotation, but could not be matched to the experimental island width at lower rotation. Simulations of TMs using resistive MHD codes such as NIMROD and PEST3 will also be presented and compared with experiments to determine the possibility for predicting TM onset by Δ' calculation. Work supported in part by the US DOE under DE-AC05-00OR22725 and DE-FC02-04ER54698.

  19. Structure, Stability and ELM Dynamics of the H-Mode Pedestal in DIII-D

    SciTech Connect

    Fenstermacher, M E; Leonard, A W; Osborne, T H; Snyder, P B; Thomas, D M; Boedo, J A; Casper, T A; Colchin, R J; Groebner, R J; Groth, M; Kempenaars, M H; Loarte, A; Saibene, G; VanZeeland, M A; Zeng, L; Xu, X Q

    2004-10-13

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

  20. The importance of matched poloidal spectra to error field correction in DIII-D

    SciTech Connect

    Paz-Soldan, C. Lanctot, M. J.; Buttery, R. J.; La Haye, R. J.; Strait, E. J.; Logan, N. C.; Park, J.-K.; Solomon, W. M.; Shiraki, D.; Hanson, J. M.

    2014-07-15

    Optimal error field correction (EFC) is thought to be achieved when coupling to the least-stable “dominant” mode of the plasma is nulled at each toroidal mode number (n). The limit of this picture is tested in the DIII-D tokamak by applying superpositions of in- and ex-vessel coil set n = 1 fields calculated to be fully orthogonal to the n = 1 dominant mode. In co-rotating H-mode and low-density Ohmic scenarios, the plasma is found to be, respectively, 7× and 20× less sensitive to the orthogonal field as compared to the in-vessel coil set field. For the scenarios investigated, any geometry of EFC coil can thus recover a strong majority of the detrimental effect introduced by the n = 1 error field. Despite low sensitivity to the orthogonal field, its optimization in H-mode is shown to be consistent with minimizing the neoclassical toroidal viscosity torque and not the higher-order n = 1 mode coupling.