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Sample records for high confinement diii-d

  1. Regime of very high confinement in the boronized DIII-D tokamak

    International Nuclear Information System (INIS)

    Jackson, G.L.; Winter, J.; Taylor, T.S.; Burrell, K.H.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Hodapp, T.; Holtrop, K.; Lazarus, E.A.; Lao, L.L.; Lippmann, S.I.; Osborne, T.H.; Petrie, T.W.; Phillips, J.; James, R.; Schissel, D.P.; Strait, E.J.; Turnbull, A.D.; West, W.P.; DIII-D Team

    1991-01-01

    Following boronization, tokamak discharges in DIII-D have been obtained with confinement times up to a factor of 3.5 above the ITER89-P L-mode scaling and 1.8 times greater than the DIII-D/JET H-mode scaling relation. Very high confinement phases are characterized by relatively high central density with n e (0)∼1x10 20 m -3 , and central ion temperatures up to 13.6 keV at moderate plasma currents (1.6 MA) and heating powers (12.5--15.3 MW). These discharges exhibit a low fraction of radiated power, P≤25%, Z eff (0) close to unity, and lower impurity influxes than comparable DIII-D discharges before boronization

  2. Impurity confinement and transport in high confinement regimes without edge localized modes on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Grierson, B. A., E-mail: bgriers@pppl.gov; Nazikian, R. M.; Solomon, W. M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Burrell, K. H.; Garofalo, A. M.; Belli, E. A.; Staebler, G. M.; Evans, T. E.; Smith, S. P.; Chrobak, C. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Fenstermacher, M. E. [Lawerence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States); McKee, G. R. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53796 (United States); Orlov, D. M. [Center for Energy Research, University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Chrystal, C. [University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States)

    2015-05-15

    Impurity transport in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] is investigated in stationary high confinement (H-mode) regimes without edge localized modes (ELMs). In plasmas maintained by resonant magnetic perturbation (RMP), ELM-suppression, and QH-mode, the confinement time of fluorine (Z = 9) is equivalent to that in ELMing discharges with 40 Hz ELMs. For selected discharges with impurity injection, the impurity particle confinement time compared to the energy confinement time is in the range of τ{sub p}/τ{sub e}≈2−3. In QH-mode operation, the impurity confinement time is shown to be smaller for intense, coherent magnetic, and density fluctuations of the edge harmonic oscillation than weaker fluctuations. Transport coefficients are derived from the time evolution of the impurity density profile and compared to neoclassical and turbulent transport models NEO and TGLF. Neoclassical transport of fluorine is found to be small compared to the experimental values. In the ELMing and RMP ELM-suppressed plasma, the impurity transport is affected by the presence of tearing modes. For radii larger than the mode radius, the TGLF diffusion coefficient is smaller than the experimental value by a factor of 2–3, while the convective velocity is within error estimates. Low levels of diffusion are observed for radii smaller than the tearing mode radius. In the QH-mode plasma investigated, the TGLF diffusion coefficient is higher inside of ρ=0.4 and lower outside of 0.4 than the experiment, and the TGLF convective velocity is more negative by a factor of approximately 1.7.

  3. Comparison of a low- to high-confinement transition theory with experimental data from DIII-D.

    Science.gov (United States)

    Guzdar, P N; Kleva, R G; Groebner, R J; Gohil, P

    2002-12-23

    From our recent theory based on the generation of shear flow and field in finite beta plasmas, the criterion for bifurcation from low to high confinement mode yields a critical parameter proportional to T(e)/square root (L(n)), where T(e) is the electron temperature and L(n) is the density scale length. The predicted threshold shows very good agreement with edge measurements on discharges undergoing low-to-high transitions in DIII-D. The observed differences in the transitions with the reversal of the toroidal magnetic field are reconciled in terms of this critical parameter. The theory also provides an explanation for pellet injection H modes in DIII-D, thereby unifying unconnected methods for accomplishing the transition.

  4. Comparison of particle confinement in the high confinement mode plasmas with the edge localized mode of the Japan Atomic Energy Research Institute Tokamak-60 Upgrade and the DIII-D tokamak

    International Nuclear Information System (INIS)

    Takenaga, H.; Mahdavi, M.A.; Baker, D.R.

    2001-01-01

    Particle confinement was compared for the high confinement mode plasmas with the edge localized mode in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [S. Ishida, JT-60 Team, Nucl. Fusion 39, 1211 (1999)] and the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] considering separate confinement times for particles supplied by neutral beam injection (NBI) (center fueling) and by recycling and gas-puffing (edge fueling). Similar dependence on the NBI power was obtained in JT-60U and DIII-D. The particle confinement time for center fueling in DIII-D was smaller by a factor of 4 in the low density discharges and by a factor of 1.8 in the high density discharges than JT-60U scaling, respectively, suggesting the stronger dependence on the density in DIII-D. The particle confinement time for edge fueling in DIII-D was comparable with JT-60U scaling in the low density discharges. However, it decreased to a much smaller value in the high density discharges

  5. Suppression of large edge localized modes with edge resonant magnetic fields in high confinement DIII-D plasmas

    International Nuclear Information System (INIS)

    Thomas, P.R.; Becoulet, M.; Evans, T.E.; Osborne, T.H.; Groebner, R.J.; Jackson, G.L.; Haye, R.J. La; Schaffer, M.J.; West, W.P.; Moyer, R.A.; Rhodes, T.L.; Rudakov, D.L.; Watkins, J.G.; Boedo, J.A.; Doyle, E.J.; Wang, G.; Zeng, L.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J.; Finken, K.H.; Harris, J.H.; Pretty, D.G.; Masuzaki, S.; Ohyabu, N.; Reimerdes, H.; Wade, M.R.

    2005-01-01

    Large divertor heat pulses due to Type-I edge localized modes (ELMs) have been eliminated reproducibly in DIII-D with small dc currents driven in a simple magnetic perturbation coil. The current required to eliminate all but a few isolated Type-I ELMs, during a coil pulse, is less than 0.4% of plasma current. Modelling shows that the perturbation fields resonate with plasma flux surfaces across most of the pedestal region (0.9 ≤ N ≤ 1.0), when q95 = 3.7±0.2 creating small remnant magnetic islands surrounded by weakly stochastic field lines. The stored energy, N , H-mode quality factor and global energy confinement time are unaltered by the magnetic perturbation. At high collisionality (ν* ∼0.5-1), there is no obvious effect of the perturbation on the edge profiles and yet ELMs are suppressed, nearly completely, for up to 9τ E . At low collisionality (ν* <0.1), there is a density pump-out and complete ELM suppression, reminiscent of the DIIID QH- mode. Other differences, specifically in the resonance condition and the magnetic fluctuations, suggest that different mechanisms are at play in the different collisionality regimes. In addition to a description and interpretation of the DIIID data, the application of this method to ELM control on other machines, such as JET and ITER will be discussed. (author)

  6. QUIESCENT H-MODE, AN ELM-FREE HIGH-CONFINEMENT MODE ON DIII-D WITH POTENTIAL FOR STATIONARY STATE OPERATION

    International Nuclear Information System (INIS)

    WEST, WP; BURRELL, KH; DeGRASSIE, JS; DOYLE, EJ; GREENFIELD, CM; LASNIER, CJ; SNYDER, PB; ZENG, L.

    2003-01-01

    OAK-B135 The quiescent H-mode (QH-mode) is an ELM-free and stationary state mode of operation discovered on DIII-D. This mode achieves H-mode levels of confinement and pedestal pressure while maintaining constant density and radiated power. The elimination of edge localized modes (ELMs) and their large divertor loads while maintaining good confinement and good density control is of interest to next generation tokamaks. This paper reports on the correlations found between selected parameters in a QH-mode database developed from several hundred DIII-D counter injected discharges. Time traces of key plasma parameters from a QH-mode discharge are shown. On DIII-D the negative going plasma current (a) indicates that the beam injection direction is counter to the plasma current direction, a common feature of all QH-modes. The D α time behavior (c) shows that soon after high powered beam heating (b) is applied, the discharge makes a transition to ELMing H-mode, then the ELMs disappear, indicating the start of the QH period that lasts for the remainder of the high power beam heating (3.5 s). Previously published work showing density and temperature profiles indicates that long-pulse, high-triangularity QH discharges develop an internal transport barrier in combination with the QH edge barrier. These discharges are known as quiescent, double-barrier discharges (QDB). The H-factor (d) and stored energy (c) rise then saturate at a constant level and the measured axial and minimum safety factors remain above 1.0 for the entire QH duration. During QDB operation the performance of the plasma can be very good, with β N *H 89L product reaching 7 for > 10 energy confinement times. These discharges show promise that a stationary state can be achieved

  7. QUIESCENT H-MODE, AN ELM-FREE HIGH-CONFINEMENT MODE ON DIII-D WITH POTENTIAL FOR STATIONARY STATE OPERATION

    Energy Technology Data Exchange (ETDEWEB)

    WEST,WP; BURRELL,KH; deGRASSIE,JS; DOYLE,EJ; GREENFIELD,CM; LASNIER,CJ; SNYDER,PB; ZENG,L

    2003-08-01

    OAK-B135 The quiescent H-mode (QH-mode) is an ELM-free and stationary state mode of operation discovered on DIII-D. This mode achieves H-mode levels of confinement and pedestal pressure while maintaining constant density and radiated power. The elimination of edge localized modes (ELMs) and their large divertor loads while maintaining good confinement and good density control is of interest to next generation tokamaks. This paper reports on the correlations found between selected parameters in a QH-mode database developed from several hundred DIII-D counter injected discharges. Time traces of key plasma parameters from a QH-mode discharge are shown. On DIII-D the negative going plasma current (a) indicates that the beam injection direction is counter to the plasma current direction, a common feature of all QH-modes. The D{sub {alpha}} time behavior (c) shows that soon after high powered beam heating (b) is applied, the discharge makes a transition to ELMing H-mode, then the ELMs disappear, indicating the start of the QH period that lasts for the remainder of the high power beam heating (3.5 s). Previously published work showing density and temperature profiles indicates that long-pulse, high-triangularity QH discharges develop an internal transport barrier in combination with the QH edge barrier. These discharges are known as quiescent, double-barrier discharges (QDB). The H-factor (d) and stored energy (c) rise then saturate at a constant level and the measured axial and minimum safety factors remain above 1.0 for the entire QH duration. During QDB operation the performance of the plasma can be very good, with {beta}{sub N}*H{sub 89L} product reaching 7 for > 10 energy confinement times. These discharges show promise that a stationary state can be achieved.

  8. Suppression of large edge localized modes with a stochastic magnetic boundary in high confinement DIII-D plasmas

    International Nuclear Information System (INIS)

    Evans, T.E.; Moyer, R.A.; Watkins, J.G.

    2005-01-01

    Large sub-millisecond heat pulses due to Type-I ELMs have been eliminated reproducibly in DIII.D for periods approaching 7 energy confinement times with small dc currents driven in a simple magnetic perturbation coil. The current required to eliminate all but a few isolated Type-I ELM impulses during a coil pulse is less than 0.4% of plasma current. Based on vacuum magnetic field line modeling, the perturbation fields resonate strongly with plasma flux surfaces across most of the pedestal region (0.9 ≤ Ψ N ≤ 1.0) when q 95 = 3.7±0.2 creating small remnant magnetic islands surrounded by weakly stochastic field lines. The stored energy, β N , H-mode quality factor and global energy confinement time are unaltered. Although some isolated ELM-like events typically occur, long periods free of large Type-I ELMs (Δt > 4-6 τ E ) have been reproduced numerous times, on multiple experimental run days including cases matching the ITER scenario 2 flux surface shape. Since large Type-I ELM impulses represent a severe constraint on the survivability of the divertor target plates in future fusion devices such as ITER, a proven method of eliminating these impulses is critical for the development of tokamak reactors. Results presented in this paper indicate that non-axisymmetric edge magnetic perturbations could be a promising option for controlling ELMs in future tokamaks such as ITER. (author)

  9. Suppression of large edge localized modes in high confinement DIII-D plasmas with a stochastic magnetic boundary

    Energy Technology Data Exchange (ETDEWEB)

    Evans, T.E. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)]. E-mail: evans@fusion.gat.com; Moyer, R.A. [University of California at San Diego, La Jolla, CA 92093-0417 (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM 87185-1129 (United States); Thomas, P.R. [Association Euratom-CEA, CEA Cadarache, F-13108, St. Paul-lez-Durance (France); Osborne, T.H. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Boedo, J.A. [University of California at San Diego, La Jolla, CA 92093-0417 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laborabory, Livermore, CA 94550 (United States); Finken, K.H. [Forschungszentrum Juelich, Institute for Plasma Physics, D52425 Juelich (Germany); Groebner, R.J. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Groth, M. [Lawrence Livermore National Laborabory, Livermore, CA 94550 (United States); Harris, J. [Australian National University, Canberra (Australia); Jackson, G.L. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Haye, R.J. La [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laborabory, Livermore, CA 94550 (United States); Schaffer, M.J. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Wang, G. [University of California, Los Angeles, California (United States); Zeng, L. [University of California, Los Angeles, California (United States)

    2005-03-01

    Large 70 Hz Type-I edge localized modes (ELMs) are converted into small 130 Hz oscillations using edge resonant magnetic perturbations (RMPs) from a coil with currents 0.4% I {sub p} in double null DIII-D plasmas. When the RMP is properly phased with respect to the background field errors, all but a few isolated ELM-like events are suppressed. The impulsive pedestal energy loss {delta}E {sub ELM}/{delta}t {sup 1/2} to the scrape-of layer is reduced a factor of 20 relative to the Type-I ELMs and the core confinement is unaffected by the perturbation field. Significant changes in the properties of the ELMs are also observed when edge RMPs are applied to lower single null plasmas but the nature of these changes are much more complex. Both lower single null and double null plasmas are being studied to determine how ELM control techniques based on the application of edge RMPs can be expected to scale to future devices such as ITER.

  10. Suppression of large edge localized modes with a stochastic magnetic boundary in high confinement DIII-D plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, P.R.; Boedo, J.A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Evans, T.E.; Osborne, T.H.; Groebner, R.J.; La Haye, R.J.; Schaffer, M.J.; West, W.P. [General Atomic, San Diego (United States); Moyer, R.A.; Rudakov, D.L. [California Univ., San Diego, La Jolla, CA (United States); Watkins, J.G. [Sandia National Labs., Albuquerque, NM (United States); Fenstermacher, M.E.; Groth, M.; Lasnier, C.J. [Lawrence Livermore National Lab., CA (United States); Finken, K.H. [FZ-Julich Euratom Association, Julich (Germany); Harris, J.H.; Pretty, D.G. [Australian National Univ., Canberra, ACT (Australia); Masuzaki, S.; Ohyabu, N. [National Institute for Fusion Science, Gifu-ken (Japan); Rhodes, T.L.; Wang, G.; Zeng, L. [California Univ., Los Angeles, CA (United States); Reimerdes, H. [Columbia Univ., New York, NY (United States); Wade, M. [Oak Ridge National Lab., TN (United States)

    2004-07-01

    This work started from the point of view that a robust way of actively suppressing type I ELMs (edge localized modes) is both mandatory and practical in ITER. We report some considerable experimental success in exploring this proposition, using a magnetic perturbation generated by the I-coils in DIII-D. The hypothesis was tested that an external magnetic perturbation, whose toroidal and poloidal spectrum is in resonance with field lines at the edge, increases transport there, reduces the pressure gradient and, thereby, eliminates the drive for the MHD instabilities that cause ELMs. Holding the pressure gradient just below the threshold at which ELMs are triggered, should bring good fusion performance without the penalty of ELMs. (authors)

  11. SUPPESSION OF LARGE EDGE LOCALIZED MODES IN HIGH CONFINEMENT DIII-D PLASMAS WITH A STOCHASTIC MAGNETIC BOUNDARY

    International Nuclear Information System (INIS)

    EVANS, TE; MOYER, RA; THOMAS, PR; WATKINS, JG; OSBORNE, TH; BOEDO, JA; FENSTERMACHER, ME; FINKEN, KH; GROEBNER, RJ; GROTH, M; HARRIS, JH; LAHAYE, RJ; LASNIER, CJ; MASUZAKI, S; OHYABU, N; PRETTY, D; RHODES, TL; REIMERDES, H; RUDAKOV, DL; SCHAFFER, MJ; WANG, G; ZENG, L.

    2003-01-01

    OAK-B135 A stochastic magnetic boundary, produced by an externally applied edge resonant magnetic perturbation, is used to suppress large edge localized modes (ELMs) in high confinement (H-mode) plasmas. The resulting H-mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H-mode transport barrier is unaffected by the stochastic boundary. The core confinement of these discharges is unaffected, despite a three-fold drop in the toroidal rotation in the plasma core. These results demonstrate that stochastic boundaries are compatible with H-modes and may be attractive for ELM control in next-step burning fusion tokamaks

  12. Suppression of large edge-localized modes in high-confinement DIII-D plasmas with a stochastic magnetic boundary.

    Science.gov (United States)

    Evans, T E; Moyer, R A; Thomas, P R; Watkins, J G; Osborne, T H; Boedo, J A; Doyle, E J; Fenstermacher, M E; Finken, K H; Groebner, R J; Groth, M; Harris, J H; La Haye, R J; Lasnier, C J; Masuzaki, S; Ohyabu, N; Pretty, D G; Rhodes, T L; Reimerdes, H; Rudakov, D L; Schaffer, M J; Wang, G; Zeng, L

    2004-06-11

    A stochastic magnetic boundary, produced by an applied edge resonant magnetic perturbation, is used to suppress most large edge-localized modes (ELMs) in high confinement (H-mode) plasmas. The resulting H mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H mode transport barrier and core confinement are unaffected by the stochastic boundary, despite a threefold drop in the toroidal rotation. These results demonstrate that stochastic boundaries are compatible with H modes and may be attractive for ELM control in next-step fusion tokamaks.

  13. Discovery of stationary operation of quiescent H-mode plasmas with net-zero neutral beam injection torque and high energy confinement on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Burrell, K. H.; Chen, X.; Garofalo, A. M.; Groebner, R. J.; Muscatello, C. M.; Osborne, T. H.; Petty, C. C.; Snyder, P. B. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Barada, K.; Rhodes, T. L.; Zeng, L. [University of California-Los Angeles, Los Angeles, California 90024 (United States); Solomon, W. M. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Yan, Z. [University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States)

    2016-05-15

    Recent experiments in DIII-D [J. L. Luxon et al., in Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] have led to the discovery of a means of modifying edge turbulence to achieve stationary, high confinement operation without Edge Localized Mode (ELM) instabilities and with no net external torque input. Eliminating the ELM-induced heat bursts and controlling plasma stability at low rotation represent two of the great challenges for fusion energy. By exploiting edge turbulence in a novel manner, we achieved excellent tokamak performance, well above the H{sub 98y2} international tokamak energy confinement scaling (H{sub 98y2} = 1.25), thus meeting an additional confinement challenge that is usually difficult at low torque. The new regime is triggered in double null plasmas by ramping the injected torque to zero and then maintaining it there. This lowers E × B rotation shear in the plasma edge, allowing low-k, broadband, electromagnetic turbulence to increase. In the H-mode edge, a narrow transport barrier usually grows until MHD instability (a peeling ballooning mode) leads to the ELM heat burst. However, the increased turbulence reduces the pressure gradient, allowing the development of a broader and thus higher transport barrier. A 60% increase in pedestal pressure and 40% increase in energy confinement result. An increase in the E × B shearing rate inside of the edge pedestal is a key factor in the confinement increase. Strong double-null plasma shaping raises the threshold for the ELM instability, allowing the plasma to reach a transport-limited state near but below the explosive ELM stability boundary. The resulting plasmas have burning-plasma-relevant β{sub N} = 1.6–1.8 and run without the need for extra torque from 3D magnetic fields. To date, stationary conditions have been produced for 2 s or 12 energy confinement times, limited only by external hardware constraints

  14. ELMs IN DIII-D HIGH PERFORMANCE DISCHARGES

    International Nuclear Information System (INIS)

    TURNBULL, A.D; LAO, L.L; OSBORNE, T.H; SAUTER, O; STRAIT, E.J; TAYLOR, T.S; CHU, M.S; FERRON, J.R; GREENFIELD, C.M; LEONARD, A.W; MILLER, R.L; SNYDER, P.B; WILSON, H.R; ZOHM, H

    2003-01-01

    A new understanding of edge localized modes (ELMs) in tokamak discharges is emerging [P.B. Snyder, et al., Phys. Plasmas, 9, 2037 (2002)], in which the ELM is an essentially ideal magnetohydrodynamic (MHD) instability and the ELM severity is determined by the radial width of the linearly unstable MHD kink modes. A detailed, comparative study of the penetration into the core of the respective linear instabilities in a standard DIII-D ELMing, high confinement mode (H-mode) discharge, with that for two relatively high performance discharges shows that these are also encompassed within the framework of the new model. These instabilities represent the key, limiting factor in extending the high performance of these discharges. In the standard ELMing H-mode, the MHD instabilities are highly localized in the outer few percent flux surfaces and the ELM is benign, causing only a small temporary drop in the energy confinement. In contrast, for both a very high confinement mode (VH-mode) and an H-mode with a broad internal transport barrier (ITB) extending over the entire core and coalesced with the edge transport barrier, the linearly unstable modes penetrate well into the mid radius and the corresponding consequences for global confinement are significantly more severe. The ELM accordingly results in an irreversible loss of the high performance

  15. Optimized profiles for improved confinement and stability in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Taylor, T.S.; St. John, H.; Turnbull, A.D.

    1995-02-01

    Simultaneous achievement of high energy confinement, τ E , and high plasma beta, β, leads to an economically attractive compact tokamak fusion reactor. High confinement enhancement, H = τ E /τ E-ITER89P = 4, and high normalized beta β N = β/(I/aB) = 6%-m-T/MA, have been obtained in DIII-D experimental discharges. These improved confinement and/or improved stability limits are observed in several DIII-D high performance operational regimes: VH-mode, high ell i H-mode, second stable core, and high beta poloidal. The authors have identified several important features of the improved performance in these discharges: details of the plasma shape, toroidal rotation or ExB flow profile, q profile and current density profile, and pressure profile. From the improved physics understanding of these enhanced performance regimes, they have developed operational scenarios which maintain the essential features of the improved confinement and which increase the stability limits using localized current profile control. The stability limit is increased by modifying the interior safety factor profile to be nonmonotonic with high central q, while maintaining the edge current density consistent with the improved transport regimes and the high edge bootstrap current. They have calculated high beta equilibria with β N = 6.5, stable to ideal n = 1 kinks and stable to ideal ballooning modes. The safety factor at the 95% flux surface is 6, the central q value is 3.9 and the minimum in q is 2.6. The current density profile is maintained by the natural profile of the bootstrap current, and a modest amount of electron cyclotron current drive

  16. HIGH PERFORMANCE STATIONARY DISCHARGES IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    Luce, T.C.; Wade, M.R.; Ferron, J.R.; Politzer, P.A.; Hyatt, A.W.; Sips, A.C.C.; Murakami, M.

    2003-01-01

    Recent experiments in the DIII-D tokamak [J.L. Luxon, Nucl. Fusion 42,614 (2002)] have demonstrated high β with good confinement quality under stationary conditions. Two classes of stationary discharges are observed--low q 95 discharges with sawteeth and higher q 95 without sawteeth. The discharges are deemed stationary when the plasma conditions are maintained for times greater than the current profile relaxation time. In both cases the normalized fusion performance (β N H 89P /q 95 2 ) reaches or exceeds the value of this parameter projected for Q fus = 10 in the International Thermonuclear Experimental Reactor (ITER) design [R. Aymar, et al., Plasma Phys. Control. Fusion 44, 519 (2002)]. The presence of sawteeth reduces the maximum achievable normalized β, while confinement quality (confinement time relative to scalings) is largely independent of q 95 . Even with the reduced β limit, the normalized fusion performance maximizes at the lowest q 95 . Projections to burning plasma conditions are discussed, including the methodology of the projection and the key physics issues which still require investigation

  17. Stability and control of resistive wall modes in high beta, low rotation DIII-D plasmas

    International Nuclear Information System (INIS)

    Garofalo, A.M.; Jackson, G.L.; Haye, R.J. La; Okabayashi, M.; Reimerdes, H.; Strait, E.J.; Ferron, J.R.; Groebner, R.J.; In, Y.; Lanctot, M.J.; Matsunaga, G.; Navratil, G.A.; Solomon, W.M.; Takahashi, H.; Takechi, M.; Turnbull, A.D.

    2007-01-01

    Recent highDIII-D (Luxon J.L. 2002 Nucl. Fusion 42 64) experiments with the new capability of balanced neutral beam injection show that the resistive wall mode (RWM) remains stable when the plasma rotation is lowered to a fraction of a per cent of the Alfven frequency by reducing the injection of angular momentum in discharges with minimized magnetic field errors. Previous DIII-D experiments yielded a high plasma rotation threshold (of order a few per cent of the Alfven frequency) for RWM stabilization when resonant magnetic braking was applied to lower the plasma rotation. We propose that the previously observed rotation threshold can be explained as the entrance into a forbidden band of rotation that results from torque balance including the resonant field amplification by the stable RWM. Resonant braking can also occur naturally in a plasma subject to magnetic instabilities with a zero frequency component, such as edge localized modes. In DIII-D, robust RWM stabilization can be achieved using simultaneous feedback control of the two sets of non-axisymmetric coils. Slow feedback control of the external coils is used for dynamic error field correction; fast feedback control of the internal non-axisymmetric coils provides RWM stabilization during transient periods of low rotation. This method of active control of the n = 1 RWM has opened access to new regimes of high performance in DIII-D. Very high plasma pressure combined with elevated q min for high bootstrap current fraction, and internal transport barriers for high energy confinement, are sustained for almost 2 s, or 10 energy confinement times, suggesting a possible path to high fusion performance, steady-state tokamak scenarios

  18. Stability in high gain plasmas in DIII-D

    International Nuclear Information System (INIS)

    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

  19. DIII-D ICRF high voltage power supply regulator upgrade

    International Nuclear Information System (INIS)

    Cary, W.P.; Burley, B.L.; Grosnickle, W.H.

    1997-11-01

    For reliable operation and component protection, of the 2 MW 30--120 MHz ICRF Amplifier System on DIII-D, it is desirable for the amplifier to respond to high VSWR conditions as rapidly as possible. This requires a rapid change in power which also means a rapid change in the high voltage power supply current demands. An analysis of the power supply's regulator dynamics was needed to verify its expected operation during such conditions. Based on this information it was found that a new regulator with a larger dynamic range and some anticipation capability would be required. This paper will discuss the system requirements, the as-delivered regulator performance, and the improved performance after installation of the new regulator system. It will also be shown how this improvement has made the amplifier perform at higher power levels more reliably

  20. Wall stabilization of high beta plasmas in DIII-D

    International Nuclear Information System (INIS)

    Taylor, T.S.; Strait, E.J.; Lao, L.L.; Turnbull, A.D.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Groebner, R.J.; La Haye, R.J.; Mauel, M.

    1995-02-01

    Detailed analysis of recent high beta discharges in the DIII-D tokamak demonstrates that the resistive vacuum vessel can provide stabilization of low n magnetohydrodynamic (MHD) modes. The experimental beta values reaching up to β T = 12.6% are more than 30% larger than the maximum stable beta calculated with no wall stabilization. Plasma rotation is essential for stabilization. When the plasma rotation slows sufficiently, unstable modes with the characteristics of the predicted open-quotes resistive wallclose quotes mode are observed. Through slowing of the plasma rotation between the q = 2 and q = 3 surfaces with the application of a non-axisymmetric field, the authors have determined that the rotation at the outer rational surfaces is most important, and that the critical rotation frequency is of the order of Ω/2π = 1 kHz

  1. DIII-D research operations

    International Nuclear Information System (INIS)

    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 ampersand D outputs to aid the Tokamak Physics Experiment (TPX) and the International Thermonuclear Experimental Reactor (ITER). Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY93 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics. The major goals of the Divertor and Boundary Physics studies are the control of impurities, efficient heat removal and understanding the strong role that the edge plasma plays in the global energy confinement of the plasma. The advanced tokamak studies initiated the investigation into new techniques for improving energy confinement, controlling particle fueling and increasing plasma beta. The major goal of the Tokamak Physics Studies is the understanding of energy and particle transport in a reactor relevant plasma

  2. Results from core-edge experiments in high Power, high performance plasmas on DIII-D

    Directory of Open Access Journals (Sweden)

    T.W. Petrie

    2017-08-01

    Full Text Available Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5 due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot” but also that heating near the slot opening is a significant source for impurity contamination of the core.

  3. Long pulse high performance discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.; Wade, M.R.; Politzer, P.A.

    2001-01-01

    Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 , indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β min >1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with β N H 89 ∼ 7 for up to 6.3 s or ∼34τ E . These discharges appear to have stationary current profiles with q min ∼1.05, in agreement with the current profile relaxation time ∼1.8 s. (author)

  4. Recent DIII-D high power heating and current drive experiments

    International Nuclear Information System (INIS)

    Simonen, T.C.; Jackson, G.L.; Lazarus, E.A.; Mahdavi, M.A.; Petrie, T.W.; Politzer, P.A.; Taylor, T.S.

    1995-01-01

    This paper describes recent DIII-D high power heating and current drive experiments. Described are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal β. ((orig.))

  5. Recent DIII-D high power heating and current drive experiments

    International Nuclear Information System (INIS)

    Simonen, T.C.; Jackson, G.L.; Mahdavi, M.A.; Petrie, T.W.; Politzer, P.A.; Taylor, T.S.; Lazarus, E.A.

    1994-02-01

    This paper describes recent DIII-D high power heating and current drive experiments. Describes are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal beta

  6. Recent DIII-D high power heating and current drive experiments

    Energy Technology Data Exchange (ETDEWEB)

    Simonen, T.C. [General Atomics, San Diego, CA (United States); Jackson, G.L. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Politzer, P.A. [General Atomics, San Diego, CA (United States); Taylor, T.S. [General Atomics, San Diego, CA (United States); DIII-D Team

    1995-01-01

    This paper describes recent DIII-D high power heating and current drive experiments. Described are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal {beta}. ((orig.)).

  7. Improved energy confinement with neon injection in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Staebler, G.M.; Jackson, G.L.; West, W.P. Groebner, R.J.; Schaffer, M.J.; Allen, S.L.; Whyte, D.G.

    1997-06-01

    In this paper the authors will report the first direct measurements of the fully stripped neon 10 + density profile in a plasma with enhanced energy confinement due to neon injection. This is made with a calibrated charge exchange recombination (CER) system. It is found that the neon 10 + density is peaked like the electron density with a slightly higher concentration towards the edge. The good news is that the neon 10 + fraction is less than 1% (normalized to the electron density). The radial electric field can also be computed from the CER measurements on DIII-D. The shear in the E x B velocity is found to exceed the maximum growth rate of the ion temperature gradient (ITG) mode over part of the profile, a condition for the suppression of turbulent transport. This agrees with the reduced power balance thermal diffusivities near the magnetic axis

  8. Comparison of H-mode pedestals in different confinement regimes in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Groebner, R J [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Leonard, A W [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Luce, T C [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Fenstermacher, M E [Lawrence Livermore National Laboratory, Livermore, California (United States); Jackson, G L [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Osborne, T H [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Thomas, D M [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States); Wade, M R [General Atomics, PO Box 85608, San Diego, California, 92186-5608 (United States)

    2006-05-15

    A survey of global performance parameters and their correlation with pedestal parameters is performed for standard H-mode, QH-mode and the enhanced confinement regimes of VH-mode, hybrid and advanced tokamak in the DIII-D tokamak. This study shows that there is a trend for global confinement quality or global beta to increase as the pedestal electron pressure or beta increases. However, there are also improvements in core confinement and beta, observed at fixed pedestal pressure or beta, which indicate that factors other than pedestal parameters also contribute to the best core performance. Several other pedestal structure parameters are found to be similar among these regimes. The scale lengths for electron pressure in the pedestal are in the range 0.8-1.6 cm at the outer midplane, most {eta}{sub e} values are in the range 1-3 in the middle of the T{sub e} pedestal and the T{sub e} and n{sub e} pedestals tend to penetrate the same distance into the plasma.

  9. Long-pulse high-performance discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.; Wade, M.R.; Politzer, P.A.

    2001-01-01

    Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance ∼10 has been sustained for >5τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and β≤5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with β N H 89 ∼7 for up to 6.3 s or ∼34 τ E . These discharges appear to be in resistive equilibrium with q min ∼1.05, in agreement with the current profile relaxation time of 1.8 s. (author)

  10. LONG-PULSE, HIGH-PERFORMANCE DISCHARGES IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    T.C. LUCE; M.R. WADE; P.A. POLITZER; S.L. ALLEN; M E. AUSTIN; D.R. BAKER; B.D. BRAY; D.P. BRENNAN; K.H. BURRELL; T.A. CASPER; M.S. CHU; J.D. De BOO; E.J. DOYLE; J.R. FERRON; A.M. GAROFALO; P.GOHIL; I.A. GORELOV; C.M. GREENFIELD; R.J. GROEBNER; W.W. HEIBRINK; C.-L. HSIEH; A.W. HYATT; R.JAYAKUMAR; J.E.KINSEY; R.J. LA HAYE; L.L. LAO; C.J. LASNIER; E.A. LAZARUS; A.W. LEONARD; Y.R. LIN-LIU; J. LOHR; M.A. MAKOWSKI; M. MURAKAMI; C.C. PETTY; R.I. PINSKER; R. PRATER; C.L. RETTIG; T.L. RHODES; B.W. RICE; E.J. STRAIT; T.S. TAYLOR; D.M. THOMAS; A.D. TURNBULL; J.G. WATKINS; W.P.WEST; K.-L. WONG

    2000-01-01

    Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance ∼10 has been sustained for >5 τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and β ∼(le) 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with β N H 89 ∼ 7 for up to 6.3 s or ∼ 34 τ E . These discharges appear to be in resistive equilibrium with q min ∼ 1.05, in agreement with the current profile relaxation time of 1.8 s

  11. Helium transport and exhaust studies in enhanced confinement regimes in DIII-D

    International Nuclear Information System (INIS)

    Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Mahdavi, M.A.; Maingi, R.; West, W.P.; Burrell, K.H.; Finkenthal, D.F.; Gohil, P.; Groebner, R.J.

    1995-02-01

    A better understanding of helium transport in the plasma core and edge in enhanced confinement regimes is now emerging from recent experimental studies on DIII-D. Overall, the results are encouraging. Significant helium exhaust (τ* He /τ E ∼ 11) has been obtained in a diverted, ELMing H-mode plasma simultaneous with a central source of helium. Detailed analysis of the helium profile evolution indicates that the exhaust rate is limited by the exhaust efficiency of the pump (∼5%) and not by the intrinsic helium transport properties of the plasma. Perturbative helium transport studies using gas puffing have shown that D He /X eff ∼1 in all confinement regimes studied to date (including H-mode and VH-mode). Furthermore, there is no evidence of preferential accumulation of helium in any of these regimes. However, measurements in the core and pumping plenum show a significant dilution of helium as it flows from the plasma core to the pumping plenum. Such dilution could be the limiting factor in the overall removal rate of helium in a reactor system

  12. Progress toward fully noninductive, high beta conditions in DIII-D

    International Nuclear Information System (INIS)

    Murakami, M.; Wade, M.R.; Greenfield, C.M.; Luce, T.C.; Ferron, J.R.; St John, H.E.; DeBoo, J.C.; Osborne, T.H.; Petty, C.C.; Politzer, P.A.; Burrell, K.H.; Gohil, P.; Gorelov, I.A.; Groebner, R.J.; Hyatt, A.W.; Kajiwara, K.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lohr, J.

    2006-01-01

    The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, β T =3.6%, normalized beta, β N =3.5, and confinement factor, H 89 =2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1 s or half of a current relaxation time (τ CR ). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2 s duration (1τ CR ) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have

  13. Experiments at high elongations in DIII-D

    International Nuclear Information System (INIS)

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

    1990-06-01

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

  14. Recent DIII-D results

    International Nuclear Information System (INIS)

    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

  15. Demonstration of high performance negative central magnetic shear discharges on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Rice, B.W.; Burrell, K.H.; Lao, L.L.

    1996-01-01

    Reliable operation of discharges with negative central magnetic shear has led to significant increases in plasma performance and reactivity in both low confinement, L-mode, and high confinement, H-mode, regimes in the DIII-D tokamak. Using neutral beam injection early in the initial current ramp, a large range of negative shear discharges have been produced with durations lasting up to 3.2 s. The total non- inductive current (beam plus bootstrap) ranges from 50% to 80% in these discharges. In the region of shear reversal, significant peaking of the toroidal rotation [f φ ∼ 30-60 kHz] and ion temperature [T i (0) ∼ 15-22 keV] profiles are observed. In high power discharges with an L-mode edge, peaked density profiles are also observed. Confinement enhancement factors up to H ≡ τ E /τ ITER-89P ∼ 2.5 with an L-mode edge, and H ∼ 3.3 in an Edge Localized Mode (ELM)-free H-mode, are obtained. Transport analysis shows both ion thermal diffusivity and particle diffusivity to be near or below standard neoclassical values in the core. Large pressure peaking in L- mode leads to high disruptivity with Β N ≡ Β T /(I/aB) ≤ 2.3, while broader pressure profiles in H- mode gives low disruptivity with Β N ≤ 4.2

  16. Experiments at high elongations in DIII-D

    International Nuclear Information System (INIS)

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

    1990-01-01

    In this paper we discuss the limitation to elongation observed in D-shaped plasmas in the DIII-D tokamak. We find that as the triangularity is increased and ell i is decreased that the n = O mode takes on an increasingly non-rigid character. Our analysis shows two aspects of the behavior: first, an increasing variation of the m/n = 1/0 component across flux surfaces and second, an increase in the relative amplitude of a m/n = 3/0 component which couples to the m/n = 1/0 component and further destabilizes the mode. In previous work we have reported on study of vertical control and the implementation of those results on DIII-D. In that study we used a single filament, with properties consistent with the radial force balance, to represent the plasma and employed an eigenmode description of the passive shell in order to allow time-ordering of the problem. The most important result of this study was that the active control coil must be positioned in the poloidal plane so as to minimize its interaction with the stabilizing shell currents. As a consequence of plasma toroidicity, these currents flow primarily in the outboard regions of the shell. Thus, control coils on the inboard side of the shell, near the midplane, are required. With such a spatial arrangement we can have radial fields from the active coil penetrating the shell on a time scale faster than the decay of the stabilizing shell currents. In accordance with these model calculations the control system for DIII-D tokamak has been modified resulting in operation to within a few percent of the ideal MHD limit for axisymmetric stability. In this work we refer to the ideal MHD limit as that of the plasma-shell system. The ideal limit can actually be reduced by a poor choice of the active control coils, however that is not the case for work discussed here. 7 refs., 6 figs

  17. The role of the radial electric field in confinement and transport in H-mode and VH-mode discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Gohil, P.; Burrell, K.H.; Groebner, R.J.; Osborne, T.H.; Doyle, E.J.; Rettig, C.L.

    1993-08-01

    Measurements of the radial electric field, E r , with high spatial and high time resolution in H-mode and VH-mode discharges in the DIII-D tokamak have revealed the significant influence of the shear in E r on confinement and transport in these discharges. These measurements are made using the DIII-D Charge Exchange Recombination (CER) System. At the L-H transition in DIII-D plasmas, a negative well-like E r profile develops just within the magnetic separatrix. A region of shear in E r results, which extends 1 to 2 cm into the plasma from the separatrix. At the transition, this region of sheared E r exhibits the greatest increase in impurity ion poloidal rotation velocity and the greatest reduction in plasma fluctuations. A transport barrier is formed in this same region of E x B velocity shear as is signified by large increases in the observed gradients of the ion temperature, the carbon density, the electron temperature and electron density. The development of the region of sheared E r , the increase in impurity ion poloidal rotation, the reduction in plasma turbulence, and the transport barrier all occur simultaneously at the L-H transition. Measurements of the radial electric field, plasma turbulence, thermal transport, and energy confinement have been performed for a wide range of plasma conditions and configurations. The results support the supposition that the progression of improving confinement at the L-H transition, into the H-mode and then into the VH-mode can be explained by the hypothesis of the suppression of plasma turbulence by the increasing penetration of the region of sheared E x B velocity into the plasma interior

  18. Fast-ion transport in qmin>2, high-β steady-state scenarios on DIII-D

    International Nuclear Information System (INIS)

    Holcomb, C. T.; Heidbrink, W. W.; Collins, C.; Ferron, J. R.; Van Zeeland, M. A.; Garofalo, A. M.; Bass, E. M.; Luce, T. C.; Pace, D. C.; Solomon, W. M.; Mueller, D.; Grierson, B.; Podesta, M.; Gong, X.; Ren, Q.; Park, J. M.; Kim, K.; Turco, F.

    2015-01-01

    Results from experiments on DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] aimed at developing high β steady-state operating scenarios with high-q min confirm that fast-ion transport is a critical issue for advanced tokamak development using neutral beam injection current drive. In DIII-D, greater than 11 MW of neutral beam heating power is applied with the intent of maximizing β N and the noninductive current drive. However, in scenarios with q min >2 that target the typical range of q 95 = 5–7 used in next-step steady-state reactor models, Alfvén eigenmodes cause greater fast-ion transport than classical models predict. This enhanced transport reduces the absorbed neutral beam heating power and current drive and limits the achievable β N . In contrast, similar plasmas except with q min just above 1 have approximately classical fast-ion transport. Experiments that take q min >3 plasmas to higher β P with q 95 = 11–12 for testing long pulse operation exhibit regimes of better than expected thermal confinement. Compared to the standard high-q min scenario, the high β P cases have shorter slowing-down time and lower ∇β fast , and this reduces the drive for Alfvénic modes, yielding nearly classical fast-ion transport, high values of normalized confinement, β N , and noninductive current fraction. These results suggest DIII-D might obtain better performance in lower-q 95 , high-q min plasmas using broader neutral beam heating profiles and increased direct electron heating power to lower the drive for Alfvén eigenmodes

  19. High temperature outgassing tests on materials used in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Holtrop, K.L.; Hansink, M.J.

    2006-01-01

    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 -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

  20. The production and confinement of runaway electrons with impurity ''killer'' pellets in DIII-D

    International Nuclear Information System (INIS)

    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 15 m -3 . Argon killer pellet modeling yields runaway densities of between 10 15 --10 16 m -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

  1. Initial results of high resolution L-H transition studies on DIII-D

    International Nuclear Information System (INIS)

    Wang, G; Rhodes, T L; Doyle, E J; Peebles, W A; Zeng, L; Burrell, K H; McKee, G R; Groebner, R J; Evans, T E

    2004-01-01

    Understanding the L-H transition in tokamaks has been an important area of research for more than two decades. High time resolution diagnostics on DIII-D allow detailed characterization of the L-H transition and, therefore, testing and benchmarking of theoretical models. An experiment was performed in DIII-D utilizing a novel, high temporal and spatial resolution reflectometer density profile system to measure densities from the SOL to the inside separatrix. Initial data analysis indicates different density profile evolution during L-H transitions in upper single-null and lower single-null divertor configuration plasmas. A detailed comparison of the density gradient and fluctuation changes is presented for these two cases

  2. Initial results of high resolution L-H transition studies on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Wang, G [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Rhodes, T L [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Peebles, W A [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Zeng, L [Department of Electrical Engineering and PSTI, University of California, Los Angeles, CA 90095 (United States); Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186 (United States); McKee, G R [University of Wisconsin-Madison, 1500 Engineering Drive, Madison, WI 53706 (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186 (United States); Evans, T E [General Atomics, PO Box 85608, San Diego, CA 92186 (United States)

    2004-05-01

    Understanding the L-H transition in tokamaks has been an important area of research for more than two decades. High time resolution diagnostics on DIII-D allow detailed characterization of the L-H transition and, therefore, testing and benchmarking of theoretical models. An experiment was performed in DIII-D utilizing a novel, high temporal and spatial resolution reflectometer density profile system to measure densities from the SOL to the inside separatrix. Initial data analysis indicates different density profile evolution during L-H transitions in upper single-null and lower single-null divertor configuration plasmas. A detailed comparison of the density gradient and fluctuation changes is presented for these two cases.

  3. Computational Study of Anomalous Transport in High Beta DIII-D Discharges with ITBs

    Science.gov (United States)

    Pankin, Alexei; Garofalo, Andrea; Grierson, Brian; Kritz, Arnold; Rafiq, Tariq

    2015-11-01

    The advanced tokamak scenarios require a large bootstrap current fraction and high β. These large values are often outside the range that occurs in ``conventional'' tokamak discharges. The GLF23, TGLF, and MMM transport models have been previously validated for discharges with parameters associated with ``conventional'' tokamak discharges. It has been demonstrated that the TGLF model under-predicts anomalous transport in high β DIII-D discharges [A.M. Garofalo et al. 2015 TTF Workshop]. In this research, the validity of MMM7.1 model [T. Rafiq et al. Phys. Plasmas 20 032506 (2013)] is tested for high β DIII-D discharges with low and high torque. In addition, the sensitivity of the anomalous transport to β is examined. It is shown that the MMM7.1 model over-predicts the anomalous transport in the DIII-D discharge 154406. In particular, a significant level of anomalous transport is found just outside the internal transport barrier. Differences in the anomalous transport predicted using TGLF and MMM7.1 are reviewed. Mechanisms for quenching of anomalous transport in the ITB regions of high-beta discharges are investigated. This research is supported by US Department of Energy.

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

    International Nuclear Information System (INIS)

    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

    2003-01-01

    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 β above the n = 1 no-wall limit, enabled by active feed-back control. The ideal wall β limit is optimized by modifying the plasma shape, current and pressure profile. Present DIII-D AT experiments operate with f BS ∼ 50%-60%, with a long-term goal of ∼ 90%. Additional current is provided by neutral beam and electron cyclotron current drive, the latter being localized well away from the magnetic axis (ρ ∼ 0.4-0.5). Guided by integrated modeling, recent experiments have produced discharges with β ∼ 3%, β N ∼ 3, f BS ∼ 55% and noninductive fraction f NI ∼ 90%. Additional control is anticipated using fast wave current drive to control the central current density

  5. Stability of DIII-D high-performance, negative central shear discharges

    Science.gov (United States)

    Hanson, J. M.; Berkery, J. W.; Bialek, J.; Clement, M.; Ferron, J. R.; Garofalo, A. M.; Holcomb, C. T.; La Haye, R. J.; Lanctot, M. J.; Luce, T. C.; Navratil, G. A.; Olofsson, K. E. J.; Strait, E. J.; Turco, F.; Turnbull, A. D.

    2017-05-01

    Tokamak plasma experiments on the DIII-D device (Luxon et al 2005 Fusion Sci. Tech. 48 807) demonstrate high-performance, negative central shear (NCS) equilibria with enhanced stability when the minimum safety factor {{q}\\text{min}} exceeds 2, qualitatively confirming theoretical predictions of favorable stability in the NCS regime. The discharges exhibit good confinement with an L-mode enhancement factor H 89  =  2.5, and are ultimately limited by the ideal-wall external kink stability boundary as predicted by ideal MHD theory, as long as tearing mode (TM) locking events, resistive wall modes (RWMs), and internal kink modes are properly avoided or controlled. Although the discharges exhibit rotating TMs, locking events are avoided as long as a threshold minimum safety factor value {{q}\\text{min}}>2 is maintained. Fast timescale magnetic feedback control ameliorates RWM activity, expanding the stable operating space and allowing access to {β\\text{N}} values approaching the ideal-wall limit. Quickly growing and rotating instabilities consistent with internal kink mode dynamics are encountered when the ideal-wall limit is reached. The RWM events largely occur between the no- and ideal-wall pressure limits predicted by ideal MHD. However, evaluating kinetic contributions to the RWM dispersion relation results in a prediction of passive stability in this regime due to high plasma rotation. In addition, the ideal MHD stability analysis predicts that the ideal-wall limit can be further increased to {β\\text{N}}>4 by broadening the current profile. This path toward improved stability has the potential advantage of being compatible with the bootstrap-dominated equilibria envisioned for advanced tokamak (AT) fusion reactors.

  6. TRANSFORMERLESS OPERATION OF DIII-D WITH HIGH BOOTSTRAP FRACTION

    International Nuclear Information System (INIS)

    POLITZER, PA; HYATT, AW; LUCE, TC; MAHDAVI, MA; MURAKAMI, M; PERKINS, FW; PRATER, R; TURNBULL, AD; CASPER, TA; FERRON, JR; JAYAKUMAR, RJ; LAHAYE, RJ; LAZARUS, EA; PETTY, CC; WADE, MR

    2003-01-01

    OAK-B135 The authors have initiated an experimental program to address some of the questions associated with operation of a tokamak with high bootstrap current fraction under high performance conditions, without assistance from a transformer. In these discharges they have maintained stationary (or slowly improving) conditions for > 2.2 s at β N ∼ β p ∼ 2.8. Significant current overdrive, with dI/dt > 50 kA/s and zero or negative voltage, is sustained for over 0.7 s. The overdrive condition is usually ended with the appearance of MHD activity, which alters the profiles and reduces the bootstrap current. Characteristically these plasmas have 65%-80% bootstrap current, 25%-30% NBCD, and 5%-10% ECCD. Fully noninductive operation is essential for steady-state tokamaks. For efficient operation, the bootstrap current fraction must be close to 100%, allowing for a small additional (∼ 10%) external current drive capability to be used for control. In such plasmas the current and pressure profiles are rightly coupled because J(r) is entirely determined by p(r) (or more accurately by the kinetic profiles). The pressure gradient in turn is determined by transport coefficients which depend on the poloidal field profile

  7. High-Z material erosion and its control in DIII-D carbon divertor

    Directory of Open Access Journals (Sweden)

    R. Ding

    2017-08-01

    Full Text Available As High-Z materials will likely be used as plasma-facing components (PFCs in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH4 injection also provides information on radial transport due to E ×B drifts and cross field diffusion. Finally, D2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.

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

    Science.gov (United States)

    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. Stability Limits of High-Beta Plasmas in DIII-D

    International Nuclear Information System (INIS)

    Strait, E.J.

    2005-01-01

    Stability at high beta is an important requirement for a compact, economically attractive fusion reactor. DIII-D experiments have shown that ideal magnetohydrodynamic (MHD) theory is an accurate predictor of the ultimate stability limits for tokamaks, and the Troyon scaling law has provided a useful approximation of ideal stability limits for discharges with 'conventional' profiles. However, variation of the discharge shape, pressure profile, and current density profile can lead to ideal MHD beta limits that differ significantly from simple Troyon scaling. The need for profiles consistent with steady-state operation places an important additional constraint on plasma stability. Nonideal effects can also be important and must be taken into account. For example, neoclassical tearing modes (NTMs), resulting from plasma resistivity and the nonlinear effects of the bootstrap current, can become unstable at beta values well below the ideal MHD limit. DIII-D experiments are now entering a new era of unprecedented control over plasma stability, including suppression of NTMs by localized current drive at the island location, and direct feedback stabilization of kink modes with a resistive wall. The continuing development of physics understanding and control tools holds the potential for stable, steady-state fusion plasmas at high beta

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

    International Nuclear Information System (INIS)

    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.; Zeeland, M.A. Van

    2006-01-01

    The absorption of fast Alfven 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. Extension of high poloidal beta scenario in DIII-D to lower q95 for steady state fusion reactor

    Science.gov (United States)

    Huang, J.; Gong, X.; Qian, J.; Ding, S.; Ren, Q.; Guo, W.; Pan, C.; Li, G.; Xia, T.; Garofalo, A.; Lao, L.; Hyatt, A.; Ferron, J.; Collins, C.; Lin, D.; McKee, G.; Rhode, T.; McClenaghan, J.; Holcomb, C.; Cui, L.; Heidbrink, W.; Zhu, Y.; Diiid Team; East Team

    2017-10-01

    DIII-D/EAST joint experiments have improved the high poloidal beta scenario with sustained large-radius internal transport barrier (ITB) extended to high plasma current Ip 1MA with q95 6.0. Slight off-axis NBCD is applied to obtain broader current density profile, ITBs can now be sustained below the previously observed βp threshold with excellent confinement (H98y2 1.8). The scenario also exhibits a local negative shear appearing with q increased at rho 0.4, which helps ITB formation and sustainment. This confirms TGLF prediction that negative magnetic shear can help recover ITB and achieve high confinement with reduced q95. Detailed analysis shows that the Shafranov shift and q profile is critical in the ITB formation at high βp regime. Supported in part by National Magnetic Confinement Fusion Program of China 2015GB102000, 2015GB110005, and US Department of Energy under DE-FC02-04ER54698.

  12. Observation of an improved energy-confinement regime in neutral-beam--heated divertor discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Burrell, K.H.; Ejima, S.; Schissel, D.P.

    1987-01-01

    Tokamak discharges using the expanded boundary divertor in the DIII-D device exhibit H-mode confinement. With neutral-beam power up to 6 MW, energy confinement remains comparable to the Ohmic value at a plasma current of 1 MA. Confinement is also independent of plasma density and toroidal field. Confinement increases with plasma current, but the exact functional dependence is, as yet, uncertain. These results show that the H mode can be achieved in a reactor-compatible open divertor configuration

  13. Fueling with edge recycling to high-density in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, A.W., E-mail: leonard@fusion.gat.com [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Elder, J.D. [University of Toronto Institute of Aerospace Studies, Toronto, Canada M3H 5T6 (Canada); Canik, J.M. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Groebner, R.J.; Osborne, T.H. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)

    2013-07-15

    Pedestal fueling through edge recycling is examined with the interpretive OEDGE code for high-density discharges in DIII-D. A high current, high-density discharge is found to have a similar radial ion flux profile through the pedestal to a lower current, lower density discharge. The higher density discharge, however, has a greater density gradient indicating a pedestal particle diffusion coefficient that scales near linear with 1/I{sub p}. The time dependence of density profile is taken into account in the analysis of a discharge with low frequency ELMs. The time-dependent analysis indicates that the inferred neutral ionization source is inadequate to account for the increase in the density profile between ELMs, implying an inward density convection, or density pinch, near the top of the pedestal.

  14. Rotational and magnetic shear stabilization of magnetohydrodynamic modes and turbulence in DIII-D high performance discharges

    International Nuclear Information System (INIS)

    Lao, L.L.; Burrell, K.H.; Casper, T.S.

    1996-08-01

    The confinement and the stability properties of the DIII-D tokamak high performance discharges are evaluated in terms of rotational and magnetic shear with emphasis on the recent experimental results obtained from the negative central magnetic shear (NCS) experiments. In NCS discharges, a core transport barrier is often observed to form inside the NCS region accompanied by a reduction in core fluctuation amplitudes. Increasing negative magnetic shear contributes to the formation of this core transport barrier, but by itself is not sufficient to fully stabilize the toroidal drift mode (trapped- electron-η i mode) to explain this formation. Comparison of the Doppler shift shear rate to the growth rate of the η i mode suggests that the large core E x B flow shear can stabilize this mode and broaden the region of reduced core transport . Ideal and resistive stability analysis indicates the performance of NCS discharges with strongly peaked pressure profiles is limited by the resistive interchange mode to low Β N < 2.3. This mode is insensitive to the details of the rotational and the magnetic shear profiles. A new class of discharges which has a broad region of weak or slightly negative magnetic shear (WNS) is described. The WNS discharges have broader pressure profiles and higher values than the NCS discharges together with high confinement and high fusion reactivity

  15. Advances towards high performance low-torque qmin > 2 operations with large-radius ITB on DIII-D

    Science.gov (United States)

    Xu, G. S.; Solomon, W. M.; Garofalo, A. M.; Ferron, J. R.; Hyatt, A. W.; Wang, Q.; Yan, Z.; McKee, G. R.; Holcomb, C. T.; EAST Team

    2015-11-01

    A joint DIII-D/EAST experiment was performed aimed at extending a fully noninductive scenario with high βP and qmin > 2 to inductive operation at lower torque and higher Ip (0.6 --> 0.8 MA) for better performance. Extremely high confinement was obtained, i.e., H98y2 ~ 2.1 at βN ~ 3, which was associated with a strong ITB at large minor radius (ρ ~ 0.7). Alfvén Eigenmodes and broadband turbulence were significantly suppressed in the core, and fast-ion confinement was improved. ITB collapses at 0.8 MA were induced by ELM-triggered n = 1 MHD modes at the ITB location, which is different from the ``relaxation oscillations'' associated with the steady-state plasmas at lower current (0.6 MA). This successful joint experiment may open up a new avenue towards high performance low-torque qmin > 2 plasmas with large-radius ITBs, which will be demonstrated on EAST in the near future. Work supported by NMCFSP 2015GB102000, 2015GB110001 and the US DOE under DE-AC02-09CH11466, DE-FC02-04ER54698, DE-FG02-89ER53296 and DE-AC52-07NA27344.

  16. Progress Toward Long Pulse, High Performance Plasmas in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    P.A. Politzer; T.C. Luce; M.E. Austin; J.R. Ferron, A.M. Garofalo; C.M. Greenfield; A.W. Hyatt; R.J. La Haye; L.L. Lao; E.A. Lazarus; M.A. Makowski; M. Murakami; C.C. Petty; R.I. Pinsker; B.W. Rice; E.J. Strait, M.R. Wade; J.G. Watkins

    2000-01-01

    A major portion of the research program of the DIII-D tokamak collaboration is devoted to the development and demonstration of high performance advanced tokamak plasmas, with profiles as close as possible to those anticipated for steady-state operation. The work during the 1999 campaign has resulted in significant progress toward this goal. High normalized performance ((beta)(sub N)(approx) 4 and(beta)(sub N) H(sub 89)(approx) 9) discharges have been sustained for up to 2 s. These plasmas are in H-mode with rapid ELMs. The most common limiting phenomena are resistive wall modes (RWMs) rather than neoclassical tearing modes (NTMs). NTMs do occur, apparently triggered by the RWMs. The observed pressure is well above the calculated beta limit without a wall, and(beta)(sub N) and gt; 4(ell)(sub i) throughout the high performance phase. The bootstrap current is estimated to be and gt;50% of the total, and measurements of the internal loop voltage show that only about 25% of the current is inductively driven. The central q profile is flat, as is the calculated bootstrap current profile, due to the absence of any localized pressure gradients. The residual inductive current is localized around r/a(approx) 0.5. To demonstrate quasi-stationary operation, it will be necessary to replace the residual inductive current with ECCD at the same minor radius. To effectively apply ECH and ECCD to these discharges, density control will be needed. Preliminary experiments using the DIII-D cryopump have reduced the density by(approx)20%. A new EC power system and a new private flux cryopump will be available for the 2000 campaign

  17. High Field Side Lower Hybrid Current Drive Launcher Design for DIII-D

    Science.gov (United States)

    Wallace, G. M.; Leccacori, R.; Doody, J.; Vieira, R.; Shiraiwa, S.; Wukitch, S. J.; Holcomb, C.; Pinsker, R. I.

    2017-10-01

    Efficient off-axis current drive scalable to reactors is a key enabling technology for a steady-state tokamak. Simulations of DIII-D discharges have identified high performance scenarios with excellent lower hybrid (LH) wave penetration, single pass absorption and high current drive efficiency. The strategy was to adapt known launching technology utilized in previous experiments on C-Mod (poloidal splitter) and Tore Supra (bi-junction) and remain within power density limits established in JET and Tore Supra. For a 2 MW source power antenna, the launcher consists of 32 toroidal apertures and 4 poloidal rows. The aperture is 60 mm x 5 mm with 1 mm septa and the peak n| | is 2.7+/-0.2 for 90□ phasing. Eight WR187 waveguides are routed from the R-1 port down under the lower cryopump, under the existing divertor, and up the central column with the long waveguide dimension along the vacuum vessel. Above the inner strike point region, each waveguide is twisted to orient the long dimension perpendicular to the vacuum vessel and splits into 4 toroidal apertures via bi-junctions. To protect the waveguide, the inner wall radius will need to increase by 2.5 cm. RF, disruption, and thermal analysis of the latest design will be presented. Work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using User Facility DIII-D, under Award Number DE-FC02-04ER54698 and by MIT PSFC cooperative agreement DE-SC0014264.

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

    Energy Technology Data Exchange (ETDEWEB)

    Truong, D. D., E-mail: dtruong@wisc.edu [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Austin, M. E. [Institute for Fusion Studies, University of Texas, Austin, Texas, 78712 (United States)

    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.

  19. A system to deposit boron films (boronization) in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Hodapp, T.R.; Jackson, G.L.; Phillips, J.; Holtrop, K.L.; Peterson, P.L.; Winters, J.

    1992-01-01

    A system has been added to the DIII-D tokamak to coat its plasma facing surfaces with a film of boron using diborane gas. The system includes special health and safety equipment for handling the diborane gas which is toxic and inflammable. The purpose f the boron film is to reduce the levels of impurity atoms in the DIII-D plasmas. Experiments following the application of the boron film in DIII-D have led to significant reductions in plasma impurity levels and the observation of a new, very high confinement regime

  20. The importance of the radial electric field (Er) on interpretation of motional Stark effect measurements of the q profile in DIII-D high performance plasmas

    International Nuclear Information System (INIS)

    Rice, B.W.; Lao, L.L.; Burrell, K.H.; Greenfield, C.M.; Lin-Liu, Y.R.

    1997-06-01

    The development of enhanced confinement regimes such as negative central magnetic shear (NCS) and VH-mode illustrates the importance of the q profile and ExB velocity shear in improving stability and confinement in tokamak plasmas. Recently, it was realized that the large values of radial electric field observed in these high performance plasmas, up to 200 kV/m in DIII-D, have an effect on the interpretation of motional Stark effect (MSE) measurements of the q profile. It has also been shown that, with additional MSE measurements, one can extract a direct measurement of E r in addition to the usual poloidal field measurement. During a recent vent on DIII-D, 19 additional MSE channels with new viewing angles were added (for a total of 35 channels) in order to descriminate between the neutral beam v b x B electric field and the plasma E r field. In this paper, the system upgrade will be described and initial measurements demonstrating simultaneous measurement of the q and E r profiles will be presented

  1. STATIONARY HIGH-PERFORMANCE DISCHARGES IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    LUCE, TC; WADE, MR; FERRON, JR; HYATT, AW; KELLMAN, AG; KINSEY, JE; LAHAYE, RJ; LASNIER, CJ; MURAKAMI, M; POLITZER, PA; SCOVILLE, JT

    2002-01-01

    A271 STATIONARY HIGH-PERFORMANCE DISCHARGES IN THE DII-D TOKAMAK. Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak under stationary conditions at relatively low plasma current (q 95 > 4). A figure of merit for fusion gain (β N H 89 /q 95 2 ) has been maintained at values corresponding to | = 10 operation in a burning plasma for > 6 s or 36τ E and 2τ R . The key element is the relaxation of the current profile to a stationary state with q min > 1. In the absence of sawteeth and fishbones, stable operation has been achieved up to the estimated no-wall β limit. Feedback control of the energy content and particle inventory allow reproducible, stationary operation. The particle inventory is controlled by gas fueling and active pumping; the wall plays only a small role in the particle balance. The reduced current lessens significantly the potential for structural damage in the event of a major disruption. In addition, the pulse length capability is greatly increased, which is essential for a technology testing phase of a burning plasma experiment where fluence (duty cycle) is important

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

    Science.gov (United States)

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

    2016-11-01

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

  3. Prospects for Off-axis Current Drive via High Field Side Lower Hybrid Current Drive in DIII-D

    Science.gov (United States)

    Wukitch, S. J.; Shiraiwa, S.; Wallace, G. M.; Bonoli, P. T.; Holcomb, C.; Park, J. M.; Pinsker, R. I.

    2017-10-01

    An outstanding challenge for an economical, steady state tokamak is efficient off-axis current drive scalable to reactors. Previous studies have focused on high field side (HFS) launch of lower hybrid waves for current drive (LHCD) in double null configurations in reactor grade plasmas. The goal of this work is to find a HFS LHCD scenario for DIII-D that balances coupling, power penetration and damping. The higher magnetic field on the HFS improves wave accessibility, which allows for lower n||waves to be launched. These waves penetrate farther into the plasma core before damping at higher Te yielding a higher current drive efficiency. Utilizing advanced ray tracing and Fokker Planck simulation tools (GENRAY+CQL3D), wave penetration, absorption and drive current profiles in high performance DIII-D H-Mode plasmas were investigated. We found LH scenarios with single pass absorption, excellent wave penetration to r/a 0.6-0.8, FWHM r/a=0.2 and driven current up to 0.37 MA/MW coupled. These simulations indicate that HFS LHCD has potential to achieve efficient off-axis current drive in DIII-D and the latest results will be presented. Work supported by U.S. Dept. of Energy, Office of Science, Office of Fusion Energy Sciences, using User Facility DIII-D, under Award No. DE-FC02-04ER54698 and Contract No. DE-FC02-01ER54648 under Scientific Discovery through Advanced Computing Initiative.

  4. High Field Side Lower Hybrid Current Drive Simulations for Off- axis Current Drive in DIII-D

    Directory of Open Access Journals (Sweden)

    Wukitch S.J.

    2017-01-01

    Full Text Available Efficient off-axis current drive scalable to reactors is a key enabling technology for developing economical, steady state tokamak. Previous studies have focussed on high field side (HFS launch of lower hybrid current drive (LHCD in double null configurations in reactor grade plasmas and found improved wave penetration and high current drive efficiency with driven current profile peaked near a normalized radius, ρ, of 0.6-0.8, consistent with advanced tokamak scenarios. Further, HFS launch potentially mitigates plasma material interaction and coupling issues. For this work, we sought credible HFS LHCD scenario for DIII-D advanced tokamak discharges through utilizing advanced ray tracing and Fokker Planck simulation tools (GENRAY+CQL3D constrained by experimental considerations. For a model and existing discharge, HFS LHCD scenarios with excellent wave penetration and current drive were identified. The LHCD is peaked off axis, ρ∼0.6-0.8, with FWHM Δρ=0.2 and driven current up to 0.37 MA/MW coupled. For HFS near mid plane launch, wave penetration is excellent and have access to single pass absorption scenarios for variety of plasmas for n||=2.6-3.4. These DIII-D discharge simulations indicate that HFS LHCD has potential to demonstrate efficient off axis current drive and current profile control in DIII-D existing and model discharge.

  5. Cooperative program to analyze heat and particle transport at high beta in DIII-D

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1990-01-01

    The objective is to collaborate with the General Atomics staff and the LLNL staff at General Atomics in the analysis of transport data from DIII-D. The Berkeley effort is integrated into the ongoing efforts at GA to help expedite progress in the fundamental understanding of transport phenomena in tokamaks

  6. DIII-D Advanced Tokamak Research Overview

    International Nuclear Information System (INIS)

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

    1999-01-01

    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 β 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

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

    International Nuclear Information System (INIS)

    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 β), 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

  8. DIII-D research operations. Annual report, October 1, 1992--September 30, 1993

    Energy Technology Data Exchange (ETDEWEB)

    La Haye, R.J. [ed.

    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.

  9. Recycling and particle control in DIII-D

    International Nuclear Information System (INIS)

    Jackson, G.L.

    1991-11-01

    Particle control of both hydrogen and impurity atoms is important in obtaining reproducible discharges with a low fraction of radiated power in the DIII-D tokamak. The main DIII-D plasma facing components are graphite tiles and Inconel. Hydrogenic species desorbed from graphite during a tokamak discharge can be a major fueling source, especially in unconditioned graphite where these species can saturate the surface regions. In this case the recycling coefficient can exceed unity, leading to an uncontrolled density rise. In addition to removing volatile hydrocarbons and oxygen, DIII-D vessel conditioning efforts have been directed at the reduction of particle fueling from the graphite tiles. Conditioning techniques include: baking to ≤ 400 degrees C, low power pulsed discharge cleaning, and glow discharges in deuterium, helium, neon, or argon. Helium glow wall conditioning, is now routinely performed before every tokamak discharge. The effects of these techniques on hydrogen recycling and impurity influxes will be presented. The Inconel walls, while not generally exposed to high heat fluxes, nevertheless represent a source of metal impurities which can lead to impurity accumulation in the discharge and a high fraction of radiated power, particularly in H-mode discharges at higher plasma currents, I p > 1.5 MA. To reduce metal influx a thin (∼100 nm) low Z film has been applied on all plasma facing surfaces in DIII-D. The application of the boron film, referred to as boronization has the additional benefit over a carbon film of further reducing the oxygen influx. Following the first boronization in DIII-D a regime of very high confinement (VH-mode) was observed, characterized by low ohmic target density, low Z eff , and low radiated power

  10. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    International Nuclear Information System (INIS)

    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

  11. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    International Nuclear Information System (INIS)

    Simonen, T.C.; Baker, D.

    1993-01-01

    The DIII-D tokamak research program is carried out by General Atomics for the U.S. Department of Energy. The DIII-D is the most flexible and best diagnosed tokamak in the world and the second largest tokamak in the U.S. The primary goal of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated in three major areas: Tokamak Physics, Divertor and Boundary Physics, and Advanced Tokamak Studies

  12. DIII-D research operations. Annual report to the Department of Energy, October 1, 1991--September 30, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D. [ed.

    1993-05-01

    The DIII-D tokamak research program is carried out by, General Atomics (GA) for the U.S. Department of Energy (DOE). The DIII-D is the most flexible tokamak in the world. The primary goal of the DIII-D tokamak research program is to provide data needed by International Thermonuclear Experimental Reactor (ITER) and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. Specific DIII-D objectives include the steady-state sustainment of plasma current as well as demonstrating techniques for microwave heating, divertor heat removal, fuel exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion with high beta and with good confinement. The DIII-D long-range plan is organized into two major thrusts; the development of an advanced divertor and the development of advanced tokamak concepts. These two thrusts have a common goal: an improved DEMO reactor with lower cost and smaller size than the present DEMO which can be extrapolated from the conventional ITER operational scenario. In order to prepare for the long-range program, in FY92 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak Studies, and Tokamak Physics.

  13. The long range DIII-D plan

    International Nuclear Information System (INIS)

    Simonen, T.C.

    1993-02-01

    The mission 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. The National Energy Strategy calls for the development of magnetic fusion as an energy option with operation of a DEMO by 2025. The DEMO will be based on nuclear technology demonstrated in ITER and the physics and engineering database established in magnetic fusion facilities during the next two decades. On the present path, based on extrapolation of current conventional operating modes, ITER is twice as large as Joint European Tokamak (JET), and DEMO, using the same logic, will be even larger. However, successful development of advanced tokamak operating modes could open the way for significantly improved confinement and stability, leading to a smaller, more commercially attractive DEMO, provided new diverter concepts are developed to handle the accompanying high divertor power density. A smaller and lower cost DEMO opens up the possibility that multiple nations, utilities, and industries could build DEMOs simultaneously and, therefore, more rapidly optimize the tokamak for commercialization. Results from experiments at DIII-D and other tokamaks indicate that plasma and divertor performance can be increased transiently beyond the baseline conceptual design of ITER. A simultaneous long pulse demonstration of such improved tokamak plasma and divertor operation for steady state would establish an advanced physics foundation for the tokamak physics experiment program, provide new operating options for ITER, and open a path to an attractive DEMO. The planned DIII-D program incorporates new theory and technology developments to extend the tokamak experimental physics database toward steady state. This research program will also continue to provide increased understanding in many areas of fusion science and technology

  14. Vertical stability, high elongation, and the consequences of loss of vertical control on DIII-D

    International Nuclear Information System (INIS)

    Kellman, A.G.; Ferron, J.R.; Jensen, T.H.; Lao, L.L.; Luxon, J.L.; Skinner, D.G.; Strait, E.J.; Reis, E.; Taylor, T.S.; Turnbull, A.D.; Lazarus, E.A.; Lister, J.B.

    1990-09-01

    Recent modifications to the vertical control system for DIII-D has enabled operation of discharges with vertical elongation κ, up to 2.5. When vertical stability is lost, a disruption follows and a large vertical force on the vacuum vessel is observed. The loss of plasma energy begins when the edge safety factor q is 2 but the current decay does not begin until q ∼1.3. Current flow on the open field lines in the plasma scrapeoff layer has been measured and the magnitude and distribution of these currents can explain the observed force on the vessel. Equilibrium calculations and simulation of this vertical displacement episode are presented. 7 refs., 4 figs

  15. High harmonic ion cyclotron heating in DIII-D: Beam ion absorption and sawtooth stabilization

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Fredrickson, E.D.; Mau, T.K.; Petty, C.C.; Pinsker, R.I.; Porkolab, M.; Rice, B.W.

    1999-01-01

    Combined neutral beam injection and fast wave heating at the fourth cyclotron harmonic produce an energetic deuterium beam ion tail in the DIII-D tokamak. When the concentration of thermal hydrogen exceeds ∼ 5%, the beam ion absorption is suppressed in favour of second harmonic hydrogen absorption. As theoretically expected, the beam absorption increases with beam ion gyro-radius; also, central absorption at the fifth harmonic is weaker than central absorption at the fourth harmonic. For central heating at the fourth harmonic, an energetic, perpendicular, beam population forms inside the q = 1 surface. The beam ion tail transiently stabilizes the sawtooth instability but destabilizes toroidicity induced Alfven eigenmodes (TAEs). Saturation of the central heating correlates with the onset of the TAEs. Continued expansion of the q = 1 radius eventually precipitates a sawtooth crash; complete magnetic reconnection is observed. (author)

  16. High spatial and temporal resolution visible spectroscopy of the plasma edge in DIII-D

    International Nuclear Information System (INIS)

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

    1990-10-01

    In DIII-D, visible spectroscopic measurements of the He II 468.6 nm and C VI 529.2 nm Doppler broadened spectral lines, resulting from charge exchange recombination interactions between beam neutral atoms and plasma ions, are performed to determine ion temperatures, and toroidal and poloidal rotation velocities. The diagnostics system comprises 32 viewing chords spanning a typical minor radius of 63 cm across the midplane, of which 16 spatial chords span 11 cm of the plasma edge just within the separatrix. A temporal resolution of 260 μs per time slice can be obtained as a result of using MCP phosphors with short decay times and fast camera readout electronics. Results from this system will be used in radial electric field comparisons with theory at the L-H transition and ion transport analysis. 6 refs., 3 figs

  17. Initial results of the high resolution edge Thomson scattering upgrade at DIII-D.

    Science.gov (United States)

    Eldon, D; Bray, B D; Deterly, T M; Liu, C; Watkins, M; Groebner, R J; Leonard, A W; Osborne, T H; Snyder, P B; Boivin, R L; Tynan, G R

    2012-10-01

    Validation of models of pedestal structure is an important part of predicting pedestal height and performance in future tokamaks. The Thomson scattering diagnostic at DIII-D has been upgraded in support of validating these models. Spatial and temporal resolution, as well as signal to noise ratio, have all been specifically enhanced in the pedestal region. This region is now diagnosed by 20 view-chords with a spacing of 6 mm and a scattering length of just under 5 mm sampled at a nominal rate of 250 Hz. When mapped to the outboard midplane, this corresponds to ~3 mm spacing. These measurements are being used to test critical gradient models, in which pedestal gradients increase in time until a threshold is reached. This paper will describe the specifications of the upgrade and present initial results of the system.

  18. Protecting Against Damage from Refraction of High Power Microwaves in the DIII-D Tokamak

    Directory of Open Access Journals (Sweden)

    Lohr John

    2017-01-01

    Full Text Available Several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps have been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.

  19. Joint DIII-D/EAST research on the development of a high poloidal beta scenario for the steady state missions of ITER and CFETR

    Science.gov (United States)

    Garofalo, A. M.; Gong, X. Z.; Ding, S. Y.; Huang, J.; McClenaghan, J.; Pan, C. K.; Qian, J.; Ren, Q. L.; Staebler, G. M.; Chen, J.; Cui, L.; Grierson, B. A.; Hanson, J. M.; Holcomb, C. T.; Jian, X.; Li, G.; Li, M.; Pankin, A. Y.; Peysson, Y.; Zhai, X.; Bonoli, P.; Brower, D.; Ding, W. X.; Ferron, J. R.; Guo, W.; Lao, L. L.; Li, K.; Liu, H.; Lyv, B.; Xu, G.; Zang, Q.

    2018-01-01

    Experimental and modeling investigations on the DIII-D and EAST tokamaks show the attractive transport and stability properties of fully noninductive, high poloidal-beta (β P ) plasmas, and their suitability for steady-state operating scenarios in ITER and CFETR. A key feature of the high-β P regime is the large-radius (ρ > 0.6) internal transport barrier (ITB), often observed in all channels (ne, Te, Ti, rotation), and responsible for both excellent energy confinement quality and excellent stability properties. Experiments on DIII-D have shown that, with a large-radius ITB, very high β N and β P values (both ≥ 4) can be reached by taking advantage of the stabilizing effect of a nearby conducting wall. Synergistically, higher plasma pressure provides turbulence suppression by Shafranov shift, leading to ITB sustainment independent of the plasma rotation. Experiments on EAST have been used to assess the long pulse potential of the high-β P regime. Using RF-only heating and current drive, EAST achieved minute-long fully noninductive steady state H-mode operation with strike points on an ITER-like tungsten divertor. Improved confinement (relative to standard H-mode) and steady state ITB features are observed with a monotonic q-profile with q min ˜ 1.5. Separately, experiments have shown that increasing the density in plasmas driven by lower hybrid wave broadens the q-profile, a technique that could enable a large radius ITB. These experimental results have been used to validate MHD, current drive, and turbulent transport models, and to project the high-β P regime to a burning plasma. These projections suggest the Shafranov shift alone will not suffice to provide improved confinement (over standard H-mode) without rotation and rotation shear. However, increasing the negative magnetic shear (higher q on axis) provides a similar turbulence suppression mechanism to Shafranov shift, and can help devices such as ITER and CFETR achieve their steady-state fusion

  20. RF high voltage performance of RF transmission line components on the DIII-D Fast Wave Current Drive (FWCD) system

    International Nuclear Information System (INIS)

    Ferguson, S.W.; Callis, R.W.; Cary, W.P.; Phelps, D.A.; Ponce, D.; Baity, F.W.; Barber, G.

    1995-01-01

    The performance of the high voltage rf components of the DIII-D Fast Wave Current Drive System (FWCD) have been evaluated under various conditions of insulator configuration, insulator material, insulating gas and gas pressure. The insulator materials that have been investigated are alumina, steatite, pyrex, quartz, and teflon. The results of this evaluation are discussed in this paper. Additionally a rf high potter was developed to aid in the evaluation of rf high voltage components. The high potter consists of a 50 Ω, 1/4 wavelength cavity with a variable position short and a 50 ohm matched tap at one end of the cavity. With this configuration rf voltages were generated in excess of 100 kVp in the frequency range 30 to 60 MHz

  1. RF high voltage performance of RF transmission line components on the DIII-D Fast Wave Current Drive (FWCD) System

    International Nuclear Information System (INIS)

    Ferguson, S.W.; Callis, R.W.; Cary, W.P.; Phelps, D.A.; Ponce, D.; Baity, F.W.; Barber, G.

    1995-12-01

    The performance of the high voltage rf components of the DIII-D Fast Wave Current Drive System (FWCD) have been evaluated under various conditions of insulator configuration, insulator material, insulating gas and gas pressure. The insulator materials that have been investigated are alumina, steatite, pyrex, quartz, and teflon. The results of this evaluation are discussed in this paper. Additionally a rf high potter was developed to aid in the evaluation of rf high voltage components. The high potter consists of a 50 Ω, 1/4 wavelength cavity with a variable position short and a 50 ohm matched tap at one end of the cavity. With this configuration rf voltages were generated in excess of 100 kVp in the frequency range 30 to 60 MHz

  2. Status and near-term plans for DIII-D

    International Nuclear Information System (INIS)

    Davis, L.G.; Callis, R.W.; Luxon, J.L.; Stambaugh, R.D.

    1987-10-01

    The DIII-D tokamak at GA Technologies began plasma operation in February of 1986 and is dedicated to the study of highly non-circular plasmas. High beta operation with enhanced energy confinement is paramount among the goals of the DIII-D research program. Commissioning of the device and facility has verified the design capability including coil and vessel loading, volt-second consumption, bakeout temperature, vessel armor, and neutral beamline thermal integrity and control systems performance. Initial experimental results demonstrate the DIII-D is capable of attaining high confinement (H-mode) discharges in a divertor configuration using modest neutral beam heating or ECH. Record values of I/sub p/aB/sub T/ have been achieved with ohmic heating as a first step toward operation at high values of toroidal beta and record values of beta have been achieved using neutral beam heating. This paper summarizes results to date and gives the near term plans for the facility. 13 refs., 6 figs., 1 tab

  3. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Brooks, N.H.; Campbell, R.B.; Fenstermacher, M.E.; Hill, D.N.; Hyatt, A.W.; Knoll, D.; Lasnier, C.J.; Lazarus, E.A.; Leonard, A.W.; Lippmann, S.I.; Mahdavi, M.A.; Maingi, R.; Meyer, W.; Moyer, R.A.; Petrie, T.W.; Porter, G.D.; Rensink, M.E.; Rognlien, T.D.; Schaffer, M.J.; Smith, J.P.; Staebler, G.M.; Stambaugh, R.D.; West, W.P.; Wood, R.D.

    1995-01-01

    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 ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while τ E remains similar 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z 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 (δ∼0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.))

  4. Fluctuations in high βp plasmas in DIII-D

    International Nuclear Information System (INIS)

    Casper, T.A.; Chu, M.S.; Gohil, C.P.

    1994-06-01

    In our investigation of improved confinement in high poloidal beta (β p = 2 to 4) advanced tokamak experiments, coincident with q 0 rising above 2, we observe the internal MHD activity to evolve from an m/n = 2/1 to a 3/1 structure consistent with the GATO code stability analysis. The plasma eventually evolves to a quiescent state at which time the stored energy increases, mostly as a result of improved particle confinement. The measured plasma pressure profiles during this time are also calculated to be stable to high-n ballooning modes consistent with operation of the core in the second stable regime. The sustained improvement in confinement is ultimately limited by our ability to control the toroidal current profile of which the bootstrap current contributes a large fraction (up to 80%)

  5. Fluctuations in high βp plasmas in DIII-D

    International Nuclear Information System (INIS)

    Casper, T.A.; Chu, M.S.; Gohil, P.

    1994-01-01

    In our investigation of improved confinement in high poloidal beta (β p = 2 to 4) advanced tokamak experiments, we observe that the internal MHD activity evolves from an m/n = 2/1 to a 3/1 structure coincident with q 0 rising above 2, and consistent with the GATO code stability analysis. The plasma eventually evolves to a quiescent state at which time the stored energy increases, mostly as a result of improved particle confinement. The bootstrap fraction rises to 80%. The measured plasma pressure profiles during this time are calculated to be stable to high-n ballooning modes consistent with operation of the core in the second stable regime. The sustained improvement in confinement is ultimately limited by our ability to control the toroidal current profile. (author) 4 refs., 6 figs

  6. Fluctuations in high βp plasmas in DIII-D

    International Nuclear Information System (INIS)

    Casper, T.A.; Chu, M.S.; Gohil, P.

    1994-07-01

    In our investigation of improved confinement in high poloidal beta (β p = 2 to 4) advanced tokamak experiments, we observe that the internal MHD activity evolves from an m/n = 2/1 to a 3/1 structure coincident with q o rising above 2, and consistent with the GATO code stability analysis. The plasma eventually evolves to a quiescent state at which time the stored energy increases, mostly as a result of improved particle confinement. The bootstrap fraction rises to 80%. The measured plasma pressure profiles during this time are calculated to be stable to high-n ballooning modes consistent with operation of the core in the second stable regime. The sustained improvement in confinement is ultimately limited by our ability to control the toroidal current profile

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

    Science.gov (United States)

    Ali, Halima; Punjabi, Alkesh

    2013-05-01

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

  8. High beta tokamak operation in DIII-D limited at low density/collisionality by resistive tearing modes

    International Nuclear Information System (INIS)

    La Haye, R.J.; Lao, L.L.; Strait, E.J.; Taylor, T.S.

    1997-01-01

    The maximum operational high beta in single-null divertor (SND) long pulse tokamak discharges in the DIII-D tokamak with a cross-sectional shape similar to the proposed International Thermonuclear Experimental Reactor (ITER) device is found to be limited by the onset of resistive instabilities that have the characteristics of neoclassically destabilized tearing modes. There is a soft limit due to the onset of an m/n=3/2 rotating tearing mode that saturates at low amplitude and a hard limit at slightly higher beta due to the onset of an m/n=2/1 rotating tearing mode that grows, slows down and locks. By operating at higher density and thus collisionality, the practical beta limit due to resistive tearing modes approaches the ideal magnetohydrodynamic (MHD) limit. (author). 15 refs, 4 figs

  9. A quantitative analysis of the effect of ELMs on H-mode thermal energy confinement in DIII-D

    International Nuclear Information System (INIS)

    Schissel, D.P.; Osborne, T.H.; Carlstrom, T.N.; Zohm, H.

    1992-06-01

    The desire to reach ignition in future tokamaks the energy confinement time critical parameter. The most promising enhanced (over L-mode) confinement regime is the H-mode, discovered on ASDEX with neutral beam heating, and then confirmed with various auxiliary heating sources on numerous machines. The knowledge of how H-mode τ E depends on different parameters is of chemical importance to the performance predictions for next generation devices. Inter-machine H-mode total and thermal energy confinement (τ th ) scalings, which are being utilized to predict ITER thermal energy confinement, have been created for discharges where the Edge Localized Mode (ELM) instability has not been present. Confinement scaling research hm concentrated on this ELM-free H-mode phase mostly owing to the difficulty of characterizing ELM behavior. To date, long pulse H-mode operation has only been achieved by utilizing ELMs to flush out unpurities and prevent radiative collapse of the discharge. Unfortunately, accompanying the ELMS is a decrease of the plasma stored energy due to the expulsion of particles near the edge of the discharge resulting in a reduction of the steep edge electron density gradient. In order to predict ITER's H-mode τ th in the presence of ELMS, an estimated 25% confinement degradation factor has been applied to the ELM-free predictions. Our work, summarized in this paper, indicates that this 25% reduction factor is too large and instead a value of approximately 15% would be more appropriate

  10. H-mode pedestal characteristics, ELMs, and energy confinement in ITER shape discharges on DIII-D

    International Nuclear Information System (INIS)

    Osborne, T.H.; Groebner, R.J.; Lao, L.L.; Leonard, A.W.; Miller, R.L.; Thomas, D.M.; Waltz, R.E.; Maingi, R.; Porter, G.D.

    1997-12-01

    The H-mode confinement enhancement factor, H, is found to be strongly correlated with the height of the edge pressure pedestal in ITER shape discharges. In discharges with Type I ELMs the pedestal pressure is set by the maximum pressure gradient before the ELM and the width of the H-mode transport barrier. The pressure gradient before Type I ELMs is found to scale as would be expected for a stability limit set by ideal ballooning modes, but with values significantly in excess of that predicted by stability code calculations. The width of the H-mode transport barrier is found to scale equally well with pedestal P(POL)(2/3) or B(POL)(1/2). The improved H value in high B(POL) discharges may be due to a larger edge pressure gradient and wider H-mode transport barrier consistent with their higher edge ballooning mode limit. Deuterium puffing is found to reduce H consistent with the smaller pedestal pressure which results from the reduced barrier width and critical pressure gradient. Type I ELM energy loss is found to be proportional to the change in the pedestal energy

  11. Comparison of turbulence measurements from DIII-D low-mode and high-performance plasmas to turbulence simulations and models

    International Nuclear Information System (INIS)

    Rhodes, T.L.; Leboeuf, J.-N.; Sydora, R.D.; Groebner, R.J.; Doyle, E.J.; McKee, G.R.; Peebles, W.A.; Rettig, C.L.; Zeng, L.; Wang, G.

    2002-01-01

    Measured turbulence characteristics (correlation lengths, spectra, etc.) in low-confinement (L-mode) and high-performance plasmas in the DIII-D tokamak [Luxon et al., Proceedings Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] show many similarities with the characteristics determined from turbulence simulations. Radial correlation lengths Δr of density fluctuations from L-mode discharges are found to be numerically similar to the ion poloidal gyroradius ρ θ,s , or 5-10 times the ion gyroradius ρ s over the radial region 0.2 θ,s or 5-10 times ρ s , an experiment was performed which modified ρ θs while keeping other plasma parameters approximately fixed. It was found that the experimental Δr did not scale as ρ θ,s , which was similar to low-resolution UCAN simulations. Finally, both experimental measurements and gyrokinetic simulations indicate a significant reduction in the radial correlation length from high-performance quiescent double barrier discharges, as compared to normal L-mode, consistent with reduced transport in these high-performance plasmas

  12. Quasistationary Plasma Predator-Prey System of Coupled Turbulence, Drive, and Sheared E ×B Flow During High Performance DIII-D Tokamak Discharges

    Science.gov (United States)

    Barada, K.; Rhodes, T. L.; Burrell, K. H.; Zeng, L.; Bardóczi, L.; Chen, Xi; Muscatello, C. M.; Peebles, W. A.

    2018-03-01

    A new, long-lived limit cycle oscillation (LCO) regime has been observed in the edge of near zero torque high performance DIII-D tokamak plasma discharges. These LCOs are localized and composed of density turbulence, gradient drives, and E ×B velocity shear damping (E and B are the local radial electric and total magnetic fields). Density turbulence sequentially acts as a predator (via turbulence transport) of profile gradients and a prey (via shear suppression) to the E ×B velocity shear. Reported here for the first time is a unique spatiotemporal variation of the local E ×B velocity, which is found to be essential for the existence of this system. The LCO system is quasistationary, existing from 3 to 12 plasma energy confinement times (˜30 - 900 LCO cycles) limited by hardware constraints. This plasma system appears to contribute strongly to the edge transport in these high performance and transient-free plasmas, as evident from oscillations in transport relevant edge parameters at LCO time scale.

  13. Stationary, high bootstrap fraction plasmas in DIII-D without inductive current control

    International Nuclear Information System (INIS)

    Politzer, P.A.; Hyatt, A.W.; Luce, T.C.; Prater, R.; Turnbull, A.D.; Ferron, J.R.; Greenfield, C.M.; La Haye, R.J.; Petty, C.C.; Perkins, F.W.; Brennan, D.P.; Lazarus, E.A.; Jayakumar, J.; Wade, M.R.

    2005-01-01

    We have initiated an experimental program to address some of the questions associated with operation of a tokamak with high bootstrap current fraction under high performance conditions, without assistance from a transformer. In these discharges stationary (or slowly improving) conditions are maintained for > 3.7 s at β N ∼ β p ≤ 3.3. The achievable current and pressure are limited by a relaxation oscillation, involving growth and collapse of an ITB at ρ ≥ 0.6. The pressure gradually increases and the current profile broadens throughout the discharge. Eventually the plasma reaches a more stable, high confinement (H89P ∼ 3) state. Characteristically these plasmas have 65%-85% bootstrap current, 15%-30% NBCD, and 0%-10% ECCD. (author)

  14. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-01-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. (In the past DIII-D operated with an open divertor.) The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix 'strike' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. (author) 5 refs., 5 figs

  15. Computer control of the high-voltage power supply for the DIII-D electron cyclotron heating system

    International Nuclear Information System (INIS)

    Clow, D.D.; Kellman, D.H.

    1992-01-01

    This paper reports on the DIII-D Electron Cyclotron Heating (ECH) high voltage power supply which is controlled by a computer. Operational control is input via keyboard and mouse, and computer/power supply interfact is accomplished with a Computer Assisted Monitoring and Control (CAMAC) system. User-friendly tools allow the design and layout of simulated control panels on the computer screen. Panel controls and indicators can be changed, added or deleted, and simple editing of user-specific processes can quickly modify control and fault logic. Databases can be defined, and control panel functions are easily referred to various data channels. User-specific processes are written and linked using Fortran, to manage control and data acquisition through CAMAC. The resulting control system has significant advantages over the hardware it emulates: changes in logic, layout, and function are quickly and easily incorporated; data storage, retrieval, and processing are flexible and simply accomplished; physical components subject to wear and degradation are minimized. In addition, the system can be expanded to multiplex control of several power supplies, each with its own database, through a single computer console

  16. Scaling of the stochastic broadening from low mn, high mn, and peeling-ballooning magnetic perturbations in the DIII-D tokamak

    Science.gov (United States)

    Zhao, Michael; Punjabi, Alkesh; Ali, Halima

    2009-11-01

    The equilibrium EFIT data for the DIII-D shot 115467 is used to construct the equilibrium generating function for magnetic field line trajectories in the DIII-D tokamak in natural canonical coordinates [A. Punjabi, and H. Ali, Phys. Plasmas 15, 122502 (2008)]. A canonical transformation is used to construct an area-preserving map for field line trajectories in the natural canonical coordinates in the DIII-D. Maps in natural canonical coordinates have the advantage that natural canonical coordinates can be inverted to calculate real space coordinates (R,Z,φ), and there is no problem in crossing the separatrix. This is not possible for magnetic coordinates [O. Kerwin, A. Punjabi, and H. Ali, Phys. Plasmas 15, 072504 (2008)]. This map is applied to calculate stochastic broadening from the low mn (m,n)=(1,1)+(1,-1); high mn (m,n)=(4,1)+(3,1); and the peeling-ballooning (m,n)=(40,10)+(30,10) magnetic perturbations. In all three cases, the scaling of the widths of stochastic layer near the X-point in the principal plane of the DIII-D deviates at most by 6% from the .5ex1 -.1em/ -.15em.25ex2 power Boozer-Rechester scaling [A. Boozer, and A. Rechester, Phys. Fluids 21, 682 (1978)]. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

  17. Physics of turbulence control and transport barrier formation in DIII-D

    International Nuclear Information System (INIS)

    Doyle, E.J.; Burrell, K.H.; Carlstrom, T.N.

    1996-10-01

    This paper describes the physical mechanisms responsible for turbulence control and transport barrier formation on DIII-D as determined from a synthesis of results from different enhanced confinement regimes, including quantitative and qualitative comparisons to theory. A wide range of DIII-D data support the hypothesis that a single underlying physical mechanism, turbulence suppression via E x B shear flow is playing an essential, though not necessarily unique, role in reducing turbulence and transport in all of the following improved confinement regimes: H-mode, VH-mode, high-ell i modes, improved performance counter-injection L-mode discharges and high performance negative central shear (NCS) discharges. DIII-D data also indicate that synergistic effects are important in some cases, as in NCS discharges where negative magnetic shear also plays a role in transport barrier formation. This work indicates that in order to control turbulence and transport it is important to focus on understanding physical mechanisms, such as E x B shear, which can regulate and control entire classes of turbulent modes, and thus control transport. In the highest performance DIII-D discharges, NCS plasmas with a VH-mode like edge, turbulence is suppressed at all radii, resulting in neoclassical levels of ion transport over most of the plasma volume

  18. Ion Bernstein wave antenna design for DIII-D

    International Nuclear Information System (INIS)

    Phelps, R.D.; Mayberry, M.J.; Pinsker, R.J.

    1989-01-01

    An array of two toroidal loop antennas has been designd and installed on the DIII-D tokamak to carry out Ion Bernstein Wave (IBW) heating experiments. The antenna will operate at the 2 MW level and provide direct excitation of the IBW over the frequency range of 30-60 MHz. This device will permit the study of coupling th IBW to divertor plasmas and will provide a menas for improving the confinement and stability of high beta plasmas through localized off-axis heating. This paper describes both the mechanical and electromagnetic design of the IBW antenna. (author). 2 refs.; 4 figs.; 1 tab

  19. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1993--September 30, 1994

    International Nuclear Information System (INIS)

    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 ampersand 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

  20. Integrated modeling of plasma ramp-up in DIII-D ITER-like and high bootstrap current scenario discharges

    Science.gov (United States)

    Wu, M. Q.; Pan, C. K.; Chan, V. S.; Li, G. Q.; Garofalo, A. M.; Jian, X.; Liu, L.; Ren, Q. L.; Chen, J. L.; Gao, X.; Gong, X. Z.; Ding, S. Y.; Qian, J. P.; Cfetr Physics Team

    2018-04-01

    Time-dependent integrated modeling of DIII-D ITER-like and high bootstrap current plasma ramp-up discharges has been performed with the equilibrium code EFIT, and the transport codes TGYRO and ONETWO. Electron and ion temperature profiles are simulated by TGYRO with the TGLF (SAT0 or VX model) turbulent and NEO neoclassical transport models. The VX model is a new empirical extension of the TGLF turbulent model [Jian et al., Nucl. Fusion 58, 016011 (2018)], which captures the physics of multi-scale interaction between low-k and high-k turbulence from nonlinear gyro-kinetic simulation. This model is demonstrated to accurately model low Ip discharges from the EAST tokamak. Time evolution of the plasma current density profile is simulated by ONETWO with the experimental current ramp-up rate. The general trend of the predicted evolution of the current density profile is consistent with that obtained from the equilibrium reconstruction with Motional Stark effect constraints. The predicted evolution of βN , li , and βP also agrees well with the experiments. For the ITER-like cases, the predicted electron and ion temperature profiles using TGLF_Sat0 agree closely with the experimental measured profiles, and are demonstrably better than other proposed transport models. For the high bootstrap current case, the predicted electron and ion temperature profiles perform better in the VX model. It is found that the SAT0 model works well at high IP (>0.76 MA) while the VX model covers a wider range of plasma current ( IP > 0.6 MA). The results reported in this paper suggest that the developed integrated modeling could be a candidate for ITER and CFETR ramp-up engineering design modeling.

  1. DIII-D physics analysis database

    International Nuclear Information System (INIS)

    Bramson, G.; Schissel, D.P.; DeBoo, J.C.; St John, H.

    1990-10-01

    Since June 1986 the DIII-D tokamak has had over 16000 discharges accumulating more than 250 Gigabytes of raw data (currently over 30 Mbytes per discharge). The centralized DIII-D databases and the associated support software described earlier provide the means to extract, analyze, store, and display reduced sets of data for specific physics issues. The confinement, stability, transition, and cleanliness databases consist of more than 7500 records of basic reduced diagnostic data datasets. Each database record corresponds to a specific snapshot in time for a selected discharge. Recently some profile datasets have been implemented. Diagnostic data are fit by a cubic spline or a parabola by the in-house ENERGY code to provide density, temperature, radiated power, effective charge (Z eff ), and rotation velocity profiles. These fits are stored in the profile datasets which are inputs for the ONETWO code which computes transport data. 3 refs., 4 figs

  2. Collaboration on DIII-D Five Year Plan

    International Nuclear Information System (INIS)

    Allen, S

    2003-01-01

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

  3. Disruptions in DIII-D

    International Nuclear Information System (INIS)

    Reiman, A.; Taylor, P.; Kellman, A.; LaHaye, R.

    1996-01-01

    We report on the results of a statistical analysis of the DIII-D disruption data base, and on an examination of a selected subset of the shots to determine the likely causes of disruptions. The statistical analysis focuses on the dependence of the disruption rate on key dimensionless parameters. We find that the disruption frequency is high at modest values of the parameters, and that it can be relatively low at operational limits. For example, the disruption frequency in an ITER relevant regime (β N /l i ∼ 2, 3 G > 0.6, where n G is the Greenwald limit) is approximately 23%. For this range of q, the disruption frequency rises only modestly to about 35% at the β limit, consistent with previous observations of a soft β limit for this q regime. For the range 6 95 G G < .9) in all q regimes we have studied. The location of the minimum moves to higher density with increasing q

  4. DIII-D research operations

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D. (ed.)

    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.

  5. DIII-D research operations

    International Nuclear Information System (INIS)

    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 ampersand D; and collaborative efforts

  6. Recent results from the DIII-D tokamak

    International Nuclear Information System (INIS)

    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

  7. Simulation of enhanced tokamak performance on DIII-D using fast wave current drive

    International Nuclear Information System (INIS)

    Grassie, J.S. de; Lin-Liu, Y.R.; Petty, C.C.; Pinsker, R.I.; Chan, V.S.; Prater, R.; John, H. St.; Baity, F.W.; Goulding, R.H.; Hoffman, D.H.

    1993-01-01

    The fast magnetosonic wave is now recognized to be a leading candidate for noninductive current drive for the tokamak reactor due to the ability of the wave to penetrate to the hot dense core region. Fast wave current drive (FWCD) experiments on DIII-D have realized up to 120 kA of rf current drive, with up to 40% of the plasma current driven noninductively. The success of these experiments at 60 MHz with a 2 MW transmitter source capability has led to a major upgrade of the FWCD system. Two additional transmitters, 30 to 120 MHz, with a 2 MW source capability each, will be added together with two new four-strap antennas in early 1994. Another major thrust of the DIII-D program is to develop advanced tokamak modes of operation, simultaneously demonstrating improvements in confinement and stability in quasi-steady-state operation. In some of the initial advanced tokamak experiments on DIII-D with neutral beam heated (NBI) discharges it has been demonstrated that energy confinement time can be improved by rapidly elongating the plasma to force the current density profile to be more centrally peaked. However, this high-l i phase of the discharge with the commensurate improvement in confinement is transient as the current density profile relaxes. By applying FWCD to the core of such a κ-ramped discharge it may be possible to sustain the high internal inductance and elevated confinement. Using computational tools validated on the initial DIII-D FWCD experiments we find that such a high-l i advanced tokamak discharge should be capable of sustainment at the 1 MA level with the upgraded capability of the FWCD system. (author) 16 refs., 3 figs., 1 tab

  8. The long range DIII-D plan

    International Nuclear Information System (INIS)

    Simonen, T.C.

    1994-01-01

    The mission 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. The National Energy Strategy calls for the development of magnetic fusion as an energy options with operation of a DEMO by 2025. The DEMO will be based on nuclear technology demonstrated in ITER and the physics and engineering database established in magnetic fusion facilities during the next two decades. On the present path, based on extrapolation of current conventional operating modes, ITER is twice as large as Joint European Tokamak (JET), and DEMO, using the same logic, will be even larger. However, successful development of advanced tokamak (AT) operating modes could open the way for significantly improved confinement and stability, leading to a smaller, more commercially attractive DEMO, provided new divertor concepts are developed to handle the accompanying high divertor power density. A smaller and lower cost DEMO opens up the possibility that multiple nations, utilities, and industries could build DEMOs simultaneously and, therefore, more rapidly optimize the tokamak for commercialization

  9. Improved charge-coupled device detectors for high-speed, charge exchange spectroscopy studies on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Burrell, K.H.; Gohil, P.; Groebner, R.J.; Kaplan, D.H.; Robinson, J.I.; Solomon, W.M.

    2004-01-01

    Charge exchange spectroscopy is one of the key ion diagnostics on the DIII-D tokamak. It allows determination of ion temperature, poloidal and toroidal velocity, impurity density, and radial electric field E r throughout the plasma. For the 2003 experimental campaign, we replaced the intensified photodiode array detectors on the central portion of the DIII-D charge exchange spectroscopy system with advanced charge-coupled device (CCD) detectors mounted on faster (f/4.7) Czerny-Turner spectrometers equipped with toroidal mirrors. The CCD detectors are improved versions of the ones installed on our edge system in 1999. The combination improved the photoelectron signal level by about a factor of 20 and the signal to noise by a factor of 2-8, depending on the absolute signal level. The new cameras also allow shorter minimum integration times while archiving to PC memory: 0.552 ms for the slower, lower-read noise (15 e) readout mode and 0.274 ms in the faster, higher-read noise (30 e) mode

  10. Advanced tokamak physics in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)

    2000-12-01

    Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)

  11. Development of a radiative divertor for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States); Campbell, R.B. [Sandia National Labs., Albuquerque, NM (United States); Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Hill, D.N. [Lawrence Livermore National Lab., CA (United States); Hyatt, A.W. [General Atomics, San Diego, CA (United States); Knoll, D.; Lasnier, C.J. [Lawrence Livermore National Lab., CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Lippmann, S.I. [General Atomics, San Diego, CA (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Maingi, R. [Oak Ridge National Lab., TN (United States); Meyer, W. [Lawrence Livermore National Lab., CA (United States); Moyer, R.A. [California Univ., Los Angeles, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Rensink, M.E. [Lawrence Livermore National Lab., CA (United States); Rognlien, T.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States); Smith, J.P. [General Atomics, San Diego, CA (United States); Staebler, G.M. [General Atomics, San Diego, CA (United States); Stambaugh, R.D. [General Atomics, San Diego, CA (United States); West, W.P. [General Atomics, San Diego, CA (United States); Wood, R.D. [Lawrence Livermore National Lab., CA (United States)

    1995-04-01

    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 ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while {tau}{sub E} remains similar 2 times ITER-89P scaling. However, n{sub e} 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}{approx}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.)).

  12. Plasma diagnostics for the DIII-D divertor upgrade (abstract)

    International Nuclear Information System (INIS)

    Hill, D.N.; Futch, A.; Buchenauer, D.; Doerner, R.; Lehmer, R.; Schmitz, L.; Klepper, C.C.; Menon, M.; Leikind, B.; Lippmann, S.; Mahdavi, M.A.; Schaffer, M.; Smith, J.; Salmonson, J.; Watkins, J.

    1990-01-01

    The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and H α measurements with TV cameras and fiber optics coupled to a high-resolution spectrometer

  13. Disruption Physics and Mitigation on DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Humphreys, D.A.; Kellman, A.G.

    2005-01-01

    The contributions of the DIII-D tokamak toward the understanding and control of disruptions are reviewed. Disruptions are found to be deterministic, and the underlying causes of disruption can therefore be predicted and avoided. With sufficiently rapid detection, possible damage from disruptions can be mitigated using an understanding of disruption phenomenology and plasma physics. Regimes of high β are readily available in DIII-D and provide access to relatively high energy density disruptions, despite DIII-D's moderate magnetic field and size. DIII-D, with all-graphite wall armor and wall conditioning between discharges, has proven highly resilient to the deleterious effects that disruptions can have on plasma operations. Simultaneously, exploitation and adaptation of DIII-D's extensive core and edge plasma diagnostic set have allowed for unique plasma measurements during disruptions. These measurements have tied into the development of several physical models used to understand aspects of disruptions, such as magnetohydrodynamic growth at the disruption onset, radiation energy balance through the thermal quench, and halo currents during the current quench. Based on this fundamental understanding, DIII-D has developed techniques to mitigate the harmful effects of disruptions by radiative dissipation of the plasma energy and extrapolated these techniques for possible use on larger devices like ITER

  14. 110GHz ECH on DIII-D

    International Nuclear Information System (INIS)

    Cary, W.P.; Allen, J.C.; Callis, R.W.; Doane, J.L.; Harris, T.E.; Moetler, C.P.; Neren, A.; Prater, P.; Rensen, D.

    1992-01-01

    This paper reports on a new high power electron cyclotron heating (ECH) system which has been introduced on DIII-D. This system is designed to operate at 110 GHz with a total output power of 2 MW. The system consists of four Varian VGT-8011 gyrotrons (output power of 500 kW), and their associated support equipment. All components have been designed for up to a 10 second pulse duration. The 110 GHz system is intended to further progress in rf current drive experiments on DIII-D when used in conjunction with the existing 60 GHz ECH (1. 6 MW) , and the 30-60 MHz ICH (2MW) systems. H-mode physics, plasma stabilization experiments and transport studies are also to be conducted at 110 GHz

  15. Currents in the DIII-D Tokamak

    Science.gov (United States)

    Azari, A.; Eidietis, N. W.

    2012-10-01

    Loss of vertical control of an elongated tokamak plasma results in a vertical displacement event (VDE) which can induce large currents on open field lines and exert high JxB forces on in-vessel components. An array of first-wall tile current monitors on DIII-D provides direct measurement of the poloidal halo currents. These measurements are analyzed to create a database of halo current magnitude and asymmetry, which are found to lie within the ranges seen by numerous other tokamaks in the ITPA Disruption Database. In addition, an analysis of halo asymmetry rotation is presented, as rotation at the resonance frequencies of in-vessel components could lead to significant amplification of the halo forces. Halo current rotation is found to be far more prevalent in old (1997-2002) DIII-D halo current data than recent data (2009), perhaps due to a change in divertor geometry over that time.

  16. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-04-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix ''strike'' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. 5 refs., 5 figs

  17. Design of DIII-D advanced divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thruston, G.

    1989-01-01

    The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffing. The Advanced Divertor has two principal components: ( 1) a toroidally symmetric baffle; and (2) a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. 2 refs., 4 figs

  18. Design of DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thurston, G.

    1989-11-01

    The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffling. The Advanced Divertor has two principal components: a toroidally symmetric baffle; and a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. All the feeds are supported from and maintain a 5 kV isolation to the vessel wall. 2 refs., 4 figs

  19. Analysis and correction of intrinsic non-axisymmetric magnetic fields in highDIII-D plasmas

    International Nuclear Information System (INIS)

    Garofalo, A.M.; La Haye, R.J.; Scoville, J.T.

    2002-01-01

    Rapid plasma toroidal rotation, sufficient for stabilization of the n=1 resistive wall mode, can be sustained by improving the axisymmetry of the toroidal magnetic field geometry of DIII-D. The required symmetrization is determined experimentally both by optimizing currents in external n=1 correction coils with respect to the plasma rotation, and by use of the n=1 magnetic feedback to detect and minimize the plasma response to non-axisymmetric fields as β increases. Both methods point to an intrinsic ∼7 G (0.03% of the toroidal field), m/n=2/1 resonant helical field at the q=2 surface as the cause of the plasma rotation slowdown above the no-wall β limit. The drag exerted by this field on the plasma rotation is consistent with the behaviour of 'slipping' in a simple induction motor model. (author)

  20. Non-inductive current drive experiments on DIII-D, and future plans

    International Nuclear Information System (INIS)

    Prater, R.; Austin, M.; Baity, F.W.; Callis, R.W.; Chiu, S.C.; DeGrassie, J.S.; Freeman, R.L.; Forest, C.B.; Goulding, R.H.; Harvey, R.W.; Hoffman, D.J.; Ikezi, H.; Lohr, J.; James, R.A.; Kupfer, K.; Lin-Liu, Y.R.; Luce, T.C.; Moeller, C.P.; Petty, C.C.; Pinsker, R.I.; Porkolab, M.; Squire, J.; Trukhin, V.

    1995-01-01

    Experiments on DIII-D (and other tokamaks) have shown that improved performance can follow from optimization of the current density profile. Increased confinement of energy and a higher limit on β have both been found in discharges in which the current density profile is modified through transient means, such as ramping of current or elongation. Peaking of the current distribution to obtain discharges with high internal inductance l i has been found to be beneficial. Alternatively, discharges with broader profiles, as in the VH mode or with high β poloidal, have shown improved performance. Non-inductive current drive is a means to access these modes of improved confinement on a steady state basis. Accordingly, experiments on non-inductive current drive are underway on the DIII-D tokamak using fast waves and electron cyclotron waves. Recent experiments on fast wave current drive have demonstrated the ability to drive up to 180kA of non-inductive current using 1.5MW of power at 60MHz, including the contribution from 1MW of ECCD and the bootstrap current. Higher power r.f. current drive systems are needed to affect strongly the current profile on DIII-D. An upgrade to the fast wave current drive system is underway to increase the total power to 6MW, using two additional antennas and two new 30-120MHz transmitters. Additionally, a 1MW prototype ECH system at 110GHz is being developed (with eventual upgrade to 10MW). With these systems, non-inductive current drive at the 1MA level will be available for experiments on profile control in DIII-D. ((orig.))

  1. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.; Bell, R.E.

    2001-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  2. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.A.; Bell, R.E.

    1999-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  3. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

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

    1999-02-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. The authors have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. They have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. They have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling they have avoided these and other processes and accessed densities > 1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density

  4. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Baylor, L.R.; Jernigan, T.

    2001-01-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. We have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. We have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. We have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling we have avoided these and other processes and accessed densities >1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density. (author)

  5. Investigation of density limit processes in DIII-D

    International Nuclear Information System (INIS)

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

    1999-01-01

    A series of experiments has been conducted in DIII-D to investigate density-limiting processes. We have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. We have shown that the critical density for MARFE onset at low edge temperature scales as I p /a 2 , i.e. similar to Greenwald scaling. We have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass' model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling we have avoided these and other processes and accessed densities > 1.5x Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density. (author)

  6. Development of burning plasma and advanced scenarios in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.

    2005-01-01

    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 ∼ 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. (author)

  7. DIII-D RESEARCH OPERATIONS ANNUAL REPORT TO THE U.S. DEPARTMENT OF ENERGY

    Energy Technology Data Exchange (ETDEWEB)

    EVANS,TE

    2003-12-01

    OAK-B135 The mission of the DIII-D research program is: ''To establish the scientific basis for the optimization of the tokamak approach to fusion energy production. The program is focused on developing the ultimate potential of the tokamak by building a better fundamental understanding of the physics of plasma confinement, stability, current drive and heating in high performance discharges while utilizing new scientific discoveries and improvements in their knowledge of these basic areas to create more efficient control systems, improved plasma diagnostics and to identify new types of enhanced operating regimes with improved stability properties. In recent years, this development path has culminated in the advanced tokamak (AT) approach. An approach that has shown substantial promise for improving both the fusion yield and the energy density of a burning plasma device. While the challenges of increasing AT plasma performance levels with greater stability for longer durations are significant, the DIII-D program has an established plan that brings together both the critical resources and the expertise needed to meet these challenges. The DIII-D research staff is comprised of about 300 individuals representing 60 institutions with many years of integrated research experience in tokamak physics, engineering and technology. The DIII-D tokamak is one of the most productive, flexible and best diagnosed magnetic fusion research devices in the world. It has significantly more flexibility than most tokamaks and continues to pioneer the development of sophisticated new plasma feedback control tools that enable the explorations of new frontiers in fusion science and engineering.

  8. DIII-D RESEARCH OPERATIONS ANNUAL REPORT October 1, 2001 through September 30, 2002

    International Nuclear Information System (INIS)

    EVANS, T.E.

    2003-01-01

    OAK-B135 The mission of the DIII-D research program is: ''To establish the scientific basis for the optimization of the tokamak approach to fusion energy production. The program is focused on developing the ultimate potential of the tokamak by building a better fundamental understanding of the physics of plasma confinement, stability, current drive and heating in high performance discharges while utilizing new scientific discoveries and improvements in their knowledge of these basic areas to create more efficient control systems, improved plasma diagnostics and to identify new types of enhanced operating regimes with improved stability properties. In recent years, this development path has culminated in the advanced tokamak (AT) approach. An approach that has shown substantial promise for improving both the fusion yield and the energy density of a burning plasma device. While the challenges of increasing AT plasma performance levels with greater stability for longer durations are significant, the DIII-D program has an established plan that brings together both the critical resources and the expertise needed to meet these challenges. The DIII-D research staff is comprised of about 300 individuals representing 60 institutions with many years of integrated research experience in tokamak physics, engineering and technology. The DIII-D tokamak is one of the most productive, flexible and best diagnosed magnetic fusion research devices in the world. It has significantly more flexibility than most tokamaks and continues to pioneer the development of sophisticated new plasma feedback control tools that enable the explorations of new frontiers in fusion science and engineering

  9. ICRH coupling in DIII-D

    International Nuclear Information System (INIS)

    Hoffman, D.J.; Baity, F.W.; Bryan, W.E.; Jaeger, E.F.; Owens, T.L.; Remsen, D.B.; Luxon, J.; Rawls, J.M.

    1986-01-01

    A 9-MW ion cyclotron resonant frequency (ICRF) experiment has been proposed to heat the Doublet III-D (DIII-D) plasma. DIII-D is a 2.2-T, 3.5-MA tokamak at GA Technologies with a major radius of 1.67 m and minor radius of 67 cm (elongation approx.2). The device was recommissioned in early 1986. The initial experimental program includes ohmic plasma and neutral beam studies; high-power rf experiments will follow in later years. Compact loop antennas (which fit completely in a 35- by 50-cm port) have been chosen to convey this power because of their inherent ease of maintenance, high efficiency, and versatility. In order to verify that the antenna will have sufficient loading, a prototype low-power (2-MW) antenna has been designed and installed. Measurements will be made through September 1986. The antenna is a cavity antenna that will operate from approximately 30 to 80 MHz with a 50-Ω match for a load resistance of approx.1 Ω. It is surrounded by a fixed graphite-covered frame and can be extended from 3 cm behind this frame to 2 cm in front. This can be used to adjust coupling to the plasma. The electrical, mechanical, and thermal characteristics of this antenna system (and its extrapolation to ignited tokamaks) are discussed. In addition to experimental exploration of coupling, we have investigated wave propagation and absorption in DIII-D by using a cold collisional plasma model in straight tokamak geometry with rotation transform. Loading and power deposition profiles as a function of frequency, density, and species mix are presented

  10. Pedestal performance dependence upon plasma shape in DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Casper, T.A.; Groebner, R.J.; Osborne, T.H.; Snyder, P.B.; Thomas, D.M.

    2007-01-01

    Higher moments of the plasma shape than triangularity are found to significantly affect the pedestal pressure and the edge localized mode (ELM) characteristics in DIII-D. The shape dependence of the pedestal pressure was experimentally examined by varying the squareness in the proposed ITER configuration while holding the triangularity fixed. Over this scan the pedestal pressure increased by ∼50% from highest squareness to lowest squareness. The variation of pedestal energy is found to be consistent with the stability analysis of the measured profiles. The ELM energy also varied with the shape to maintain a nearly constant fraction of the pedestal energy. Stability analysis using model shapes and pressure profiles indicates that much of the advantage of high triangularity for high pedestal pressure can be achieved in lower triangularity shapes by optimizing squareness and/or the distance of the secondary upper separatrix from the primary separatrix. In high beta discharges an increase in pedestal pressure is observed with higher global stored energy. The greatest pedestal pressure increase is at low squareness due to an increase in both the pressure gradient stability limit and the width of the pedestal. The variation in pedestal pressure with squareness was also used to optimize 'hybrid' discharges in DIII-D where a lower pedestal pressure was required for an improved overall performance. In the 'hybrid' regime low squareness resulted in a high pedestal pressure with large infrequent ELMs that eventually triggered an internal 2/1 tearing mode that locked, resulting in a disruption. At higher squareness the pedestal pressure was reduced with smaller and more rapid ELMs, resulting in the maintenance of a steady beneficial internal 3/2 tearing mode and good confinement. For all the cases studied, an increase in the pedestal width at low squareness appears to be a significant factor in the increase in the total pedestal pressure

  11. ACHIEVING AND SUSTAINING STEADY-STATE ADVANCED TOKAMAK CONDITIONS ON DIII-D

    International Nuclear Information System (INIS)

    WADE, MR; MURAKAMI, M; BRENNAN, DP; CASPER, TA; FERRON, JR; GAROFALO, AM; GREENFIELD, CM; HYATT, AW; JAYAKUMAR, R; KINSEY, JE; LAHAYE, RJ; LAO, LL; LAZARUS, EA; LOHR, J; LUCE, TC; PETTY, CC; POLITZER, PA; PRATER, R; STRAIT, EJ; TURNBULL, AD; WATKINS, JG; WEST, WP

    2002-01-01

    Recent experiments on the DIII-D tokamak have demonstrated the feasibility of sustaining advanced tokamak conditions that combine high fusion power density (β > 4%), high bootstrap current fraction (f BS ∼ 65%), and high non-inductive current fractions (f NI ∼ 85%) for several energy confinement times. The duration of such conditions is limited only by resistive relaxation of the current density profile. Modeling studies indicate that the application of off-axis ECCD will be able to maintain a favorable current density profile for several seconds

  12. Achieving and sustaining steady-state advanced tokamak conditions on DIII-D

    International Nuclear Information System (INIS)

    Wade, M.R.; Murakami, M.; Brennan, D.P.

    2003-01-01

    Recent experiments on the DIII-D tokamak have demonstrated the feasibility of sustaining advanced tokamak conditions that combine high fusion power density (β > 4%), high bootstrap current fraction (f BS ∼ 65%), and high non-inductive current fractions (f NI ∼85%) for several energy confinement times. The duration of such conditions is limited only by resistive relaxation of the current density profile. Modeling studies indicate that the application of off-axis ECCD will be able to maintain a favorable current density profile for several seconds. (author)

  13. Advances in integrated plasma control on DIII-D

    International Nuclear Information System (INIS)

    Walker, M.L.; Ferron, J.R.; Hahn, S.H.; Humphreys, D.A.; In, Y.; Johnson, R.D.; Kim, J.S.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Welander, A.S.; Xiao, B.

    2007-01-01

    The DIII-D advanced tokamak physics program requires extremely high performance from the DIII-D plasma control system, including simultaneous accurate regulation of plasma shape, stored energy, density and divertor characteristics, as well as coordinated suppression of magnetohydrodynamic instabilities. To satisfy these demanding control requirements, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another, simulation of control action against off-line and actual machine control platforms, and optimization through iteration of the design-test loop. The present work describes progress in development of physics models and development and experimental application of new model-based plasma controllers on DIII-D. We also describe the development of the control software, hardware, and model-based control algorithms for the superconducting EAST and KSTAR tokamaks

  14. Fast wave current drive on DIII-D

    International Nuclear Information System (INIS)

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

    1995-01-01

    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 γ = 0.4 x 10 18 T eo (keV) [A/m 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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

    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. (paper)

  16. New DIII-D tokamak plasma control system

    International Nuclear Information System (INIS)

    Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C.M.; Pinsker, R.I.; Lazarus, E.A.

    1992-09-01

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

  17. Metastable beta limit in DIII-D

    International Nuclear Information System (INIS)

    La Haye, R.J.; Callen, J.D.; Gianakon, T.A.

    1997-06-01

    The long-pulse, slowly evolving single-null divertor (SND) discharges in DIII-D with H-mode, ELMs, and sawteeth are found to be limited significantly below (factor of 2) the predicted ideal limit β N = 4l i by the onset of tearing modes. The tearing modes are metastable in that they are explained by the neoclassical bootstrap current (high β θ ) destabilization of a seed island which occurs even if Δ' θ , there is a region of the modified Rutherford equation such that dw/dt > 0 for w larger than a threshold value; the plasma is metastable, awaiting the critical perturbation which is then amplified to the much larger saturated island. Experimental results from a large number of tokamaks indicate that the high beta operational envelope of the tokamak is well defined by ideal magnetohydrodynamic (MHD) theory. The highest beta values achieved have historically been obtained in fairly short pulse discharges, often <1-2 sawteeth periods and < 1-2 energy replacement times. The maximum operational beta in single-null divertor (SND), long-pulse discharges in DIII-D with a cross-sectional shape similar to the proposed ITER tokamak is found to be limited significantly below the threshold for ideal instabilities by the onset of resistive MHD instabilities

  18. DIII-D research program progress

    Energy Technology Data Exchange (ETDEWEB)

    Stambaugh, R.D.

    1990-11-01

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

  19. Mitigation of divertor heat flux by high-frequency ELM pacing with non-fuel pellet injection in DIII-D

    Directory of Open Access Journals (Sweden)

    A. Bortolon

    2017-08-01

    Full Text Available Experiments have been conducted on DIII-D investigating high repetition rate injection of non-fuel pellets as a tool for pacing Edge Localized Modes (ELMs and mitigating their transient divertor heat loads. Effective ELM pacing was obtained with injection of Li granules in different H-mode scenarios, at frequencies 3–5 times larger than the natural ELM frequency, with subsequent reduction of strike-point heat flux (Bortolon et al., Nucl. Fus., 56, 056008, 2016. However, in scenarios with high pedestal density (∼6 ×1019m−3, the magnitude of granule triggered ELMs shows a broad distribution, in terms of stored energy loss and peak heat flux, challenging the effectiveness of ELM mitigation. Furthermore, transient heat-flux deposition correlated with granule injections was observed far from the strike-points. Field line tracing suggest this phenomenon to be consistent with particle loss into the mid-plane far scrape-off layer, at toroidal location of the granule injection.

  20. Interpretation of changes in diffusive and non-diffusive transport in the edge plasma during pedestal buildup following a low-high transition in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W. M.; Sayer, M.-H.; Floyd, J.-P. [Georgia Tech, Atlanta, Georgia 30332 (United States); Groebner, R. J. [General Atomics, San Diego, California 92186 (United States)

    2013-01-15

    The evolution of diffusive and non-diffusive transport during pedestal buildup following a low-high (L-H) transition has been interpreted from a particle-momentum-energy balance analysis of the measured density, temperature, and rotation velocity profiles in the plasma edge (0.82<{rho}<1.0) of a DIII-D [Luxon, Nucl. Fusion 42, 614 (2002)] discharge. In the discharge examined, there was an edge-localized-mode-free period of more than 600 ms following the L-H transition, and the majority of edge pedestal development occurred within the first 100 ms following the L-H transition. There appears to be a spatio-temporal correlation among the measured toroidal and poloidal rotation, the formation of a negative well in the measured radial electric field, the creation of a large inward particle pinch, the calculated intrinsic rotation due to ion orbit loss, and the measured formation of steep gradients in density and temperature in the outer region ({rho}>0.95) of the edge pedestal.

  1. Characteristics of the scrape-off layer in DIII-D high-performance negative central magnetic shear discharges

    Energy Technology Data Exchange (ETDEWEB)

    Lasnier, C.J. [General Atomics, San Diego, CA (United States); Maingi, R. [General Atomics, San Diego, CA (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Allen, S.L. [General Atomics, San Diego, CA (United States); Buchenauer, D.A. [General Atomics, San Diego, CA (United States); Burrell, K.H. [General Atomics, San Diego, CA (United States); Casper, T.A. [General Atomics, San Diego, CA (United States); Cuthbertson, J.W. [General Atomics, San Diego, CA (United States); Fenstermacher, M.E. [General Atomics, San Diego, CA (United States); Hill, D.N. [General Atomics, San Diego, CA (United States); Jong, R.A. [General Atomics, San Diego, CA (United States); Lao, L.L. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [General Atomics, San Diego, CA (United States); Moyer, R.A. [General Atomics, San Diego, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [General Atomics, San Diego, CA (United States); Rice, B.W. [General Atomics, San Diego, CA (United States); Stallard, B.W. [General Atomics, San Diego, CA (United States); Taylor, T.S. [General Atomics, San Diego, CA (United States); Watkins, J.G. [General Atomics, San Diego, CA (United States)

    1997-02-01

    In this paper we present measurements of the global power and particle balance in the high-performance phase of negative central magnetic shear (NCS) discharges and compare with reference VH-mode discharges. The principal differences observed are that NCS has a much lower fraction of the total input power flowing into the boundary, less core radiation, and larger rate of stored energy increase as a fraction of total power. Scrape-off layer (SOL) temperature and divertor heat flux profiles, and radiation profiles at the midplane, are similar to VH-mode. Due to the good core particle confinement and efficient fueling by neutral beam injection (NBI), with little gas puffing, the gas load on the walls and the recycling are very low during the NCS discharges. This results in a rate of density rise relative to beam fueling at the L to H transition time which is 1/3 of the value for VH transitions, which is in turn 1/2 that for L-to-ELMing-H-mode transitions. (orig.).

  2. DIII-D DATA MANAGEMENT

    International Nuclear Information System (INIS)

    McHARG, B.B; BURUSS, J.R. Jr.; FREEMAN, J.; PARKER, C.T.; SCHACHTER, J.; SCHISSEL, D.P.

    2001-08-01

    OAK-B135 The DIII-D tokamak at the DIII-D National Fusion Facility routinely acquires ∼ 500 Megabytes of raw data per pulse of the experiment through a centralized data management system. It is expected that in FY01, nearly one Terabyte of data will be acquired. In addition there are several diagnostics, which are not part of the centralized system, which acquire hundreds of megabytes of raw data per pulse. There is also a growing suite of codes running between pulses that produce analyzed data, which add ∼ 10 Megabytes per pulse with total disk usage of about 100 Gigabytes. A relational database system has been introduced which further adds to the overall data load. In recent years there has been an order of magnitude increase in magnetic disk space devoted to raw data and a Hierarchical Storage Management system (HSM) was implemented to allow 7 x 24 unattended access to raw data. The management of all of the data is a significant and growing challenge as the quantities of both raw and analyzed data are expected to continue to increase in the future. This paper will examine the experiences of the approaches that have been taken in management of the data and plans for the continued growth of the data quantity

  3. Characterization of wall conditions in DIII-D

    International Nuclear Information System (INIS)

    Holtrop, K.L.; Jackson, G.L.; Kellman, A.G.; Lee, R.L.; West, W.P.; Wood, R.D.; Whyte, D.G.

    1996-10-01

    Wall conditioning in DIII-D is one of the most important factors in achieving reproducible high confinement discharges. For example, the very high confinement mode (VH-mode) was only discovered after boronization, a CVD technique to deposit a thin boron film over the entire surface of the tokamak. In order to evaluate wall conditions and provide a data base to correlate these wall conditions with tokamak discharge performance, a series of nominally identical reference VH-mode discharges (1.6 MA, 2.1 T, double-null diverted) were taken at various times during a series of experimental campaigns with evolving wall conditions. These reference discharges have allowed a quantitative determination of how the wall conditions have evolved. For instance, core carbon and oxygen levels in the VH-mode phase remains at historically low levels during the 1995 run year and there was also a steady decrease in the oxygen levels at plasma initiation during this period. The authors discuss the long term changes in low Z impurities and the effect of wall conditioning techniques such as boronization and baking on these impurities. In addition, the evolution of the deuterium recycling rates will be discussed

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

    International Nuclear Information System (INIS)

    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

  5. Plasma boundary experiments on DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, P.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.K.; Owen, L.; Matthews, G.

    1990-01-01

    A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p T < or approx.10.7%, P(auxiliary)< or approx.20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions. (orig.)

  6. Plasma boundary experiments on DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, T.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.; Owen, L.; Matthews, G.

    1990-06-01

    A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p approx-lt 3 MA, β T approx-lt 10.7%, P(auxiliary) approx-lt 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma surface interactions. 41 refs., 8 figs., 1 tab

  7. Alfven Eigenmode Control in DIII-D

    Science.gov (United States)

    Hu, W.; Olofsson, E.; Welander, A.; van Zeeland, M.; Collins, C.; Heidbrink, W.

    2017-10-01

    Alfven eigenmodes (AE) driven by fast ions from neutral beam and ion cyclotron heating are common in present day tokamak plasmas and are expected to be destabilized by alpha particles in future burning plasma experiments. Because these waves have been shown to cause loss and redistribution of fast ions which can impact plasma performance and potentially device integrity, developing control techniques for AEs is of paramount importance. In the DIII-D plasma control system, spectral analysis of real-time ECE data is used as a monitor of AE amplitude, frequency, and location. These values are then used for feedback control of the neutral beam power to control Alfven waves and reduce fast ion loss. This work describes tests of AE control experiments in the current ramp up phase, during which multiple Alfven eigenmodes are typically unstable and fast ion confinement is degraded significantly. Comparisons of neutron emission and confined fast ion profiles with and without active AE control will be made. Work supported by the U.S. Dept. of Energy under Award Number DE-FC02-04ER54698.

  8. Advanced tokamak research in DIII-D

    International Nuclear Information System (INIS)

    Greenfield, C M; Murakami, M; Ferron, J R

    2004-01-01

    Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with β N ≤ 3.5 and β T ≤ 3.6% sustained for up to 1 s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by

  9. Boundary perturbations coupled to core 3/2 tearing modes on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Tobias, B; Yu, L; Domier, C W; Luhmann, N C Jr; Austin, M E; Paz-Soldan, C; Turnbull, A D; Classen, I G J

    2013-01-01

    High confinement (H-mode) discharges on the DIII-D tokamak are routinely subject to the formation of long-lived, non-disruptive magnetic islands that degrade confinement and limit fusion performance. Simultaneous, 2D measurement of electron temperature fluctuations in the core and edge regions allows for reconstruction of the radially resolved poloidal mode number spectrum and phase of the global plasma response associated with these modes. Coherent, n = 2 excursions of the plasma boundary are found to be the result of coupling to an n = 2, kink-like mode which arises locked in phase to the 3/2 island chain. This coupling dictates the relative phase of the displacement at the boundary with respect to the tearing mode. This unambiguous phase relationship, for which no counter-examples are observed, is presented as a test for modeling of the perturbed fields to be expected outside the confined plasma. (paper)

  10. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.; Buzhinskij, O.I.; Opimach, I.V.

    1998-08-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point (OSP) of two divertor plasma conditions: attached (T e > 40 eV) ELMing plasmas, and detached (T e 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood. In the attached cases, physical sputtering (with enhancement from self-sputtering and oblique incidence) is dominant, and the effective sputtering yield, Y, is greater than 10%. In ELM-free discharges, the total OSP net erosion rate is equal to the rate of carbon accumulation in the core plasma. For the detached divertor cases, the cold incident plasma eliminates physical sputtering. Attempts to measure chemically eroded hydrocarbon molecules spectroscopically indicate an upper limit of Y ≤ 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/exposure-year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates at the OSP of an attached plasma (∼ 10 microm/s > 1,000x erosion rate of aligned surfaces). Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  11. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

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

    1998-05-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point of two divertor plasma conditions: (1) attached (Te > 40 eV) ELMing plasmas and (2) detached (Te 10 cm/year, even with incident heat flux 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood and that effective sputtering yields are > 10%. In ELM-free discharges, this erosion rate can account for the rate of carbon accumulation in the core plasma. Divertor plasma detachment eliminates physical sputtering, while spectroscopically measured chemical erosion yields are also found to be low (Y(C/D + ) ≤ 2.0 x 10 -3 ). This leads to suppression of net erosion at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates (∼ 10 microm/s) at the OSP of an attached plasma. Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  12. The DIII-D cryogenic system upgrade

    International Nuclear Information System (INIS)

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

    1993-10-01

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

  13. The DIII-D cryogenic system upgrade

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-10-01

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

  14. The upgrade of the DIII-D EC system using 120 GHz ITER gyrotrons

    International Nuclear Information System (INIS)

    Callis, R.W.; Lohr, J.; Gorelov, I.A.; Ponce, D.; Kajiwara, K.; Tooker, J.F.

    2005-01-01

    The planned growth in the EC system on DIII-D over the next few years requires the installation of two depressed collector gyrotrons, a high voltage power supply, two low loss transmission lines, and the required support equipment. This new DIII-D EC equipment could be made identical to the ITER EC system requirements. By building the DIII-D hardware to the ITER specifications, it will allow ITER to gain beneficial prototyping experience on a working tokamak, prior to committing to building the hardware for delivery to ITER

  15. Simulation of DIII-D Flat q Discharges

    International Nuclear Information System (INIS)

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

    2004-01-01

    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 β, 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 min remains high and at large radius, ρ ∼ 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

  16. Demonstration of ITER operational scenarios on DIII-D

    International Nuclear Information System (INIS)

    Doyle, E.J.; DeBoo, J.C.; Ferron, J.R.; Jackson, G.L.; Luce, T.C.; Osborne, T.H.; Politzer, P.A.; Groebner, R.J.; Hyatt, A.W.; La Haye, R.J.; Petrie, T.W.; Petty, C.C.; Murakami, M.; Park, J.-M.; Reimerdes, H.; Budny, R.V.; Casper, T.A.; Holcomb, C.T.; Challis, C.D.; McKee, G.R.

    2010-01-01

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

  17. Demonstration of ITER Operational Scenarios on DIII-D

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  18. Density limit studies on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R. [Oak Ridge National Lab., TN (United States); Mahdavi, M.A.; Petrie, T.W. [General Atomics, San Diego, CA (United States)] [and others

    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}.

  19. Density limit studies on DIII-D

    International Nuclear Information System (INIS)

    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 ∼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 e sep,det ) at detachment onset (partial detachment) increases with the scrape-off layer heating power, P heat , i.e., n e sep,det ∼ P heat 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 heat 0.7

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

    International Nuclear Information System (INIS)

    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, βτ 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

  1. Studies of high-δ (baffled) and low-δ (open) pumped divertor operation on DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Fenstermacher, M.E.; Greenfield, C.M.

    1998-08-01

    The authors report new experimental results with the RDP-OB (Radiative Divertor Project-outer baffle) and cryopump in both upper single-null (USN) and double-null (DN) ELMing H-mode discharges. The baffled divertor reduced the core ionization (∼2--2.5x), in reasonable agreement with predictions from UEDGE/DEGAS modeling (∼3.75x). The upper cryopump achieved density control of n e /n gw ∼ 0.22 (line density/Greenwald density) with Z eff ∼ 2 in high-δ plasmas. The measured exhaust is comparable to the lower pump, except at lower core electron densities (n e 19 m -3 ). Efficient impurity exhaust was obtained with deuterium SOL flow. Preliminary experiments with DN operation has shown that the particle exhaust to the upper pump depends on the up/down magnetic balance. Preliminary experiments indicate that the DN exhaust is roughly 40--50% of the USN exhaust at n e ∼ 4 x 10 19 m -3

  2. Computer control of the high-voltage power supply for the DIII-D Electron Cyclotron Heating system

    International Nuclear Information System (INIS)

    Clow, D.D.; Kellman, D.H.

    1991-10-01

    The D3-D Electron Cyclotron Heating (ECH) high voltage power supply is controlled by a computer. Operational control is input via keyboard and mouse, and computer/power supply interface is accomplished with a Computer Assisted Monitoring and Control (CAMAC) system. User-friendly tools allow the design and layout of simulated control panels on the computer screen. Panel controls and indicators can be changed, added or deleted, and simple editing of user-specific processes can quickly modify control and fault logic. Databases can be defined, and control panel functions are easily referred to various data channels. User-specific processes are written and linked using Fortran, to manage control and data acquisition through CAMAC. The resulting control system has significant advantages over the hardware it emulates: changes in logic, layout, and function are quickly and easily incorporated; data storage, retrieval, and processing are flexible and simply accomplished, physical components subject to wear and degradation are minimized. In addition, the system can be expanded to multiplex control of several power supplied, each with its own database, through a single computer and console. 5 refs., 4 figs., 1 tab

  3. Increasing the Tokamak Pressure Limit: Tearing Mode Experiments in DIII-D

    International Nuclear Information System (INIS)

    La Haye, R.J.

    2005-01-01

    Since its reconfiguration in 1986, DIII-D has performed a number of experiments involving resistive magnetohydrodynamic (MHD) stability. These were and are directed to understand the conditions in which confinement and beta reducing tearing mode islands form, how to avoid them, and if unavoidable, how to stabilize them. Coils for correction of toroidal nonaxisymmetry have been developed to avoid error field locked mode islands. Basic classical tearing mode stability physics has been confirmed with a state-of-the-art ensemble of profile diagnostics, MHD equilibrium reconstruction, and stability code analysis. Neoclassical tearing mode thresholds and seeding are now much better understood with future large higher field devices expected to be 'metastable'. DIII-D is the leader in sophisticated real-time alignment of stabilizing electron cyclotron current drive on otherwise unstable rational surfaces. In all, DIII-D experiments are showing how higher stable beta with good confinement can be maintained without tearing mode islands limiting the plasma pressure

  4. Features and Initial Results of the DIII-D Advanced Tokamak Radiative Divertor

    International Nuclear Information System (INIS)

    R.C. O'Neill; A.S. Bozek; M.E. Friend; C.B. Baxi; E.E. Reis; M.A. Mahdavi; D.G. Nilson; S.L. Allen; W.P. West

    1999-01-01

    The Radiative Divertor Program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryobelium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma. The baffles, which consist of water-cooled panels that allow for the attachment of tiles of various sizes and shapes, house gas puff systems. The intent of the puffing systems is to inject gas in and around the divertor to minimize the heat flux on specific areas on the divertor and its components. The reduction of the heat flux on the divertor minimizes the impurities that are generated from excess heat on divertor components, specifically tiles. Experiments involving the gas puff systems and the divertor structures have shown the heat flux can be spread over a large area of the divertor, reducing the peak heat flux in specific areas. The three diverters also incorporate a variety of diagnostic tools such as halo current monitors, magnetic probes and thermocouples to monitor certain plasma characteristics as well as determine the effectiveness of the cryopumps and baffle configurations. The diverters were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo currents and toroidal induced currents. Incorporated also

  5. CURRENT DRIVE AND PRESSURE PROFILE MODIFICATION WITH ELECTRON CYCLOTRON POWER IN DIII-D QUIESCENT DOUBLE BARRIER EXPERIMENTS

    International Nuclear Information System (INIS)

    CASPER, TA; BURRELL, KH; DOYLE, EJ; GOHIL, P; GREENFIELD, CM; GROEBNER, RJ; JAYAKUMAR, RJ; MAKOWSKI, MA; RHODES, TL; WEST, WP

    2003-01-01

    OAK-B135 High confinement mode (H-mode) operation is a leading scenario for burning plasma devices due to its inherently high energy-confinement characteristics. The quiescent H-mode (QH-mode) offers these same advantages with the additional attraction of more steady edge conditions where the highly transient power loads due to edge localized mode (ELM) activity is replaced by the steadier power and particle losses associated with an edge harmonic oscillation (EHO). With the addition of an internal transport barrier (ITB), the capability is introduced for independent control of both the edge conditions and the core confinement region giving potential control of fusion power production for an advanced tokamak configuration. The quiescent double barrier (QDB) conditions explored in DIII-D experiments exhibit these characteristics and have resulted in steady plasma conditions for several confinement times (∼ 26 τ E ) with moderately high stored energy, β N H 89 ∼ 7 for 10 τ E

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

    International Nuclear Information System (INIS)

    PETTY, C.C.; PRATER, R.; LUCE, T.C.; ELLIS, R.A.; HARVEY, R.W.; KINSEY, J.E.; LAO, L.L.; LOHR, J.; MAKOWSKI, M.A.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    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.

    2015-01-01

    Recent efforts have shown that helicon waves (fast waves at > 20ω 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lau, C., E-mail: lauch@ornl.gov; Blazevski, D.; Green, D. L.; Murakami, M.; Park, J. M. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN (United States); Jaeger, E. F.; Berry, L. A. [XCEL Engineering, Inc., 1066 Commerce Park Dr., Oak Ridge, TN (United States); Bertelli, N. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Pinsker, R. I.; Prater, R. [General Atomics, San Diego, CA (United States)

    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.

  9. Dust Studies in DIII-D and TEXTOR

    International Nuclear Information System (INIS)

    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.

    2009-01-01

    Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. 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 ∼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 (∼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.

  10. VUV Spectroscopy in DIII-D Divertor

    International Nuclear Information System (INIS)

    Alkesh Punjabi; Nelson Jalufka

    2004-01-01

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

  11. INTEGATED ADVANCED TOKAMAK OPERATION ON DIII-D

    International Nuclear Information System (INIS)

    WADE, M.R.; MURAKAMI, M.; LUCE, T.C.; FERRON, J.R.; PETTY, C.C.; BRENNEN, D.P.; GAROFALO, A.M.; GREENFIELD, C.M.; HYATT, A.W.; JAYAKUMAR, R.; KINSEY, J.E.; La HAYE, R.J.; LAO, L.L.; LOHR, J.; POLITZER, P.A.; PRATER, R.; STRAIT, E.J.; WATKINS, J.G.

    2002-01-01

    Recent experiments on DIII-D have demonstrated the ability to sustain plasma conditions that integrate and sustain the key ingredients of Advanced Tokamak (AT) operation: high β with 1.5 min min > 2.0, plasmas with β ∼ 2.9% and 90% of the plasma current driven non-inductively have been sustained for nearly 2 s (limited only by the duration of the ECCD pulse). Negative central magnetic shear is produced by the ECCD, leading to the formation of a weak internal transport barrier even in the presence of Type I ELMs. Separate experiments have demonstrated the ability to sustain a steady current density profile using ECCD for periods as long as 1 s with β = 3.3% and > 90% of the current driven non-inductively. In addition, stable operation well above the ideal no-wall β limit has been sustained for several energy confinement times with the duration only limited by resistive relaxation of the current profile to an unstable state. Stability analysis indicates that the experimental β limit depends on the degree to which the no-wall limit can be exceeded and weakly on the actual no-wall limit. Achieving the necessary density levels required for adequate ECCD efficiency requires active divertor exhaust and reducing the wall inventory buildup prior to the high performance phase. Simulation studies indicate that the successful integration of high β operation with current profile control consistent with these experimental results should result in high β, fully non-inductive plasma operation

  12. Neutral beam current drive scaling in DIII-D

    International Nuclear Information System (INIS)

    Porter, G.D.; Bhadra, D.K.; Burrell, K.H.

    1989-03-01

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

  13. SELF-CONSISTENT,INTEGRATED,ADVANCED TOKAMAK OPERATION ON DIII-D

    International Nuclear Information System (INIS)

    WADE, MR; MURAKAMI, M; LUCE, TC; FERRON, JR; PETTY, CC; BRENNAN, DP; GAROFALO, AM; GREENFIELD, CM; HYATT, AW; JAYAKUMAR, R; LAHAYE, RJ; LAO, LL; LOHR, J; POLITZER, PA; PRATER, R; STRAIT, EJ

    2002-01-01

    Recent experiments on DIII-D have demonstrated the ability to sustain plasma conditions that integrate and sustain the key ingredients of Advanced Tokamak (AT) operation: high β with q min >> 1, good energy confinement, and high current drive efficiency. Utilizing off-axis (ρ 0.4) electron cyclotron current drive (ECCD) to modify the current density profile in a plasma operating near the no-wall ideal stability limit with q min > 2.0, plasmas with β = 2.9% and 90% of the plasma current driven non-inductively have been sustained for nearly 2 s (limited only by the duration of the ECCD pulse). Separate experiments have demonstrated the ability to sustain a steady current density profile using ECCD for periods as long as 1 s with β = 3.3% and > 90% of the current driven non-inductively

  14. Dust Studies in DIII-D and TEXTOR

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  16. Understanding and Control of Transport in Advanced Tokamak Regimes in DIII-D

    International Nuclear Information System (INIS)

    C.M. Greenfield; J.C. DeBoo; T.C. Luce; B.W. Stallard; E.J. Synakowski; L.R. Baylor; K.H. Burrell; T.A. Casper; E.J. Doyle; D.R. Ernst; J.R. Ferron; P. Gohil; R.J. Groebner; L.L. Lao; M. Makowski; G.R. McKee; M. Murakami; C.C. Petty; R.I. Pinsker; P.A. Politzer; R. Prater; C.L. Rettig; T.L. Rhodes; B.W. Rice; G.L. Schmidt; G.M. Staebler; E.J. Strait; D.M. Thomas; M.R. Wade

    1999-01-01

    Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection takes place when the heating power exceeds a threshold value of about 9 MW, contrasting to CO-NBI injection, where P threshold N H 89 = 9 for 16 confinement times has been accomplished in a discharge combining an ELMing H-mode edge and an ITB, and exhibiting ion thermal transport down to 2-3 times neoclassical. The microinstabilities usually associated with ion thermal transport are predicted stable, implying that another mechanism limits performance. High frequency MHD activity is identified as the probable cause

  17. Experiment and Modeling of ITER Demonstration Discharges in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Park, Jin Myung; Doyle, E. J.; Ferron, J.R.; Holcomb, C.T.; Jackson, G.L.; Lao, L.L.; Luce, T.C.; Owen, Larry W.; Murakami, Masanori; Osborne, T.H.; Politzer, P.A.; Prater, R.; Snyder, P.B.

    2011-01-01

    DIII-D is providing experimental evaluation of 4 leading ITER operational scenarios: the baseline scenario in ELMing H-mode, the advanced inductive scenario, the hybrid scenario, and the steady state scenario. The anticipated ITER shape, aspect ratio and value of I/αB were reproduced, with the size reduced by a factor of 3.7, while matching key performance targets for β N and H 98 . Since 2008, substantial experimental progress was made to improve the match to other expected ITER parameters for the baseline scenario. A lower density baseline discharge was developed with improved stationarity and density control to match the expected ITER edge pedestal collisionality (ν* e ∼ 0.1). Target values for β N and H 98 were maintained at lower collisionality (lower density) operation without loss in fusion performance but with significant change in ELM characteristics. The effects of lower plasma rotation were investigated by adding counter-neutral beam power, resulting in only a modest reduction in confinement. Robust preemptive stabilization of 2/1 NTMs was demonstrated for the first time using ECCD under ITER-like conditions. Data from these experiments were used extensively to test and develop theory and modeling for realistic ITER projection and for further development of its optimum scenarios in DIII-D. Theory-based modeling of core transport (TGLF) with an edge pedestal boundary condition provided by the EPED1 model reproduces T e and T i profiles reasonably well for the 4 ITER scenarios developed in DIII-D. Modeling of the baseline scenario for low and high rotation discharges indicates that a modest performance increase of ∼ 15% is needed to compensate for the expected lower rotation of ITER. Modeling of the steady-state scenario reproduces a strong dependence of confinement, stability, and noninductive fraction (f NI ) on q 95 , as found in the experimental I p scan, indicating that optimization of the q profile is critical to simultaneously achieving the

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

    International Nuclear Information System (INIS)

    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.

    2006-01-01

    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 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 min by 5-20 ms. There is no significant low-frequency magnetohydrodynamic activity prior to or just after the crossing of the integer q 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

  19. Quiescent double barrier regime in the DIII-D tokamak.

    Science.gov (United States)

    Greenfield, C M; Burrell, K H; DeBoo, J C; Doyle, E J; Stallard, B W; Synakowski, E J; Fenzi, C; Gohil, P; Groebner, R J; Lao, L L; Makowski, M A; McKee, G R; Moyer, R A; Rettig, C L; Rhodes, T L; Pinsker, R I; Staebler, G M; West, W P

    2001-05-14

    A new sustained high-performance regime, combining discrete edge and core transport barriers, has been discovered in the DIII-D tokamak. Edge localized modes (ELMs) are replaced by a steady oscillation that increases edge particle transport, thereby allowing particle control with no ELM-induced pulsed divertor heat load. The core barrier resembles those usually seen with a low (L) mode edge, without the degradation often associated with ELMs. The barriers are separated by a narrow region of high transport associated with a zero crossing in the E x B shearing rate.

  20. Disruption studies in DIII-D

    International Nuclear Information System (INIS)

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

    1996-09-01

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

  1. Developing physics basis for the snowflake divertor in the DIII-D tokamak

    Science.gov (United States)

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

    2018-03-01

    Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard

  2. DIII-D edge physics database

    International Nuclear Information System (INIS)

    Jong, R.A.; Porter, G.D.; Hill, D.N.; Buchenauer, D.A.; Bramson, G.

    1992-03-01

    We have developed an edge-physics database containing data for the plasma in the divertor region and the scrape-off layer (SOL) for the DIII-D tokamak. The database provides many of the parameters necessary to model the power flow to the divertor and other plasma processes in the plasma edge. It will also facilitate the analysis of DIII-D data for comparison with other divertor tokamaks. In addition to the core plasma parameters, edge-specific data are included in this database. Initial results using the database show good agreement between the pressure profiles measured by the Langmuir probes and those determined from the Thomson data for the inner strike point, but not for the outer strike point region. We also find that the ratio of separatrix density to average core density, as well as the in/out asymmetry in the SOL power at the divertor in DIII-D do not agree with values currently assumed in modeling the International Thermonuclear Experimental Reactor (ITER)

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

    International Nuclear Information System (INIS)

    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 -2 ). Results of measurements made on the pumping characteristics for D 2 , H 2 , and Ar are discussed

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

    International Nuclear Information System (INIS)

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

    2004-03-01

    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 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

  5. Advances in Integrated Plasma Control on DIII-D

    International Nuclear Information System (INIS)

    Walker, M.L.; Ferron, J.R.; Humphreys, D.A.

    2006-01-01

    The DIII-D experimental program in advanced tokamak (AT) physics requires extremely high performance from the DIII-D plasma control system (PCS) [B.G.Penaflor, et al., 4 th IAEA Tech. Mtg on Control and Data Acq., San Diego, CA (2003)], including simultaneous and highly accurate regulation of plasma shape, stored energy, density, and divertor characteristics, as well as coordinated suppression of magnetohydrodynamic instabilities. To satisfy these demanding control requirements, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and optimization through iteration of the design-test loop. The present work describes progress in development of physics models and development and experimental application of several new model-based plasma controllers on DIII-D. We discuss experimental use of advanced shape control algorithms containing nonlinear techniques for improving control of steady state plasmas, model-based controllers for optimal rejection of edge localized mode disturbances during resistive wall mode stabilization, model-based controllers for neoclassical tearing mode stabilization, including methods for maximizing stabilization effectiveness with substantial constraints on available power, model-based integrated control of plasma rotation and beta, and initial experience in development of model-based controllers for advanced tokamak current profile modification. The experience gained from DIII-D has been applied to the development of control systems for the EAST and KSTAR tokamaks. We describe the development of the control software, hardware, and model-based control algorithms for these superconducting tokamaks, with emphasis on relevance of

  6. ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

    International Nuclear Information System (INIS)

    HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

    2002-01-01

    A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  9. Deposition of deuterium and metals on divertor tiles in the DIII--D tokamak

    International Nuclear Information System (INIS)

    Walsh, D.S.; Doyle, B.L.; Jackson, G.L.

    1992-01-01

    Hydrogen recycling and impurity influx are important issues in obtaining high confinement discharges in the DIII--D tokamak. To reduce metallic impurities in DIII--D, 40% of the wall area, including the highest heat flux zones, have been covered with graphite tiles. However, erosion, redeposition, and hydrogen retention in the tiles, as well as metal transport from the remaining Inconel walls, can lead to enhanced recycling and impurity influx. Hydrogen and metal retention in divertor floor tiles have been measured using external ion beam analysis techniques following four campaigns where tiles were exposed to several thousand tokamak discharges. The areal density of deuterium retained following exposure to tokamak plasmas was measured with external nuclear reaction analysis. External proton-induced x-ray emission analysis was used to measure the areal densities of metallic impurities deposited upon the divertor tiles either by sputtering of metallic components during discharges or as contamination during tile fabrication. Measurements for both deuterium and metallic impurities were taken on both the tile surfaces which face the operating plasma and the surfaces on the sides of the tiles which form the small gaps separating each of the tiles in the divertor. The highest areal densities of both deuterium (from 2 to 8 x 10 18 atoms/cm 2 ) and metals (from 0.2 to 1 x 10 18 atoms/cm 2 ) were found on the plasma-facing surface near the inner strike point region of each set of divertor tiles. Significant deposits, extending as far as 1 cm from the plasma-facing surface and containing up to 40% of the total divertor deposition, were also observed on the gap-forming surfaces of the tiles

  10. ELECTRON CYCLOTRON CURRENT DRIVE IN DIII-D: EXPERIMENT AND THEORY

    International Nuclear Information System (INIS)

    PRATER, R; PETTY, CC; LUCE, TC; HARVEY, RW; CHOI, M; LAHAYE, RJ; LIN-LIU, Y-R; LOHR, J; MURAKAMI, M; WADE, MR; WONG, K-L

    2003-01-01

    A271 ELECTRON CYCLOTRON CURRENT DRIVE IN DIII-D: EXPERIMENT AND THEORY. Experiments on the DIII-D tokamak in which the measured off-axis electron cyclotron current drive has been compared systematically to theory over a broad range of parameters have shown that the Fokker-Planck code CQL3D provides an excellent model of the relevant current drive physics. This physics understanding has been critical in optimizing the application of ECCD to high performance discharges, supporting such applications as suppression of neoclassical tearing modes and control and sustainment of the current profile

  11. Structural design of the DIII-D radiative divertor

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  12. Reduction of recycling in DIII-D by degassing and conditioning of the graphite tiles

    International Nuclear Information System (INIS)

    Jackson, G.L.; Taylor, T.S.; Allen, S.L.

    1988-05-01

    Reduced recycling, reduced edge neutral pressure, improved density control, and improved discharge reproducibility have been achieved in the DIII-D tokamak by in situ helium conditioning of the graphite tiles. An improvement in energy confinement has been observed in hydrogen discharges with hydrogen beam injection after helium preconditioning. After the graphite wall coverage in DIII-D was increased to 40%, helium glow wall conditioning, routinely applied before each tokamak discharge, has been necessary to reduce recycling and obtain H-mode. The utilization of helium glow wall conditioning was an important factor in the achievement of an ohmic H-mode, i.e. no auxillary heating, with significant improvement in ohmic energy confinement. 16 refs., 8 figs

  13. Automated Calculation of DIII-D Neutral Beam Availability

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  14. BETA SCALING OF TRANSPORT ON THE DIII-D TOKAMAK: IS TRANSPORTELECTROSTATIC OR ELECTROMAGNETIC?

    International Nuclear Information System (INIS)

    PETTY, C.C; LUCE, T.C; McDONALD, D.C; MANDREKAS, J; WADE, M.R; CANDY, J; CORDEY, J.G; DROZDOV, V; EVANS, T.E; FERRON, J.R; GROEBNER, R.J; HYATT, A.W; JACKSON, G.L; LA HAYE, R.J; OSBORNE, T.H; WALTZ, R.E.

    2003-01-01

    Determining the scaling of transport with (β), the ratio of the plasma kinetic pressure to the magnetic pressure, helps to differentiate between various proposed theories of turbulent transport since mechanisms that are primarily electrostatic show little change in transport with increasing β, while primarily electromagnetic mechanisms generally have a strong unfavorable β scaling. Experiments on the DIII-D tokamak have measured the β scaling of heat transport with all of the other dimensionless parameters held constant in high confinement mode (H-mode) plasmas with edge localized modes (ELMs). A four point scan varied β from 30% to 85% of the ideal ballooning stability limit (normalized beta from 1.0 to 2.8) and found no change in the normalized confinement time, i.e., Bτ th ∞ β -0.01 ± 0.09. The measured thermal diffusivities, normalized to the Bohm diffusion coefficient, also did not vary during the β can to within the experimental uncertainties, whereas the normalized helium particle transport decreased with increasing β. The H-mode pedestal β varied in concert with the core β and showed no signs of saturation. This weak, possibly non-existent, β scaling of transport favors primarily electrostatic mechanisms such as E x B transport, and is in marked disagreement with the strong unfavorable β dependence contained in empirical scaling relations derived from multi-machine H-mode confinement databases

  15. Fast wave current drive on DIII-D

    International Nuclear Information System (INIS)

    deGrassie, J.S.; Petty, C.C.; Pinsker, R.I.; Forest, C.B.; Ikezi, H.; Prater, R.; Baity, F.W.; Callis, R.W.; Cary, W.P.; Chiu, S.C.; Doyle, E.J.; Ferguson, S.W.; Hoffman, D.J.; Jaeger, E.F.; Kim, K.W.; Lee, J.H.; Lin-Liu, Y.R.; Murakami, M.; ONeill, R.C.; Porkolab, M.; Rhodes, T.L.; Swain, D.W.

    1996-01-01

    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 γ=0.4x10 18 T e0 (keV) [A/m 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 clear evidence for a toroidally directed wave with antenna phasing set for current drive. copyright 1996 American Institute of Physics

  16. Recent results from DIII-D and future plans

    International Nuclear Information System (INIS)

    Simonen, T.

    1992-01-01

    This paper summarizes recent DIII-D tokamak experimental results, describes new hardware being implemented to carry out the DIII-D 1990's tokamak research program, and discusses their implications for engineering designs for next generation tokamaks, such as ITER

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

    International Nuclear Information System (INIS)

    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 inclose 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 β 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 DT = 0.32

  18. Joint DIII-D/EAST Experiments Toward Steady State AT Demonstration

    Science.gov (United States)

    Garofalo, A. M.; Meneghini, O.; Staebler, G. M.; van Zeeland, M. A.; Gong, X.; Ding, S.; Qian, J.; Ren, Q.; Xu, G.; Grierson, B. A.; Solomon, W. M.; Holcomb, C. T.

    2015-11-01

    Joint DIII-D/EAST experiments on fully noninductive operation at high poloidal beta have demonstrated several attractive features of this regime for a steady-state fusion reactor. Very large bootstrap fraction (>80 %) is desirable because it reduces the demands on external noninductive current drive. High bootstrap fraction with an H-mode edge results in a broad current profile and internal transport barriers (ITBs) at large minor radius, leading to high normalized energy confinement and high MHD stability limits. The ITB radius expands with higher normalized beta, further improving both stability and confinement. Electron density ITB and large Shafranov shift lead to low AE activity in the plasma core and low anomalous fast ion losses. Both the ITB and the current profile show remarkable robustness against perturbations, without external control. Supported by US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466 & DE-AC52-07NA27344 & by NMCFSP under contracts 2015GB102000 and 2015GB110001.

  19. Global Alfven Eigenmodes in DIII-D

    International Nuclear Information System (INIS)

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

    1992-06-01

    Global Alfven modes, such as the Toroidicity-Induced Alfven Eigenmode (TAE), pose a serious threat for strongly-heated tokamaks since they can result in saturation of the achievable beam β at moderate levels and they may also cause serious α-particle losses in future ignited devices. The DIII-D tokamak has a unique capability for study of the resonant excitation of these instabilities by energetic beam ions. TAE modes have now been observed in DIII-D over a wide range of operating conditions, including both circular cross-section and elongated (κ ∼ 1.8) discharges. Equilibrium reconstructions of several representative discharges, using all available external magnetic and internal profile data, have been done and analyzed in detail. The computed real mode frequencies of the TAE modes are in good agreement with the experimentally observed mode frequencies and differ significantly from the estimated kinetic ballooning mode frequencies. The TAE calculations include coupling to the Alfven and acoustic continuum branches of the MHD spectrum and generally indicate that the simplified circular cross-section, large aspect-ratio assumptions made in analytic calculations are poor approximations to the actual TAE mode structures. In particular, the global TAE modes are almost always coupled to one or more continuum branches by toroidicity, poloidal shaping, and finite β effects. Estimates of the various resonant excitation and damping mechanisms, including continuum damping, have been made and the total is found to be in reasonable agreement with the experimental threshold

  20. Review of DIII-D H-Mode Density Limit Studies

    International Nuclear Information System (INIS)

    Maingi, R.; Mahdavi, M.A.

    2005-01-01

    Density limit studies over the past 10 yr on DIII-D have successfully identified several processes that limit plasma density in various operating modes. The recent focus of these studies has been on maintenance of the high-density operational window with good H-mode level energy confinement. We find that detachment and onset of multifaceted axisymmetric radiation from the edge (MARFE), fueling efficiency, particle confinement, and magnetohydrodynamic activity can impose density limits in certain regimes. By studying these processes, we have devised techniques with either pellets or gas fueling and divertor pumping to achieve line average density above Greenwald scaling, relying on increasing the ratio of pedestal to separatrix density, as well as density profile peaking. The scaling of several of these processes to next-step devices (e.g., the International Thermonuclear Experimental Reactor) has indicated that sufficiently high pedestal density can be achieved with conventional fueling techniques while ensuring divertor partial detachment needed for heat flux reduction. One density limit process requiring further study is neoclassical tearing mode (NTM) onset, and techniques for avoidance/mitigation of NTMs need additional development in present-day devices operated at high density

  1. RECENT DEVELOPMENTS ON THE 110 GHz ELECTRON CYCLOTRON INSTATLLATION ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PONCE, D.; CALLIS, R.W.; CARY, W.P.; FERRON, J.R.; GREEN, M.; GRUNLOH, H.J.; GORELOV, Y.; LOHR, J.; ELLIS, R.A.

    2002-01-01

    OAK A271 RECENT DEVELOPMENTS ON THE 110 GHZ ELECTRON CYCLOTRON INSTALLATION ON THE DIII-D TOKAMAK. Significant improvements are being implement4ed to the capability of the 110 GHz electron cyclotron system on the DIII-D tokamak. Chief among these is the addition of the fifth and sixth 1 MW class gyrotrons, increasing the power available for auxiliary heating and current drive by nearly 60%. These tubes use artificially grown diamond rf output windows to obtain high power with long pulse capability. The beams from these tubes are nearly Gaussian, facilitating coupling to the waveguide. A new fully articulating dual launcher capable of high speed spatial scanning has been designed and tested. The launcher has two axis independent steering for each waveguide. the mirrors can be rotated at up to 100 o /s. A new feedback system linking the DIII-D Plasma Control System (PCS) with the gyrotron beam voltage waveform generators permits real-time feedback control of some plasma properties such as electron temperature. The PCS can use a variety of plasma monitors to generate its control signal, including electron cyclotron emission and Mirnov probes. Electron cyclotron heating and electron cyclotron current drive (ECH and ECCD) were used during this year's DIII-D experimental campaign to control electron temperature, density, and q profiles, induce an ELM-free H-mode, and suppress the m=2/n=1 neoclassical tearing mode. The new capabilities have expanded the role of EC systems in tokamak plasma control

  2. Toroidally Resolved Structure of Divertor Heat Flux in RMP H-mode Discharges on DIII-D

    International Nuclear Information System (INIS)

    Jakubowski, M.W.; Evans, T.E.; Fenstermacher, M.E.; Lasnier, C.J.; Wolf, R.C.; Baylor, Larry R.; Boedo, J.A.; Burrell, K.H.; DeGrassie, J.S.; Gohil, P.; Mordijck, S.; Laengner, R.; Leonard, A.W.; Moyer, R.A.; Petrie, T.W.; Petty, C.C.; Pinsker, R.I.; Rhodes, T.L.; Schaffer, M.J.; Schmitz, O.; Snyder, P.B.; Stoschus, H.; Osborne, T.H.; Orlov, D.M.; Unterberg, Ezekial A.; Watkins, J.G.

    2011-01-01

    As shown on DIII-D edge localized modes (ELMs) can be either completely eliminated or mitigated with resonant magnetic perturbation (RMP) fields. Two infrared cameras, separated 105 degrees toroidally, were used to make simultaneous measurements of ELM heat loads with high frame rates. Without the RMP fields ELMs display a variety of different heat load dynamics and a range of toroidal variability that is characteristic of their 3D structure. Comparing radial averages there is no asymmetry between two toroidal locations. With RMP-mitigated ELMs, the variability in the radially averaged power loads is significantly reduced and toroidal asymmetries in power loads are introduced. In addition to RMP ELM suppression scenarios an RMP scenario with only very small ELMs and very good confinement has been achieved.

  3. Progress on advanced tokamak and steady-state scenario development on DIII-D and NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, California 90095 (United States); Garofalo, A M [Columbia University, New York, New York 10027 (United States); Greenfield, C M [General Atomics, San Diego, California 92186-5608 (United States); Kaye, S M [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Menard, J E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Murakami, M [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Sabbagh, S A [Columbia University, New York, New York 10027 (United States); Austin, M E [University of Texas-Austin, Austin, Texas 78712 (United States); Bell, R E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Burrell, K H [General Atomics, San Diego, California 92186-5608 (United States); Ferron, J R [General Atomics, San Diego, California 92186-5608 (United States); Gates, D A [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Groebner, R J; Hyatt, A W; Luce, T C; Petty, C C; Wade, M R; Waltz, R E [General Atomics, San Diego, California 92186-5608 (United States); Jayakumar, R J [Lawrence Livermore National Lab., Livermore, California 94550 (United States); Kinsey, J E [Lehigh Univ., Bethlehem, Pennsylvania 18015 (United States); LeBlanc, B P [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); McKee, G R [Univ. of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Okabayashi, M [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Peng, Y-K M [Oak Ridge National Lab., Oak Ridge, Tennessee 37831 (United States); Politzer, P A [General Atomics, San Diego, California 92186-5608 (United States); Rhodes, T L [Dept. of Electrical Engineering and PSTI, Univ. of California, Los Angeles, California 90095 (United States)

    2006-12-15

    Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f{sub BS} {>=} 60%, q{sub 95} {approx} 4-5 and G {identical_to}{beta}{sub N}H{sub scaling}/q{sub 95}{sup 2} {>=}0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E ? B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at {beta}{sub N} {>=} 4 with q{sub min} {>=} 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. {beta}{sub N} = 4, H{sub 89} = 2.5, with f{sub BS} > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q {>=} 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

  4. Disruption mitigation studies in DIII-D

    International Nuclear Information System (INIS)

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  7. Investigation of Physical Processes Limiting Plasma Density in DIII--D

    Science.gov (United States)

    Maingi, R.

    1996-11-01

    Understanding the physical processes which limit operating density is crucial in achieving peak performance in confined plasmas. Studies from many of the world's tokamaks have indicated the existence(M. Greenwald, et al., Nucl. Fusion 28) (1988) 2199 of an operational density limit (Greenwald limit, n^GW_max) which is proportional to the plasma current and independent of heating power. Several theories have reproduced the current dependence, but the lack of a heating power dependence in the data has presented an enigma. This limit impacts the International Thermonuclear Experimental Reactor (ITER) because the nominal operating density for ITER is 1.5 × n^GW_max. In DIII-D, experiments are being conducted to understand the physical processes which limit operating density in H-mode discharges; these processes include X-point MARFE formation, high core recycling and neutral pressure, resistive MHD stability, and core radiative collapse. These processes affect plasma properties, i.e. edge/scrape-off layer conduction and radiation, edge pressure gradient and plasma current density profile, and core radiation, which in turn restrict the accessible density regime. With divertor pumping and D2 pellet fueling, core neutral pressure is reduced and X-point MARFE formation is effectively eliminated. Injection of the largest-sized pellets does cause transient formation of divertor MARFEs which occasionally migrate to the X-point, but these are rapidly extinguished in pumped discharges in the time between pellets. In contrast to Greenwald et al., it is found that the density relaxation time after pellets is largely independent of the density relative to the Greenwald limit. Fourier analysis of Mirnov oscillations indicates the de-stabilization and growth of rotating, tearing-type modes (m/n= 2/1) when the injected pellets cause large density perturbations, and these modes often reduce energy confinement back to L-mode levels. We are examining the mechanisms for de

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

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

    2004-03-01

    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

  10. Lower hybrid current drive for edge current density modification in DIII-D: Final status report

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Porkolab, M.

    1993-01-01

    Application of Lower Hybrid (LH) Current Drive (CD) in the DIII-D tokamak has been studied at LLNL, off and on, for several years. The latest effort began in February 1992 in response to a letter from ASDEX indicating that the 2.45 GHz, 3 MW system there was available to be used on another device. An initial assessment of the possible uses for such a system on DIII-D was made and documented in September 1992. Multiple meetings with GA personnel and members of the LH community nationwide have occurred since that time. The work continued through the submission of the 1995 Field Work Proposals in March 1993 and was then put on hold due to budget limitations. The purpose of this document is to record the status of the work in such a way that it could fairly easily be restarted at a future date. This document will take the form of a collection of Appendices giving both background and the latest results from the FY 1993 work, connected by brief descriptive text. Section 2 will describe the final workshop on LHCD in DIII-D held at GA in February 1993. This was an open meeting with attendees from GA, LLNL, MIT and PPPL. Summary documents from the meeting and subsequent papers describing the results will be included in Appendices. Section 3 will describe the status of work on the use of low frequency (2.45 GHZ) LH power and Parametric Decay Instabilities (PDI) for the special case of high dielectric in the edge regions of the DIII-D plasma. This was one of the critical issues identified at the workshop. Other potential issues for LHCD in the DIII-D scenarios are: (1) damping of the waves on fast ions from neutral beam injection, (2) runaway electrons in the low density edge plasma, (3) the validity of the WKB approximation used in the ray-tracing models in the steep edge density gradients

  11. Plasma flow in the DIII-D divertor

    International Nuclear Information System (INIS)

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

    1998-07-01

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

  12. Engineering design of a Radiative Divertor for DIII-D

    International Nuclear Information System (INIS)

    Smith, J.P.; Allen, S.L.; Anderson, P.M.; Baxi, C.B.; Chin, E.; Fenstermacher, M.E.; Hill, D.N.; Hollerbach, M.A.; Hyatt, A.W.; Junge, R.; Mahdavi, M.A.; Porter, G.D.; Redler, K.; Reis, E.E.; Schaffer, M.J.; Sevier, D.L.; Stambaugh, R.D.

    1995-01-01

    A new divertor called the Radiative Divertor is presently being designed for the DIII-D tokamak. Input from tokamak experiments and modeling form the basis for the new design. The Radiative Divertor is intended to reduce the heat flux on the divertor plates by dispersing the power with radiation. Gas puffing experiments in the current open divertor have shown a reduction of the divertor heat flux with either deuterium or impurity puffing. However, either the plasma density (D 2 ) or the core Z eff (impurities) increases in these experiments. The radiative divertor uses a slot structure to isolate the divertor plasma region from the area surrounding the core plasma. Modeling has shown that the Radiative Divertor hardware will provide better baffling and particle control and thereby minimize the effect of the gas puffing in the divertor region on the plasma core. In addition, the Radiative Divertor structure will allow density control in plasma shapes with high triangularity (>0.8) required for advanced tokamak operation. The divertor structure allows for operation in either double or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. Biasing is an integral part of the design and is based on experience at the Tokamak de Varennes (TdeV) and DIII-D. Boron nitride tiles electrically insulate the inner and outer strike points and a low current electrode is used to apply a radial electric field to the scrape-off layer. TdeV has shown that biasing can provide particle and impurity control. The design is extremely flexible, and will allow physics studies of the effect of slot width and height. This is extremely important, as the amount of chamber volume needed for the divertor in future machines such as International Thermonuclear Experiment Reactor (ITER) and Tokamak Physics Experiment (TPX) must be determined. (orig./WL)

  13. Effects of divertor geometry and pumping on plasma performance on DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Hill, D.N.; Porter, G.D.

    1997-06-01

    This paper reports the status of an ongoing investigation to discern the influence of the divertor and plasma geometry on the confinement of both ELM-free and ELMing discharges in DIII-D. The ultimate goal is to achieve a high-performance core plasma which coexists with an advanced divertor plasma. The divertor plasma must reduce the heat flux to acceptable levels; the current technique disperses the heat flux over a wide area by radiation (a radiative divertor). To date, we have obtained our best performance in double-null (DN) high-triangularity (δ ∼ 0.8) ELM-free discharges. As discussed in detail elsewhere, there are several advantages for both the core and divertor plasma with highly-shaped DN operation. Previous radiative-divertor experiments with D 2 injection in DN high-δ ELMing H-mode have shown that this configuration is more sensitive to gas puffing (τ decreases). Moving the X-point away from the target plate (to ∼15 cm above the plate) decreases this sensitivity. Preliminary measurements also indicate that gas puffing reduces the divertor heat flux but does not reduce the plasma pressure along the field line. The up/down heat flux balance can be varied magnetically (by changing the distance between the separatrices), with a slight magnetic imbalance required to balance the heat flux. The overall mission of the Radiative Divertor Project (RDP) is to install a fully pumped and baffled high-δ DN divertor. To date, however, both the DIII-D divertor diagnostics and pump were optimized for lower single-null (LSN) low-δ (δ∼ 0.4) plasmas, so much of the divertor physics has been performed in LSN; these results are discussed in Section 2. As part of the first phase of the RDP, we have installed a new high-δ USN divertor baffle and pump; these results are discussed in Section 3. Both divertor and core parameters are discussed in each case

  14. 13C-Tracer Experiments in DIII-D Preliminary to Thermal Oxidation Experiments to Understand Tritium Recovery in DIII-D, JET, C-Mod, and MAST

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  15. System upgrades to the DIII-D facility

    International Nuclear Information System (INIS)

    Kellman, A.G.

    2007-01-01

    Major upgrades to the DIII-D facility have been performed that significantly enhance the capability of both the DIII-D device and the entire facility. The most significant of these include the rotation of a neutral beam line, installation of a new lower divertor, and a significant set of new and enhanced diagnostics. The upgrades and initial results are presented in this paper

  16. Automated Fault Detection for DIII-D Tokamak Experiments

    International Nuclear Information System (INIS)

    Walker, M.L.; Scoville, J.T.; Johnson, R.D.; Hyatt, A.W.; Lee, J.

    1999-01-01

    An automated fault detection software system has been developed and was used during 1999 DIII-D plasma operations. The Fault Identification and Communication System (FICS) executes automatically after every plasma discharge to check dozens of subsystems for proper operation and communicates the test results to the tokamak operator. This system is now used routinely during DIII-D operations and has led to an increase in tokamak productivity

  17. Alfv?nic Instabilities and Fast Ion Transport in the DIII-D Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Van Zeeland, M; Heidbrink, W; Nazikian, R; Austin, M; Berk, H; Gorelenkov, N; Holcomb, C; Kramer, G; Lohr, J; Luo, Y; Makowski, M; McKee, G; Petty, C; Prater, R; Solomon, W; White, R

    2008-10-14

    Neutral beam injection into reversed magnetic shear DIII-D plasmas produces a variety of Alfvenic activity including Toroidicity and Ellipticity induced Alfven Eigenmodes (TAE/EAE, respectively) and Reversed Shear Alfven Eigenmodes (RSAE) as well as their spatial coupling. These modes are typically studied during the discharge current ramp phase when incomplete current penetration results in a high central safety factor and strong drive due to multiple higher order resonances. During this same time period Fast-Ion D{sub {alpha}} (FIDA) spectroscopy shows that the central fast ion profile is flattened, the degree of which depends on the Alfven eigenmode amplitude. Interestingly, localized electron cyclotron heating (ECH) near the mode location stabilizes RSAE activity and results in significantly improved fast ion confinement relative to discharges with ECH deposition on axis. In these discharges, RSAE activity is suppressed when ECH is deposited near the radius of the shear reversal point and enhanced with deposition near the axis. To simulate the observed neutral beam ion redistribution, NOVA calculations of the 3D eigenmode structures are matched with experimental measurements and used in combination with the ORBIT guiding center following code. For fixed frequency eigenmodes, it is found that ORBIT calculations cannot explain the observed beam ion transport with experimentally measured mode amplitudes. Possible explanations are considered including recent simulation results incorporating eigenmodes with time dependent frequencies.

  18. Magnetic Flux Conversion in the DIII-D Steady-State Hybrid Scenario

    Science.gov (United States)

    Taylor, N. Z.; Luce, T. C.; La Haye, R. J.; Petty, C. C.; Nazikian, R.

    2017-10-01

    The hybrid is a promising high confinement scenario for ITER. The broader current profile aids discharge sustainment by raising qmin > 1 thereby avoiding sawtooth-triggered 2/1 tearing modes. In DIII-D hybrid scenario discharges, the rate of poloidal magnetic energy consumption is more than the rate of energy flow from the poloidal field coils. This is evidence that there is a conversion of toroidal flux to poloidal flux, which may be responsible for the anomalous broadening of the current profile known as flux pumping. The rate of poloidal flux being provided and consumed was tracked with coil and kinetic flux states. During long stationary intervals (1.5 seconds) with constant stored magnetic energy, a significant flux state deficit rate >10 mV was observed. The inequality in the evolution of the flux states was observed in hybrids that were 100% non-inductive and with successful RMP ELM suppression. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC05-06OR23100.

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

    Science.gov (United States)

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

    2015-11-01

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

  20. Alfvenic Instabilities and Fast Ion Transport in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Van Zeeland, M.; Heidbrink, W.; Nazikian, R.; Austin, M.; Berk, H.; Gorelenkov, N.; Holcomb, C.; Kramer, G.; Lohr, J.; Luo, Y.; Makowski, M.; McKee, G.; Petty, C.; Prater, R.; Solomon, W.; White, R.

    2008-01-01

    Neutral beam injection into reversed magnetic shear DIII-D plasmas produces a variety of Alfvenic activity including Toroidicity and Ellipticity induced Alfven Eigenmodes (TAE/EAE, respectively) and Reversed Shear Alfven Eigenmodes (RSAE) as well as their spatial coupling. These modes are typically studied during the discharge current ramp phase when incomplete current penetration results in a high central safety factor and strong drive due to multiple higher order resonances. During this same time period Fast-Ion D α (FIDA) spectroscopy shows that the central fast ion profile is flattened, the degree of which depends on the Alfven eigenmode amplitude. Interestingly, localized electron cyclotron heating (ECH) near the mode location stabilizes RSAE activity and results in significantly improved fast ion confinement relative to discharges with ECH deposition on axis. In these discharges, RSAE activity is suppressed when ECH is deposited near the radius of the shear reversal point and enhanced with deposition near the axis. To simulate the observed neutral beam ion redistribution, NOVA calculations of the 3D eigenmode structures are matched with experimental measurements and used in combination with the ORBIT guiding center following code. For fixed frequency eigenmodes, it is found that ORBIT calculations cannot explain the observed beam ion transport with experimentally measured mode amplitudes. Possible explanations are considered including recent simulation results incorporating eigenmodes with time dependent frequencies

  1. Status of DIII-D plasma control

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  2. Dimensionally similar discharges with central rf heating on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Luce, T.C.; Pinsker, R.I.

    1993-04-01

    The scaling of L-mode heat transport with normalized gyroradius is investigated on the DIII-D tokamak using central rf heating. A toroidal field scan of dimensionally similar discharges with central ECH and/or fast wave heating show gyro-Bohm-like scaling both globally and locally. The main difference between these restats and those using NBI heating on DIII-D is that with rf heating the deposition profile is not very sensitive to the plasma density. Therefore central heating can be utilized for both the low-B and high-B discharges, whereas for NBI the power deposition is decidedly off-axis for the high-B discharge (i.e., high density)

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    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.

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

    International Nuclear Information System (INIS)

    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 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 ∼1 MW

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

    International Nuclear Information System (INIS)

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

    1999-01-01

    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 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 ∼1 MW. (c) 1999 American Institute of Physics

  7. Neoclassical tearing modes in DIII-D and calculations of the stabilizing effects of localized electron cyclotron current drive

    International Nuclear Information System (INIS)

    Prater, R.; La Haye, R. J.; Lin-Liu, Y. R.; Lohr, J.; Perkins, F. W.; Bernabei, S.; Wong, K.-L.; Harvey, R. W.

    1999-01-01

    Neoclassical tearing modes are found to limit the achievable beta in many high performance discharges in DIII-D. Electron cyclotron current drive within the magnetic islands formed as the tearing mode grows has been proposed as a means of stabilizing these modes or reducing their amplitude, thereby increasing the beta limit by a factor around 1.5. Some experimental success has been obtained previously on Asdex-U. Here we examine the parameter range in DIII-D in which this effect can best be studied

  8. Feedback Control of Resistive Wall Modes in Slowly Rotating DIII-D Plasmas

    Science.gov (United States)

    Okabayashi, M.; Chance, M. S.; Takahashi, H.; Garofalo, A. M.; Reimerdes, H.; in, Y.; Chu, M. S.; Jackson, G. L.; La Haye, R. J.; Strait, E. J.

    2006-10-01

    In slowly rotating plasmas on DIII-D, the requirement of RWM control feedback have been identified, using a MHD code along with measured power supply characteristics. It was found that a small time delay is essential for achieving high beta if no rotation stabilization exists. The overall system delay or the band pass time constant should be in the range of 0.4 of the RWM growth time. Recently the control system was upgraded using twelve linear audio amplifiers and a faster digital control system, reducing the time-delay from 600 to 100 μs. The advantage has been clearly observed when the RWMs excited by ELMs were effectively controlled by feedback even if the rotation transiently slowed nearly to zero. This study provides insight on stability in the low- rotation plasmasw with balanced NBI in DIII-D and also in ITER.

  9. Operational upgrades to the DIII-D 60 GHz electron cyclotron resonant heating system

    International Nuclear Information System (INIS)

    Harris, T.E.; Cary, W.P.

    1993-10-01

    One of the primary components of the DIII-D radio frequency (rf) program over the past seven years has been the 60 GHz electron cyclotron resonant heating (ECRH) system. The system now consists of eight units capable of operating and controlling eight Varian VGE-8006 60 GHz, 200 kW gyrotrons along with their associated waveguide components. This paper will discuss the operational upgrades and the overall system performance. Many modifications were instituted to enhance the system operation and performance. Modifications discussed in this paper include an improved gyrotron tube-fault response network, a computer controlled pulse-timing and sequencing system, and an improved high-voltage power supply control interface. The discussion on overall system performance will include operating techniques used to improve system operations and reliability. The techniques discussed apply to system start-up procedures, operating the system in a conditioning mode, and operating the system during DIII-D plasma operations

  10. A remote control room at DIII-D

    International Nuclear Information System (INIS)

    Abla, G.; Schissel, D.P.; Penaflor, B.G.; Wallace, G.

    2008-01-01

    This paper describes a remote control room built at DIII-D to support remote participation activities of DIII-D research staff. In order to create a persistent, efficient, and reliable remote participation environment for DIII-D scientists, a remote control room has been built in a 640-ft 2 dedicated area. The purpose of this room is to experiment and define a remote control room framework that can facilitate the remote participation needs of current and future fusion experiments such as ITER. A variety of hardware equipment has been installed and several remote participation and collaboration technologies have been deployed. Objectivity and practical consideration has been the key while designing the room and deploying the technologies. Although, the DIII-D remote control room is still a work in progress and new software tools are being implemented, it has been already useful for a number of international remote participation activities. For example, it has been used for remote support of the EAST Tokamak in China during the start up operation and proven effective for other collaborative experiment activities. The description of the remote control room design is given along with technologies deployed for remote collaboration needs. We will also discuss our recent experiences involving the DIII-D remote control room as well as future plans for improvements

  11. Measurement of runaway electron energy distribution function during high-Z gas injection into runaway electron plateaus in DIII-D

    International Nuclear Information System (INIS)

    Hollmann, E. M.; Moyer, R. A.; Rudakov, D. L.; Parks, P. B.; Eidietis, N. W.; Paz-Soldan, C.; Commaux, N.; Shiraki, D.; Austin, M. E.; Lasnier, C. J.

    2015-01-01

    The evolution of the runaway electron (RE) energy distribution function f ε during massive gas injection into centered post-disruption runaway electron plateaus has been reconstructed. Overall, f ε is found to be much more skewed toward low energy than predicted by avalanche theory. The reconstructions also indicate that the RE pitch angle θ is not uniform, but tends to be large at low energies and small θ ∼ 0.1–0.2 at high energies. Overall power loss from the RE plateau appears to be dominated by collisions with background free and bound electrons, leading to line radiation. However, the drag on the plasma current appears to be dominated by collisions with impurity ions in most cases. Synchrotron emission appears not to be significant for overall RE energy dissipation but may be important for limiting the peak RE energy

  12. Fast wave current drive experiment on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.; Harvey, R.W.; Lohr, J.; Luce, T.C.; Mayberry, M.J.; Prater, R.; Porkolab, M.; Baity, F.W.; Goulding, R.H.; Hoffman, J.D.; James, R.A.; Kawashima, H.

    1992-06-01

    One method of radio-frequency heating which shows theoretical promise for both heating and current drive in tokamak plasmas is the direct absorption by electrons of the fast Alfven wave (FW). Electrons can directly absorb fast waves via electron Landau damping and transit-time magnetic pumping when the resonance condition ω - κ parallele υ parallele = O is satisfied. Since the FW accelerates electrons traveling the same toroidal direction as the wave, plasma current can be generated non-inductively by launching FW which propagate in one toroidal direction. Fast wave current drive (FWCD) is considered an attractive means of sustaining the plasma current in reactor-grade tokamaks due to teh potentially high current drive efficiency achievable and excellent penetration of the wave power to the high temperature plasma core. Ongoing experiments on the DIII-D tokamak are aimed at a demonstration of FWCD in the ion cyclotron range of frequencies (ICRF). Using frequencies in the ICRF avoids the possibility of mode conversion between the fast and slow wave branches which characterized early tokamak FWCD experiments in the lower hybrid range of frequencies. Previously on DIII-D, efficient direct electron heating by FW was found using symmetric (non-current drive) antenna phasing. However, high FWCD efficiencies are not expected due to the relatively low electron temperatures (compared to a reactor) in DIII-D

  13. Customizable scientific web-portal for DIII-D nuclear fusion experiment

    International Nuclear Information System (INIS)

    Abla, G; Kim, E N; Schissel, D P

    2010-01-01

    Increasing utilization of the Internet and convenient web technologies has made the web-portal a major application interface for remote participation and control of scientific instruments. While web-portals have provided a centralized gateway for multiple computational services, the amount of visual output often is overwhelming due to the high volume of data generated by complex scientific instruments and experiments. Since each scientist may have different priorities and areas of interest in the experiment, filtering and organizing information based on the individual user's need can increase the usability and efficiency of a web-portal. DIII-D is the largest magnetic nuclear fusion device in the US. A web-portal has been designed to support the experimental activities of DIII-D researchers worldwide. It offers a customizable interface with personalized page layouts and list of services for users to select. Each individual user can create a unique working environment to fit his own needs and interests. Customizable services are: real-time experiment status monitoring, diagnostic data access, interactive data analysis and visualization. The web-portal also supports interactive collaborations by providing collaborative logbook, and online instant announcement services. The DIII-D web-portal development utilizes multi-tier software architecture, and Web 2.0 technologies and tools, such as AJAX and Django, to develop a highly-interactive and customizable user interface.

  14. Customizable scientific web-portal for DIII-D nuclear fusion experiment

    Energy Technology Data Exchange (ETDEWEB)

    Abla, G; Kim, E N; Schissel, D P, E-mail: abla@fusion.gat.co [General Atomics, PO Box 85608, San Diego, California 92186-5608 (United States)

    2010-04-01

    Increasing utilization of the Internet and convenient web technologies has made the web-portal a major application interface for remote participation and control of scientific instruments. While web-portals have provided a centralized gateway for multiple computational services, the amount of visual output often is overwhelming due to the high volume of data generated by complex scientific instruments and experiments. Since each scientist may have different priorities and areas of interest in the experiment, filtering and organizing information based on the individual user's need can increase the usability and efficiency of a web-portal. DIII-D is the largest magnetic nuclear fusion device in the US. A web-portal has been designed to support the experimental activities of DIII-D researchers worldwide. It offers a customizable interface with personalized page layouts and list of services for users to select. Each individual user can create a unique working environment to fit his own needs and interests. Customizable services are: real-time experiment status monitoring, diagnostic data access, interactive data analysis and visualization. The web-portal also supports interactive collaborations by providing collaborative logbook, and online instant announcement services. The DIII-D web-portal development utilizes multi-tier software architecture, and Web 2.0 technologies and tools, such as AJAX and Django, to develop a highly-interactive and customizable user interface.

  15. Customizable scientific web-portal for DIII-D nuclear fusion experiment

    Science.gov (United States)

    Abla, G.; Kim, E. N.; Schissel, D. P.

    2010-04-01

    Increasing utilization of the Internet and convenient web technologies has made the web-portal a major application interface for remote participation and control of scientific instruments. While web-portals have provided a centralized gateway for multiple computational services, the amount of visual output often is overwhelming due to the high volume of data generated by complex scientific instruments and experiments. Since each scientist may have different priorities and areas of interest in the experiment, filtering and organizing information based on the individual user's need can increase the usability and efficiency of a web-portal. DIII-D is the largest magnetic nuclear fusion device in the US. A web-portal has been designed to support the experimental activities of DIII-D researchers worldwide. It offers a customizable interface with personalized page layouts and list of services for users to select. Each individual user can create a unique working environment to fit his own needs and interests. Customizable services are: real-time experiment status monitoring, diagnostic data access, interactive data analysis and visualization. The web-portal also supports interactive collaborations by providing collaborative logbook, and online instant announcement services. The DIII-D web-portal development utilizes multi-tier software architecture, and Web 2.0 technologies and tools, such as AJAX and Django, to develop a highly-interactive and customizable user interface.

  16. Recent developments on the 110 GHz electron cyclotron installation on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Ponce, D.; Callis, R.W.; Cary, W.P.; Ferron, J.R.; Green, M.; Grunloh, H.J.; Gorelov, Y.; Lohr, J.; Ellis, R.A.

    2003-01-01

    Significant improvements are being implemented to the capability of the 110 GHz electron cyclotron system on the DIII-D tokamak. Chief among these is the addition of the fifth and sixth 1 MW class gyrotrons, increasing the power available for auxiliary heating and current drive by nearly 60%. These tubes use artificially grown diamond r.f. output windows to obtain high power with long pulse capability. The beams from these tubes are nearly Gaussian, facilitating coupling to the waveguide. A new fully articulating dual launcher capable of high speed spatial scanning has been designed and tested. The launcher has two axis independent steering for each waveguide. The mirrors can be rotated at up to 100 deg./s. A new feedback system linking the DIII-D Plasma Control System (PCS) with the gyrotron beam voltage waveform generators permits real-time feedback control of some plasma properties such as electron temperature. The PCS can use a variety of plasma monitors to generate its control signal, including electron cyclotron emission and Mirnov probes. Electron cyclotron heating and electron cyclotron current drive were used during this year's DIII-D experimental campaign to control electron temperature, density, and q profiles, induce an ELM-free H-mode, and suppress the m=2/n=1 neoclassical tearing mode. The new capabilities have expanded the role of EC systems in tokamak plasma control

  17. AORSA full wave calculations of helicon waves in DIII-D and ITER

    Science.gov (United States)

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

    2018-06-01

    Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D, FNSF, and DEMO tokamaks. Previous ray tracing modeling using GENRAY predicts strong single pass absorption and current drive in the mid-radius region on DIII-D in high beta tokamak discharges. The full wave code AORSA, which is valid to all order of Larmor radius and can resolve arbitrary ion cyclotron harmonics, has been used to validate the ray tracing technique. If the scrape-off-layer (SOL) is ignored in the modeling, AORSA agrees with GENRAY in both the amplitude and location of driven current for DIII-D and ITER cases. These models also show that helicon current drive can possibly be an efficient current drive actuator for ITER. Previous GENRAY analysis did not include the SOL. AORSA has also been used to extend the simulations to include the SOL and to estimate possible power losses of helicon waves in the SOL. AORSA calculations show that another mode can propagate in the SOL and lead to significant (~10%–20%) SOL losses at high SOL densities. Optimizing the SOL density profile can reduce these SOL losses to a few percent.

  18. Development of a new error field correction coil (C-coil) for DIII-D

    International Nuclear Information System (INIS)

    Robinson, J.I.; Scoville, J.T.

    1995-12-01

    The C-coil recently installed on the DIII-D tokamak was developed to reduce the error fields created by imperfections in the location and geometry of the existing coils used to confine, heat, and shape the plasma. First results from C-coil experiments include stable operation in a 1.6 MA plasma with a density less than 1.0 x 10 13 cm -3 , nearly a factor of three lower density than that achievable without the C-coil. The C-coil has also been used in magnetic braking of the plasma rotation and high energy particle confinement experiments. The C-coil system consists of six individual saddle coils, each 60 degree wide toroidally, spanning the midplane of the vessel with a vertical height of 1.6 m. The coils are located at a major radius of 3.2 m, just outside of the toroidal field coils. The actual shape and geometry of each coil section varied somewhat from the nominal dimensions due to the large number of obstructions to the desired coil path around the already crowded tokamak. Each coil section consists of four turns of 750 MCM insulated copper cable banded with stainless steel straps within the web of a 3 in. x 3 in. stainless steel angle frame. The C-coil structure was designed to resist peak transient radial forces (up to 1,800 Nm) exerted on the coil by the toroidal and ploidal fields. The coil frames were supported from existing poloidal field coil case brackets, coil studs, and various other structures on the tokamak

  19. DIII-D Integrated plasma control solutions for ITER and next-generation tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Ferron, J.R.; Hyatt, A.W.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; In, Y.

    2008-01-01

    Plasma control design approaches and solutions developed at DIII-D to address its control-intensive advanced tokamak (AT) mission are applicable to many problems facing ITER and other next-generation devices. A systematic approach to algorithm design, termed 'integrated plasma control,' enables new tokamak controllers to be applied operationally with minimal machine time required for tuning. Such high confidence plasma control algorithms are designed using relatively simple ('control-level') models validated against experimental response data and are verified in simulation prior to operational use. A key element of DIII-D integrated plasma control, also required in the ITER baseline control approach, is the ability to verify both controller performance and implementation by running simulations that connect directly to the actual plasma control system (PCS) that is used to operate the tokamak itself. The DIII-D PCS comprises a powerful and flexible C-based realtime code and programming infrastructure, as well as an arbitrarily scalable hardware and realtime network architecture. This software infrastructure provides a general platform for implementation and verification of realtime algorithms with arbitrary complexity, limited only by speed of execution requirements. We present a complete suite of tools (known collectively as TokSys) supporting the integrated plasma control design process, along with recent examples of control algorithms designed for the DIII-D PCS. The use of validated physics-based models and a systematic model-based design and verification process enables these control solutions to be directly applied to ITER and other next-generation tokamaks

  20. System control and data acquisition of the two new FWCD RF systems at DIII-D

    International Nuclear Information System (INIS)

    Harris, T.E.; Allen, J.C.; Cary, W.P. Petty, C.C.

    1995-10-01

    The Fast Wave Current Drive (FWCD) system at DIII-D has increased its available radio frequency (RF) power capabilities with the addition of two new high power transmitters along with their associated transmission line systems. A Sun Sparc-10 workstation, functioning as the FWCD operator console, is being used to control transmitter operating parameters and transmission line tuning parameters, along with acquiring data and making data available for integration into the DIII-D data acquisition system. Labview, a graphical user interface application, is used to manage and control the above processes. This paper will discuss the three primary branches of the FWCD computer control system: transmitter control, transmission line tuning control, and FWCD data acquisition. The main control program developed uses VXI, GPIB, CAMAC, Serial, and Ethernet protocols to blend the three branches together into one cohesive system. The control of the transmitters utilizes VXI technology to communicate with the transmitter's digital interface. A GPIB network allows for communication with various instruments and CAMAC crate controllers. CAMAC crates are located at each phase-shifter/stub-tuner station and are used to digitize transmission line parameters along with transmission line fault detection during RF transmission. The phase-shifter/stub-tuner stations are located through out the DIII-D facility and are controlled from the FWCD operator console via the workstation's Serial port. The Sun workstation has an Ethernet connection allowing for the utilization of the DIII-D data acquisition open-quotes Open Systemclose quotes architecture and of course providing communication with the rest of the world

  1. Upgraded divertor Thomson scattering system on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Glass, F., E-mail: glassf@fusion.gat.com; Carlstrom, T. N.; Du, D.; Taussig, D. A.; Boivin, R. L. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); McLean, A. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States)

    2016-11-15

    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 (T{sub e} in the range of 0.5 eV–2 keV, n{sub e} in the range of 5 × 10{sup 18}–1 × 10{sup 21} m{sup 3}) for both low T{sub e} in detachment and high T{sub e} measurement up beyond the separatrix.

  2. Overview of Recent DIII-D Experimental Results

    Science.gov (United States)

    Fenstermacher, Max

    2015-11-01

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

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

    Science.gov (United States)

    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.

  4. Model-based control of the resistive wall mode in DIII-D: A comparison study

    International Nuclear Information System (INIS)

    Dalessio, J.; Schuster, E.; Humphreys, D.A.; Walker, M.L.; In, Y.; Kim, J.-S.

    2009-01-01

    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. One of the approaches for RWM stabilization, referred to as magnetic control, uses feedback control to produce magnetic fields opposing the moving field that accompanies the growth of the mode. These fields are generated by coils arranged around the tokamak. One problem with RWM control methods used in present experiments is that they predominantly use simple non-model-based proportional-derivative (PD) controllers requiring substantial derivative gain for stabilization, which implies a large response to noise and perturbations, leading to a requirement for high peak voltages and coil currents, usually leading to actuation saturation and instability. Motivated by this limitation, current efforts in DIII-D include the development of model-based RWM controllers. The General Atomics (GA)/Far-Tech DIII-D RWM model represents the plasma surface as a toroidal current sheet and characterizes the wall using an eigenmode approach. Optimal and robust controllers have been designed exploiting the availability of the RWM dynamic model. The controllers are tested through simulations, and results are compared to present non-model-based PD controllers. This comparison also makes use of the μ structured singular value as a measure of robust stability and performance of the closed-loop system.

  5. Software development on the DIII-D control and data acquisition computers

    International Nuclear Information System (INIS)

    Penaflor, B.G.; McHarg, B.B. Jr.; Piglowski, D.

    1997-11-01

    The various software systems developed for the DIII-D tokamak have played a highly visible and important role in tokamak operations and fusion research. Because of the heavy reliance on in-house developed software encompassing all aspects of operating the tokamak, much attention has been given to the careful design, development and maintenance of these software systems. Software systems responsible for tokamak control and monitoring, neutral beam injection, and data acquisition demand the highest level of reliability during plasma operations. These systems made up of hundreds of programs totaling thousands of lines of code have presented a wide variety of software design and development issues ranging from low level hardware communications, database management, and distributed process control, to man machine interfaces. The focus of this paper will be to describe how software is developed and managed for the DIII-D control and data acquisition computers. It will include an overview and status of software systems implemented for tokamak control, neutral beam control, and data acquisition. The issues and challenges faced developing and managing the large amounts of software in support of the dynamic and everchanging needs of the DIII-D experimental program will be addressed

  6. Modeling and experimental studies of the DIII-D neutral beam system

    Energy Technology Data Exchange (ETDEWEB)

    Crowley, B., E-mail: crowleyb@fusion.gat.com; Rauch, J.; Scoville, J.T.

    2015-10-15

    Highlights: • The issues surrounding proposals to increase neutral beam power are evaluated. • A tetrode version of the DIII-D ion source is modeled. • A neutralization efficiency of the DIII-D neutral beam is measured. • A power loading model of the neutral beam line is presented. - Abstract: In this paper, we present the results of beam physics experimental and modeling efforts aimed at learning from and building on the experience of the DIII-D off-axis neutral beam upgrade and other neutral beam system upgrades such as those at JET. The modeling effort includes electrostatic accelerator modeling (using a Poisson solver), gas dynamics modeling for the neutralizer and beam transport models for the beamline. Experimentally, spectroscopic and calorimetric techniques are used to evaluate the system performance. We seek to understand and ameliorate problems such as anomalous power deposition, originating from misdirected or excessively divergent beam particles, on a number of beamline components. We qualitatively and quantitatively evaluate possible project risks such as neutralization efficiency deficit and high voltage hold off associated with increasing the beam energy up to 105 keV.

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

    International Nuclear Information System (INIS)

    O'NEIL, RC; STAMBAUGH, RD

    2002-01-01

    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

  8. Cooperative program on DIII-D (FY93)

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1994-01-01

    This is a proposal to continue support of the authors cooperative research program on DIII-D, under Department of Energy contract DE-FG03-89ER51116. The proposal describes work carried out recently in support of DIII-D data analysis and modeling, with a focus on divertors, edge physics and transport phenomena linking edge and core physics. Proposed work will continue to focus on edge physics, instabilities, the further development of codes to model the plasma, and data analysis in support of related experimental work

  9. Electron cyclotron current drive experiments on DIII-D

    International Nuclear Information System (INIS)

    James, R.A.; Giruzzi, G.; Gentile, B. de; Rodriguez, L.; Fyaretdinov, A.; Gorelov, Yu.; Trukhin, V.; Harvey, R.; Lohr, J.; Luce, T.C.; Matsuda, K.; Politzer, P.; Prater, R.; Snider, R.; Janz, S.

    1990-05-01

    Electron Cyclotron Current Drive (ECCD) experiments on the DIII-D tokamak have been performed using 60 GHz waves launched from the high field side of the torus. Preliminary analysis indicates rf driven currents between 50 and 100 kA in discharges with total plasma currents between 200 and 500 kA. These are the first ECCD experiments with strong first pass absorption, localized deposition of the rf power, and τ E much longer than the slowing-down time of the rf generated current carriers. The experimentally measured profiles for T e , η e and Z eff are used as input for a 1D transport code and a multiply-ray, 3D ray tracing code. Comparisons with theory and assessment of the influence of the residual electric field, using a Fokker-Planck code, are in progress. The ECH power levels were between 1 and 1.5 MW with pulse lengths of about 500 msec. ECCD experiments worldwide are motivated by issues relating to the physics and technical advantages of the use of high frequency rf waves to drive localized currents. ECCD is accomplished by preferentially heating electrons moving in one toroidal direction, reducing their collisionality and thereby producing a non-inductively driven toroidal current. 6 refs., 4 figs

  10. Electron cyclotron current drive experiments on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    James, R.A. (Lawrence Livermore National Lab., CA (USA)); Giruzzi, G.; Gentile, B. de; Rodriguez, L. (Association Euratom-CEA, Centre d' Etudes Nucleaires de Cadarache, 13 - Saint-Paul-les-Durance (France)); Fyaretdinov, A.; Gorelov, Yu.; Trukhin, V. (Kurchatov Inst. of Atomic Energy, Moscow (USSR)); Harvey, R.; Lohr, J.; Luce, T.C.; Matsuda, K.; Politzer, P.; Prater, R.; Snider, R. (General Atomics, San Di

    1990-05-01

    Electron Cyclotron Current Drive (ECCD) experiments on the DIII-D tokamak have been performed using 60 GHz waves launched from the high field side of the torus. Preliminary analysis indicates rf driven currents between 50 and 100 kA in discharges with total plasma currents between 200 and 500 kA. These are the first ECCD experiments with strong first pass absorption, localized deposition of the rf power, and {tau}{sub E} much longer than the slowing-down time of the rf generated current carriers. The experimentally measured profiles for T{sub e}, {eta}{sub e} and Z{sub eff} are used as input for a 1D transport code and a multiply-ray, 3D ray tracing code. Comparisons with theory and assessment of the influence of the residual electric field, using a Fokker-Planck code, are in progress. The ECH power levels were between 1 and 1.5 MW with pulse lengths of about 500 msec. ECCD experiments worldwide are motivated by issues relating to the physics and technical advantages of the use of high frequency rf waves to drive localized currents. ECCD is accomplished by preferentially heating electrons moving in one toroidal direction, reducing their collisionality and thereby producing a non-inductively driven toroidal current. 6 refs., 4 figs.

  11. Operations Studies of the Gyrotrons on DIII-D

    Science.gov (United States)

    Storment, Stephen; Lohr, John; Cengher, Mirela; Gorelov, Yuri; Ponce, Dan; Torrezan, Antonio

    2017-10-01

    The gyrotrons are high power vacuum tubes used in fusion research to provide high power density heating and current drive in precisely localized areas of the plasma. Despite the increasing experience with both the manufacture and operation of these devices, individual gyrotrons with similar design and manufacturing processes can exhibit important operational differences in terms of generated rf power, efficiency and lifetime. This report discusses differences in the performance of several gyrotrons in operation at DIII-D and presents the results of a series of measurements that could lead to improved the performance of single units based on a better understanding of the causes of these differences. The rf power generation efficiency can be different from gyrotron to gyrotron. In addition, the power loading of the collector can feature localized hot spots, where the collector can locally be close to the power deposition limits. Measurements of collector power loading provide maps of the power deposition and can provide understanding of the effect of modulation of the output rf beam on the total loading, leading to improved operational rules increasing the safety margins for the gyrotrons under different operational scenarios. Work supported by US DOE under DE-FC02-04ER54698.

  12. Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

    Science.gov (United States)

    Si, H.; Guo, H. Y.; Covele, B.; Leonard, A. W.; Watkins, J. G.; Thomas, D.; Ding, R.

    2018-05-01

    One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment from 1.18× {{10}19} {{m}-3} to 0.88× {{10}19} {{m}-3} . Moreover, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of 0.67× {{10}19} {{m}-3} , thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.

  13. Optimization of DIII-D discharges to avoid AE destabilization

    Science.gov (United States)

    Varela, Jacobo; Spong, Donald; Garcia, Luis; Huang, Juan; Murakami, Masanori

    2017-10-01

    The aim of the study is to analyze the stability of Alfven Eigenmodes (AE) perturbed by energetic particles (EP) during DIII-D operation. We identify the optimal NBI operational regimes that avoid or minimize the negative effects of AE on the device performance. We use the reduced MHD equations to describe the linear evolution of the poloidal flux and the toroidal component of the vorticity in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles, including the effect of the acoustic modes. We add the Landau damping and resonant destabilization effects using a closure relation. We perform parametric studies of the MHD and AE stability, taking into account the experimental profiles of the thermal plasma and EP, also using a range of values of the energetic particles β, density and velocity as well the effect of the toroidal couplings. We reproduce the AE activity observed in high poloidal β discharge at the pedestal and reverse shear discharges. This material based on work is supported both by the U.S. Department of Energy, Office of Science, under Contract DE-AC05-00OR22725 with UT-Battelle, LLC. Research sponsored in part by the Ministerio de Economia y Competitividad of Spain under the project.

  14. DIII-D Neutral Beam control system operator interface

    International Nuclear Information System (INIS)

    Harris, J.J.; Campbell, G.L.

    1993-10-01

    A centralized graphical user interface has been added to the DIII-D Neutral Beam (NB) control systems for status monitoring and remote control applications. This user interface provides for automatic data acquisition, alarm detection and supervisory control of the four NB programmable logic controllers (PLC) as well as the Mode Control PLC. These PLCs are used for interlocking, control and status of the NB vacuum pumping, gas delivery, and water cooling systems as well as beam mode status and control. The system allows for both a friendly user interface as well as a safe and convenient method of communicating with remote hardware that formerly required interns to access. In the future, to enable high level of control of PLC subsystems, complete procedures is written and executed at the touch of a screen control panel button. The system consists of an IBM compatible 486 computer running the FIX DMACS trademark for Windows trademark data acquisition and control interface software, a Texas Instruments/Siemens communication card and Phoenix Digital optical communications modules. Communication is achieved via the TIWAY (Texas Instruments protocol link utilizing both fiber optic communications and a copper local area network (LAN). Hardware and software capabilities will be reviewed. Data and alarm reporting, extended monitoring and control capabilities will also be discussed

  15. ELM-Induced Plasma Wall Interactions in DIII-D

    International Nuclear Information System (INIS)

    Rudakov, D.L.; Boedo, J.A.; Yu, J.H.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Hollmann, E.M.; Lasnier, C.J.; McLean, A.G.; Moyer, R.A.; Stangeby, P.C.; Tynan, G.R.; Wampler, W.R.; Watkins, J.G.; West, W.P.; Wong, C.C.; Zeng, L.; Bastasz, R.J.; Buchenauer, D.; Whaley, J.

    2008-01-01

    Intense transient fluxes of particles and heat to the main chamber components induced by edge localized modes (ELMs) are of serious concern for ITER. In DIII-D, plasma interaction with the outboard chamber wall is studied using Langmuir probes and optical diagnostics including a fast framing camera. Camera data shows that ELMs feature helical filamentary structures localized at the low field side of the plasma and aligned with the local magnetic field. During the nonlinear phase of an ELM, multiple filaments are ejected from the plasma edge and propagate towards the outboard wall with velocities of 0.5-0.7 km/s. When reaching the wall, filaments result in 'hot spots'--regions of local intense plasma-material interaction (PMI) where the peak incident particle and heat fluxes are up to 2 orders of magnitude higher than those between ELMs. This interaction pattern has a complicated geometry and is neither toroidally nor poloidally symmetric. In low density/collisionality H-mode discharges, PMI at the outboard wall is almost entirely due to ELMs. In high density/collisionality discharges, contributions of ELMs and inter-ELM periods to PMI at the wall are comparable. A Midplane Material Evaluation Station (MiMES) has been recently installed in order to conduct in situ measurements of erosion/redeposition at the outboard chamber wall, including those caused by ELMs

  16. Progress Towards Increased Understanding and Control of Internal Transport Barriers on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Doyle, E. J. [University of California, Los Angeles; Greenfield, C. M. [General Atomics; Austin, M. E. [University of Texas, Austin; Baylor, Larry R [ORNL; Burrell, K. H. [General Atomics; Casper, T. A. [Lawrence Livermore National Laboratory (LLNL); DeBoo, J. C. [General Atomics; Ernst, D. R. [Princeton Plasma Physics Laboratory (PPPL); Fenzi, C. [University of Wisconsin, Madison; Gohil, P. [General Atomics; Groebner, R. J. [General Atomics; Heidbrink, W. W. [University of California, Irvine; Jackson, G. L. [General Atomics; Jernigan, Thomas C [ORNL; Kinsey, J. E. [Lehigh University, Bethlehem, PA; Lao, L. L. [General Atomics; Makowski, M. A. [Lawrence Livermore National Laboratory (LLNL); Mckee, G. R. [University of Wisconsin, Madison; Murakami, Masanori [ORNL; Peebles, W. A. [University of California, Los Angeles; Prater, R. [General Atomics; Rettig, C. L. [University of California, Los Angeles; Rhodes, T. L. [University of California, Los Angeles; Rost, J. C. [Massachusetts Institute of Technology (MIT); Staebler, G. M. [General Atomics; Stallard, B. W. [Lawrence Livermore National Laboratory (LLNL); Strait, E. J. [General Atomics; Synakowski, E. J. [Princeton Plasma Physics Laboratory (PPPL); Thomas, D. M. [General Atomics; Wade, Mickey R [ORNL; Waltz, R. E. [General Atomics; Zeng, L. [University of California, Los Angeles

    2001-01-01

    Substantial progress has been made towards both understanding and control of internal transport barriers (ITBs) on DIII-D, resulting in the discovery of a new sustained high performance operating mode termed the Quiescent Double-Barrier (QDB) regime. The QDB regime combines core transport barriers with a quiescent, ELM-free H-mode edge (termed QH-mode), giving rise to separate (double) core and edge transport barriers. The core and edge barriers are mutually compatible and do not merge, resulting in broad core profiles with an edge pedestal. The QH-mode edge is characterized by ELM-free behavior with continuous multiharmonic MHD activity in the pedestal region, and has provided density and impurity control for 3.5 s (>20 τE) with divertor pumping. QDB plasmas are long-pulse high-performance candidates, having maintained a βNH89 product of 7 for 5 energy confinement times (Ti ≤ 16 keV, βN ≤ 2.9, H89 ≤ 2.4, τE ≤ 150 ms, DD neutron rate Sn ≤ 4x1015 s-1). The QDB regime has only been obtained in counter-NBI discharges (injection anti-parallel to plasma current) with divertor pumping. Other results include successful expansion of the ITB radius using (separately) both impurity injection and counter-NBI, and the formation of ITBs in the electron thermal channel using both ECH and strong negative central shear (NCS) at high power. These results are interpreted within a theoretical framework in which turbulence suppression is the key to ITB formation and control, and a decrease in core turbulence is observed in all cases of ITB formation.

  17. Progress towards increased understanding and control of internal transport barriers (ITBs) on DIII-D

    International Nuclear Information System (INIS)

    Doyle, E.J.; Greenfield, C.M.; Austin, M.E.

    2001-01-01

    Substantial progress has been made towards both understanding and control of internal transport barriers (ITBs) on DIII-D, resulting in the discovery of a new sustained high performance operating mode termed the Quiescent Double-Barrier (QDB) regime. The QDB regime combines core transport barriers with a quiescent, ELM-free H-mode edge (termed QH-mode), giving rise to separate (double) core and edge transport barriers. The core and edge barriers are mutually compatible and do not merge, resulting in broad core profiles with an edge pedestal. The QH-mode edge is characterized by ELM-free behavior with continuous multiharmonic MHD activity in the pedestal region, and has provided density and impurity control for 3.5 s (>20 τ E ) with divertor pumping. QDB plasmas are long-pulse high-performance candidates, having maintained a β N H 89 product of 7 for 5 energy confinement times (T i ≤16 keV, β N ≤2.9, H 89 ≤2.4, τ E ≤150 ms, DD neutron rate S n ≤4x10 15 s -1 ). The QDB regime has only been obtained in counter-NBI discharges (injection anti-parallel to plasma current) with divertor pumping. Other results include successful expansion of the ITB radius using (separately) both impurity injection and counter-NBI, and the formation of ITBs in the electron thermal channel using both ECH and strong negative central shear (NCS) at high power. These results are interpreted within a theoretical framework in which turbulence suppression is the key to ITB formation and control, and a decrease in core turbulence is observed in all cases of ITB formation. (author)

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

    International Nuclear Information System (INIS)

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

    1990-09-01

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

  19. The DIII-D Computing Environment: Characteristics and Recent Changes

    International Nuclear Information System (INIS)

    McHarg, B.B. Jr.

    1999-01-01

    The DIII-D tokamak national fusion research facility along with its predecessor Doublet III has been operating for over 21 years. The DIII-D computing environment consists of real-time systems controlling the tokamak, heating systems, and diagnostics, and systems acquiring experimental data from instrumentation; major data analysis server nodes performing short term and long term data access and data analysis; and systems providing mechanisms for remote collaboration and the dissemination of information over the world wide web. Computer systems for the facility have undergone incredible changes over the course of time as the computer industry has changed dramatically. Yet there are certain valuable characteristics of the DIII-D computing environment that have been developed over time and have been maintained to this day. Some of these characteristics include: continuous computer infrastructure improvements, distributed data and data access, computing platform integration, and remote collaborations. These characteristics are being carried forward as well as new characteristics resulting from recent changes which have included: a dedicated storage system and a hierarchical storage management system for raw shot data, various further infrastructure improvements including deployment of Fast Ethernet, the introduction of MDSplus, LSF and common IDL based tools, and improvements to remote collaboration capabilities. This paper will describe this computing environment, important characteristics that over the years have contributed to the success of DIII-D computing systems, and recent changes to computer systems

  20. Multi-scale transport in the DIII-D ITER baseline scenario with direct electron heating and projection to ITER

    Science.gov (United States)

    Grierson, B. A.; Staebler, G. M.; Solomon, W. M.; McKee, G. R.; Holland, C.; Austin, M.; Marinoni, A.; Schmitz, L.; Pinsker, R. I.; DIII-D Team

    2018-02-01

    Multi-scale fluctuations measured by turbulence diagnostics spanning long and short wavelength spatial scales impact energy confinement and the scale-lengths of plasma kinetic profiles in the DIII-D ITER baseline scenario with direct electron heating. Contrasting discharge phases with ECH + neutral beam injection (NBI) and NBI only at similar rotation reveal higher energy confinement and lower fluctuations when only NBI heating is used. Modeling of the core transport with TGYRO using the TGLF turbulent transport model and NEO neoclassical transport reproduces the experimental profile changes upon application of direct electron heating and indicates that multi-scale transport mechanisms are responsible for changes in the temperature and density profiles. Intermediate and high-k fluctuations appear responsible for the enhanced electron thermal flux, and intermediate-k electron modes produce an inward particle pinch that increases the inverse density scale length. Projection to ITER is performed with TGLF and indicates a density profile that has a finite scale length due to intermediate-k electron modes at low collisionality and increases the fusion gain. For a range of E × B shear, the dominant mechanism that increases fusion performance is suppression of outward low-k particle flux and increased density peaking.

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    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. DIII-D research advancing the scientific basis for burning plasmas and fusion energy

    Science.gov (United States)

    W. M. SolomonThe DIII-D Team

    2017-10-01

    The DIII-D tokamak has addressed key issues to advance the physics basis for ITER and future steady-state fusion devices. In work related to transient control, magnetic probing is used to identify a decrease in ideal stability, providing a basis for active instability sensing. Improved understanding of 3D interactions is emerging, with RMP-ELM suppression correlated with exciting an edge current driven mode. Should rapid plasma termination be necessary, shattered neon pellet injection has been shown to be tunable to adjust radiation and current quench rate. For predictive simulations, reduced transport models such as TGLF have reproduced changes in confinement associated with electron heating. A new wide-pedestal variant of QH-mode has been discovered where increased edge transport is found to allow higher pedestal pressure. New dimensionless scaling experiments suggest an intrinsic torque comparable to the beam-driven torque on ITER. In steady-state-related research, complete ELM suppression has been achieved that is relatively insensitive to q 95, having a weak effect on the pedestal. Both high-q min and hybrid steady-state plasmas have avoided fast ion instabilities and achieved increased performance by control of the fast ion pressure gradient and magnetic shear, and use of external control tools such as ECH. In the boundary, experiments have demonstrated the impact of E× B drifts on divertor detachment and divertor asymmetries. Measurements in helium plasmas have found that the radiation shortfall can be eliminated provided the density near the X-point is used as a constraint in the modeling. Experiments conducted with toroidal rings of tungsten in the divertor have indicated that control of the strike-point flux is important for limiting the core contamination. Future improvements are planned to the facility to advance physics issues related to the boundary, transients and high performance steady-state operation.

  3. Thermostating highly confined fluids.

    Science.gov (United States)

    Bernardi, Stefano; Todd, B D; Searles, Debra J

    2010-06-28

    In this work we show how different use of thermostating devices and modeling of walls influence the mechanical and dynamical properties of confined nanofluids. We consider a two dimensional fluid undergoing Couette flow using nonequilibrium molecular dynamics simulations. Because the system is highly inhomogeneous, the density shows strong fluctuations across the channel. We compare the dynamics produced by applying a thermostating device directly to the fluid with that obtained when the wall is thermostated, considering also the effects of using rigid walls. This comparison involves an analysis of the chaoticity of the fluid and evaluation of mechanical properties across the channel. We look at two thermostating devices with either rigid or vibrating atomic walls and compare them with a system only thermostated by conduction through vibrating atomic walls. Sensitive changes are observed in the xy component of the pressure tensor, streaming velocity, and density across the pore and the Lyapunov localization of the fluid. We also find that the fluid slip can be significantly reduced by rigid walls. Our results suggest caution in interpreting the results of systems in which fluid atoms are thermostated and/or wall atoms are constrained to be rigid, such as, for example, water inside carbon nanotubes.

  4. Recent experimental studies of edge and internal transport barriers in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Gohil, P; Baylor, L R; Burrell, K H; Casper, T A; Doyle, E J; Greenfield, C M; Jernigan, T C; Kinsey, J E; Lasnier, C J; Moyer, R A; Murakami, M; Rhodes, T L; Rudakov, D L; Staebler, G M; Wang, G; Watkins, J G; West, W P; Zeng, L

    2003-01-01

    Results from recent experiments on the DIII-D tokamak have revealed many important details on transport barriers at the plasma edge and in the plasma core. These experiments include: (a) the formation of the H-mode edge barrier directly by pellet injection; (b) the formation of a quiescent H-mode edge barrier (QH-mode) which is free from edge localized modes, but which still exhibits good density and radiative power control; (c) the formation of multiple transport barriers, such as the quiescent double barrier (QDB) which combines an internal transport barrier with the quiescent H-mode edge barrier. Results from the pellet-induced H-mode experiments indicate that: (a) the edge temperature (electron or ion) does not need to attain a critical value for the formation of the H-mode barrier, (b) pellet injection leads to an increased gradient in the radial electric field, E r , at the plasma edge; (c) the experimentally determined edge parameters at barrier transition are well below the predictions of several theories on the formation of the H-mode barrier, (d) pellet injection can lower the threshold power required to form the H-mode barrier. The quiescent H-mode barrier exhibits good density control as the result of continuous magnetohydrodynamic activity at the plasma edge called the edge harmonic oscillation (EHO). The EHO enhances the edge particle transport whilst maintaining a good energy transport barrier. The ability to produce multiple barriers in the QDB regime has led to long duration, high-performance plasmas with β N H 89 values of 7 for up to 10 times the confinement time. Density profile control in the plasma core of QDB plasmas has been demonstrated using on-axis electron cyclotron heating

  5. Validation of the kinetic-turbulent-neoclassical theory for edge intrinsic rotation in DIII-D

    Science.gov (United States)

    Ashourvan, Arash; Grierson, B. A.; Battaglia, D. J.; Haskey, S. R.; Stoltzfus-Dueck, T.

    2018-05-01

    In a recent kinetic model of edge main-ion (deuterium) toroidal velocity, intrinsic rotation results from neoclassical orbits in an inhomogeneous turbulent field [T. Stoltzfus-Dueck, Phys. Rev. Lett. 108, 065002 (2012)]. This model predicts a value for the toroidal velocity that is co-current for a typical inboard X-point plasma at the core-edge boundary (ρ ˜ 0.9). Using this model, the velocity prediction is tested on the DIII-D tokamak for a database of L-mode and H-mode plasmas with nominally low neutral beam torque, including both signs of plasma current. Values for the flux-surface-averaged main-ion rotation velocity in the database are obtained from the impurity carbon rotation by analytically calculating the main-ion—impurity neoclassical offset. The deuterium rotation obtained in this manner has been validated by direct main-ion measurements for a limited number of cases. Key theoretical parameters of ion temperature and turbulent scale length are varied across a wide range in an experimental database of discharges. Using a characteristic electron temperature scale length as a proxy for a turbulent scale length, the predicted main-ion rotation velocity has a general agreement with the experimental measurements for neutral beam injection (NBI) powers in the range PNBI balanced—but high powered—NBI, the net injected torque through the edge can exceed 1 Nm in the counter-current direction. The theory model has been extended to compute the rotation degradation from this counter-current NBI torque by solving a reduced momentum evolution equation for the edge and found the revised velocity prediction to be in agreement with experiment. Using the theory modeled—and now tested—velocity to predict the bulk plasma rotation opens up a path to more confidently projecting the confinement and stability in ITER.

  6. Surface impurity removal from DIII-D graphite tiles by boron carbide grit blasting

    International Nuclear Information System (INIS)

    Lee, R.L.; Hollerbach, M.A.; Holtrop, K.L.; Kellman, A.G.; Taylor, P.L.; West, W.P.

    1993-11-01

    During the latter half of 1992, the DIII-D tokamak at General Atomics (GA) underwent several modifications of its interior. One of the major tasks involved the removal of accumulated metallic impurities from the surface of the graphite tiles used to line the plasma facing surfaces inside of the tokamak. Approximately 1500 graphite tiles and 100 boron nitride tiles from the tokamak were cleaned to remove the metallic impurities. The cleaning process consisted of several steps: the removed graphite tiles were permanently marked, surface blasted using boron carbide (B 4 C) grit media (approximately 37 μm. diam.), ultrasonically cleaned in ethanol to remove loose dust, and outgassed at 1000 degrees C. Tests were done using, graphite samples and different grit blaster settings to determine the optimum propellant and abrasive media pressures to remove a graphite layer approximately 40-50 μm deep and yet produce a reasonably smooth finish. EDX measurements revealed that the blasting technique reduced the surface Ni, Cr, and Fe impurity levels to those of virgin graphite. In addition to the surface impurity removal, tritium monitoring was performed throughout the cleaning process. A bubbler system was set up to monitor the tritium level in the exhaust gas from the grit blaster unit. Surface wipes were also performed on over 10% of the tiles. Typical surface tritium concentrations of the tiles were reduced from about 500 dpm/100 cm 2 to less than 80 dpm/100 cm 2 following the cleaning. This tile conditioning, and the installation of additional graphite tiles to cover a high fraction of the metallic plasma facing surfaces, has substantially reduced metallic impurities in the plasma discharges which has allowed rapid recovery from a seven-month machine opening and regimes of enhanced plasma energy confinement to be more readily obtained. Safety issues concerning blaster operator exposure to carcinogenic metals and radioactive tritium will also be addressed

  7. Overview of recent experimental results from the DIII-D advanced tokamak program

    International Nuclear Information System (INIS)

    Allen, S.L.

    2001-01-01

    The goals of DIII-D Advanced Tokamak (AT) experiments are to investigate and optimize the upper limits of energy confinement and MHD stability in a tokamak plasma, and to simultaneously maximize the fraction of non-inductive current drive. Significant overall progress has been made in the past 2 years, as the performance figure of merit β N H 89P of 9 has been achieved in ELMing H-mode for over 16 τ E without sawteeth. We also operated at β N ∼7 for over 35 τ E or 3 τ R , with the duration limited by hardware. Real-time feedback control of β (at 95% of the stability boundary), optimizing the plasma shape (e.g., δ, divertor strike- and X-point, double/single null balance), and particle control (n e /n GW ∼0.3, Z eff N H 89P of 7. The QDB regime has been obtained to date only with counter neutral beam injection. Further modification and control of internal transport barriers (ITBs) has also been demonstrated with impurity injection (broader barrier), pellets, and ECH (strong electron barrier). The new Divertor-2000, a key ingredient in all these discharges, provides effective density, impurity and heat flux control in the high-triangularity plasma shapes. Discharges at n e /n GW ∼1.4 have been obtained with gas puffing by maintaining the edge pedestal pressure; this operation is easier with Divertor-2000. We are developing several other tools required for AT operation, including real-time feedback control of resistive wall modes (RWMs) with external coils, and control of neoclassical tearing modes (NTMs) with electron cyclotron current drive (ECCD). (author)

  8. Particle control in DIII-D with helium glow discharge conditioning

    International Nuclear Information System (INIS)

    Jackson, G.L.; Taylor, T.S.; Taylor, P.L.

    1990-01-01

    Helium glow discharge conditioning of DIII-D is routinely used before every tokamak discharge to desorb hydrogen from the graphite tiles, which are the plasma facing surfaces for the floor, inner wall and top of the vessel. In addition to reducing hydrogen fuelling of the plasma by the graphite surfaces, helium glow discharges are also effective in removing low-Z impurities, primarily in the form of carbon monoxide and hydrocarbons, and this has permitted higher current divertor operation and more rapid recovery from tokamak disruptions. Since the implementation of repetitive helium glow wall conditioning, the parameter space in which tokamak discharges in DIII-D can be obtained has been expanded to include the first observations of limiter H-mode confinement, the Ohmic H-mode with periods of up to 150 ms that are free of edge localized modes, more reliable low q operation with volume averaged beta of up to 9.3%, improved control over locked modes and plasma discharges at lower electron density. (author). 37 refs, 12 figs, 1 tab

  9. Prospects for Edge Current Density Determination Using LIBEAM on DIII-D

    International Nuclear Information System (INIS)

    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

    2000-01-01

    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 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 (ρ = 0.9-1.1). Using Zeeman polarization spectroscopy of the 2S-2P lithium resonance line emission from the DIII-D LIBEAM, measurements of the various field components may be made to the necessary precision in exactly the region of interest to these studies. Because of the negligible Stark mixing of the relevant atomic levels, this method of determining j(r) is insensitive to the large local electric fields typically found in enhanced confinement (H-mode) edges, and thus avoids an ambiguity common to Motional Stark Effect (MSE) measurements of B. Key issues for utilizing this technique include good beam quality, an optimum viewing geometry, and a suitable optical pre-filter to isolate the polarized emission line. A prospective diagnostic system for the DIII-D AT program will be described

  10. FWCD (fast wave current drive) and ECCD (electron cyclotron current drive) experiments on DIII-D

    International Nuclear Information System (INIS)

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

    1994-01-01

    Fast wave current drive and electron cyclotron current drive experiments have been performed on the DIII-D tokamak as part of the advanced tokamak program. The goal of this program is to develop techniques for controlling the profile of the current density in order to access regimes of improved confinement and stability. The experiments on fast wave current drive used a four strap antenna with 90deg phasing between straps. A decoupler was used to help maintain the phasing, and feedback control of the plasma position was used to keep the resistive loading constant. RF pickup loops demonstrate that the directivity of the antenna is as expected. Plasma currents up to 0.18 MA were driven by 1.5 MW of fast wave power. Electron cyclotron current drive experiments at 60 GHz have shown 0.1 MA of plasma current driven by 1 MW of power. New fast wave and electron cyclotron heating systems are in development for DIII-D, so that the goals of the advanced tokamak program can be carried out. (author)

  11. Stability of negative central magnetic shear discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    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, β N = β (I/aB) -1 ≤ 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, β 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

  12. Radial transport effects on ECCD in the TCV and DIII-D tokamaks and on Ohmic discharges in the MST RFP

    International Nuclear Information System (INIS)

    Harvey, R.W.; Sauter, O.; Nikkola, P.; Prater, R.; O'Connell, R.; Forest, C.B.

    2003-01-01

    The comprehensive CQL3D Fokker-Planck/Quasilinear simulation code has been benchmarked against experiment over a wide range of electron cyclotron conditions in the DIII-D tokamak (C.C. Petty et al., 14. Topical Conf. on RF Power in Plasmas, 2002). The same code, in disagreement with experiment, gives 560 kA of ECCD for a well documented, completely ECCD-driven, 100 kA TCV shot [O. Sauter et al, PRL, 2000]. Recent work (R.W. Harvey et al, Phys. Rev. Lett., 2002) has resolved the differences as due to radial transport at a level closely consistent with ITER scaling. Transport does not substantially affect DIII-D ECCD, but at similar ECH power has an overwhelming effect on the much smaller TCV. The transport is consistent with electrostatic-type diffusion (D ρρ constant in velocity space) and not with a magnetic-type diffusion (D ρρ ∝ |v || |). Fokker-Planck simulation of Ohmic reversed field pinch (RFP) discharges in the MST device reveals transport velocity dependence stronger than |v || |) will give agreement with current and soft X-ray spectra in standard discharges, but in the higher confinement, current profile controlled PPCD discharges, transport is again electrostatic-like. This is consistent with the object of PPCD, which is to replace magnetic turbulence driven current with auxiliary CD to improve transport. The tokamak and high-confinement RFP results mutually reinforce the constant-in-velocity-space 'electrostatic-type turbulence' conclusion. The steady-state energy and toroidal current are governed by the same radial transport equation. (authors)

  13. RADIAL TRANSPORT EFFECTS ON ECCD IN THE TCV AND DIII-D TOKAMAKS AND ON OHMIC DISCHARGES IN THE MST RFP

    International Nuclear Information System (INIS)

    HARVEY, R.W.; SAUTER, O.; PRATER, R.; NIKKOLA, P.; O'CONNELL, R.; FOREST, C.B.

    2002-01-01

    The comprehensive CQL3D Fokker-Planck/Quasilinear simulation code has been benchmarked against experiment over a wide range of electron cyclotron conditions in the DIII-D tokamak (C.C. Petty et al., 14th Topical Conf. on RF Power in Plasmas, 2002). The same code, in disagreement with experiment, gives 560 kA of ECCD for a well documented, completely ECCD-driven, 100 kA TCV shot [O. Sauter et al, PRL, 2000]. Recent work (R.W. Harvey et al, Phys. Rev. Lett., 2002) has resolved the differences as due to radial transport at a level closely consistent with ITER scaling. Transport does not substantially affect DIII-D ECCD, but at similar ECH power has an overwhelming effect on the much smaller TCV. The transport is consistent with electrostatic-type diffusion (D ρρ constant in velocity-space) and not with a magnetic-type diffusion (D ρρ ∝ |v(parallel)|). Fokker-Planck simulation of Ohmic reversed field pinch (RFP) discharges in the MST device reveals transport velocity dependence stronger than |v(parallel)| will give agreement with current and soft X-ray spectra in standard discharges, but in the higher confinement, current profile controlled PPCD discharges, transport is again electrostatic-like. This is consistent with the object of PPCD, which is to replace magnetic turbulence driven current with auxiliary CD to improve transport. The tokamak and high-confinement RFP results mutually reinforce the constant-in-velocity-space ''electrostatic-type turbulence'' conclusion. The steady-state energy and toroidal current are governed by the same radial transport equation

  14. Divertor plasma studies on DIII-D: Experiment and modeling

    International Nuclear Information System (INIS)

    West, W.P.; Brooks, N.H.; Allen, S.L.

    1996-09-01

    In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process

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

    International Nuclear Information System (INIS)

    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 (Βτ) - 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

  16. Effects of neutrals on plasma rotation in DIII-D

    International Nuclear Information System (INIS)

    Monier-Garbet, P.; Burrell, K.H.; Hinton, F.L.; Kim, J.; Garbet, X.; Groebner, R.J.

    1997-01-01

    Friction due to charge exchange with cold neutral atoms in the edge is investigated as a candidate to govern the poloidal rotation in the edge of a tokamak plasma. The Hirshman and Sigmar neoclassical moment approach is used to determine the rotation velocities of the main plasma ions and of one impurity species, when charge exchange friction is included. It is found that the poloidal rotation of the main plasma ions is controlled by charge exchange friction with neutrals. The impurity ion poloidal rotation is governed by the balance between the impurity viscous force and the main-ion-impurity-ion friction force. The results of the calculation are compared with the measurements obtained in the edge of a DIII-D high (H) mode plasma, using charge exchange recombination (CER) spectroscopy. It is found that the measured main ion poloidal rotation can be accurately predicted by the neoclassical theory including the effect of neutrals, assuming a neutral density n > = 3 x 10 17 m -3 at the separatrix, decreasing exponentially inside the plasma with an e-folding length of 0.012 m, and peaking near the X point region with a poloidal peaking parameter y ≡ n > 2 >/ n B 2 > = 1.5. However, for the impurity ions, the neoclassical theory including a single impurity charge state, and regardless of the effect of the neutrals, gives a prediction that has the correct sign, but whose value is a factor of 5 or 6 different from the experimental value. (author). 12 refs, 7 figs, 1 tab

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  18. Divertor particle exhaust and wall inventory on DIII-D

    International Nuclear Information System (INIS)

    Maingi, R.; Jackson, G.L.; Mahdavi, M.A.; Schaffer, M.J.; Wade, M.R.; Mioduszewski, P.K.; Hogan, J.T.; Klepper, C.C.; Haas, G.

    1995-01-01

    Many tokamaks achieve optimum plasma performance by achieving low recycling; various wall conditioning techniques including helium glow discharge cleaning (HeGDC) are routinely applied to help achieve low recycling. Many of these techniques allow strong, transient wall pumping, but they may not be effective for long-pulse tokamaks, such as the International Thermonuclear Experimental Reactor (ITER), the Tokamak Physics Experiment (TPX), Tore Supra Continu, and JT-60SU. Continuous particle exhaust using an in-situ pumping scheme may be effective for wall inventory control in such devices. Recent particle balance experiments on the Tore Supra and DIII-D tokamaks demonstrated that the wall particle inventory could be reduced during a given discharge by use of continuous particle exhaust. In this paper the authors report the first results of wall inventory control and good performance with the in-situ DIII-D cryopump, replacing the HeGDC normally applied between discharges

  19. Performance of the DIII-D neutral beam injection system

    International Nuclear Information System (INIS)

    Kim, J.; Callis, R.W.; Colleraine, A.P.; Cummings, J.; Glad, A.S.; Gootgeld, A.M.; Haskovec, J.S.; Hong, R.; Kellman, D.H.; Langhorn, A.R.

    1987-01-01

    During the upgrade of the Doublet III tokamak, the neutral beam injection system as also modified to accommodate long pulse sources and to utilize the larger entrance apertures to the torus vessel. All four beamlines on DIII-D are now in operation with a total of eight common long pulse sources. These have exhibited easier conditioning and good reproducibility. Performance results of the beamlines and supporting systems are presented, and the observed beam properties are discussed

  20. RESEARCH PROGRESS AND HARDWARE SYSTEMS AT DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    PETERSEN,P.I; THE DIII-D TEAM

    2003-10-01

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

  1. Pumping Characteristics of the DIII-D Cryopump

    International Nuclear Information System (INIS)

    A.S. Bozek; C.B. Baxi; R.W. Callis; M.A. Mahdavi; R.C. O'Neill; E.E. Reis

    1999-01-01

    Beginning in 1992, the first of the DIII-D divertor baffles and cryocondensation pumps was installed. This open divertor configuration, located on the outermost floor of the DIII-D vessel, includes a cryopump with a predicted pumping speed of 50,000 ell/s excluding obstructions such as support hardware. Taking the pump structural and support characteristics into consideration, the corrected pumping speed for D 2 is 30,000 ell/s [1]. In 1996, the second divertor baffle and cryopump were installed. This closed divertor structure, located on the outermost ceiling of the DIII-D vessel, has a cryopump with a predicted pumping speed of 32,000 ell/s. In the fall of 1999, the third divertor baffle and cryopump will be installed. This divertor structure will be located on the 45 o angled corner on the innermost ceiling of the DIII-D vessel, known as the private flux region of the plasma configuration. With hardware supports factored into the pumping speed calculation, the private flux cryopump is expected to have a pumping speed of 15,000 ell/s. There was question regarding the effectiveness of the private flux cryopump due to the close proximity of the private flux baffle. This led to a conductance calculation study of the impact of rotating the cryopump aperture by 180 o to allow for greater particle and gas exhaust into the cryopump's helium panel. This study concluded that the cost and schedule impact of changing the private flux cryopump orientation and design did not warrant the possible 20% (3,000 ell/s) increase in pumping ability gained by rotating the cryopump aperture 180 o . The comparison of pumping speed of the first two cryocondensation pumps with the measured results will be presented as well as the calculation of the pumping speed for the private flux cryopump now being installed

  2. Effects of low-Z and high-Z impurities on divertor detachment and plasma confinement

    Directory of Open Access Journals (Sweden)

    H.Q. Wang

    2017-08-01

    Full Text Available The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss power to the l-H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection with up to 50% of the injected power radiating in the core cools the pedestal and core plasmas, thus significantly degrading the confinement. As for Ne seeding, medium confinement with H98∼0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N2 seeded plasmas, radiation is predominately confined in the boundary plasma, which leads to less effect on the confinement and pedestal. In the case of strong N2 gas puffing, the confinement recovers during the detachment, from ∼20% reduction at the onset of the detachment to greater than unity comparable to that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.

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

    International Nuclear Information System (INIS)

    Politzer, P.A.

    2005-01-01

    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 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 ∼80% bootstrap current in stationary discharges without inductive current drive. The remaining current is ∼20% NBCD. This is achieved at β N [approximately equal to] β p > 3, but at relatively high q 95 (∼10). In lower q 95 Advanced Tokamak plasmas, f bs ∼ 0.6 has been reached in essentially noninductive plasmas. The phenomenology of high β p and β 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 β, limiting the achievable average β 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

  4. Upgrade of the DIII-D RF systems

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  5. Divertor extreme ultraviolet (EUV) survey spectroscopy in DIII-D

    Science.gov (United States)

    McLean, Adam; Allen, Steve; Ellis, Ron; Jarvinen, Aaro; Soukhanovskii, Vlad; Boivin, Rejean; Gonzales, Eduardo; Holmes, Ian; Kulchar, James; Leonard, Anthony; Williams, Bob; Taussig, Doug; Thomas, Dan; Marcy, Grant

    2017-10-01

    An extreme ultraviolet spectrograph measuring resonant emissions of D and C in the lower divertor has been added to DIII-D to help resolve an 2X discrepancy between bolometrically measured radiated power and that predicted by boundary codes for DIII-D, JET and ASDEX-U. With 290 and 450 gr/mm gratings, the DivSPRED spectrometer, an 0.3 m flat-field McPherson model 251, measures ground state transitions for D (the Lyman series) and C (e.g., C IV, 155 nm) which account for >75% of radiated power in the divertor. Combined with Thomson scattering and imaging in the DIII-D divertor, measurements of position, temperature and fractional power emission from plasma components are made and compared to UEDGE/SOLPS-ITER. Mechanical, optical, electrical, vacuum, and shielding aspects of DivSPRED are presented. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698 and DE-AC52-07NA27344, and by the LLNL Laboratory Directed R&D Program, project #17-ERD-020.

  6. Engineering design of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

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

    1995-10-01

    A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The Radiative Divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the Radiative Divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The Radiative Divertor design is very flexible, and will allow physics studies of the effects of slot width and length. Radiative Divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The Radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D Advanced Divertor program form a foundation for the design work on the Radiative Divertor

  7. Overview of the DIII-D program computer systems

    International Nuclear Information System (INIS)

    McHarg, B.B. Jr.

    1997-11-01

    Computer systems pervade every aspect of the DIII-D National Fusion Research program. This includes real-time systems acquiring experimental data from data acquisition hardware; cpu server systems performing short term and long term data analysis; desktop activities such as word processing, spreadsheets, and scientific paper publication; and systems providing mechanisms for remote collaboration. The DIII-D network ties all of these systems together and connects to the ESNET wide area network. This paper will give an overview of these systems, including their purposes and functionality and how they connect to other systems. Computer systems include seven different types of UNIX systems (HP-UX, REALIX, SunOS, Solaris, Digital UNIX, Ultrix, and IRIX), OpenVMS systems (both BAX and Alpha), MACintosh, Windows 95, and more recently Windows NT systems. Most of the network internally is ethernet with some use of FDDI. A T3 link connects to ESNET and thus to the Internet. Recent upgrades to the network have notably improved its efficiency, but the demand for bandwidth is ever increasing. By means of software and mechanisms still in development, computer systems at remote sites are playing an increasing role both in accessing and analyzing data and even participating in certain controlling aspects for the experiment. The advent of audio/video over the interest is now presenting a new means for remote sites to participate in the DIII-D program

  8. Interprocess communication within the DIII-D plasma control system

    International Nuclear Information System (INIS)

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

    1999-06-01

    The DIII-D tokamak fusion research experiment's real-time digital plasma control system (PCS) is a complex and ever evolving system. During a plasma experiment, it is tasked with some of the most crucial functions at DIII-D. Key responsibilities of the PCS involve sub-system control, data acquisition/storage, and user interface. To accomplish these functions, the PCS is broken down into individual components (both software and hardware), each capable of handling a specific duty set. Constant interaction between these components is necessary prior, during and after a standard plasma cycle. Complicating the matter even more is that some components, mostly those which deal with user interaction, may exist remotely, that is to say they are not part of the immediate hardware which makes up the bulk of the PCS. The four main objectives of this paper are to (1) present a brief outline of the PCS hardware/software and how they relate to each other; (2) present a brief overview of a standard DIII-D plasma cycle (a shot); (3) using three sets of PCS sub-systems, describe in more detail the communication processes; and (4) evaluate the benefits and drawbacks of said systems

  9. Analysis of plasma coupling with the prototype DIII-D ICRF antenna

    International Nuclear Information System (INIS)

    Ryan, P.M.; Hoffman, D.J.; Bigelow, T.S.; Baity, F.W.; Gardner, W.L.; Mayberry, M.J.; Rothe, K.E.

    1988-01-01

    Coupling to plasma in the H-mode is essential to the success of future ignited machines such as CIT. To ascertain voltage and current requirements for high-power second harmonic heating (2 MW in a 35- by 50-cm port), coupling to the DIII-D tokamak with a prototype compact loop antenna has been measured. The results show good loading for L-mode and limiter plasmas, but coupling 2 MW to an H-mode plasma demands voltages and currents near the limit of present technology. We report the technological analysis and progress that allow coupling of these power densities. 5 refs., 4 figs

  10. Analysis of plasma coupling with the prototype DIII-D ICRF antenna

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, P.M.; Hoffman, D.J.; Bigelow, T.S.; Baity, F.W.; Gardner, W.L.; Mayberry, M.J.; Rothe, K.E.

    1988-01-01

    Coupling to plasma in the H-mode is essential to the success of future ignited machines such as CIT. To ascertain voltage and current requirements for high-power second harmonic heating (2 MW in a 35- by 50-cm port), coupling to the DIII-D tokamak with a prototype compact loop antenna has been measured. The results show good loading for L-mode and limiter plasmas, but coupling 2 MW to an H-mode plasma demands voltages and currents near the limit of present technology. We report the technological analysis and progress that allow coupling of these power densities. 5 refs., 4 figs.

  11. Measurement of local, internal magnetic fluctuations via cross-polarization scattering in the DIII-D tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Barada, K., E-mail: kshitish@ucla.edu; Rhodes, T. L.; Crocker, N. A.; Peebles, W. A. [University of California-Los Angeles, P.O. Box 957099, Los Angeles, California 90095 (United States)

    2016-11-15

    We present new measurements of internal magnetic fluctuations obtained with a novel eight channel cross polarization scattering (CPS) system installed on the DIII-D tokamak. Measurements of internal, localized magnetic fluctuations provide a window on an important physics quantity that we heretofore have had little information on. Importantly, these measurements provide a new ability to challenge and test linear and nonlinear simulations and basic theory. The CPS method, based upon the scattering of an incident microwave beam into the opposite polarization by magnetic fluctuations, has been significantly extended and improved over the method as originally developed on the Tore Supra tokamak. A new scattering geometry, provided by a unique probe beam, is utilized to improve the spatial localization and wavenumber range. Remotely controllable polarizer and mirror angles allow polarization matching and wavenumber selection for a range of plasma conditions. The quasi-optical system design, its advantages and challenges, as well as important physics validation tests are presented and discussed. Effect of plasma beta (ratio of kinetic to magnetic pressure) on both density and magnetic fluctuations is studied and it is observed that internal magnetic fluctuations increase with beta. During certain quiescent high confinement operational regimes, coherent low frequency modes not detected by magnetic probes are detected locally by CPS diagnostics.

  12. The Resistive Wall Mode Feedback Control System on DIII-D

    International Nuclear Information System (INIS)

    J.T. Scoville; D.H. Kellman; S.G.E. Pronko; A. Nerem; R. Hatcher; D. O'Neill; G. Rossi; M. Bolha

    1999-01-01

    One of the primary instabilities limiting the performance of the plasma in advanced tokamak operating regimes is the resistive wall mode (RWM) [1]. The most common RWM seen in the DIII-D tokamak is originated by an n=1 ideal external kink mode which, in the presence of a resistive wall, is converted to a slowly growing RWM. The mode causes a reduction in plasma rotation, a loss of stored energy, and sometimes leads to plasma disruption. It routinely limits the performance of a tokamak operating near reactor relevant parameter levels. A system designed to actively control the RWM has recently been installed on the DIII-D tokamak for the control of low m n=1 modes. In initial experiments, the control system has been capable of delaying the onset of RWMs in energetic discharges for several hundred milliseconds. The feedback control system consists of detector coils connected via control software to high power current amplifiers driving the excitation coils. The three pairs of excitation coils are each driven by a current amplifier and a DC power supply. The control signal is derived from a set of six sensor coils that measure radial flux as low as one Gauss. The signals are digitally processed by realtime software in the DIII-D Plasma Control System (PCS) to create a command that is sent to the current amplifier, with a cycle time of approximately 100 micros. The amplifiers, designed and fabricated by Robicon Corporation to a specification developed by PPPL and GA, are bipolar devices capable of ±5 kA at 300 V, with an operating bandwidth of approximately 800 Hz (-3 dB)

  13. Pedestal width and ELM size identity studies in JET and DIII-D; implications for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Beurskens, M N A; Lomas, P; Saarelma, S; Balboa, I; Flanagan, J; Giroud, C; Kempenaars, M [EURATOM/UKAEA Fusion Association, Culham Sc. Centre, Abingdon, OX14 3DB (United Kingdom); Osborne, T H; Groebner, R; Leonard, A; Snyder, P B; Bray, B [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Horton, L D [JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Frassinetti, L [Association EURATOM-VR, Alfven Laboratory, School of Electrical Engineering, KTH, Stockholm (Sweden); Nunes, I [Centro de Fusao Nuclear, Associacao EURATOM-IST, Lisboa (Portugal); Crombe, K [Department of Applied Physics, Ghent University, Rozier 44, 9000 Gent (Belgium); Giovannozzi, E [Associazione EURATOM-ENEA Sulla Fusione, Consorzio RFX Padova (Italy); Kohen, N [Association EURATOM-CEA, CEA/DSM/DRFC-Cadarache 13108, St Paul Durance (France); Loarte, A [ITER Organization, CS 90 046, F-13067 Saint Paul lez Durance Cedex (France); Loennroth, J, E-mail: Marc.Beurskens@jet.u [Association EURATOM-Tekes, Helsinki University of Technology (Finland)

    2009-12-15

    The dependence of the H-mode edge transport barrier width on normalized ion gyroradius (rho* = rho/a) in discharges with type I ELMs was examined in experiments combining data for the JET and DIII-D tokamaks. The plasma configuration as well as the local normalized pressure (beta), collisionality (nu*), Mach number and the ratio of ion and electron temperature at the pedestal top were kept constant, while rho* was varied by a factor of four. The width of the steep gradient region of the electron temperature (T{sub e}) and density (n{sub e}) pedestals normalized to machine size showed no or only a weak trend with rho*. A rho{sup 1/2} or rho{sup 1} dependence of the pedestal width, given by some theoretical predictions, is not supported by the current experiments. This is encouraging for the pedestal scaling towards ITER as it operates at lower rho* than existing devices. Some differences in pedestal structure and ELM behaviour were, however, found between the devices; in the DIII-D discharges, the n{sub e} and T{sub e} pedestal were aligned at high rho* but the n{sub e} pedestal shifted outwards in radius relative to T{sub e} as rho* decreases, while on JET the profiles remained aligned while rho* was scanned by a factor of two. The energy loss at an ELM normalized to the pedestal energy increased from 10% to 40% as rho* increased by a factor of two in the DIII-D discharges but no such variation was observed in the case of JET. The measured pedestal pressures and widths were found to be consistent with the predictions from modelling based on peeling-ballooning stability theory, and are used to make projections towards ITER

  14. Absolute calibration of a SPRED [Spectrometer Recording Extended Domain] EUV [extreme ultraviolet] spectrograph for use on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Wood, R.D.; Allen, S.L.

    1988-01-01

    We have performed an absolute intensity calibration of a SPRED multichannel EUV spectrograph using synchrotron radiation from the NBS SURF-II electron storage ring. The calibration procedure and results for both a survey grating (450 g/mm) and a high-resolution (2100 g/mm) grating are presented. The spectrograph is currently in use on the DIII-D tokamak with a tangential line-of-sight at the plasma midplane. Data is first acquired and processed by a microcomputer; the absolute line intensities are then sent to the DIII-D database for comparison with data from other diagnostics. Representative data from DIII-D plasma operations will be presented. 6 refs., 3 figs., 1 tab

  15. Development of a Closed Loop Simulator for Poloidal Field Control in DIII-D

    International Nuclear Information System (INIS)

    J.A. Leuer; M.L. Walker; D.A. Humphreys; J.R. Ferron; A. Nerem; B.G. Penaflor

    1999-01-01

    The design of a model-based simulator of the DIII-D poloidal field system is presented. The simulator is automatically configured to match a particular DIII-D discharge circuit. The simulator can be run in a data input mode, in which prior acquired DIII-D shot data is input to the simulator, or in a stand-alone predictive mode, in which the model operates in closed loop with the plasma control system. The simulator is used to design and validate a multi-input-multi-output controller which has been implemented on DIII-D to control plasma shape. Preliminary experimental controller results are presented

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. Dependence of the DIII-D beta limit on the current profile

    Energy Technology Data Exchange (ETDEWEB)

    Strait, E.J.; Chu, M.S.; Ferron, J.R.; Lao, L.L.; Osborne, T.H.; Taylor, T.S.; Turnbull, A.D. (General Atomics, San Diego, CA (United States)); Lazarus, E.A. (Oak Ridge National Lab., TN (United States))

    1991-01-01

    The maximum beta values achieved in DIII-D are not fully described by the simple scaling law [beta][sub max][proportional to]I/aB. There is, in addition, a dependence on the form of the current profile as parameterized by the safety factor q and internal inductance l[sub i]. The maximum experimentally achieved value of normalized beta [beta][sub N] = [beta]/(I/aB) varies from 3.5 at low safety factor q (q[sub 95]<3) to 5 at higher values of q. At low q, discharges are terminated by disruptions at high [beta][sub N] and at both the low and high l[sub i] boundaries of the stable range. These disruptions are attributed to external and global kink modes. At higher q, such disruptions are much less frequent, and beta is limited by slowly growing resistive modes, fishbones, and possibly by ballooning modes. At each value of q, the maximum beta tends to increase with internal inductance l[sub i]. A numerical study of kink mode stability has shown a similar trend for optimized pressure profiles. These observations have suggested a new scaling law for the operational beta limit: [beta][sub max]=4l[sub i](I/aB), which fits the DIII-D data well. (author) 13 refs., 4 figs.

  18. DEMONSTRATION IN THE DIII-D TOKAMAK OF AN ALTERNATE BASELINE SCENARIO FOR ITER AND OTHER BURNING PLASMA EXPERIMENTS

    International Nuclear Information System (INIS)

    LUCE, T.C.; WADE, M.R.; FERRON, J.R.; HYATT, A.W.; KELLMAN, A.G.; KINSEY, J.E.; LAHAY, R.J.; LASNIER, C.J.; MURAKAMI, M.; POLITZER, P.A.; SCOVILLE, J.T.

    2002-01-01

    OAK A271 DEMONSTRATION IN THE DIII-D TOKAMAK OF AN ALTERNATE BASELINE SCENARIO FOR ITER AND OTHER BURNING PLASMA EXPERIMENTS. Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak in stationary conditions with relatively low plasma current (q 95 > 4). A figure of merit for fusion gain Β N H 89 /q 95 2 has been maintained at values corresponding to Q = 10 operation in a burning plasma for > 6 s or 36 τ E and 2 τ R . The key element is the relaxation of the current profile to a stationary state with q min > 1, which allows stable operation up to the no-wall ideal β limit. These plasmas maintain particle balance by active pumping rather than transient wall conditions. The reduced current lessens significantly the potential for structural damage in the event of a major disruption

  19. Real-time identification of the resistive-wall-mode in DIII-D with Kalman filter ELM discrimination

    International Nuclear Information System (INIS)

    Edgell, D.H.; Fransson, C.M.; Humphreys, D.A.; Ferron, J.R.; Garofalo, A.M.; Kim, J.S.; La Haye, R.J.; Okabayashi, M.; Reimerdes, H.; Strait, E.J.; Turnbull, A.D.

    2004-01-01

    The resistive-wall-mode (RWM) is a major performance-limiting instability in present-day tokamaks. Active control and stabilization of the mode will almost certainly be essential for the success of advanced tokamaks and for the economic viability of tokamak fusion reactors. High performance tokamak plasmas often experience edge-localized-modes (ELMs) which can interfere with RWM identification and control. If the RWM control scheme reacts to an ELM the RWM may be driven unstable instead of controlled. An algorithm for real-time identification of the RWM with discrimination of ELMs in the DIII-D tokamak has been developed using a combination of matched filter and Kalman filter methods. The algorithm has been implemented in DIII-D's real-time plasma control system (PCS) and is available to drive active mode control schemes

  20. Signal processing techniques for lithium beam polarimetry on DIII-D

    International Nuclear Information System (INIS)

    Thomas, D. M.; Leonard, A. W.

    2006-01-01

    On the DIII-D tokamak the LIBEAM diagnostic provides precise measurements of the local magnetic field direction by combined polarimetry/ spectroscopy of the Zeeman-split 2S-2P lithium resonance line. Using these measurements we are able to determine the behavior of the edge toroidal current density j φ (r), a parameter of critical interest for edge stability and performance. For a successful measurement, analysis of the polarization state of the spectrally filtered fluorescence must be done with high precision in the presence of nonideal filtering, beam intensity evolution, and dynamically varying background light. This is accomplished by polarization modulation of the collected emission, followed by digital demodulation at various harmonics of the modulation frequency. Either lock-in or fast Fourier transform techniques can be used to determine the various Stokes parameters and reconstruct the field directions based on accurate spatial and polarization efficiency calibrations. Details of the specific techniques used to analyze various DIII-D discharges are described, along with a discussion of the present limitations and some possible avenues towards improving the analysis

  1. Measurements and modeling of intra-ELM tungsten sourcing and transport in DIII-D

    Science.gov (United States)

    Abrams, T.; Leonard, A. W.; Thomas, D. M.; McLean, A. G.; Makowski, M. A.; Wang, H. Q.; Unterberg, E. A.; Briesemeister, A. R.; Rudakov, D. L.; Bykov, I.; Donovan, D.

    2017-10-01

    Intra-ELM tungsten erosion profiles in the DIII-D divertor, acquired via W I spectroscopy with high temporal and spatial resolution, are consistent with SDTrim.SP sputtering modeling using measured ion saturation currents and impact energies during ELMs as input and an ad-hoc 2% C2+ impurity flux. The W sputtering profile peaks close to the OSP both during and between ELMs in the favorable BT direction. In reverse BT the W source peaks close to the OSP between ELMs but strongly broadens and shifts outboard during ELMs, heuristically consistent with radially outward ion transport via ExB drifts. Ion impact energies during ELMs (inferred taking the ratio of divertor heat flux to the ion saturation current) are found to be approximately equal to Te,ped, lower than the 4*Te,ped value predicted by the Fundamenski/Moulton free streaming model. These impact energies imply both D main ions and C impurities contribute strongly to W sputtering during ELMs on DIII-D. This work represents progress towards a predictive model to link upstream conditions (i.e., pedestal height) and SOL impurity levels to the ELM-induced W impurity source at both the strike-point and far-target regions in the ITER divertor. Correlations between ELM size/frequency and SOL W fluxes measured via a midplane deposition probe will also be presented. Work supported by US DOE under DE-FC02-04ER54698.

  2. Active and passive spectroscopic imaging in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Van Zeeland, M A; Brooks, N H; Burrell, K H; Groebner, R J; Hyatt, A W; Luce, T C; Wade, M R; Yu, J H; Pablant, N; Heidbrink, W W; Solomon, W M

    2010-01-01

    Wide-angle, 2D imaging of Doppler-shifted, Balmer alpha (D α ) emission from high energy injected neutrals, charge exchange recombination (CER) emission from neutral beam interaction with thermal ions and fully stripped impurity ions and visible bremsstrahlung (VB) from the core of DIII-D plasmas has been carried out. Narrowband interference filters were used to isolate the specific wavelength ranges of visible radiation for detection by a tangentially viewing, fast-framing camera. Measurements of the D α emission from fast neutrals injected into the plasma from the low field side reveal the vertical distribution of the beam, its divergence and the variation in its radial penetration with density. Modeling of this emission using both a full Monte Carlo collisional radiative code as well as a simple beam attenuation code coupled to Atomic Data and Analysis Structure emissivity lookup tables yields qualitative agreement, however the absolute magnitudes of the emissivities in the predicted distribution are larger than those measured. Active measurements of carbon CER brightness are in agreement with those made independently along the beam midplane using DIII-D's multichordal, CER spectrometer system, confirming the potential of this technique for obtaining 2D profiles of impurity density. Passive imaging of VB, which can be inverted to obtain local emissivity profiles, is compared with measurements from both a calibrated filter/photomultiplier array and the standard multichordal CER spectrometer system.

  3. Investigation of runaway electron dissipation in DIII-D using a gamma ray imager

    Science.gov (United States)

    Lvovskiy, A.; Paz-Soldan, C.; Eidietis, N.; Pace, D.; Taussig, D.

    2017-10-01

    We report the findings of a novel gamma ray imager (GRI) to study runaway electron (RE) dissipation in the quiescent regime on the DIII-D tokamak. The GRI measures the bremsstrahlung emission by RE providing information on RE energy spectrum and distribution across a poloidal cross-section. It consists of a lead pinhole camera illuminating a matrix of BGO detectors placed in the DIII-D mid-plane. The number of detectors was recently doubled to provide better spatial resolution and additional detector shielding was implemented to reduce un-collimated gamma flux and increase single-to-noise ratio. Under varying loop voltage, toroidal magnetic field and plasma density, a non-monotonic RE distribution function has been revealed as a result of the interplay between electric field, synchrotron radiation and collisional damping. A fraction of the high-energy RE population grows forming a bump at the RE distribution function while synchrotron radiation decreases. A possible destabilizing effect of Parail-Pogutse instability on the RE population will be also discussed. Work supported by the US DOE under DE-FC02-04ER54698.

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

    Science.gov (United States)

    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.

  5. Alfven eigenmode observations on DIII-D via two-colour CO2 interferometry

    International Nuclear Information System (INIS)

    Zeeland, M A van; Kramer, G J; Nazikian, R; Berk, H L; Carlstrom, T N; Solomon, W M

    2005-01-01

    Measurements are presented of toroidicity-induced (TAEs) and reverse shear (RSAEs) Alfven eigenmodes made using the standard two-colour CO 2 interferometer on DIII-D modified for increased bandwidth. Typical values of the effective line-integrated density perturbation in DIII-D are found to be d(nL)/nL ∼ 10 -3 , and comparisons are made with NOVA calculations as well as magnetic measurements. There is a strong difference in the measured power spectrum between vertical and radial chords through the plasma. On average, vertical views are characterized by a larger line-integrated density perturbation due to TAEs than radial chords. Radial chords, however, can be used much more reliably than vertical chords to identify the presence of RSAEs in the plasma-a result found to be due to the radially localized nature of these modes. In general, the apparent amplitude of the observed modes for both TAE and RSAE is found to be highly dependent on viewing location. (letter to the editor)

  6. Active and passive spectroscopic imaging in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Van Zeeland, M A; Brooks, N H; Burrell, K H; Groebner, R J; Hyatt, A W; Luce, T C; Wade, M R [General Atomics, PO Box 85608 San Diego, CA 92186-5608 (United States); Yu, J H; Pablant, N [University of California-San Diego, 9500 Gilman Drive, La Jolla, CA 92093 (United States); Heidbrink, W W [University of California-Irvine, 4129 Frederick Reines Hall, Irvine, CA 92697 (United States); Solomon, W M, E-mail: vanzeeland@fusion.gat.co [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2010-04-15

    Wide-angle, 2D imaging of Doppler-shifted, Balmer alpha (D{sub a}lpha) emission from high energy injected neutrals, charge exchange recombination (CER) emission from neutral beam interaction with thermal ions and fully stripped impurity ions and visible bremsstrahlung (VB) from the core of DIII-D plasmas has been carried out. Narrowband interference filters were used to isolate the specific wavelength ranges of visible radiation for detection by a tangentially viewing, fast-framing camera. Measurements of the D{sub a}lpha emission from fast neutrals injected into the plasma from the low field side reveal the vertical distribution of the beam, its divergence and the variation in its radial penetration with density. Modeling of this emission using both a full Monte Carlo collisional radiative code as well as a simple beam attenuation code coupled to Atomic Data and Analysis Structure emissivity lookup tables yields qualitative agreement, however the absolute magnitudes of the emissivities in the predicted distribution are larger than those measured. Active measurements of carbon CER brightness are in agreement with those made independently along the beam midplane using DIII-D's multichordal, CER spectrometer system, confirming the potential of this technique for obtaining 2D profiles of impurity density. Passive imaging of VB, which can be inverted to obtain local emissivity profiles, is compared with measurements from both a calibrated filter/photomultiplier array and the standard multichordal CER spectrometer system.

  7. The control of divertor carbon erosion/redeposition in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Whyte, D.G.; West, W.P.; Wong, C.P.C.

    2001-01-01

    The DIII-D tokamak has demonstrated an operational scenario where the graphite-covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplished using a low temperature (detached) divertor plasma that eliminates physical sputtering. Likewise, the carbon source rate arising from chemical erosion is found to be very low in the detached divertor. Near strikepoint regions, the rate of carbon deposition is ∼3 cm/burn-year, with a corresponding hydrogenic codeposition rate >1kg/m 2 /burn-year; rates both problematic for steady-state fusion reactors. The carbon net deposition rate in the divertor is consistent with carbon arriving from the core plasma region. Carbon influx from the main wall is measured to be relatively large in the high-density detached regime and is of sufficient magnitude to account for the deposition rate in the divertor. Divertor redeposition is therefore determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D with detachment, no significant reduction is found in the core plasma carbon density, illustrating the importance of non-divertor erosion and the complex coupling between erosion/redeposition and impurity plasma transport. (author)

  8. Plasma rotation and rf heating in DIII-D

    International Nuclear Information System (INIS)

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

    1999-05-01

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

  9. Design of the vacuum control system for DIII-D

    International Nuclear Information System (INIS)

    Campbell, G.L.; Callis, R.W.; Haskovec, J.S.; Heckman, E.J.; Moore, C.D.; Scoville, J.T.

    1986-01-01

    The vacuum control and instrumentation for the DIII-D upgrade was designed using a new large programmable controller with color graphic operator interfaces and intelligent distributed devices. Remote, optically isolated input and output is used as well as optical isolation for the operator and programming consoles. Gate valves between experimental equipment and the vacuum vessel are interlocked for machine safety by an intelligent interface based upon a commercially available microcontroller card. Complete automatic operation with capability for remote operator intervention was one goal of this design effort. The design of the system with emphasis on the graphics, optical isolation and microcontroller implementation will be discussed

  10. A DESIGN RETROSPECTIVE OF THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    LUXON, J.L

    2001-06-01

    OAK-B135 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 program. This paper gives an integrated picture of the facility and its capabilities

  11. Plasma rotation and rf heating in DIII-D

    International Nuclear Information System (INIS)

    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.

    1999-01-01

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

  12. Measurement and Modelling of Tearing Mode Stability for Steady-State Plasmas in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Turco, F; Luce, T; Ferron, J; Petty, C; Politzer, P; Turnbull, A; Brennan, D; Murakami, M; LoDestro, L; Pearlstein, L; Casper, T; Jayakumar, R; Holcomb, C

    2009-06-23

    High-beta, quasi-steady state scenarios represent a fundamental step towards the performance required for future fusion reactors. In DIII-D steady-state scenario discharges, the normalized beta {beta}{sub N} {triple_bond} {beta}(%) {center_dot} a(m) {center_dot} B{sub T}(T)/I{sub p}(MA) (where {beta} is the ratio of the plasma pressure to the magnetic field pressure, {alpha} the plasma minor radius, B{sub T} the toroidal magnetic field and I{sub p} the plasma current) exceeds the no-wall ideal kink beta limit. The performance of this scenario is limited by the onset of an n = 1 tearing mode, which appears on the resistive evolution time-scale (1-2 s) at constant pressure and causes both a loss of confinement and a radial redistribution of the current density from which the available current drive sources cannot recover. It is routinely observed that the injection of electron cyclotron current drive (ECCD), with a broad deposition localized around {rho} {approx} 0.35, can prevent the mode from appearing. It must be noted that this is not a case of a direct stabilization due to the interaction with the mode's rational surface. These variations of the scenario are illustrated in Fig. 1, where the total injected power [neutral beam injection (NBI) and ECCD], {beta}{sub N} and the n = 1 magnetic perturbation at the outer wall are shown. In case (a), the onset of the n = 1 mode is observed when the EC power is not present or if it is stopped before the end of the high {beta} phase, whereas in case (b) the difference is pointed out between broad and narrow current deposition (with the narrow deposition case becoming unstable). The current density profile evolution and the MHD modes of several sets of significant discharges with and without ECCD (at different locations) have been analyzed, using motional Stark effect (MSE) spectroscopy measurements for the former and edge magnetic probes measurements, toroidal rotation profiles and fast electron cyclotron emission

  13. Measurement and Modelling of Tearing Mode Stability for Steady-State Plasmas in DIII-D

    International Nuclear Information System (INIS)

    Turco, F.; Luce, T.; Ferron, J.; Petty, C.; Politzer, P.; Turnbull, A.; Brennan, D.; Murakami, M.; LoDestro, L.; Pearlstein, L.; Casper, T.; Jayakumar, R.; Holcomb, C.

    2009-01-01

    High-beta, quasi-steady state scenarios represent a fundamental step towards the performance required for future fusion reactors. In DIII-D steady-state scenario discharges, the normalized beta β N (triple b ond) β(%) · a(m) · B T (T)/I p (MA) (where β is the ratio of the plasma pressure to the magnetic field pressure, α the plasma minor radius, B T the toroidal magnetic field and I p the plasma current) exceeds the no-wall ideal kink beta limit. The performance of this scenario is limited by the onset of an n = 1 tearing mode, which appears on the resistive evolution time-scale (1-2 s) at constant pressure and causes both a loss of confinement and a radial redistribution of the current density from which the available current drive sources cannot recover. It is routinely observed that the injection of electron cyclotron current drive (ECCD), with a broad deposition localized around ρ ∼ 0.35, can prevent the mode from appearing. It must be noted that this is not a case of a direct stabilization due to the interaction with the mode's rational surface. These variations of the scenario are illustrated in Fig. 1, where the total injected power (neutral beam injection (NBI) and ECCD), β N and the n = 1 magnetic perturbation at the outer wall are shown. In case (a), the onset of the n = 1 mode is observed when the EC power is not present or if it is stopped before the end of the high β phase, whereas in case (b) the difference is pointed out between broad and narrow current deposition (with the narrow deposition case becoming unstable). The current density profile evolution and the MHD modes of several sets of significant discharges with and without ECCD (at different locations) have been analyzed, using motional Stark effect (MSE) spectroscopy measurements for the former and edge magnetic probes measurements, toroidal rotation profiles and fast electron cyclotron emission (ECE) data for the latter. One equilibrium based on EFIT reconstruction (1) with kinetic

  14. CONFINEMENT OF HIGH TEMPERATURE PLASMA

    Science.gov (United States)

    Koenig, H.R.

    1963-05-01

    The confinement of a high temperature plasma in a stellarator in which the magnetic confinement has tended to shift the plasma from the center of the curved, U-shaped end loops is described. Magnetic means are provided for counteracting this tendency of the plasma to be shifted away from the center of the end loops, and in one embodiment this magnetic means is a longitudinally extending magnetic field such as is provided by two sets of parallel conductors bent to follow the U-shaped curvature of the end loops and energized oppositely on the inside and outside of this curvature. (AEC)

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

    International Nuclear Information System (INIS)

    STAMBAUGH, RD

    2003-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Allen, S.L.; Calderon, M.O.; Ellis, R.M.

    1988-01-01

    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

  18. A phase contrast interferometer on DIII-D

    International Nuclear Information System (INIS)

    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 x 10 7 cm -3 with a 1 MHz bandwidth can be resolved. The diagnostic employs a 7.6 cm diameter CO 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

  19. Tritium in the DIII-D carbon tiles

    International Nuclear Information System (INIS)

    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. The ground-fault detection system for DIII-D

    International Nuclear Information System (INIS)

    Scoville, J.T.; Petersen, P.I.

    1987-10-01

    This paper presents a discussion of the ground-fault detection systems on the DIII-D tokamak. The subsystems that must be monitored for an inadvertent ground include the toroidal and poloidal coil systems, the vacuum vessel, and the coil support structures. In general, one point of each coil is tied to coil/power supply ground through a current limiting resistor. For ground protection the current through this resistor is monitored using a dynamically feedback balanced Hall probe transducer from LEM Industries. When large inductive currents flow in closed loops near the tokamak, the result is undesirable magnetic error fields in the plasma region and noise generation on signal cables. Therefore, attention must be paid to avoid closed loops in the design of the coil and vessel support structure. For DIII-D a concept of dual insulating breaks and a single-point ground for all structure elements was used to satisfy this requirement. The integrity of the support structure is monitored by a system which continuously attempts to couple a variable frequency waveform onto these single-point grounds. The presence of an additional ground completes the circuit resulting in current flow. A Rogowski coil is then used to track the unwanted ground path in order to eliminate it. Details of the ground fault detection circuitry, and a description of its operation will be presented. 2 refs., 7 figs

  1. A fast scanning probe for DIII--D

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. The back transition and hysteresis effects in DIII-D

    International Nuclear Information System (INIS)

    Thomas, D.M.; Groebner, R.J.; Burrell, K.H.; Osborne, T.H.; Carlstrom, T.N.

    1997-09-01

    The back transition from H-mode to L-mode has been studied on DIII-D as a part of the investigation of the L-H transition power threshold scaling. Based on a density-dependent scaling for the H-mode power threshold, ITER will require substantial hysteresis in this parameter to remain in H-mode as n e rises. Defining the hysteresis in terms of the ratio of sustaining to threshold power, P HL /P LH may need to be as small as 50% for ITER. Operation of DIII-D at injection powers slightly above the H-mode threshold results in an oscillatory behavior with multiple forward-backward transitions in the course of a discharge. These discharges represent a unique system for studying various control parameters that may influence the H↔L state transition. Careful analysis of the power flow through the edge gives values for the sustaining power which are well below the corresponding threshold powers (P HL /P LH = 35--70%), indicating substantial hysteresis can be achieved in this parameter. Studies of other control parameter candidates such as edge temperature during the back transitions are less clear: the amount of hysteresis seen in these parameters, if any, is primarily dependent on the nature (ELMing, ELM-free) of the parent H-state

  3. UEDGE code comparisons with DIII-D bolometer data

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Schissel, D.P.

    1998-09-01

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

  5. Comprehensive energy transport scalings derived from DIII-D similarity experiments

    International Nuclear Information System (INIS)

    Petty, C.C.; Luce, T.C.; Baity, F.W.

    1998-12-01

    The dependences of heat transport on the dimensionless plasma physics parameters has been measured for both L-mode and H-mode plasmas on the DIII-D tokamak. Heat transport in L-mode plasmas has a gyroradius scaling that is gyro-Bohm-like for electrons and worse than Bohm-like for ions, with no measurable beta or collisionality dependence; this corresponds to having an energy confinement time that scales like τ E ∝ n 0.5 P -0.5 . H-mode plasmas have gyro-Bohm-like scaling of heat transport for both electrons and ions, weak beta scaling, and moderate collisionality scaling. In addition, H-mode plasmas have a strong safety factor scaling (χ ∼ q 2 ) at all radii. Combining these four dimensionless parameter scalings together gives an energy confinement time scaling for H-mode plasmas like τ E ∝ B -1 ρ -3.15 β 0.03 v -0.42 q 95 -1.43 ∝ I 0.84 B 0.39 n 0.18 P -0.41 L 2.0 , which is similar to empirical scalings derived from global confinement databases

  6. Comprehensive energy transport scalings derived from DIII-D similarity experiments

    International Nuclear Information System (INIS)

    Petty, C.C.; Luce, T.C.; Baity, F.W.

    1999-01-01

    The dependences of heat transport on the dimensionless plasma physics parameters has been measured for both L-mode and H-mode plasmas on the DIII-D tokamak. Heat transport in L-mode plasmas has a gyroradius scaling that is gyro-Bohm-like for electrons and worse than Bohm-like for ions, with no measurable beta or collisionality dependence; this corresponds to having an energy confinement time that scales like τ E ∝ n 0.5 P -0.5 . H-mode plasmas have gyro-Bohm-like scaling of heat transport for both electrons and ions, weak beta scaling, and moderate collisionality scaling. In addition, H-mode plasmas have a strong safety factor scaling (χ ∼ q 2 ) at all radii. Combining these four dimensionless parameter scalings together gives an energy confinement time scaling for H-mode plasmas like τ E ∝ B -1 ρ -3.15 β 0.03 ν -0.42 q 95 -1.43 ∝ I 0.84 B 0.39 n 0.18 P -0.41 L 2.0 , which is similar to empirical scalings derived from global confinement databases. (author)

  7. Comprehensive energy transport scalings derived from DIII-D similarity experiments

    International Nuclear Information System (INIS)

    Petty, C.C.; Luce, T.C.; Baker, D.R.

    2001-01-01

    The dependences of heat transport on the dimensionless plasma physics parameters has been measured for both L-mode and H-mode plasmas on the DIII-D tokamak. Heat transport in L-mode plasmas has a gyroradius scaling that is gyro-Bohm-like for electrons and worse than Bohm-like for ions, with no measurable beta or collisionality dependence; this corresponds to having an energy confinement time that scales like τ E ∝n 0.5 P -0.5 . H-mode plasmas have gyro-Bohm-like scaling of heat transport for both electrons and ions, weak beta scaling, and moderate collisionality scaling. In addition, H-mode plasmas have a strong safety factor scaling (χ∼q 2 ) at all radii. Combining these four dimensionless parameter scalings together gives an energy confinement time scaling for H-mode plasmas like τ E ∝ B -1 ρ -3.15 β 0.03 ν -0.42 q 95 -1.43 ∝ I 0.84 B 0.39 n 0.18 P -0.41 L 2.0 , which is similar to empirical scalings derived from global confinement databases. (author)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  10. Design and Control of Small Neutral Beam Arc Chamber for Investigations of DIII-D Neutral Beam Failure During Helium Operation

    Science.gov (United States)

    Fremlin, Carl; Beckers, Jasper; Crowley, Brendan; Rauch, Joseph; Scoville, Jim

    2017-10-01

    The Neutral Beam system on the DIII-D tokamak consists of eight ion sources using the Common Long Pulse Source (CLPS) design. During helium operation, desired for research regarding the ITER pre-nuclear phase, it has been observed that the ion source arc chamber performance steadily deteriorates, eventually failing due to electrical breakdown of the insulation. A significant investment of manpower and time is required for repairs. To study the cause of failure a small analogue of the DIII-D neutral beam arc chamber has been constructed. This poster presents the design and analysis of the arc chamber including the PLC based operational control system for the experiment, analysis of the magnetic confinement and details of the diagnostic suite. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

  11. Enhanced performance discharges in the DIII-D tokamak with lithium wall conditioning

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, G.L. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [General Atomics, San Diego, CA (United States)]|[Oak Ridge National Laboratory, Oak Ridge, TN (United States); Navratil, G.A. [General Atomics, San Diego, CA (United States)]|[Columbia University, New York, NY (United States); Bastasz, R. [General Atomics, San Diego, CA (United States)]|[Sandia National Laboratories, Livermore, CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States); Garnier, D.T. [General Atomics, San Diego, CA (United States)]|[Massachusetts Institute of Technology, Cambridge, MA (United States); Holtrop, K.L. [General Atomics, San Diego, CA (United States); Phillips, J.C. [General Atomics, San Diego, CA (United States); Marmar, E.S. [General Atomics, San Diego, CA (United States)]|[Massachusetts Institute of Technology, Cambridge, MA (United States); Taylor, T.S. [General Atomics, San Diego, CA (United States); Thomas, D.M. [General Atomics, San Diego, CA (United States); Wampler, W.R. [General Atomics, San Diego, CA (United States)]|[Sandia National Laboratories, Albuquerque, NM (United States); Whyte, D.G. [General Atomics, San Diego, CA (United States)]|[INRS - Energie et Materiaux, Varennes, Que. (Canada); West, W.P. [General Atomics, San Diego, CA (United States)

    1997-02-01

    Lithium wall conditioning has been used in a recent campaign evaluating high performance negative central shear (NCS) discharges. During this campaign, the highest values of stored energy (4.4 MJ), neutron rate (2.4 x 10{sup 16}/s), and nT{sub i}{tau} (7 x 10{sup 20} m{sup -3} keV s) achieved to date in DIII-D were obtained. High performance NCS discharges were achieved prior to beginning lithium conditioning, but it is clear that shot reproducibility and performance were improved by lithium conditioning. Central and edge oxygen concentrations were reduced after lithium conditioning. Lithium conditioning, consisting of up to four pellets injected at the end of the preceding discharge, allowed the duration of the usual inter-shot helium glow discharge to be reduced and reproducible high auxiliary power discharges, P{sub NBI}{<=}22 MW, were obtained with plasma currents up to 2.4 MA. (orig.).

  12. Impurity enrichment and radiative enhancement using induced SOL flow in DIII-D

    International Nuclear Information System (INIS)

    Wade, M.R.; West, W.P.; Wood, R.D.

    1998-07-01

    Experiments on DIII-D have demonstrated the efficacy of using induced scrap-off-layer (SOL) flow to preferentially enrich impurities in the divertor plasma. This SOL floe 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 modest SOL flow is as high as 17. Using this induced SOL flow technique and argon injection, radiative plasmas have been produced that combine high radiation losses (P rad /P input > 70%), low core fuel dilution (Z eff E > 1.0 τ E,ITER93H )

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

    Science.gov (United States)

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

    2010-11-01

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

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

    International Nuclear Information System (INIS)

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

    1999-06-01

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

  15. PC application in DIII-D neutral beam operation

    International Nuclear Information System (INIS)

    Gladd, A.S.

    1986-01-01

    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 online shot-to-shot control and analysis. The PC hardware was configured to interface all four neutral beam host mini-computers, support multi-tasking, 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 FORTRAN and BASIC. 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 online control and analysis

  16. Magnetic Transport Barriers in the DIII-D Tokamak

    Science.gov (United States)

    Kessler, J.; Volpe, F.; Evans, T. E.; Ali, H.; Punjabi, A.

    2009-11-01

    Large overlapping magnetic islands generate chaotic fields. However, a previous work [1] showed that second or third order perturbations of special topology and strength can also generate magnetic diffusion ``barriers" in the middle of stochastic regions. In the present study, we numerically assess their experimental feasibility at DIII-D. For this, realistic I- and C-coils perturbations are superimposed on the equilibrium field and puncture plots are generated with a field-line tracer. A criterion is defined for the automatic recognition of barriers and successfully tested on earlier symplectic maps in magnetic coordinates. The criterion is systematically applied to the new puncture plots in search for dependencies, e.g. upon the edge safety factor q95, which might be relevant to edge localized mode (ELM) stability, as well as to assess the robustness of barriers against fluctuations of the plasma parameters and coil currents. 8pt [1] H. Ali and A. Punjabi, Plasma Phys. Control. Fusion 49, 1565 (2007).

  17. 2-D tomography with bolometry in DIII-D

    International Nuclear Information System (INIS)

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

    1994-07-01

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

  18. Motional stark effect upgrades on DIII-D

    International Nuclear Information System (INIS)

    Rice, B.W.; Nilson, D.G.; Wroblewski, D.

    1994-04-01

    The measurement and control of the plasma current density profile (or q profile) is critical to the advanced tokamak program on DIII-D. A complete understanding of the stability and transport properties of advanced operating regimes requires detail poloidal field measurements over the entire plasma radius from the core to the edge. In support of this effort, the authors have recently completed an upgrade of the existing MSE diagnostic, increasing the number of channels from 8 to 16. A new viewing geometry has been added to the outer edge of the plasma which improves the radial resolution in this region from 10 cm to < 4 cm. This view requires the use of a reflector that has been designed to minimize polarization amplitude and phase effects. Vacuum-compatible polarizers have also been added to the instrument for in-situ calibration. Future use of the MSE diagnostic for feedback control of the q profile will also be discussed

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

    International Nuclear Information System (INIS)

    Glad, A.S.

    1986-01-01

    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 fortran and basIc. 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

  20. Optimized Baking of the DIII-D Vessel

    International Nuclear Information System (INIS)

    Anderson, P.M.; Kellman, A.G.

    1999-01-01

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

  1. Physics analysis database for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Schissel, D.P.; Bramson, G.; DeBoo, J.C.

    1986-01-01

    The authors report on a centralized database for handling reduced data for physics analysis implemented for the DIII-D tokamak. Each database record corresponds to a specific snapshot in time for a selected discharge. Features of the database environment include automatic updating, data integrity checks, and data traceability. Reduced data from each diagnostic comprises a dedicated data bank (a subset of the database) with quality assurance provided by a physicist. These data banks will be used to create profile banks which will be input to a transport code to create a transport bank. Access to the database is initially through FORTRAN programs. One user interface, PLOTN, is a command driven program to select and display data subsets. Another user interface, PROF, compares and displays profiles. The database is implemented on a Digital Equipment Corporation VAX 8600 running VMS

  2. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  3. Recent DIII-D neutral beam calibration results

    International Nuclear Information System (INIS)

    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

  4. Task III: UCSD/DIII-D/Textor FY-97-98 Accomplishments

    International Nuclear Information System (INIS)

    Boedo, J.A.

    2000-01-01

    OAK (B204) Task III: UCSD/DIII-D/Textor FY-97-98 Accomplishments. A comprehensive report on the physics of pump limiters and particularly, the characterization of ALT-II, was published in Nuclear Fusion, bringing the project to a closure. The performance of the toroidal pump limiter was characterized under full auxiliary heating of 7 MW of NBI and ICRH and full pumping, as stated in the project milestones. Relevant highlights are: (1) Pumping with ALT-II allows for density control. (2) The achieved exhaust efficiency is 4% during NBI operation and near 2% during OH or ICRH operation. (3) We have shown that an exhaust efficiency of 2% is sufficient to satisfy the ash removal requirements of fusion reactors. (4) The plasma particle efflux and the pumped flux both increase with density and heating power. (5) The particle confinement time is less than the energy confinement time by a factor of 4. In summary, pumped belt limiters could provide the density control and ash exhaust requirements of fusion reactors

  5. Monte Carlo impurity transport modeling in the DIII-D transport

    International Nuclear Information System (INIS)

    Evans, T.E.; Finkenthal, D.F.

    1998-04-01

    A description of the carbon transport and sputtering physics contained in the Monte Carlo Impurity (MCI) transport code is given. Examples of statistically significant carbon transport pathways are examined using MCI's unique tracking visualizer and a mechanism for enhanced carbon accumulation on the high field side of the divertor chamber is discussed. Comparisons between carbon emissions calculated with MCI and those measured in the DIII-D tokamak are described. Good qualitative agreement is found between 2D carbon emission patterns calculated with MCI and experimentally measured carbon patterns. While uncertainties in the sputtering physics, atomic data, and transport models have made quantitative comparisons with experiments more difficult, recent results using a physics based model for physical and chemical sputtering has yielded simulations with about 50% of the total carbon radiation measured in the divertor. These results and plans for future improvement in the physics models and atomic data are discussed

  6. Plasma interactions with the outboard chamber wall in DIII-D

    International Nuclear Information System (INIS)

    Rudakov, D.L.; Boedo, J.A.; Yu, J.H.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Hollmann, E.M.; Lasnier, C.J.; McLean, A.G.; Moyer, R.A.; Stangeby, P.C.; Tynan, G.R.; Wampler, W.R.; Watkins, J.G.; West, W.P.; Wong, C.P.C.; Bastasz, R.J.; Buchenauer, D.; Whaley, J.

    2009-01-01

    Erosion of the main chamber plasma-facing components is of concern for ITER. Plasma interaction with the outboard chamber wall is studied in DIII-D using Langmuir probes and optical diagnostics. Fast camera data shows that edge localized modes (ELMs) feature helical filamentary structures propagating towards the outboard wall. Upon reaching the wall, filaments result in regions of local intense plasma-material interaction (PMI) where peak incident particle and heat fluxes are up to two orders of magnitude higher than those between ELMs. In low density/collisionality H-mode discharges, PMI at the outboard wall is almost entirely due to ELMs. A moderate change of the gap between the separatrix and the outer wall strongly affects PMI intensity at the wall. Material samples exposed near the outboard wall showed net carbon deposition in high-density discharges (near the Greenwald limit) and tendency towards net erosion in lower density discharges (∼0.45 of the Greenwald limit).

  7. Improved fueling and transport barrier formation with pellet injection from different locations on DIII-D

    International Nuclear Information System (INIS)

    Baylor, L.R.; Jernigan, T.C.; Gohil, P.

    2001-01-01

    Pellet injection has been employed on DIII-D from different injection locations to optimize the mass deposition for density profile control and internal transport barrier formation. Transport barriers have been formed deep in the plasma core with central mass deposition from high field side (HFS) injected pellets and in the edge with pellets that trigger L-mode to H-mode transitions. Pellets injected from all locations can trigger the H-mode transition, which depends on the edge density gradient created and not on the radial extent of the pellet deposition. Pellets injected from inside the magnetic axis from the inner wall or vertical port lead to stronger central mass deposition than pellets injected from the low field side (LFS) and thus yield deeper more efficient fueling. (author)

  8. Energy and particle transport in the radiative divertor plasmas of DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Allen, S.L.; Brooks, N.H.

    1997-06-01

    It has been argued that divertor energy transport dominated by parallel electron thermal conduction, or q parallel = -kT 5/2 2 dT e /ds parallel, leads to severe localization of the intense radiating region and ultimately limits the fraction of energy flux that can be radiated before striking the divertor target. This is due to the strong T 5/2 e dependence of electron heat conduction which results in very short spatial scales of the T e gradient at high power densities and low temperatures where deuterium and impurities radiate most effectively. However, we have greatly exceeded this constraint on DIII-D with deuterium gas puffing which reduces the peak heat flux to the divertor plate a factor of 5 while distributing the divertor radiation over a long length

  9. Implementation of reflectometry as a standard density profile diagnostic on DIII-D

    International Nuclear Information System (INIS)

    Zeng, L.; Doyle, E. J.; Luce, T. C.; Peebles, W. A.

    2001-01-01

    The profile reflectometer system on the DIII-D tokamak has been significantly upgraded in order to improve time coverage, data quality, and profile availability. The performance of the reflectometer system, which utilizes continuous frequency modulated (FMCW) radar techniques, has been improved as follows: First, a new PC-based data acquisition system has been installed, providing higher data sampling rates and larger memory depth. The higher sampling rate enables use of faster frequency sweeps of the FMCW microwave source, improving time resolution, and increasing profile accuracy. The larger memory depth enables longer data records, so that profiles can now be obtained throughout 5 s discharges at 100 Hz profile measurement rates, while continuous sampling at 10 MHz is available for 1 s for high time resolution physics studies. Second, an initial automated between-shots profile analysis capability is now available. Third, availability of the profiles to end users has been significantly improved

  10. SAFETY FACTOR SCALING OF ENERGY TRANSPORT IN L-MODE PLASMAS ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PETTY, C.C.; KINSEY, J.E.; LUCE, T.C.

    2003-01-01

    OAK-B135 The scaling of energy transport with safety factor (q) at fixed magnetic shear has been measured on the DIII-D tokamak [Nucl. Fusion 42, 614 (2002)] for low confinement (L) mode discharges. At constant density, temperature, and toroidal magnetic field strength, such that the toroidal dimensionless parameters other than q are held fixed, the one-fluid thermal diffusivity is found to scale like χ ∝ q 0.84±0.15 , with the ion channel having a stronger q dependence than the electron channel in the outer half of the plasma. The measured q scaling is in good agreement with the predicted scaling by the GLF23 transport model for the ion temperature gradient and trapped electron modes, but it is significantly weaker than the inferred scaling from empirically-derived confinement scaling relations

  11. Beta-induced Alfven-acoustic Eigenmodes in NSTX and DIII-D Driven by Beam Ions

    International Nuclear Information System (INIS)

    Gorelenkov, N.N.; Van Zeeland, M.A.; Berk, H.L.; Crocker, N.A.; Darrow, D.; Fredrickson, E.; Fu, G.-Y.; Heidbrink, W.W.; Menard, J.; Nazikian, R.

    2009-01-01

    Kinetic theory and experimental observations of a special class of energetic particle driven instabilities called here Beta-induced Alfven-Acoustic Eigenmodes (BAAE) are reported confirming previous results [N.N. Gorelenkov H.L. Berk, N.A. Crocker et. al. Plasma Phys. Control. Fusion 49 B371 (2007)] The kinetic theory is based on the ballooning dispersion relation where the drift frequency effects are retained. BAAE gaps are recovered in kinetic theory. It is shown that the observed certain low-frequency instabilities on DIII-D [J.L. Luxon, Nucl. Fusion 42 614 (2002)] and National Spherical Torus Experiment [M. Ono, S.M. Kaye, Y.-K M. Peng et. al., Nucl. Fusion 40 3Y 557 (2000)] are consistent with their identification as BAAEs. BAAEs deteriorated the fast ion confinement in DIII-D and can have a similar effect in next-step fusion plasmas, especially if excited together with multiple global Toroidicity-induced shear Alfven Eigenmode (TAE) instabilities. BAAEs can also be used to diagnose safety factor profiles, a technique known as magnetohydrodynamic spectroscopy

  12. Beta-induced Alfven-acousti Eigenmodes in NSTX and DIII-D Driven by Beam Ions

    Energy Technology Data Exchange (ETDEWEB)

    Gorelenkov, N. N.; Van Zeeland, M. A.; Berk, H. L.; Crocker, N. A.; Darrow, D.; Fredrickson, E.; Fu, G. Y.; Heidbrink, W. W.; Menard, J.; Nazikian, R.

    2009-03-06

    Kinetic theory and experimental observations of a special class of energetic particle driven instabilities called here Beta-induced Alfven-Acoustic Eigenmodes (BAAE) are reported confirming previous results [N.N. Gorelenkov H.L. Berk, N.A. Crocker et. al. Plasma Phys. Control. Fusion 49 B371 (2007)] The kinetic theory is based on the ballooning dispersion relation where the drift frequency effects are retained. BAAE gaps are recovered in kinetic theory. It is shown that the observed certain low-frequency instabilities on DIII-D [J.L. Luxon, Nucl. Fusion 42 614 (2002)] and National Spherical Torus Experiment [M. Ono, S.M. Kaye, Y.-K M. Peng et. al., Nucl. Fusion 40 3Y 557 (2000)] are consistent with their identification as BAAEs. BAAEs deteriorated the fast ion confinement in DIII-D and can have a similar effect in next-step fusion plasmas, especially if excited together with multiple global Toroidicity-induced shear Alfven Eigenmode (TAE) instabilities. BAAEs can also be used to diagnose safety factor profiles, a technique known as magnetohydrodynamic spectroscopy.

  13. An advanced plasma control system for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J.; Lazarus, E.

    1991-11-01

    An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Development of this system is expected to lead to control system technology appropriate for use on future tokamaks such as ITER and BPX. The digital system will allow for increased precision in shape control through real time adjustment of the control algorithm to changes in the shape and discharge parameters such as β p , ell i and scrape-off layer current. The system will be used for research on real time optimization of discharge performance for disruption avoidance, current and pressure profile control, optimization of rf antenna loading, or feedback on heat deposition patterns through divertor strike point position control, for example. Shape control with this system is based on linearization near a target shape of the controlled parameters as a function of the magnetic diagnostic signals. This digital system is unique in that it is designed to have the speed necessary to control the unstable vertical motion of highly elongated tokamak discharges such as those produced in DIII-D and planned for BPX and ITER. a 40 MHz Intel i860 processor is interfaced to up to 112 channels of analog input signals. The commands to the poloidal field coils can be updated at 80 μs intervals for the control of vertical position with a delay between sampling of the analog signal and update of the command of less than 80 μs

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

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  15. Thermal deposition analysis during disruptions on DIII-D using infrared scanners

    International Nuclear Information System (INIS)

    Lee, R.L.; Hyatt, A.W.; Kellman, A.G.; Taylor, P.L.; Lasnier, C.J.

    1995-12-01

    The DIII-D tokamak generates plasma discharges with currents up to 3 MA and auxiliary input power up to 20 MW from neutral beams and 4 MW from radio frequency systems. In a disruption, a rapid loss of the plasma current and internal thermal energy occurs and the energy is deposited onto the torus graphite wall. Quantifying the spatial and temporal characteristics of the heat deposition is important for engineering and physics-related issues, particularly for designing future machines such as ITER. Using infrared scanners with a time resolution of 120 micros, measurements of the heat deposition onto the all-graphite walls of DIII-D during two types of disruptions have been made. Each scanner contains a single point detector sensitive to 8--12 microm radiation, allowing surface temperatures from 20 C to 2,000 C to be measured. A zinc selenide window that transmits in the infrared is used as the vacuum window. Views of the upper and lower divertor regions and the centerpost provide good coverage of the first wall for single and double null divertor discharges. During disruptions, the thermal energy is not deposited evenly onto the inner surface of the tokamak, but is deposited primarily in the divertor region when operating diverted discharges. Analysis of the heat deposition during a radiative collapse disruption of a 1.5 MA discharge revealed power densities of 300--350 MW/m 2 in the divertor region. During the thermal quench of the disruption, the energy deposited onto the divertor region was more than 70% of the stored thermal energy in the discharge prior to the disruption. The spatial distribution and temporal behavior of power deposition during high β disruptions will also be presented

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

    Energy Technology Data Exchange (ETDEWEB)

    Baker, D. [ed.

    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.

  17. Engineering, installation, testing, and initial operation of the DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

    Andersen, P.M.; Baxi, C.B.; Reis, E.E.; Schaffer, M.J.; Smith, J.P.

    1990-09-01

    The Advanced Divertor (AD) for General Atomics tokamak, DIII-D, was installed in the summer of 1990. The AD has enabled two classes of physics experiments to be run: divertor biasing and divertor baffling. Both are new experiments for DIII-D. The AD has two principal components: (1) a continuous ring electrode; and (2) a toroidally symmetric baffle. The tokamak can be run in bias baffle or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D control system. The paper covers design, analysis, fabrication, installation, instrumentation, testing, initial operation, and future plans for the Advanced Divertor from an engineering viewpoint. 2 refs., 5 figs

  18. A Comparison of Fueling with Deuterium Pellet Injection from Different Locations on the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Baylor, L.R.; Combs, S.K.; Gohil, P.; Houlberg, W.A.; Hsieh, C.; Jernigan, T.C.; Parks, P.B.

    1999-01-01

    Initial pellet injection experiments on DIII-D with high field side (HFS) injection have demonstrated that deeper pellet fuel deposition is possible even with HFS injected pellets that are significantly slower than pellets injected from the low field side (LFS) (outer midplane) location. A radial displacement of the pellet mass shortly after or during the ablation process is consistent with the observed mass deposition profiles measured shortly after injection. Vertical injection inside the magnetic axis shows some improvement in fueling efficiency over LFS injection and may provide an optimal injection location for fueling with high speed pellets

  19. Fabrication of a 1200 kg Ingot of V-4Cr-4Ti for the DIII-D Radiative Divertor Program

    International Nuclear Information System (INIS)

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

    1998-01-01

    Vanadium chromium titanium alloys are attractive materials for fusion reactors because of their high temperature capability and their potential for low neutron active and rapid activation decay. A V-4Cr-4Ti alloy has been selected in the U.S. as the current leading candidate vanadium alloy for future use in fusion reactor structural applications. General Atomics (GA), in conjunction with the Department of Energy's (DOE) DIII-D Program, is carrying out a plan for the utilization of this vanadium alloy in the DIII-D tokamak. The plan will culminate in the fabrication, installation, and operation of a V-4Ti alloy structure in the DIII-D Radiative Divertor (RD) upgrade. The deployment of vanadium alloy will provide a meaningful step in the development and technology acceptance of this advanced material for future fusion power devices. Under a GA contract and material specification, an industrial scale 1200 kg heat (ingot) of a V-4Cr-4Ti alloy has been produced and converted into product forms by Wah Chang of Albany, Oregon (WCA). To assure the proper control of minor and trace impurities which affect the mechanical and activation behavior of this vanadium alloy, selected lots of raw vanadium base metal were processed by aluminothermic reduction of high purity vanadium oxide, and were then electron beam melted into two high purity vanadium ingots. The ingots were then consolidated with high purity Cr and Ti, and double vacuum-arc melted to obtain a 1200 kg V-4Cr-4Ti alloy ingot. Several billets were extruded from the ingot, and were then fabricated into plate, sheet, and rod at WCA. Tubing was subsequently processed from plate material. The chemistry and fabrication procedures for the product forms were specified on the basis of experience and knowledge gained from DOE Fusion Materials Program studies on previous laboratory scale heats and a large scale ingot (500 kg)

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

    International Nuclear Information System (INIS)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  2. Neutron sawtooth behavior in the PLT, DIII-D, and TFTR tokamaks

    International Nuclear Information System (INIS)

    Lovberg, J.A.; Heidbrink, W.W.; Strachan, J.D.; Zaveryaev, V.S.

    1988-10-01

    The effect of the sawtooth instability on the 2.5 MeV neutron emission in the PLT, DIII-D, and TFTR tokamaks is studied. In thermonuclear plasmas, the instability typically results in a 20% reduction in emission. The time evolution of the thermonuclear neutron signal suggests that the sawtooth crash consists of four phases. First, the electron density profile flattens rapidly (in /approximately/30μsec on PLT) but, in some cases, there is little associated change in neutron emission, suggesting that most reacting ions remain confined in the sawtooth region but do not completely mix. After the electron sawtooth, the ions continue to mix, resulting in a /approximately/10% reduction in neutron emission in /approximately/0.5 msec. The emission then decays more slowly during the final two phases. Thermalization of reacting ions on a /approximately/3/tau//sub ii/ time scale accounts for only /approximately/20% of the slow drop. Most of the slow drop seems to be caused by loss of ion energy from the mixing region (an ion heat pulse). 36 refs., 15 figs., 1 tabs

  3. Far SOL transport and main wall plasma interaction in DIII-D

    International Nuclear Information System (INIS)

    Rudakov, D.L.; Boedo, J.A.; Moyer, R.A.; Doerner, R.P.; Hollmann, E.M.; Krasheninnikov, S.I.; Pigarov, A.Yu.; Stangeby, P.C.; McLean, A.G.; Watkins, J.G.; Wampler, W.R.; Whyte, D.G.; McKee, G.R.; Zeng, L.; Wang, G.; Brooks, N.H.; Evans, T.E.; Leonard, A.W.; Mahdavi, M.A.; West, W.P.; Wong, C.P.C.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J.

    2005-01-01

    Far scrape-off layer (SOL) and near-wall plasma parameters in DIII-D depend strongly on the discharge parameters and confinement regime. In L-mode discharges cross-field transport increases with the average discharge density and flattens far SOL profiles, thus increasing plasma-wall contact. In H-mode between edge localized modes (ELMs), plasma-wall contact is generally weaker than in L-mode. During ELMs plasma fluxes to the wall increase to, or above the L-mode levels. Depending on the discharge conditions ELMs are responsible for 30-90% of the ion flux to the outboard chamber wall. Cross-field fluxes in far SOL are dominated by large amplitude intermittent transport events that may propagate all the way to the outer wall and cause sputtering. A Divertor Material Evaluation System (DiMES) probe containing samples of several ITER-relevant materials including carbon, beryllium and tungsten was exposed to a series of upper single null (USN) discharges as a proxy to measure the first wall erosion. (author)

  4. Control of plasma poloidal shape and position in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

    1997-11-01

    Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    Fast beam-ion losses were studied in DIII-D in the presence of a scaled mockup of two Test Blanket Modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam-ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot which is predicted to be different among the various codes.

  6. Methane penetration in DIII-D ELMing H-mode plasmas

    International Nuclear Information System (INIS)

    West, W.P.; Lasnier, C.J.; Whyte, D.G.; Isler, R.C.; Evans, T.E.; Jackson, G.L.; Rudakov, D.; Wade, M.R.; Strachan, J.

    2003-01-01

    Carbon penetration into the core plasma during midplane and divertor methane puffing has been measured for DIII-D ELMing H-mode plasmas. The methane puffs are adjusted to a measurable signal, but global plasma parameters are only weakly affected (line average density, e > increases by E , drops by 6+ density profiles in the core measured as a function of time using charge exchange recombination spectroscopy. The methane penetration factor is defined as the difference in the core content with the puff on and puff off, divided by the carbon confinement time and the methane puffing rate. In ELMing H-mode discharges with ion ∇B drift direction into the X-point, increasing the line averaged density from 5 to 8x10 19 m -3 dropped the penetration factor from 6.6% to 4.6% for main chamber puffing. The penetration factor for divertor puffing was below the detection limit (<1%). Changing the ion ∇B drift to away from the X-point decreased the penetration factor by more than a factor of five for main chamber puffing

  7. Enhanced computational infrastructure for data analysis at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

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

    2000-01-01

    Recently a number of enhancements to the computer hardware infrastructure have been implemented at the DIII-D National Fusion Facility. Utilizing these improvements to the hardware infrastructure, software enhancements are focusing on streamlined analysis, automation, and graphical user interface (GUI) systems to enlarge the user base. The adoption of the load balancing software package LSF Suite by Platform Computing has dramatically increased the availability of CPU cycles and the efficiency of their use. Streamlined analysis has been aided by the adoption of the MDSplus system to provide a unified interface to analyzed DIII-D data. The majority of MDSplus data is made available in between pulses giving the researcher critical information before setting up the next pulse. Work on data viewing and analysis tools focuses on efficient GUI design with object-oriented programming (OOP) for maximum code flexibility. Work to enhance the computational infrastructure at DIII-D has included a significant effort to aid the remote collaborator since the DIII-D National Team consists of scientists from nine national laboratories, 19 foreign laboratories, 16 universities, and five industrial partnerships. As a result of this work, DIII-D data is available on a 24x7 basis from a set of viewing and analysis tools that can be run on either the collaborators' or DIII-D's computer systems. Additionally, a web based data and code documentation system has been created to aid the novice and expert user alike

  8. Effect of heating scheme on SOL width in DIII-D and EAST

    Directory of Open Access Journals (Sweden)

    L. Wang

    2017-08-01

    Full Text Available Joint DIII-D/EAST experiments in the radio-frequency (RF heated H-mode scheme with comparison to that of neutral beam (NB heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH H-mode plasmas.

  9. INVESTIGATION OF MAIN-CHAMBER AND DIVERTOR RECYCING IN DIII-D USING TANGENTIALLY VIEWING CID CAMERAS

    International Nuclear Information System (INIS)

    GROTH, M.; PORTER, G.D.; PETRIE, T.W.; FENSTERMACHER, M.E.; BROOKS, N.H.

    2003-01-01

    OAK-B135 Measurements of the D α emission profiles from the divertor and main chamber region in DIII-D, performed in low-density L-mode, and low and high-density ELMy H-mode plasmas imply that core plasma fueling occurs through the divertor channel. Emission profiles of carbon, combined with UEDGE modeling of the L-mode plasmas, also suggests that chemical sputtering of carbon from the flux surface adjacent to the inner divertor walls, and temperature gradient forces in the scrape-off layer, determine the carbon content of the inner scrape-off layer

  10. ICRF [ion cyclotron range of frequencies] coupling on DIII-D and the implications on ICRF technology development

    International Nuclear Information System (INIS)

    Hoffman, D.J.; Baity, F.W.; Mayberry, M.J.; Swain, D.W.

    1987-01-01

    Low-power coupling tests have been carried out with a prototype ion cyclotron range of frequencies (ICRF) compact loop antenna on the DIII-D tokamak. Plasma load resistance values higher than originally calculated are measured in ohmic and L-mode, beam-heated plasmas. Load resistance decreases by a factor of ∼2 in H-mode operation. When edge localized modes (ELMs) occur, the antenna loading increases transiently to several ohms. Results indicate that fast-wave ICRF antenna coupling characteristics are highly sensitive to changes in the edge plasma profiles associated with the H-mode regime

  11. Development of a tokamak plasma optimized for stability and confinement

    International Nuclear Information System (INIS)

    Politzer, P.A.

    1995-02-01

    Design of an economically attractive tokamak fusion reactor depends on producing steady-state plasma operation with simultaneous high energy density (β) and high energy confinement (τ E ); either of these, by itself, is insufficient. In operation of the DIII-D tokamak, both high confinement enhancement (H≡ τ E /τ ITER-89P = 4) and high normalized β (β N ≡ β/(I/aB) = 6%-m-T/MA) have been obtained. For the present, these conditions have been produced separately and in transient discharges. The DIII-D advanced tokamak development program is directed toward developing an understanding of the characteristics which lead to high stability and confinement, and to use that understanding to demonstrate stationary, high performance operation through active control of the plasma shape and profiles. The authors have identified some of the features of the operating modes in DIII-D that contribute to better performance. These are control of the plasma shape, control of both bulk plasma rotation and shear in the rotation and Er profiles, and particularly control of the toroidal current profiles. In order to guide their future experiments, they are developing optimized scenarios based on their anticipated plasma control capabilities, particularly using fast wave current drive (on-axis) and electron cyclotron current drive (off-axis). The most highly developed model is the second-stable core VH-mode, which has a reversed magnetic shear safety factor profile [q(O) = 3.9, q min = 2.6, and q 95 = 6]. This model plasma uses profiles which the authors expect to be realizable. At β N ≥ 6, it is stable to n=l kink modes and ideal ballooning modes, and is expected to reach H ≥ 3 with VH-mode-like confinement

  12. The effect of electron cyclotron heating on density fluctuations at ion and electron scales in ITER baseline scenario discharges on the DIII-D tokamak

    Science.gov (United States)

    Marinoni, A.; Pinsker, R. I.; Porkolab, M.; Rost, J. C.; Davis, E. M.; Burrell, K. H.; Candy, J.; Staebler, G. M.; Grierson, B. A.; McKee, G. R.; Rhodes, T. L.; The DIII-D Team

    2017-12-01

    Experiments simulating the ITER baseline scenario on the DIII-D tokamak show that torque-free pure electron heating, when coupled to plasmas subject to a net co-current beam torque, affects density fluctuations at electron scales on a sub-confinement time scale, whereas fluctuations at ion scales change only after profiles have evolved to a new stationary state. Modifications to the density fluctuations measured by the phase contrast imaging diagnostic (PCI) are assessed by analyzing the time evolution following the switch-off of electron cyclotron heating (ECH), thus going from mixed beam/ECH to pure neutral beam heating at fixed βN . Within 20 ms after turning off ECH, the intensity of fluctuations is observed to increase at frequencies higher than 200 kHz in contrast, fluctuations at lower frequency are seen to decrease in intensity on a longer time scale, after other equilibrium quantities have evolved. Non-linear gyro-kinetic modeling at ion and electron scales scales suggest that, while the low frequency response of the diagnostic is consistent with the dominant ITG modes being weakened by the slow-time increase in flow shear, the high frequency response is due to prompt changes to the electron temperature profile that enhance electron modes and generate a larger heat flux and an inward particle pinch. These results suggest that electron heated regimes in ITER will feature multi-scale fluctuations that might affect fusion performance via modifications to profiles.

  13. An interior vessel viewing system for DIII-D

    International Nuclear Information System (INIS)

    Senior, R.

    1989-11-01

    It was anticipated that there could be damage to the interior walls of the vacuum vessel during operations of the DIII-D tokamak. A method of viewing the inside of the vessel from the outside was required, that would allow the interior walls to be inspected visually for damage and to locate any debris resulting from operations. A miniature closed circuit television color camera system was developed which could be inserted into one of several ports of the vessel during a 'clean' vent, i.e., vented to inert gas. The system has pan, tilt and zoom capability and carries its own lighting. The use of this system allows a quick assessment of the condition of the vessel to be made under 'clean' vent conditions. This precludes the need for the permit process and manned entry into the vessel which would allow air inside the vessel. A permanent record of the inspection can then be made on video tape. The design and configuration of this camera system is presented and its use as a diagnostic tool discussed. 2 refs., 5 figs

  14. The DIII-D Tokamak trouble report database

    International Nuclear Information System (INIS)

    Petersen, P.I.; Miller, S.M.

    1992-01-01

    Operation of the DIII-D tokamak at General Atomics involves many groups which work on the various subsystems. To overview and speed the solution to trouble or problem areas that limit machine availability, a common trouble report system was established. The TROUBLE database automates the recording of trouble reports and eases analysis of problem areas. It contains information on equipment affected, description of problem, cause of problem, solution to problem, and machine downtime (if any). It was created using S1032 from Compuserve Data Technologies and runs on a VAX 8650. The data is used to find the major problem areas so they can be solved and improve the tokamak availabilty. The data is available to Idaho National Engineering Laboratory (INEL). They are using the data with data from other tokamaks to develop a Fusion Failure Experience Data Collection. The authors' experience is that a few failures are often the cause of a major part of the downtime. In this paper, the authors will discuss these failures and the actions taken to correct them. The data base also will be used to determine the preventive maintenance schedule for different components

  15. The DIII-D 3 MW, 110 GHz ECH System

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  16. Computerized operation of the DIII-D neutral beams

    International Nuclear Information System (INIS)

    Glad, A.S.; Tooker, J.F.

    1986-01-01

    Operation of the DIII-D neutral beams utilizes computerized control to provide routine tokamak beam heating shots and an effective method for automatic ion source operation. Computerized control reduces operational complexity, thus providing consistent reliability and availability of beams and a significant reduction in the the costs of routine operation. The objectives in implementing computerized control for operation were: (1) to improve operator efficiency for controlling multiple beam lines and increasing beam availability through standard procedures, (2) to provide a simplified scheme that operators and coordinators can construct and maintain, and (3) to provide a single integrated mechanism for both tokamak operation and automatic source conditioning. The years of experience in operating neutral beams at Doublet III provided the data necessary to meet the objectives. The method for computerized control consisted of three integrated functions: (1) a structured command language was implemented to provide the mechanism for automatically sequencing beams, (2) a historical file was constructed from the operational parameters to characterize the ion source, and consists of data from approximately 100,000 beam shots, and (3) procedures were developed integrating the language to the historical file for normal operation and source conditioning. This paper describes the method for sequencing beams automatically, the structure of the historical data file, and the procedures which integrate the historical data with tokamak operation and automatic source conditioning

  17. Fast wave current drive in DIII-D

    International Nuclear Information System (INIS)

    Petty, C.C.; Callis, R.W.; Chiu, S.C.; deGrassie, J.S.; Forest, C.B.; Freeman, R.L.; Gohil, P.; Harvey, R.W.; Ikezi, H.; Lin-Liu, Y.-R.

    1995-02-01

    The non-inductive current drive from fast Alfven waves launched by a directional four-element antenna was measured in the DIII-D tokamak. The fast wave frequency (60 MHz) was eight times the deuterium cyclotron frequency at the plasma center. An array of rf pickup loops at several locations around the torus was used to verify the directivity of the four-element antenna. Complete non-inductive current drive was achieved using a combination of fast wave current drive (FWCD) and electron cyclotron current drive (ECCD) in discharges for which the total plasma current was inductively ramped down from 400 to 170 kA. For discharges with steady plasma current, up to 110 kA of FWCD was inferred from an analysis of the loop voltage, with a maximum non-inductive current (FWCD, ECCD, and bootstrap) of 195 out of 310 kA. The FWCD efficiency increased linearly with central electron temperature. For low current discharges, the FWCD efficiency was degraded due to incomplete fast wave damping. The experimental FWCD was found to agree with predictions from the CURRAY ray-tracing code only when a parasitic loss of 4% per pass was included in the modeling along with multiple pass damping

  18. Gamma ray imager on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.; Van Zeeland, M. A.; Watkins, M. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Cooper, C. M. [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California-San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Riso, V. [State University of New York-Buffalo, 12 Capen Hall, Buffalo, New York 14260-1660 (United States)

    2016-04-15

    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.

  19. An overview of the DIII-D program

    International Nuclear Information System (INIS)

    Luxon, J.L.

    1996-10-01

    The DIII-D program focuses on developing fusion physics in an integrated program of tokamak concept improvement. The intent is both to support the present ITER physics R and D and to develop more efficient concepts for the later phases of ITER and eventual power plants. Progress in this effort can be best summarized by recent results for a diverted deuterium discharge with negative central shear which reached a performance level of Q DT = 0.32. The ongoing development of the tools needed to carry out this program of understanding and optimization continues to be crucial to its success. Control of the plasma cross-sectional shape and the internal distributions of plasma current, density, and rotation has been essential to optimizing plasma performance. Advanced divertor concepts provide edge power and particle control for future devices such as ITER and provide techniques to help manage the edge power and particle flows for advanced tokamak concepts. New divertor diagnostics and improved modeling are developing excellent divertor understanding. Many of the plasma physics issues being posed by ITER are being addressed. Scrapeoff layer power flow is being characterized to provide an accurate basis for the design of reactor devices. Ongoing studies of the density limit focus on identifying ways in which ITER can achieve the required densities in excess of the Greenwald limit. Better understanding of disruptions is crucial to the design of future reactors

  20. State transitions, hysteresis, and control parameters on DIII-D

    International Nuclear Information System (INIS)

    Thomas, D.M.; Groebner, R.J.; Carlstrom, T.N.; Osborne, T.H.; Petrie, T.W.

    1998-07-01

    The theory of turbulence decorrelation by ExB velocity shear is the leading candidate to explain the changes in turbulence and transport that are seen at the plasma edge at the L to H transition. Based on this, a key question is: What are the conditions or control parameters needed to begin the formation of the E r shear layer and thus trigger the L to H transition? On the DIII-D tokamak, the authors are attacking this question both through direct tests of the various theories and by trying to gain insight into the fundamental physics by investigating the control parameters which have a major effect on the power threshold. In this paper the authors describe results of studies on oscillating discharges where the plasma transitions continuously between L and H states. By following the dynamics of the plasma state through the forward and back transitions, they can represent the evolution of various control parameter candidates as a trajectory in various parametric spaces. The shape of these control curves can illustrate the specific nonlinearities governing the L-H transition problem, and under the proper conditions may be interpreted in the context of various phase-transition based models. In particular, the hysteresis exhibited in the various curves may help to clarify causality (what are the critical parameters) and may serve as tests of the models, given sufficient experimental accuracy. At present they are looking at T e , E r and ballooning/diamagnetic parameters as possible control parameter candidates

  1. Visible spectroscopy in the DIII-D divertor

    International Nuclear Information System (INIS)

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

    1996-06-01

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

  2. Multivariable shape control development on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

    1997-11-01

    In this paper, the authors describe recent work on plasma shape and position control at DIII-D. This control consists of two equally challenging problems--the problem of identifying what the plasma actually looks like in real time, i.e. measuring the parameters to be controlled, and the task of determining the feedback algorithm which best controls these plasma parameters in a multiple input-output system. Recent implementation of the EFIT plasma equilibrium reconstruction algorithm in real time code which produces a new equilibrium estimate every 1.5 ms seems to solve the longstanding problem of obtaining sufficiently accurate plasma shape and position estimation. Stabilization of the open-loop unstable vertical motion is also viewed as a solved problem. The primary remaining problem appears to be how best to command the power supplies to achieve a desired shaping control response. They will describe the effort to understand and apply linearized models of plasma evolution to development and implementation of multivariable plasma controllers

  3. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Porter, G.D.; Wood, R.D.; Allen, S.L.; Boedo, J.; Brooks, N.H.; Evans, T.E.; Fenstermacher, M.E.; Hill, D.N.; Isler, R.C.; Lasnier, C.J.; Lehmer, R.D.; Mahdavi, M.A.; Maingi, R.; Moyer, R.A.; Petrie, T.W.; Schaffer, M.J.; Wade, M.R.; Watkins, J.G.; West, W.P.; Whyte, D.G.

    1998-01-01

    The radiation of divertor heat flux on DIII-D [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low-Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction-dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE [T. Rognlien, J. L. Milovich, M. E. Rensink, and G. D. Porter, J. Nucl. Mater. 196 endash 198, 347 (1992)] has reproduced many of the observed experimental features. copyright 1998 American Institute of Physics

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

    International Nuclear Information System (INIS)

    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

  5. Simulation of density fluctuations before the L-H transition for Hydrogen and Deuterium plasmas in the DIII-D tokamak using the BOUT++ code

    Science.gov (United States)

    Wang, Y. M.; Xu, X. Q.; Yan, Z.; Mckee, G. R.; Grierson, B. A.; Xia, T. Y.; Gao, X.

    2018-02-01

    A six-field two-fluid model has been used to simulate density fluctuations. The equilibrium is generated by experimental measurements for both Deuterium (D) and Hydrogen (H) plasmas at the lowest densities of DIII-D low to high confinement (L-H) transition experiments. In linear simulations, the unstable modes are found to be resistive ballooning modes with the most unstable mode number n  =  30 or k_θρ_i˜0.12 . The ion diamagnetic drift and E× B convection flow are balanced when the radial electric field (E r ) is calculated from the pressure profile without net flow. The curvature drift plays an important role in this stage. Two poloidally counter propagating modes are found in the nonlinear simulation of the D plasma at electron density n_e˜1.5×1019 m-3 near the separatrix while a single ion mode is found in the H plasma at the similar lower density, which are consistent with the experimental results measured by the beam emission spectroscopy (BES) diagnostic on the DIII-D tokamak. The frequency of the electron modes and the ion modes are about 40 kHz and 10 kHz respectively. The poloidal wave number k_θ is about 0.2 cm -1 (k_θρ_i˜0.05 ) for both ion and electron modes. The particle flux, ion and electron heat fluxes are  ˜3.5-6 times larger for the H plasma than the D plasma, which makes it harder to achieve H-mode for the same heating power. The change of the atomic mass number A from 2 to 1 using D plasma equilibrium make little difference on the flux. Increase the electric field will suppress the density fluctuation. The electric field scan and ion mass scan results show that the dual-mode results primarily from differences in the profiles rather than the ion mass.

  6. COMPARISON OF ELM PULSE PROPAGATION IN THE DIII-D SOL AND DIVERTORS WITH AN ION CONVECTION MODEL

    International Nuclear Information System (INIS)

    FENSTERMACHER, ME; PORTER, GD; LEONARD, AW; BROOKS, NH; BOEDO, JA; COLCHIN, RJ; GRAY, DS; GROEBNER, RJ; GROTH, M; HOGAN, JT; HOLLMANN, EM; LASNIER, CJ; OSBORNE, TH; PETRIE, TW; RUDAKOV, DL; SNYDER, PB; TAKAHASHI, H; WATKINS, JG; ZENG, L; DIII-D TEAM

    2003-01-01

    OAK-B135 Results from dedicated ELM experiments, performed in DIII-D with fast diagnostics to measure the evolution of Type-I ELM effects in the SOL and divertor, are compared with a simple ion convection model and with initial time-dependent UEDGE simulations. Delays between ELM effects observed in the inner versus the outer divertor regions in the experiments scale, as a function of density, with the difference in ion convection time along field lines from the outer midplane to the divertor targets. The ELM perturbation was modeled as an instantaneous radially uniform increase of diffusion coefficients from the top of the pedestal to the outer SOL. The perturbation was confined to a low field side poloidal zone ± 40 o from the outer midplane. The delays in the simulations are similar to those observed in the experiments

  7. Scaling of ELM and H-mode pedestal characteristics in ITER shape discharges in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Osborne, T.H.; Groebner, R.J.; Lao, L.L.; Leonard, A.W.; Miller, R.L.; Thomas, D.M.; Waltz, R.E.; Maingi, R.; Porter, G.D.

    1997-07-01

    The authors have shown a correlation between the H-mode pressure pedestal height and the energy confinement enhancement in ITER shape discharges on DIII-D which is consistent with the behavior of H in different ELM classes. The width of the steep gradient region was found to equally well fit the scalings δ/R ∝ (ρ POL /R) 2/3 and δ/R ∝ (β POL PED /R) 1/2 . The normalized pressure gradient α MHD was found to be relatively constant just before a type I ELM. An estimate of T PED for ITER gave 1 to 5 keV. They also estimate ΔE ELM ≅ 26 MJ for ITER. They identified a distinct class of type III ELM at low density which may play a role in setting H at powers near the H-mode threshold power

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  9. Helium Exhaust Studies in H-Mode Discharges in the DIII-D Tokamak Using an Argon-Frosted Divertor Cryopump

    International Nuclear Information System (INIS)

    Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Mahdavi, M.A.; Maingi, R.; West, W.P.; Brooks, N.H.; Burrell, K.H.; Groebner, R.J.; Jackson, G.L.; Klepper, C.C.; Laughon, G.; Menon, M.M.; Mioduszewski, P.K.

    1995-01-01

    The first experiments demonstrating exhaust of thermal helium in a diverted, H-mode deuterium plasma have been performed on the DIII-D tokamak. The helium, introduced via gas puffing, is observed to reach the plasma core, and then is readily removed from the plasma with a time constant of ∼10--20 energy-confinement times by an in-vessel cryopump conditioned with argon frosting. Detailed analysis of the helium profile evolution suggests that the exhaust rate is limited by the exhaust efficiency of the pump (∼5%) and not by the intrinsic helium-transport properties of the plasma

  10. Measurement of deuterium density profiles in the H-mode steep gradient region using charge exchange recombination spectroscopy on DIII-D.

    Science.gov (United States)

    Haskey, S R; Grierson, B A; Burrell, K H; Chrystal, C; Groebner, R J; Kaplan, D H; Pablant, N A; Stagner, L

    2016-11-01

    Recent completion of a thirty two channel main-ion (deuterium) charge exchange recombination spectroscopy (CER) diagnostic on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] enables detailed comparisons between impurity and main-ion temperature, density, and toroidal rotation. In a H-mode DIII-D discharge, these new measurement capabilities are used to provide the deuterium density profile, demonstrate the importance of profile alignment between Thomson scattering and CER diagnostics, and aid in determining the electron temperature at the separatrix. Sixteen sightlines cover the core of the plasma and another sixteen are densely packed towards the plasma edge, providing high resolution measurements across the pedestal and steep gradient region in H-mode plasmas. Extracting useful physical quantities such as deuterium density is challenging due to multiple photoemission processes. These challenges are overcome using a detailed fitting model and by forward modeling the photoemission using the FIDASIM code, which implements a comprehensive collisional radiative model.

  11. Development of improved methods for remote access of DIII-D data and data analysis

    International Nuclear Information System (INIS)

    Greene, K.L.; McHarg, B.B. Jr.

    1997-11-01

    The DIII-D tokamak is a national fusion research facility. There is an increasing need to access data from remote sites in order to facilitate data analysis by collaborative researchers at remote locations, both nationally and internationally. In the past, this has usually been done by remotely logging into computers at the DIII-D site. With the advent of faster networking and powerful computers at remote sites, it is becoming possible to access and analyze data from anywhere in the world as if the remote user were actually at the DIII-D site. The general mechanism for accessing DIII-D data has always been via the PTDATA subroutine. Substantial enhancements are being made to that routine to make it more useful in a non-local environment. In particular, a caching mechanism is being built into PTDATA to make network data access more efficient. Studies are also being made of using Distributed File System (DFS) disk storage in a Distributed Computing Environment (DCE). A data server has been created that will migrate, on request, shot data from the DIII-D environment into the DFS environment

  12. Magnetic barriers and their q95 dependence at DIII-D

    Science.gov (United States)

    Volpe, F. A.; Kessler, J.; Ali, H.; Evans, T. E.; Punjabi, A.

    2012-05-01

    It is well known that externally generated resonant magnetic perturbations (RMPs) can form islands in the plasma edge. In turn, large overlapping islands generate stochastic fields, which are believed to play a role in the avoidance and suppression of edge localized modes (ELMs) at DIII-D. However, large coalescing islands can also generate, in the middle of these stochastic regions, KAM surfaces effectively acting as ‘barriers’ against field-line dispersion and, indirectly, particle diffusion. It was predicted in Ali and Punjabi (2007 Plasma Phys. Control. Fusion 49 1565-82) that such magnetic barriers can form in piecewise analytic DIII-D plasma equilibria. In this work, the formation of magnetic barriers at DIII-D is corroborated by field-line tracing calculations using experimentally constrained EFIT (Lao et al 1985 Nucl. Fusion 25 1611) DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. According to these calculations, the occurrence and location of magnetic barriers depend on the edge safety factor q95. It was thus suggested that magnetic barriers might contribute to narrowing the edge stochastic layer and play an indirect role in the RMPs failing to control ELMs for certain values of q95. The analysis of DIII-D discharges where q95 was varied, however, does not show anti-correlation between barrier formation and ELM suppression.

  13. Magnetic barriers and their q95 dependence at DIII-D

    International Nuclear Information System (INIS)

    Volpe, F.A.; Kessler, J.; Ali, H.; Punjabi, A.; Evans, T.E.

    2012-01-01

    It is well known that externally generated resonant magnetic perturbations (RMPs) can form islands in the plasma edge. In turn, large overlapping islands generate stochastic fields, which are believed to play a role in the avoidance and suppression of edge localized modes (ELMs) at DIII-D. However, large coalescing islands can also generate, in the middle of these stochastic regions, KAM surfaces effectively acting as ‘barriers’ against field-line dispersion and, indirectly, particle diffusion. It was predicted in Ali and Punjabi (2007 Plasma Phys. Control. Fusion 49 1565–82) that such magnetic barriers can form in piecewise analytic DIII-D plasma equilibria. In this work, the formation of magnetic barriers at DIII-D is corroborated by field-line tracing calculations using experimentally constrained EFIT (Lao et al 1985 Nucl. Fusion 25 1611) DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. According to these calculations, the occurrence and location of magnetic barriers depend on the edge safety factor q 95 . It was thus suggested that magnetic barriers might contribute to narrowing the edge stochastic layer and play an indirect role in the RMPs failing to control ELMs for certain values of q 95 . The analysis of DIII-D discharges where q 95 was varied, however, does not show anti-correlation between barrier formation and ELM suppression. (paper)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  15. ECH system developments including the design of an intelligent fault processor on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Ponce, D.; Lohr, J.; Tooker, J.F.; O'Neill, R.C.; Moeller, C.P.; Doane, J.L.; Noraky, S.; Dubovenko, K.; Gorelov, Y.A.; Cengher, M.; Penaflor, B.G.; Ellis, R.A.

    2011-01-01

    A new generation fault processor is in development which is intended to increase fault handling flexibility and reduce the number of incomplete DIII-D shots due to gyrotron faults. The processor, which is based upon a field programmable gate array device, will analyze signals for aberrant operation and ramp down high voltage to try to avoid hard faults. The processor will then attempt to ramp back up to an attainable operating point. The new generation fault processor will be developed during an expansion of the electron cyclotron heating (ECH) areas that will include the installation of a depressed collector gyrotron and associated equipment. Existing systems will also be upgraded. Testing of real-time control of the ECH launcher poloidal drives by the DIII-D plasma control system will be completed. The ECH control system software will be upgraded for increased scalability and to increase operator productivity. Resources permitting, all systems will receive an extra layer of interlocks for the filament and magnet power supplies, added shielding for the tank electronics, programmable filament boost shape for long pulses, and electronics upgrades for the installation of the advanced fault processor.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  17. Design and implementation of a user-friendly interface for DIII-D neutral beam automated operation

    International Nuclear Information System (INIS)

    Phillips, J.; Colleraine, A.P.; Hong, R.; Kim, J.; Lee, R.L.; Wight, J.J.

    1989-12-01

    The operational interface to the DIII-D neutral beam system, in use for the past 10 years, consisted of several interactive devices that the operator used to sequence neutral beam conditioning and plasma heating shots. Each of four independent MODCOMP Classic control computers (for four DIII-D beamlines) included a touch screen, rotary knobs, an interactive dual port terminal, and a keyboard to selectively address each of five display screens. Most of the hardware had become obsolete and repair was becoming increasingly expensive. It was clear that the hardware could be replaced with current equipment, while improving the ergonomics of control. Combined with an ongoing effort to increase the degree of automated operation and its reliability, a single microcomputer-based interface for each of the four neutral beam MODCOMP Classic control computers was developed, effectively replacing some twenty pieces of hardware. Macintosh II microcomputers were selected, with 1 megabyte of RAM and ''off-the-shelf'' input/output (I/O) consisting of a mouse, serial ports, and two monochrome high-resolution video monitors. The software is written in PASCAL and adopts standard Macintosh ''window'' techniques. From the Macintosh interface to the MODCOMP Classic, the operator can control the power supply setpoints, adjust ion source timing and synchronization, call up waveform displays on the Grinnell color display system, view the sequencing of procedures to ready a neutral beam shot, and add operator comments to an automated shot logging system. 3 refs., 2 figs

  18. OEDGE modeling for the planned tungsten ring experiment on DIII-D

    Directory of Open Access Journals (Sweden)

    J.D. Elder

    2017-08-01

    Full Text Available The OEDGE code is used to model tungsten erosion and transport for experiments with toroidal rings of high-Z metal tiles in the DIII-D tokamak. Such modeling is needed for both experimental and diagnostic design to have estimates of the expected core and edge tungsten density and to understand the various factors contributing to the uncertainties in these calculations. OEDGE simulations are performed using the planned experimental magnetic geometries and plasma conditions typical of both L-mode and inter-ELM H-mode discharges in DIII-D. OEDGE plasma reconstruction based on specific representative discharges for similar geometries is used to determine the plasma conditions applied to tungsten plasma impurity simulations. A new model for tungsten erosion in OEDGE was developed which imports charge-state resolved carbon impurity fluxes and impact energies from a separate OEDGE run which models the carbon production, transport and deposition for the same plasma conditions as the tungsten simulations. These values are then used to calculate the gross tungsten physical sputtering due to carbon plasma impurities which is then added to any sputtering by deuterium ions; tungsten self-sputtering is also included. The code results are found to be dependent on the following factors: divertor geometry and closure, the choice of cross-field anomalous transport coefficients, divertor plasma conditions (affecting both tungsten source strength and transport, the choice of tungsten atomic physics data used in the model (in particular ionization rate for W-atoms, and the model of the carbon flux and energy used for calculating the tungsten source due to sputtering. Core tungsten density is found to be of order 1015m−3 (excluding effects of any core transport barrier and with significant variability depending on the other factors mentioned with density decaying into the scrape off layer. For the typical core density in the plasma conditions examined of 2 to 4

  19. Simultaneous Feedback Control of Plasma Rotation and Stored Energy on the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Scoville, J.T.; Ferron, J.R.; Humphreys, D.A.; Walker, M.L.

    2006-01-01

    One of the major modifications made to the DIII-D tokamak during the 2005 Long Torus Opening was the rotation of one of the four two-source neutral beam injection systems. Prior to this modification, all beams injected power with a component in the same direction as the usual plasma current ('' co-injection ''). Starting in early 2006, two of the seven beams inject with a component in the opposite direction ('' counter-injection ''). This new capability allows, for the first time, a partial decoupling of the injected energy and momentum during neutral beam heating experiments. An immediate advantage of mixed co- and counter-injection beams is the capability to control the plasma rotation velocity. High beta plasmas can now be studied over a wide range of the plasma rotation velocity. The stabilizing effect of rotation on the resistive wall mode (RWM), for example, can be directly compared to the stabilization achieved by external feedback coils. This is an advantage over previous techniques to control plasma rotation, such as magnetic braking, which have had only limited success. We describe development and implementation of a model-based control algorithm for simultaneous regulation of plasma rotation and beta. The model includes the two relevant plasma states (plasma rotation and stored energy), and describes the dynamic effects of the relevant actuators on those states. The actuators include the applied beam torque and beam power, which depend on the amount of co and counter-injected beams. Implementation of the model-based control within the plasma control system (PCS) [B.G. Penaflor, et al, '' Current Status of DIII-D Plasma Control System Computer Upgrades,'' Fusion Eng. and Design 71 (2004) 47] requires real-time measurements of the plasma rotation, obtained from the charge exchange recombination (CER) diagnostic, and stored energy calculated by the real-time EFIT equilibrium reconstruction. Details of this model and its development, and a comparison with

  20. DIII-D research operations annual report to the U.S. Department of Energy, October 1, 1996 through September 30, 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-10-01

    The main goals of the DIII-D experiments in 1997 were, by extending and integrating the understanding of fusion science, to make progress in the tokamak concept improvements as delineated in the DIII-D Long Range Plan and to make substantial contributions to urgently needed R and D for the ITER Engineering Design Activity. For these purposes, the authors modified the top divertor to include pumping with baffling of high triangularity shaped plasmas and brought into operation two megawatt-level-gyrotrons for electron cyclotron heating (ECH) and off-axis current drive. The elements of the DIII-D experimental program and its objectives are organized into five topical areas: Stability and Disruption Physics, Transport and Turbulence Physics, Divertor and Boundary Physics, Wave-Particle Physics, and Integrated Fusion Science and Innovative Concept Improvement. The resulting DIII-D fusion science accomplishments are described in detail in this report. This year was characterized by a number of important activities, most notably, two 110 GHz ECH gyrotrons were installed and commissioned, the upper RDP cryopump and baffle was installed, and the ohmic heating coil lead was successfully reinforced to allow return to the design coil configuration and an increase to 7.5 V-s next year. Real-time ``Isoflux`` plasma control was implemented to control the shape and position of the plasma. This system solves the MHD equilibrium equation in real time to accurately determine the location of the plasma boundary. At the same time, the authors were able to improve their safety record with three minor accidents and no lost time accidents. The staff available for operations tasks was substantially reduced owing to recent budget reductions and this impacted a number of activities.

  1. DIII-D research operations annual report to the U.S. Department of Energy, October 1, 1996 through September 30, 1997

    International Nuclear Information System (INIS)

    1998-10-01

    The main goals of the DIII-D experiments in 1997 were, by extending and integrating the understanding of fusion science, to make progress in the tokamak concept improvements as delineated in the DIII-D Long Range Plan and to make substantial contributions to urgently needed R and D for the ITER Engineering Design Activity. For these purposes, the authors modified the top divertor to include pumping with baffling of high triangularity shaped plasmas and brought into operation two megawatt-level-gyrotrons for electron cyclotron heating (ECH) and off-axis current drive. The elements of the DIII-D experimental program and its objectives are organized into five topical areas: Stability and Disruption Physics, Transport and Turbulence Physics, Divertor and Boundary Physics, Wave-Particle Physics, and Integrated Fusion Science and Innovative Concept Improvement. The resulting DIII-D fusion science accomplishments are described in detail in this report. This year was characterized by a number of important activities, most notably, two 110 GHz ECH gyrotrons were installed and commissioned, the upper RDP cryopump and baffle was installed, and the ohmic heating coil lead was successfully reinforced to allow return to the design coil configuration and an increase to 7.5 V-s next year. Real-time ''Isoflux'' plasma control was implemented to control the shape and position of the plasma. This system solves the MHD equilibrium equation in real time to accurately determine the location of the plasma boundary. At the same time, the authors were able to improve their safety record with three minor accidents and no lost time accidents. The staff available for operations tasks was substantially reduced owing to recent budget reductions and this impacted a number of activities

  2. Extending the capabilities of the DIII-D Plasma Control System for worldwide fusion research collaborations

    International Nuclear Information System (INIS)

    Penaflor, B.G.; Ferron, J.R.; Walker, M.L.; Humphreys, D.A.; Leuer, J.A.; Piglowski, D.A.; Johnson, R.D.; Xiao, B.J.; Hahn, S.H.; Gates, D.A.

    2009-01-01

    This paper will discuss the recent enhancements which have been made to the DIII-D Plasma Control System (PCS) in order to further extend its usefulness as a shared tool for worldwide fusion research. The PCS developed at General Atomics is currently being used in a number of fusion research experiments worldwide, including the DIII-D Tokamak Facility in San Diego, and most recently the KSTAR Tokamak in South Korea. A number of enhancements have been made to support the ongoing needs of the DIII-D Tokamak in addition to meeting the needs of other PCS users worldwide. Details of the present PCS hardware and software architecture along with descriptions of the latest enhancements will be given.

  3. Performance of V-4Cr-4Ti material exposed to DIII-D tokamak environment

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    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.

  4. Performance, diagnostics, controls and plans for the gyrotron system on the DIII-D tokamak

    Directory of Open Access Journals (Sweden)

    Ponce D.M.

    2012-09-01

    Full Text Available The DIII-D ECH complex is being upgraded with three new depressed collector gyrotrons. The performance of the existing system has been very good. As more gyrotrons having higher power are added to the system, diagnostics of gyrotron operation, optimization of the performance and qualification of components for higher power become more important. A new FPGA-based gyrotron control system is being installed, additional capabilities for rapid real time variation of the rf injection angles by the DIII-D Plasma Control System are being tested and infrastructure enhancements are being completed. Longer term plans continue to include ECH as a major component in the DIII-D heating and current drive capabilities.

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

    International Nuclear Information System (INIS)

    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 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

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

    International Nuclear Information System (INIS)

    Budny, R.V.; Candy, J.; Waltz, R.E.

    2005-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  9. Real-time, vibration-compensated CO2 interferometer operation on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Ahlgren, D.R.; Crosbie, J.

    1988-01-01

    A multichannel, two-color, quadrature heterodyne interferometer is used to measure the line density in the DIII-D tokamak. The unique feature of this real-time vibration-compensated interferometer is the combination of high speed (1 MHz), high resolution (2π/256), and wide range ( +- 8193 fringes). Quadrature phase information from a CO 2 laser (10.6 μm) and a He--Ne laser (0.63 μm) are digitized with high-speed (6 MHz) flash digitizers. Zero crossings of the signals are counted with digital circuitry yielding quarter fringe resolution with a 4-MHz bandwidth. Further fringe resolution of 1/256 is provided at 350 kHz by a PROM which uses the digital signals as input to a look-up table. Analog line density is presently available at 80 kHz with a system noise equivalent phase shift of +- 2/256. Error monitoring is provided for low signal amplitude and exceeding the maximum fringe rate. In addition, a method to prevent coating of in-vessel mirrors due to plasma and vessel wall cleaning discharges has been developed

  10. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1990--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Simonen, T.C.; Evans, T.E. (eds.)

    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.

  11. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1990--September 30, 1991

    International Nuclear Information System (INIS)

    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

  12. Analysis of toroidal rotation data for the DIII-D tokamak

    International Nuclear Information System (INIS)

    St John, H.; Stroth, U.; Burrell, K.H.; Groebner, R.J.; DeBoo, J.C.; Gohil, P.

    1989-01-01

    Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Our results are based on numerical inversions using the transport code ONETWO, modified to account for the radial diffusion of toroidal angular momentum. 13 refs., 4 figs

  13. Analysis of toroidal rotation data for the DIII-D tokamak

    International Nuclear Information System (INIS)

    John, H.St.; Burrell, K.H.; Groebner, R.; DeBoo, J.; Gohil, P.

    1989-01-01

    Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner et al. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Similar studies have been previously reported for Doublet III, ASDEX, TFTR, JET and other tokamaks. (author) 13 refs., 4 figs

  14. Scrape-off layer transport and deposition studies in DIII-D

    International Nuclear Information System (INIS)

    Groth, M.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Porter, G. D.; Rensink, M. E.; Rognlien, T. D.; Boedo, J. A.; Rudakov, D. L.; Brooks, N. H.; Groebner, R. J.; Leonard, A. W.; West, W. P.; Elder, J. D.; McLean, A. G.; Lisgo, S.; Stangeby, P. C.; Wampler, W. R.; Watkins, J. G.; Whyte, D. G.

    2007-01-01

    Trace 13 CH 4 injection experiments into the main scrape-off layer (SOL) of low density L-mode and high-density H-mode plasmas have been performed in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to mimic the transport and deposition of carbon arising from a main chamber sputtering source. These experiments indicated entrainment of the injected carbon in plasma flow in the main SOL, and transport toward the inner divertor. Ex situ surface analysis showed enhanced 13 C surface concentration at the corner formed by the divertor floor and the angled target plate of the inner divertor in L-mode; in H-mode high surface concentration was found both at the corner and along the surface bounding the private flux region inboard of the outer strike point. Interpretative modeling was made consistent with these experimental results by imposing a parallel carbon ion flow in the main SOL toward the inner target, and a radial pinch toward the separatrix. Predictive modeling carried out to better understand the underlying plasma transport processes suggests that the deuterium flow in the main SOL is related to the degree of detachment of the inner divertor leg. These simulations show that carbon ions are entrained with the deuteron flow in the main SOL via frictional coupling, but higher charge-state carbon ions may be suspended upstream of the inner divertor X-point region due to balance of the friction force and the ion temperature gradient force

  15. Physics and Control of Locked Modes in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Volpe, Francesco

    2017-01-01

    This Final Technical Report summarizes an investigation, carried out under the auspices of the DOE Early Career Award, of the physics and control of non-rotating magnetic islands (''locked modes'') in tokamak plasmas. Locked modes are one of the main causes of disruptions in present tokamaks, and could be an even bigger concern in ITER, due to its relatively high beta (favoring the formation of Neoclassical Tearing Mode islands) and low rotation (favoring locking). For these reasons, this research had the goal of studying and learning how to control locked modes in the DIII-D National Fusion Facility under ITER-relevant conditions of high pressure and low rotation. Major results included: the first full suppression of locked modes and avoidance of the associated disruptions; the demonstration of error field detection from the interaction between locked modes, applied rotating fields and intrinsic errors; the analysis of a vast database of disruptive locked modes, which led to criteria for disruption prediction and avoidance.

  16. Effects of plasma shape and profiles on edge stability in DIII-D

    International Nuclear Information System (INIS)

    Lao, L.L.; Chan, V.S.; Chen, L.

    1998-12-01

    The results of recent experimental and theoretical studies concerning the effects of plasma shape and current and pressure profiles on edge instabilities in DIII-D are presented. Magnetic oscillations with toroidal mode number n ∼ 2--9 and a fast growth time γ -1 = 20--150 micros are often observed prior to the first giant type 1 ELM in discharges with moderate squareness. High n ideal ballooning second stability access encourages edge instabilities by facilitating the buildup of the edge pressure gradient and bootstrap current density which destabilize the intermediate to low n modes. Analysis suggests that discharges with large edge pressure gradient and bootstrap current density are more unstable to n > 1 modes. Calculations and experimental results show that ELM amplitude and frequency can be varied by controlling access to the second ballooning stability regime at the edge through variation of the squareness of the discharge shape. A new method is proposed to control edge instabilities by reducing access to the second ballooning stability regime at the edge using high order local perturbation of the plasma shape in the outboard bad curvature region

  17. Physics and Control of Locked Modes in the DIII-D Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Volpe, Francesco [Columbia Univ., New York, NY (United States). Dept. of Applied Physics and Applied Mathematics

    2017-01-30

    This Final Technical Report summarizes an investigation, carried out under the auspices of the DOE Early Career Award, of the physics and control of non-rotating magnetic islands (“locked modes”) in tokamak plasmas. Locked modes are one of the main causes of disruptions in present tokamaks, and could be an even bigger concern in ITER, due to its relatively high beta (favoring the formation of Neoclassical Tearing Mode islands) and low rotation (favoring locking). For these reasons, this research had the goal of studying and learning how to control locked modes in the DIII-D National Fusion Facility under ITER-relevant conditions of high pressure and low rotation. Major results included: the first full suppression of locked modes and avoidance of the associated disruptions; the demonstration of error field detection from the interaction between locked modes, applied rotating fields and intrinsic errors; the analysis of a vast database of disruptive locked modes, which led to criteria for disruption prediction and avoidance.

  18. Effects of plasma shape and profiles on edge stability in DIII-D

    International Nuclear Information System (INIS)

    Lao, L.L.; Ferron, J.R.; Miller, R.L.

    2001-01-01

    The results of recent experimental and theoretical studies concerning the effects of plasma shape and current and pressure profiles on edge instabilities in DIII-D are presented. Magnetic oscillations with toroidal mode number n∼2-9 and a fast growth time γ -1 =20-150μs are often observed prior to the first giant type I ELM in discharges with moderate squareness. High n ideal ballooning second stability access encourages edge instabilities by facilitating the buildup of the edge pressure gradient and bootstrap current density which destabilize the intermediate to low n modes. Analysis suggests that discharges with large edge pressure gradient and bootstrap current density are more unstable to n>1 modes. Calculations and experimental results show that ELM amplitude and frequency can be varied by controlling access to the second ballooning stability regime at the edge through variation of the squareness of the discharge shape. A new method is proposed to control edge instabilities by reducing access to the second ballooning stability regime at the edge using high order local perturbation of the plasma shape in the outboard bad curvature region. (author)

  19. Application of ECH to the Study of Transport in ITER Baseline Scenario-like Discharges in DIII-D

    Directory of Open Access Journals (Sweden)

    Pinsker R.I.

    2015-01-01

    Full Text Available Recent DIII-D experiments in the ITER Baseline Scenario (IBS have shown strong increases in fluctuations and correlated reduction of confinement associated with entering the electron-heating-dominated regime with strong electron cyclotron heating (ECH. The addition of 3.2 MW of 110 GHz EC power deposited at ρ∼0.42 to IBS discharges with ∼3 MW of neutral beam injection causes large increases in low-k and medium-k turbulent density fluctuations observed with Doppler backscatter (DBS, beam emission spectroscopy (BES and phase-contrast imaging (PCI diagnostics, correlated with decreases in the energy, particle, and momentum confinement times. Power balance calculations show the electron heat diffusivity χe increases significantly in the mid-radius region 0.4<ρ<0.8, which is roughly the same region where the DBS and BES diagnostics show the increases in turbulent density fluctuations. Confinement of angular momentum is also reduced during ECH. Studies with the TGYRO transport solver show that the model of turbulent transport embodied in the TGLF code quantitatively reproduces the measured transport in both the neutral beam (NB-only and in the NB plus EC cases. A simple model of the decrease in toroidal rotation with EC power is set forth, which exhibits a bifurcation in the rotational state of the discharge.

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

    International Nuclear Information System (INIS)

    Coda, S.; Porkolab, M.

    1994-05-01

    A novel CO 2 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 (rvec n approx-gt 10 9 cm -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

  1. Investigation of collisional effects within the bending magnet region of a DIII-D neutral beamline

    International Nuclear Information System (INIS)

    Kessler, D.N.; Hong, R.; Kellman, D.H.

    1993-10-01

    The region between the pole faces of the DIII-D neutral beamline residual ion bending magnets is an area of transient high gas pressure which may cause beam defocusing and increased heating of beamline internal components due to collisional effects. An investigation of these effects helps in understanding residual ion trajectories and in providing information for studying in the beamline capability for operation with increased pulse duration. Examination of collisional effects, and of the possible existence of space charge blow-up, was carried out by injecting deuterium gas into the region between the magnet pole faces with rates varying from 0 to 18 torr-ell/sec. Thermocouple and waterflow calorimetry data were taken to measure the beamline component heating and beam powder deposition on the magnet pole shields, magnet louvers, ion dump, beam collimators, and calorimeter. Data was also taken at gas flow rates varying from 0 to 25 torr-ell/sec into the neutralizer cell and is compared with the magnet region gas injection data obtained. Results show that both collisional effects and space charge blow-up play a role in magnet region component heating and that neutralizer gas flow sufficiently reduces component heating without incurring unacceptable power losses through collisional effects

  2. Thermal design, analysis, and experimental verification for a DIII-D cryogenic pump

    International Nuclear Information System (INIS)

    Baxi, C.B.; Anderson, P.; Langhorn, A.; Schaubel, K.; Smith, J.

    1991-01-01

    As part of the advanced divertor program, it is planned to install a 50 m 3 /s capacity cryopump for particle removal in the DIII-D tokamak. The cryopump will be located in the outer bottom corner of the vacuum vessel. The pump will consist of a surface at liquid helium temperature (helium panel) with a surface area of about 1 m 2 , a surface at liquid nitrogen temperature (nitrogen shield) to reduce radiation heat load on the helium panel, and a secondary shield around the nitrogen shield. The cryopump design poses a number of thermal hydraulic problems such as estimation of heat loads on helium and nitrogen panels, stability of the two-phase helium flow, performance of the pump components during high temperature bakeout, and cooldown performance of the helium panel from ambient temperatures. This paper presents the thermal analysis done to resolve these issues. A prototypic experiment performed at General Atomics verified the analysis and increased the confidence in the design. The experimental results are also summarized in this paper. (orig.)

  3. Fast wave current drive in neutral beam heated plasmas on DIII-D

    International Nuclear Information System (INIS)

    Petty, C.C.; Forest, C.B.; Pinsker, R.I.

    1997-04-01

    The physics of non-inductive current drive and current profile control using the fast magnetosonic wave has been demonstrated on the DIII-D tokamak. In non-sawtoothing discharges formed by neutral beam injection (NBI), the radial profile of the fast wave current drive (FWCD) was determined by the response of the loop voltage profile to co, counter, and symmetric antenna phasings, and was found to be in good agreement with theoretical models. The application of counter FWCD increased the magnetic shear reversal of the plasma and delayed the onset of sawteeth, compared to co FWCD. The partial absorption of fast waves by energetic beam ions at high harmonics of the ion cyclotron frequency was also evident from a build up of fast particle pressure near the magnetic axis and a correlated increase in the neutron rate. The anomalous fast particle pressure and neutron rate increased with increasing NBI power and peaked when a harmonic of the deuterium cyclotron frequency passed through the center of the plasma. The experimental FWCD efficiency was highest at 2 T where the interaction between the fast waves and the beam ions was weakest; as the magnetic field strength was lowered, the FWCD efficiency decreased to approximately half of the maximum theoretical value

  4. OEDGE modeling of the DIII-D double null (CH4)-C-13 puffing experiment

    International Nuclear Information System (INIS)

    Elder, J.D.; Wampler, W.R.; McLean, A.G.; Stangeby, P.C.; Allen, S.L.; Bray, B.D.; Brooks, N.H.; Leonard, A.W.; Unterberg, Ezekial A.; Watkins, J.G.

    2011-01-01

    Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. (CH4)-C-13 was injected into the secondary outer divertor to simulate sputtering there. The majority of the C-13 found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall (CH4)-C-13 injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240 degrees C, is incomplete. The principal features of the measured C-13 deposition pattern have been replicated by the OEDGE interpretive code.

  5. Dissipation of post-disruption runaway electron plateaus by shattered pellet injection in DIII-D

    Science.gov (United States)

    Shiraki, D.; Commaux, N.; Baylor, L. R.; Cooper, C. M.; Eidietis, N. W.; Hollmann, E. M.; Paz-Soldan, C.; Combs, S. K.; Meitner, S. J.

    2018-05-01

    We report on the first demonstration of dissipation of fully avalanched post-disruption runaway electron (RE) beams by shattered pellet injection in the DIII-D tokamak. Variation of the injected species shows that dissipation depends strongly on the species mixture, while comparisons with massive gas injection do not show a significant difference between dissipation by pellets or by gas, suggesting that the shattered pellet is rapidly ablated by the relativistic electrons before significant radial penetration into the runaway beam can occur. Pure or dominantly neon injection increases the RE current dissipation through pitch-angle scattering due to collisions with impurity ions. Deuterium injection is observed to have the opposite effect from neon, reducing the high-Z impurity content and thus decreasing the dissipation, and causing the background thermal plasma to completely recombine. When injecting mixtures of the two species, deuterium levels as low as  ∼10% of the total injected atoms are observed to adversely affect the resulting dissipation, suggesting that complete elimination of deuterium from the injection may be important for optimizing RE mitigation schemes.

  6. Turbulence imaging and applications using beam emission spectroscopy on DIII-D (invited)

    Science.gov (United States)

    McKee, G. R.; Fenzi, C.; Fonck, R. J.; Jakubowski, M.

    2003-03-01

    Two-dimensional measurements of density fluctuations are obtained in the radial and poloidal plane of the DIII-D tokamak with the Beam Emission Spectroscopy (BES) diagnostic system. The goals are to visualize the spatial structure and time evolution of turbulent eddies, as well as to obtain the 2D statistical properties of turbulence. The measurements are obtained with an array of localized BES spatial channels configured to image a midplane region of the plasma. 32 channels have been deployed, each with a spatial resolution of about 1 cm in the radial and poloidal directions, thus providing measurements of turbulence in the wave number range 0movies have broad application to a wide variety of fundamental turbulence studies: imaging of the highly complex, nonlinear turbulent eddy interactions, measurement of the 2D correlation function, and S(kr,kθ) wave number spectra, and direct measurement of the equilibrium and time-dependent turbulence flow field. The time-dependent, two-dimensional turbulence velocity flow-field is obtained with time-delay-estimation techniques.

  7. Update on the DIII-D ECH system: experiments, gyrotrons, advanced diagnostics, and controls

    Directory of Open Access Journals (Sweden)

    Lohr John

    2017-01-01

    Full Text Available The ECH system on DIII-D is continuing to be upgraded, while simultaneously being operated nearly daily for plasma experiments. The latest major hardware addition is a new 117.5 GHz gyrotron, which generated 1.7 MW for short pulses during factory testing. A new gyrotron control system based on Field Programmable Gate Array (FPGA technology with very high speed system data acquisition has significantly increased the flexibility and reliability of individual gyrotron operation. We have improved the performance of the fast mirror scanning, both by increasing the scan speeds and by adding new algorithms for controlling the aiming using commands generated by the Plasma Control System (PCS. The system is used for transport studies, ELM control, current profile control, non-inductive current generation, suppression of MHD modes, startup assist, plasma density control, and other applications. A program of protective measures, which has been in place for more than two years, has eliminated damage to hardware and diagnostics caused by overdense operation. Other activities not directly related to fusion research have used the ECH system to test components, study methods for improving production of semiconductor junctions and materials, and test the feasibility of using ground based microwave systems to power satellites into orbit.

  8. Update on the DIII-D ECH system: experiments, gyrotrons, advanced diagnostics, and controls

    Science.gov (United States)

    Lohr, John; Brambila, Rigoberto; Cengher, Mirela; Gorelov, Yuri; Grosnickle, William; Moeller, Charles; Ponce, Dan; Torrezan, Antonio; Ives, Lawrence; Reed, Michael; Blank, Monica; Felch, Kevin; Parisuaña, Claudia; LeViness, Alexandra

    2017-08-01

    The ECH system on DIII-D is continuing to be upgraded, while simultaneously being operated nearly daily for plasma experiments. The latest major hardware addition is a new 117.5 GHz gyrotron, which generated 1.7 MW for short pulses during factory testing. A new gyrotron control system based on Field Programmable Gate Array (FPGA) technology with very high speed system data acquisition has significantly increased the flexibility and reliability of individual gyrotron operation. We have improved the performance of the fast mirror scanning, both by increasing the scan speeds and by adding new algorithms for controlling the aiming using commands generated by the Plasma Control System (PCS). The system is used for transport studies, ELM control, current profile control, non-inductive current generation, suppression of MHD modes, startup assist, plasma density control, and other applications. A program of protective measures, which has been in place for more than two years, has eliminated damage to hardware and diagnostics caused by overdense operation. Other activities not directly related to fusion research have used the ECH system to test components, study methods for improving production of semiconductor junctions and materials, and test the feasibility of using ground based microwave systems to power satellites into orbit.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy

  12. Advances in the operation of the DIII-D neutral beam computer systems

    International Nuclear Information System (INIS)

    Phillips, J.C.; Busath, J.L.; Penaflor, B.G.; Piglowski, D.; Kellman, D.H.; Chiu, H.K.; Hong, R.M.

    1998-02-01

    The DIII-D neutral beam system routinely provides up to 20 MW of deuterium neutral beam heating in support of experiments on the DIII-D tokamak, and is a critical part of the DIII-D physics experimental program. The four computer systems previously used to control neutral beam operation and data acquisition were designed and implemented in the late 1970's and used on DIII and DIII-D from 1981--1996. By comparison to modern standards, they had become expensive to maintain, slow and cumbersome, making it difficult to implement improvements. Most critical of all, they were not networked computers. During the 1997 experimental campaign, these systems were replaced with new Unix compliant hardware and, for the most part, commercially available software. This paper describes operational experience with the new neutral beam computer systems, and new advances made possible by using features not previously available. These include retention and access to historical data, an asynchronously fired ''rules'' base, and a relatively straightforward programming interface. Methods and principles for extending the availability of data beyond the scope of the operator consoles will be discussed

  13. 60 MHz fast wave current drive experiment for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Mayberry, M.J.; Chiu, S.C.; Porkolab, M.; Chan, V.; Freeman, R.; Harvey, R.; Pinsker, R. (General Atomics, San Diego, CA (USA))

    1989-07-01

    The DIII-D facility provides an opportunity to test fast wave current drive appoach. Efficient FWCD is achieved by direct electron absorption due to Landa damping and transit time magnetic pumping. To avoid competing damping mechamisms we seek to maximize the single-pass asorption of the fast waves by electrons. (AIP)

  14. Data analysis software tools for enhanced collaboration at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

    Schachter, J.; Peng, Q.; Schissel, D.P.

    2000-01-01

    Data analysis at the DIII-D National Fusion Facility is simplified by the use of two software packages in analysis codes. The first is 'GAPlotObj', an IDL-based object-oriented library used in visualization tools for dynamic plotting. GAPlotObj gives users the ability to manipulate graphs directly through mouse and keyboard-driven commands. The second software package is 'MDSplus', which is used at DIII-D as a central repository for analyzed data. GAPlotObj and MDSplus reduce the effort required for a collaborator to become familiar with the DIII-D analysis environment by providing uniform interfaces for data display and retrieval. Two visualization tools at DIII-D that benefit from them are 'ReviewPlus' and 'EFITviewer'. ReviewPlus is capable of displaying interactive 2D and 3D graphs of raw, analyzed, and simulation code data. EFITviewer is used to display results from the EFIT analysis code together with kinetic profiles and machine geometry. Both bring new possibilities for data exploration to the user, and are able to plot data from any fusion research site with an MDSplus data server

  15. Comparison of calculated neutral beam shine through with measured shine-through in DIII-D

    International Nuclear Information System (INIS)

    Chiu, H.K.; Hong, R.

    1997-11-01

    A comparison of the calculated shine through of neutral particle beams in the DIII-D plasma to measured values inferred from the target temperature rise is reported. This provides an opportunity to verify the shine through calculations and makes them more reliable in those cases where the shine through can not be measured. The DIII-D centerpost neutral beam target tiles are safe-guarded against excessive beam shine-through by pyrometry and thermocouple (TC) arrays on the tiles. Shine-through beam power is calculated from the measured temperature changes reported by the target tile TC array. These measurements are performed at the beginning of each operational year at DIII-D. Theoretically, the beam energy deposited into the plasma can be expressed as a function of the change in beam density. Neutral beam energy deposition in plasma (of known density) is inferred by comparing the results of a series of shine-through measurements for the 1997 campaign at DIII-D to the expected shine-through given by theory

  16. Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

    2000-12-01

    Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

  17. Tools for remote collaboration on the DIII-D national fusion facility

    International Nuclear Information System (INIS)

    McHarg, B.B. Jr.; Greenwood, D.

    1999-01-01

    The DIII-D national fusion facility, a tokamak experiment funded by the US Department of Energy and operated by General Atomics (GA), is an international resource for plasma physics and fusion energy science research. This facility has a long history of collaborations with scientists from a wide variety of laboratories and universities from around the world. That collaboration has mostly been conducted by travel to and participation at the DIII-D site. Many new developments in the computing and technology fields are now facilitating collaboration from remote sites, thus reducing some of the needs to travel to the experiment. Some of these developments include higher speed wide area networks, powerful workstations connected within a distributed computing environment, network based audio/video capabilities, and the use of the world wide web. As the number of collaborators increases, the need for remote tools become important options to efficiently utilize the DIII-D facility. In the last two years a joint study by GA, Princeton Plasma Physics Laboratory (PPPL), Lawrence Livermore National Laboratory (LLNL), and Oak Ridge National Laboratory (ORNL) has introduced remote collaboration tools into the DIII-D environment and studied their effectiveness. These tools have included the use of audio/video for communication from the DIII-D control room, the broadcast of meetings, use of inter-process communication software to post events to the network during a tokamak shot, the creation of a DCE (distributed computing environment) cell for creating a common collaboratory environment, distributed use of computer cycles, remote data access, and remote display of results. This study also included sociological studies of how scientists in this environment work together as well as apart. (orig.)

  18. Observation of the L-H confinement bifurcation triggered by a turbulence-driven shear flow in a tokamak plasma.

    Science.gov (United States)

    Yan, Z; McKee, G R; Fonck, R; Gohil, P; Groebner, R J; Osborne, T H

    2014-03-28

    Comprehensive 2D turbulence and eddy flow velocity measurements on DIII-D demonstrate a rapidly increasing turbulence-driven shear flow that develops ∼100  μs prior to the low-confinement (L mode) to high-confinement (H mode) transition and appears to trigger it. These changes are localized to a narrow layer 1-2 cm inside the magnetic boundary. Increasing heating power increases the Reynolds stress, the energy transfer from turbulence to the poloidal flow, and the edge flow shearing rate that then exceeds the decorrelation rate, suppressing turbulence and triggering the transition.

  19. Interpretive modeling of simple-as-possible-plasma discharges on DIII-D using the OEDGE code

    International Nuclear Information System (INIS)

    Stangeby, P.C.; Elder, J.D.; Boedo, J.A.; Bray, B.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Isler, R.C.; Lao, L.L.; Lisgo, S.; Porter, G.D.; Reiter, D.; Rudakov, D.L.; Watkins, J.G.; West, W.P.; Whyte, D.G.

    2003-01-01

    Recently a number of major, unanticipated effects have been reported in tokamak edge research raising the question of whether we understand the controlling physics of the edge. This report is on the first part - here focused on the outer divertor - of a systematic study of the simplest possible edge plasma - no ELMs, no detachment, etc. - for a set of 10 repeat, highly diagnosed, single-null, divertor discharges in DIII-D. For almost the entire, extensive data set so far evaluated, the matches of experiment and model are so close as to imply that the controlling processes at the outer divertor for these simple plasma conditions have probably been correctly identified and quantitatively characterized in the model. The principal anomaly flagged so far relates to measurements of T e near the target, potentially pointing to a deficiency in our understanding of sheath physics in the tokamak environment

  20. CONTROL SYSTEM FOR THE LITHIUM BEAM EDGE PLASMA CURRENT DENSITY DIAGNOSTIC ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PEAVY, J.J.; CARY, W.P; THOMAS, D.M; KELLMAN, D.H.; HOYT, D.M; DELAWARE, S.W.; PRONKO, S.G.E.; HARRIS, T.E.

    2004-03-01

    OAK-B135 An edge plasma current density diagnostic employing a neutralized lithium ion beam system has been installed on the DIII-D tokamak. The lithium beam control system is designed around a GE Fanuc 90-30 series PLC and Cimplicity(reg s ign) HMI (Human Machine Interface) software. The control system operates and supervises a collection of commercial and in-house designed high voltage power supplies for beam acceleration and focusing, filament and bias power supplies for ion creation, neutralization, vacuum, triggering, and safety interlocks. This paper provides an overview of the control system, while highlighting innovative aspects including its remote operation, pulsed source heating and pulsed neutralizer heating, optimizing beam regulation, and beam ramping, ending with a discussion of its performance

  1. Demonstration in the DIII-D tokamak of an alternate baseline scenario for ITER and other burning plasma experiments

    International Nuclear Information System (INIS)

    Luce, T.C.; Ferron, J.R.; Wade, M.R.

    2003-01-01

    Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak in stationary conditions with relatively low plasma current (q 95 > 4). A figure of merit for fusion gain β N H 89 / q 95 2 2 has been maintained at values corresponding to Q = 10 operation in a burning plasma for >6 s or 36 τ E and 2 τ R . The key element is the relaxation of the current profile to a stationary state with q min > 1, which allows stable operation up to the no-wall ideal β limit. These plasmas maintain particle balance by active pumping rather than transient wall conditioning. The reduced current lessens significantly the potential for structural damage in the event of a major disruption. (author)

  2. Characterisation of the ELM synchronized H-mode edge pedestal in ASDEX upgrade and DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Philip A.; Wolfrum, Elisabeth; Guenter, Sibylle; Kurzan, Bernd; Zohm, Hartmut [Max Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Groebner, Rich; Osborne, Tom H.; Ferron, John; Snyder, Philip B. [General Atomics, San Diego, CA (United States); Dunne, Mike G. [Department of Physics, University College Cork, Association Euratom-DCU, Cork (Ireland); Collaboration: ASDEX Upgrade Team; DIII-D Team

    2011-07-01

    The results of a large database of edge pedestal data from type-I ELMy H-mode discharges from ASDEX Upgrade and DIII-D are presented. The data from high resolution edge diagnostics of both devices is analysed with the same analysis code in order to avoid systematic differences. Furthermore, sophisticated equilibrium reconstructions are used to asses uncertainties which arise during mapping from 2D real space coordinates to 1D flux coordinates. ELM synchronization allows the study of the pedestal structure at the ELM stability boundary. The pedestal is characterized by its top value, the gradient and the width. A large parameter range is covered by the two devices. Over this parameter range the profile shape of edge electron density differs from that of the temperature, irrespective of the device. However, the resulting electron pressure profile shape remains similar for all analysed H-Mode discharges.

  3. Momentum confinement at low torque

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, W M [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); De Grassie, J S [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Budny, R [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Kinsey, J E [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Kramer, G J [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Luce, T C [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M A [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Mikkelsen, D [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Nazikian, R [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Petty, C C [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Politzer, P A [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Scott, S D [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Zeeland, M A Van [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Zarnstorff, M C [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

    2007-12-15

    Momentum confinement was investigated on DIII-D as a function of applied neutral beam torque at constant normalized beta {beta}{sub N}, by varying the mix of co (parallel to the plasma current) and counter neutral beams. Under balanced neutral beam injection (i.e. zero total torque to the plasma), the plasma maintains a significant rotation in the co-direction. This 'intrinsic' rotation can be modeled as being due to an offset in the applied torque (i.e. an 'anomalous torque'). This anomalous torque appears to have a magnitude comparable to one co neutral beam source. The presence of such an anomalous torque source must be taken into account to obtain meaningful quantities describing momentum transport, such as the global momentum confinement time and local diffusivities. Studies of the mechanical angular momentum in ELMing H-mode plasmas with elevated q{sub min} show that the momentum confinement time improves as the torque is reduced. In hybrid plasmas, the opposite effect is observed, namely that momentum confinement improves at high torque/rotation. GLF23 modeling suggests that the role of E x B shearing is quite different between the two plasmas, which may help to explain the different dependence of the momentum confinement on torque.

  4. COMPLETE SUPPRESSION OF THE M/N = 2/1 NEOCLASSICAL TEARING MODE USING RADIALLY LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D AND THE REQUIREMENTS FOR ITER

    International Nuclear Information System (INIS)

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

    2003-01-01

    A271 COMPLETE SUPPRESSION OF THE M/N = 2/1 NEOCLASSICAL TEARING MODE USING RADIALLY LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D AND THE REQUIREMENTS FOR ITER. DIII-D experiments demonstrate the first real-time feedback control of the relative location of a narrow beam of microwaves to completely suppress and eliminate a growing tearing mode at the q = 2 surface. long wavelength tearing modes such as the m/n = 2/1 instability are particularly deleterious to tokamak operation. Confinement is seriously degraded by the island, plasma rotation can cease (mode-lock) and disruption can occur. The neoclassical tearing mode (NTM) becomes unstable due to the presence of a helically-perturbed bootstrap current and can be stabilized by replacing the missing bootstrap current in the island O-point by precisely located co-electron cyclotron current drive (ECCD). The optimum position is found when the DIII-D plasma control system (PCS) is put into a search and suppress mode that makes small radial shifts (in about 1 cm steps) in the ECCD location based on minimizing the Mirnov amplitude. Requirements for ITER are addressed

  5. ELECTRON TEMPERATURE FLUCTUATIONS AND CROSS-FIELD HEAT TRANSPORT IN THE EDGE OF DIII-D

    International Nuclear Information System (INIS)

    RUDAKOV, DL; BOEDO, JA; MOYER, RA; KRASENINNIKOV, S; MAHDAVI, MA; McKEE, GR; PORTER, GD; STANGEBY, PC; WATKINS, JG; WEST, WP; WHYTE, DG.

    2003-01-01

    OAK-B135 The fluctuating E x B velocity due to electrostatic turbulence is widely accepted as a major contributor to the anomalous cross-field transport of particles and heat in the tokamak edge and scrape-off layer (SOL) plasmas. This has been confirmed by direct measurements of the turbulent E x B transport in a number of experiments. Correlated fluctuations of the plasma radial velocity v r , density n, and temperature T e result in time-average fluxes of particles and heat given by (for electrons): Equation 1--Λ r ES = r > = 1/B varφ θ ; Equation 2--Q r ES = e (tilde v) r > ∼ 3/2 kT e Λ r ES + 3 n e /2 B varφ e (tilde E) θ > Q conv + Q cond . The first term in Equation 2 is referred to as convective and the second term as conductive heat flux. Experimental determination of fluxes given by Equations 1 and 2 requires simultaneous measurements of the density, temperature and poloidal electric field fluctuations with high spatial and temporal resolution. Langmuir probes provide most readily available (if not the only) tool for such measurements. However, fast measurements of electron temperature using probes are non-trivial and are not always performed. Thus, the contribution of the T e fluctuations to the turbulent fluxes is usually neglected. Here they report results of the studies of T e fluctuations and their effect on the cross-field transport in the SOL of DIII-D

  6. Commissioning of the long-pulse fast wave current drive antennas for DIII-D

    International Nuclear Information System (INIS)

    Baity, F.W.; Barber, G.C.; Goulding, R.H.; Hoffman, D.J.; DeGrassie, J.S.; Pinsker, R.I.; Petty, C.C.; Cary, W.

    1995-01-01

    Two new four-element fast wave current drive antennas have been installed on DIII-D. These antennas are designed for 10-s pulses at 2 MW each in the frequency range of 30 to 120 MHz. Each element comprises two poloidal segments fed in parallel in order to optimize plasma coupling at the upper end of the frequency range. The antennas are mounted on opposite sides of the vacuum vessel, in ports designated 0 degrees and 180 degrees after their toroidal angle. Each antenna array is fed by a single transmitter. The power is first split two ways by means of a 3-dB hybrid coupler, then each of these lines feeds a resonant loop connecting a pair of array elements. The power transfer during asymmetric phasing is shunted between resonant loops by a decoupler. The resonant loops are fitted with line stretchers so that multiple frequencies of operation are possible without reconfiguring the transmission line. Commissioning of these antennas has been underway since June 1994. Several deficiencies in the transmission line system were uncovered during initial vacuum conditioning, including problems with the transmission line insulators and with the drive rods for the variable elements. The former was solved by replacing the original alumina insulators, and the latter has been avoided during operation to date by positioning the tuners to avoid high voltage appearing on the drive rods. A modified design for the drive rods will be implemented before RF operations resume operation June 1995. New transmitters were procured from ABB for the new antennas and were installed in parallel with the antenna installation. During initial vacuum conditioning of the antenna in the 180 degree port a fast digital oscilloscope was used to try to pinpoint the location of arcing by a time-of-flight technique and to develop an understanding of the typical arc signature in the system

  7. A cryocondensation pump for the DIII-D Advanced Divertor Program

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.; Reis, E.; Sevier, L.

    1992-03-01

    A cryocondensation pump was designed for the baffle chamber of General Atomics DIII-D tokamak and will be installed in the fall of 1992. The purpose of the pump is to study plasma density control by pumping the divertor. The pump is toroidally continuous, approximately 10 m long and located in the lower outer corner of the vacuum chamber of the machine. It consists of a 1 m 2 liquid helium-cooled surface surrounded by a liquid nitrogen-cooled shield to limit the heat load on the helium-cooled surface. The liquid nitrogen-cooled surface is surrounded by a radiation/particle shield to prevent energetic particles from impacting and releasing condensed water molecules. A thermal enhancement coating was applied to the nitrogen shell to lower the maximum temperature of the shell. The coating is non-continuous to keep the toroidal electrical resistance high. The whole pump is supported off the water-cooled vacuum vessel wall. Supports for the pump were designed to accommodate the thermal differences between the 4 K helium surface, the 77 K nitrogen shells, and the 300 K vacuum vessel supporting the pump and to provide a low heat leak structural support. Disruption loading on the pump was analyzed and a finite element structural analysis of the pump was completed. A testing program was completed to evaluate coating techniques to enhance heat transfer and emissivity of the various surfaces. Fabrication tests were performed to determine the best method of attaching the liquid nitrogen flow tubes to their shield surfaces and to determine the best alternative to fabricating the different shells of the pump. A prototype sector of the pump was built to verify fabrication and assembly techniques

  8. COMPLETE SUPPRESSION OF THE m=2/n-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D

    International Nuclear Information System (INIS)

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

    2003-01-01

    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 θ /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

  9. Energy confinement of high-density tokamaks

    NARCIS (Netherlands)

    Schüller, F.C.; Schram, D.C.; Coppi, B.; Sadowski, W.

    1977-01-01

    Neoclassical ion heat conduction is the major energy loss mechanism in the center of an ohmically heated high-d. tokamak discharge (n>3 * 1020 m-3). This fixes the mutual dependence of plasma quantities on the axis and leads to scaling laws for the poloidal b and energy confinement time, given the

  10. Lower-hybrid counter current drive for edge current density modification in DIII-D

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Nevins, W.M.; Porkolab, M.; Bonoli, P.T.; Harvey, R.W.

    1994-01-01

    Each of the Advanced Tokamak operating modes in DIII-D is thought to have a distinctive current density profile. So far these modes have only been achieved transiently through experiments which ramp the plasma current and shape. Extension of these modes to steady state requires non-inductive current profile control, e.g., with lower hybrid current drive (LHCD). Calculations of LHCD have been done for DIII-D using the ACCOME and CQL3D codes, showing that counter driven current at the plasma edge can cancel some of the undesirable edge bootstrap current and potentially extend the VH-mode. Results will be presented for scenarios using 2.45 GHz LH waves launched from both the midplane and off-axis ports. The sensitivity of the results to injected power, n e and T e , and launched wave spectrum will also be shown

  11. Edge density fluctuation diagnostic for DIII-D using lithium beams: 1992 annual report

    International Nuclear Information System (INIS)

    Thomas, D.M.

    1994-01-01

    During the past several months the Lithium beam diagnostic was commissioned of DIII-D and began yielding useful information. The author developed the remote control and monitoring of the ion source operation and beam formation and focussing, and integrated the control system and data acquisition into the DIII-D operating system. Several detector types were fabricated, and fluorescence data were collected using several differing detector arrangements. Beam-gas measurements were conducted to analyze the intrinsic beam fluctuations and stability. Fluorescence data was then obtained on a number of Tokamak discharges under varying discharge conditions. Analysis of this initial data is proceeding but has already yielded some interesting features. These include changes in the edge plasma density behavior during the l- to h-transition, disruptions, and edge localized modes (ELMs). Based on the quality of data obtained the author proceeded with the design and construction of the full 16-channel detection system which will be completed and tested shortly

  12. Handling and archiving of magnetic fusion data at DIII-D

    International Nuclear Information System (INIS)

    VanderLaan, J.F.; Miller, S.; McHarg, B.B. Jr.; Henline, P.A.

    1995-10-01

    Recent modifications to the computer network at DIII-D enhance the collection and distribution of newly acquired and archived experimental data. Linked clients and servers route new data from diagnostic computers to centralized mass storage and distribute data on demand to local and remote workstations and computers. Capacity for data handling exceeds the upper limit of DIII-D Tokamak data production of about 4 GBytes per day. Network users have fast access to new data stored on line. An interactive program handles requests for restoration of data archived off line. Disk management procedures retain selected data on line in preference to other data. Redundancy of all components on the archiving path from the network to magnetic media has prevented loss of data. Older data are rearchived as dictated by limited media life

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

    International Nuclear Information System (INIS)

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

    2010-01-01

    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 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.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    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. (paper)

  15. Experimental survey of the L-H transition conditions in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Gohil, P.; Watkins, J.C.

    1994-01-01

    We present the global analysis of a recent survey of the H-mode power threshold in DIII-D using D o → D + NBI after boronization of the vacuum vessel. Single parameter scans of B T , I p , density, and plasma shape have been carried out on the DIII-D tokamak for neutral beam heated single-null and double-null diverted plasmas. In single-null discharges, the power threshold is found to increase approximately linearly with B T and n e but remains independent of I p . In double-null discharges, the power threshold is found to be approximately independent of both B T and n e . Various shape parameters such as plasma-wall gaps had only a weak effect on the power threshold. Imbalancing the double null configuration resulted in a large increase in the threshold power

  16. Effects of particle exhaust on neutral compression ratios in DIII-D

    International Nuclear Information System (INIS)

    Colchin, R.J.; Maingi, R.; Wade, M.R.; Allen, S.L.; Greenfield, C.M.

    1998-08-01

    In this paper, neutral particles in DIII-D are studied via their compression in the plenum and via particle exhaust. The compression of gas in the plena is examined in terms of the magnetic field configuration and wall conditions. DIII-D compression ratios are observed in the range from 1 to ≥ 1,000. Particle control ultimately depends on the exhaust of neutrals via plenum or wall pumping. Wall pumping or outgassing is calculated by means of a detailed particle balance throughout individual discharges, and its effect on particle control is discussed. It is demonstrated that particle control through wall conditioning leads to lower normalized densities. A two-region model shows that the gas compression ratio (C div = divertor plenum neutral pressure/torus neutral pressure) can be interpreted in relation to gas flows in the torus and divertor including the pumping speed of the plenum cryopumps, plasma pumping, and the pumping or outgassing of the walls

  17. A decade of DIII-D research. Final report for the period of work, October 1, 1989--September 30, 1998

    International Nuclear Information System (INIS)

    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

  18. A decade of DIII-D research. Final report for the period of work, October 1, 1989--September 30, 1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  19. Initial assessment of lower hybrid current drive in DIII-D

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Porkolab, M.

    1992-01-01

    In this report we discuss the choice of rf frequency for LHCD in DIII-D, and its consequences for the operating regimes. We also discuss some Brambilla code results for particular launcher geometry, as well as edge density requirements for efficient coupling by the launcher. We present results of the ACCOME code modelling for selected frequency spectra. Finally, a summary is given, as well as suggestions for future studies

  20. 2 MW 110 GHz ECH heating system for DIII-D

    International Nuclear Information System (INIS)

    Moeller, C.; Prater, R.; Callis, R.; Remsen, D.; Doane, J.; Cary, W.; Phelps, R.; Tupper, M.

    1990-09-01

    A 2 MW 110 GHz ECH system using Varian 0.5 MW gyrotrons is under construction for use on the DIII-D tokamak by late 1991. Most of the components are being design and fabricated at General Atomics, including the gyrotron tanks, superconducting magnets, and transmission line. These components are intended for operation with 10 second pulses and, in the future, with 1 MW gyrotrons. 6 refs., 5 figs

  1. Dependence of {beta} {center_dot} {tau} on plasma shape in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Lazarus, E.A. [Oak Ridge National Lab., TN (United States)

    1993-12-31

    In this paper we discuss the observed variation in plasma performance with plasma shape, in particular, we shall compare single and double null diverted plasmas. The product {beta} {center_dot} {tau} has been used as a figure-of-merit for comparing different toroidal magnetic configurations. Here we shall use it as the figure-of-merit for comparing differing configurations within the DIII-D tokamak. (author) 5 refs., 5 figs.

  2. Dependence of β·τ on plasma shape in DIII-D

    International Nuclear Information System (INIS)

    Lazarus, E.A.

    1993-05-01

    In this paper we discuss the observed variation in plasma performance with plasma shape, in particular, we shall compare single and double null diverted plasmas. The product β·τ has been used as a figure-of-merit for comparing different toroidal magnetic configurations. Here we shall use it as the figure-of-merit for comparing differing configurations within the DIII-D tokamak

  3. Characterization and Modification of Edge-Driven Instabilities in the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Ferron, J.R.; Lao, L.L.; Osborne, T.H.; Strait, E.J.; Turnbull, A.D.; Miller, R.L.; Taylor, T.S.; Doyle, E.J.; Rice, B.W.; Zhang, C.; Chen, L.; Baylor, L.R.; Murakami, M.; Wade, M.R.

    1999-01-01

    The character of edge localized modes (ELMs) and the height of the edge pressure pedestal in DIII-D tokamak H-mode discharges have been modified by varying the discharge shape (triangularity and squareness) and the safety factor, increasing the edge radiation, and injecting deuterium pellets. Changes in the ELM frequency and amplitude, and the magnitude of the edge pressure gradient, and changes in the calculated extent of the region of access to the ballooning mode second stability regime are observed

  4. First results on fast wave current drive in advanced tokamak discharges in DIII-D

    International Nuclear Information System (INIS)

    Prater, R.; Cary, W.P.; Baity, F.W.

    1995-07-01

    Initial experiments have been performed on the DIII-D tokamak on coupling, direct electron heating, and current drive by fast waves in advanced tokamak discharges. These experiments showed efficient central heating and current drive in agreement with theory in magnitude and profile. Extrapolating these results to temperature characteristic of a power plant (25 keV) gives current drive efficiency of about 0.3 MA/m 2

  5. A neural network for the analysis of DIII-D charge exchange recombination data

    International Nuclear Information System (INIS)

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

    1994-01-01

    A neural network of the multiple-layer perceptron (MLP) type, named CERNEUR, has been created for the task of analysing the charge exchange recombination data from DIII-D for the purpose of providing control-room ion temperatures and rotation velocities between shots and, in the future, to provide initial guesses for the standard curve-fitting code. CERNEUR provides very useful 'control-room' in-between shot analysis of the rotation velocity and ion temperature profiles. (author)

  6. Recent improvements to the DIII-D neutral beam instrumentation and control system

    International Nuclear Information System (INIS)

    Kellman, D.H.; Hong, R.

    1997-11-01

    The DIII-D neutral beam (NB) instrumentation and control (I and C) system provides for operational control and synchronization of the eight DIII-D neutral beam injection systems, as well as for pertinent data acquisition and safety interlocking. Recently, improvements were made to the I and C system. With the replacement of the NB control computers, new signal interfacing was required to accommodate the elimination of physical operator panels, in favor of graphical user interface control pages on computer terminal screens. The program in the mode control (MC) programmable logic controller (PLC), which serves as a logic-processing interface between the NB control computers and system hardware, was modified to improve the availability of NB heating of DIII-D plasmas in the event that one or more individual beam systems suddenly become unavailable while preparing for a tokamak experimental shot sequences. An upgraded computer platform was adopted for the NB control system operator interface and new graphical user interface pages were developed to more efficiently display system status data. A failure mode of the armor tile infrared thermometers (pyrometers), which serve to terminate beam pulsing if beam shine-through overheats wall thermal shielding inside the DIII-D tokamak, was characterized such that impending failures can be detected and repairs effected to mitigate beam system down-time. The hardware that controls gas flow to the beamline neutralizer cells was upgraded to reduce susceptibility to electromagnetic interference (EMI), and interlocking was provided to terminate beam pulsing in the event of insufficient neutralizer gas flow. Motivation, implementation, and results of these improvements are presented

  7. Thermal-stress analysis and testing of DIII-D armor tiles

    International Nuclear Information System (INIS)

    Baxi, C.B.; Anderson, P.M.; Reis, E.E.; Smith, J.P.; Smith, P.D.; Croesmann, C.; Watkins, J.; Whitley, J.

    1987-10-01

    It is planned to install about 1500 new armor tiles in the DIII-D tokamak. The armor tiles currently installed in DIII-D are made by brazing Poco AXF-5Q graphite onto Inconel X-750 stock. A small percentage of these have failed by breakage of graphite. These failures were believed to be related to significant residual stress remaining in graphite after brazing. Hence, an effort was undertaken to improve the design with all-graphite tiles. Three criteria must be satisfied by the armor tiles and the hardware used to attach the tiles to the vessel walls: tiles should not structurally fail, peak tile temperature must be less than 2500 K, and peak vessel stresses must be below acceptable levels. A number of alternate design concepts were first analyzed with the two-dimensional finite element codes TOPAZ2D and NIKE2D. Promising designs were optimized for best parameters such as thicknesses, etc. The two best designs were further analyzed for thermal stresses with the three-dimensional codes P/THERMAL and P/STRESS. Prototype tiles of a number of materials were fabricated by GA and tested at the Plasma Materials Test Facility of the Sandia National Laboratory at Albuquerque. The tests simulated the heat flux and cooling conditions in DIII-D. This paper describes the 2-D and 3-D thermal stress analyses, the test results and logic which led to the selected design of the DIII-D armor tiles. 5 refs., 7 figs., 3 tabs

  8. Extending DIII-D Neutral Beam Modulated Operations with a Camac Based Total on Time Interlock

    International Nuclear Information System (INIS)

    Baggest, D.S.; Broesch, J.D.; Phillips, J.C.

    1999-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    Theoretical work, computation, and results from TCV [J. Decker “Effect of density fluctuations on ECCD in ITER and TCV,” EPJ Web of Conf. 32, 01016 (2012)] suggest that density fluctuations in the edge region of a tokamak plasma can cause broadening of the ECH deposition profile. In this paper, a GUI tool is presented which is used for analysis of ECH deposition as a first step towards looking for this broadening, which could explain effects seen in previous DIII-D ECH transport studies [K.W. Gentle “Electron energy transport inferences from modulated electron cyclotron heating in DIII-D,” Phys. Plasmas 13, 012311 (2006)]. By applying an FFT to the T 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

  10. Non-linear instability of DIII-D to error fields

    International Nuclear Information System (INIS)

    La Haye, R.J.; Scoville, J.T.

    1991-10-01

    Otherwise stable DIII-D discharges can become nonlinearly unstable to locked modes and disrupt when subjected to resonant m = 2, n = 1 error field caused by irregular poloidal field coils, i.e. intrinsic field errors. Instability is observed in DIII-D when the magnitude of the radial component of the m = 2, n = 1 error field with respect to the toroidal field is B r21 /B T of about 1.7 x 10 -4 . The locked modes triggered by an external error field are aligned with the static error field and the plasma fluid rotation ceases as a result of the growth of the mode. The triggered locked modes are the precursors of the subsequent plasma disruption. The use of an ''n = 1 coil'' to partially cancel intrinsic errors, or to increase them, results in a significantly expanded, or reduced, stable operating parameter space. Precise error field measurements have allowed the design of an improved correction coil for DIII-D, the ''C-coil'', which could further cancel error fields and help to avoid disruptive locked modes. 6 refs., 4 figs

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

    International Nuclear Information System (INIS)

    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

  12. Symplectic homoclinic tangles of the ideal separatrix of the DIII-D from type I ELMs

    Science.gov (United States)

    Punjabi, Alkesh; Ali, Halima

    2012-10-01

    The ideal separatrix of the divertor tokamaks is a degenerate manifold where both the stable and unstable manifolds coincide. Non-axisymmetric magnetic perturbations remove the degeneracy; and split the separatrix manifold. This creates an extremely complex topological structure, called homoclinic tangles. The unstable manifold intersects the stable manifold and creates alternating inner and outer lobes at successive homoclinic points. The Hamiltonian system must preserve the symplectic topological invariance, and this controls the size and radial extent of the lobes. Very recently, lobes near the X-point have been experimentally observed in MAST [A. Kirk et al, PRL 108, 255003 (2012)]. We have used the DIII-D map [A. Punjabi, NF 49, 115020 (2009)] to calculate symplectic homoclinic tangles of the ideal separatrix of the DIII-D from the type I ELMs represented by the peeling-ballooning modes (m,n)=(30,10)+(40,10). The DIII-D map is symplectic, accurate, and is in natural canonical coordinates which are invertible to physical coordinates [A. Punjabi and H. Ali, POP 15, 122502 (2008)]. To our knowledge, we are the first to symplectically calculate these tangles in physical space. Homoclinic tangles of separatrix can cause radial displacement of mobile passing electrons and create sheared radial electric fields and currents, resulting in radial flows, drifts, differential spinning, and reduction in turbulence, and other effects. This work is supported by the grants DE-FG02-01ER54624 and DE-FG02-04ER54793.

  13. Comparison of discharges with core transport barriers on DIII-D and JET

    International Nuclear Information System (INIS)

    Luce, T.C.; Alper, B.; Challis, C.D.

    1997-07-01

    The basic phenomenology of discharges with core transport barriers is the same for DIII-D and JET. The limitations on performance in both cases are well described by MHD stability calculations. Since the discharge behavior of the two machines is so similar, it seems reasonable to apply a simple parameterization of fusion performance which describes well the best performance discharges on DIII-D. The highest fusion performance shot on JET has Q DD = 3.1 10 -3 at 3.2 MA. Scaling from the highest Q DD DIII-D single-null discharge would predict Q DD = 4.2 10 -3 for JET. Raising the plasma current to 4.0 MA would increase the projection to 6.6 10 -3 . Realization of such performance would require significant effort to develop lower q plasmas with an H-mode edge. Because the performance is so closely tied to the current profile, this class of discharges also shows significant potential for steady state if current profile control can be demonstrated

  14. Progress towards a predictive model for pedestal height in DIII-D

    International Nuclear Information System (INIS)

    Groebner, R.J.; Leonard, A.W.; Snyder, P.B.; Osborne, T.H.; Petty, C.C.; Maggi, C.F.; Fenstermacher, M.E.; Owen, L.W.

    2009-01-01

    Recent DIII-D pedestal studies provide improved characterization of pedestal scaling for comparison with models. A new pedestal model accurately predicts the maximum achieved pedestal width and height in type I ELMing discharges over a large range of DIII-D operational space, including ITER demonstration discharges. The model is a combination of the peeling-ballooning theory for the MHD stability limits on the pedestal with a simple pedestal width scaling in which the width is proportional to the square root of the pedestal poloidal beta. Width scalings based on the ion toroidal or poloidal gyroradius are much poorer descriptions of DIII-D data. A mass scaling experiment in H and D provides support for a poloidal beta scaling and is not consistent with an ion poloidal gyroradius scaling. Studies of pedestal evolution during the inter-ELM cycle provide evidence that both the pedestal width and height increase during pedestal buildup. Model studies with a 1D kinetic neutrals calculation show that the temporal increase in density width cannot be explained in terms of increased neutral penetration depth. These studies show a correlation of pedestal width with both the square root of the pedestal poloidal beta and the square root of the pedestal ion temperature during the pedestal buildup.

  15. Data Analysis Software Tools for Enhanced Collaboration at the DIII-D National Fusion Facility

    International Nuclear Information System (INIS)

    Schachter, J.; Peng, Q.; Schissel, D.P.

    1999-01-01

    Data analysis at the DIII-D National Fusion Facility is simplified by the use of two software packages in analysis codes. The first is GAP1otObj, an IDL-based object-oriented library used in visualization tools for dynamic plotting. GAPlotObj gives users the ability to manipulate graphs directly through mouse and keyboard-driven commands. The second software package is MDSplus, which is used at DIED as a central repository for analyzed data. GAPlotObj and MDSplus reduce the effort required for a collaborator to become familiar with the DIII-D analysis environment by providing uniform interfaces for data display and retrieval. Two visualization tools at DIII-D that benefit from them are ReviewPlus and EFITviewer. ReviewPlus is capable of displaying interactive 2D and 3D graphs of raw, analyzed, and simulation code data. EFITviewer is used to display results from the EFIT analysis code together with kinetic profiles and machine geometry. Both bring new possibilities for data exploration to the user, and are able to plot data from any fusion research site with an MDSplus data server

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

    Energy Technology Data Exchange (ETDEWEB)

    Brookman, M. W., E-mail: brookmanmw@fusion.gat.com; Austin, M. E. [Institute for Fusion Studies, University of Texas at Austin, MS 13-505, 3483 Dunhill St, San Diego, CA 92121-1200 (United States); Petty, C. C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States)

    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. Modeling of Synergy Between 4th and 6th Harmonic Absorptions of Fast Waves on Injected Beams in DIII-D Tokamak

    International Nuclear Information System (INIS)

    Choi, M.; Pinsker, R. I.; Chan, V. S.; Muscatello, C. M.; Jaeger, E. F.

    2011-01-01

    In recent moderate to high harmonic fast wave heating and current drive experiments in DIII-D, a synergy effect was observed when the 6 th harmonic 90 MHz fast wave power is applied to the plasma preheated by neutral beams and the 4 th harmonic 60 MHz fast wave. In this paper, we investigate how the synergy can occur using ORBIT-RF coupled with AORSA. Preliminary simulations suggest that damping of 4 th harmonic FW on beam ions accelerates them above the injection energy, which may allow significant damping of 6 th harmonic FW on beam ion tails to produce synergy.

  18. Momentum Confinement at Low Torque

    International Nuclear Information System (INIS)

    Solomon, W.M.; Burrell, K.H.; deGrassie, J.S.; Budny, R.; Groebner, R.J.; Heidbrink, W.W.; Kinsey, J.E.; Kramer, G.J.; Makowski, M.A.; Mikkelsen, D.; Nazikian, R.; Petty, C.C.; Politzer, P.A.; Scott, S.D.; Van Zeeland, M.A.; Zarnstorff, M.C.

    2007-01-01

    Momentum confinement was investigated on DIII-D as a function of applied neutral beam torque at constant normalized β N , by varying the mix of co (parallel to the plasma current) and counter neutral beams. Under balanced neutral beam injection (i.e. zero total torque to the plasma), the plasma maintains a significant rotation in the co-direction. This 'intrinsic' rotation can be modeled as being due to an offset in the applied torque (i.e. an 'anomalous torque'). This anomalous torque appears to have a magnitude comparable to one co-neutral beam source. The presence of such an anomalous torque source must be taken into account to obtain meaningful quantities describing momentum transport, such as the global momentum confinement time and local diffusivities. Studies of the mechanical angular momentum in ELMing H-mode plasmas with elevated q min show that the momentum confinement time improves as the torque is reduced. In hybrid plasmas, the opposite effect is observed, namely that momentum confinement improves at high torque/rotation. The relative importance of E x B shearing between the two is modeled using GLF23 and may suggest a possible explanation.

  19. Divertor plasma physics experiments on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Allen, S.L.; Evans, T.E.

    1996-10-01

    In this paper we present an overview of the results and conclusions of our most recent divertor physics and development work. Using an array of new divertor diagnostics we have measured the plasma parameters over the entire divertor volume and gained new insights into several divertor physics issues. We present direct experimental evidence for momentum loss along the field lines, large heat convection, and copious volume recombination during detachment. These observations are supported by improved UEDGE modeling incorporating impurity radiation. We have demonstrated divertor exhaust enrichment of neon and argon by action of a forced scrape off layer (SOL) flow and demonstrated divertor pumping as a substitute for conventional wall conditioning. We have observed a divertor radiation zone with a parallel extent that is an order of magnitude larger than that estimated from a 1-D conduction limited model of plasma at coronal equilibrium. Using density profile control by divertor pumping and pellet injection we have attained H-mode confinement at densities above the Greenwald limit. Erosion rates of several candidate ITER plasma facing materials are measured and compared with predictions of a numerical model

  20. Role of E x B Shear and Magnetic Shear in the Formation of Transport Barriers in DIII-D

    International Nuclear Information System (INIS)

    Burrell, K.H.

    2005-01-01

    Development of the E x B shear stabilization model to explain the formation of transport barriers in magnetic confinement devices is a major achievement of fusion research. This concept has the universality needed to explain the H-mode edge transport barriers seen in limiter and divertor tokamaks, stellarators, and mirror machines; the broader edge transport barrier seen in VH-mode plasmas; and the core transport barriers formed in tokamaks with low or negative magnetic shear. These examples of confinement improvement are of considerable physical interest; it is not often that a system self-organizes to reduce transport when an additional source of free energy is applied to it. The transport decrease associated with E x B velocity shear is also of great practical benefit to fusion research. The fundamental physics involved in transport reduction is the effect of E x B shear on the growth, radial extent, and phase correlation of turbulent eddies in the plasma. The same basic transport reduction process can be operational in various portions of the plasma because there are a number of ways to change the radial electric field E r . An important theme in this area is the synergistic effect of E x B velocity shear and magnetic shear. Although the E x B velocity shear appears to have an effect on broader classes of microturbulence, magnetic shear can mitigate some potentially harmful effects of E x B velocity shear and facilitate turbulence stabilization. The experimental results on DIII-D and other devices are generally consistent with the basic theoretical models

  1. Observation of Dust in DIII-D Divertor and SOL by Visible Imaging

    International Nuclear Information System (INIS)

    Rudakov, D L; West, W P; Groth, M; Yu, J H; Wong, C C; Boedo, J A; Brooks, N H; Evans, T E; Fenstermacher, M E; Hollmann, E M; Hyatt, A W; Lasnier, C J; McLean, A G; Moyer, R A; Pigarov, A; Smirnov, R; Solomon, W M; Watkins, J G

    2007-01-01

    Dust is commonly found in fusion devices. Though generally of no concern in the present day machines, dust may pose serious safety and operational concerns for ITER. Micron-size dust usually dominates the samples collected from tokamaks. During a plasma discharge micron-size dust particles can become highly mobile and travel over distances of a few meters. Once inside the plasma, dust particles heat up to over 3000 K and emit thermal radiation that can be detected by visible imaging techniques. Observations of naturally occurring and artificially introduced dusts have been performed in DIII-D divertor and scrape-off layer (SOL) using standard frame rate CMOS cameras, a gated-intensified CID camera, and a fast-framing CMOS camera. In the first 2-3 plasma discharges after a vent with personnel entry inside the vacuum vessel ('dirty vent') dust levels were quite high with thousands of particles observed in each discharge. Individual particles moving at velocities of up to a few hundred m/s and breakup of larger particles into pieces were observed. After about 15 discharges dust was virtually gone during the stationary portion of a discharge, and appeared at much reduced levels during the plasma initiation and termination phases. After a few days of plasma operations (about 70 discharges) dust levels were further reduced to just a few observed events per discharge except in discharges with current disruptions that produced significant amounts of dust. An injection of a few milligram of micron-size (6 micron median diameter) carbon dust into a high-power lower single-null ELMing H-mode discharge with strike points swept across the lower divertor floor was performed. A significant increase of the core carbon radiation was observed for about 250 ms after the injection, as the total radiated power increased twofold. Dust particles from the injection were observed by the fast framing camera in the outboard SOL near the midplane. The amount of dust observed by the fast

  2. Comparison of heat flux measurement techniques during the DIII-D metal ring campaign

    Science.gov (United States)

    Barton, J. L.; Nygren, R. E.; Unterberg, E. A.; Watkins, J. G.; Makowski, M. A.; Moser, A.; Rudakov, D. L.; Buchenauer, D.

    2017-12-01

    The heat fluxes expected in the ITER divertor raise concerns about the damage tolerances of tungsten, especially due to thermal transients caused by edge localized modes (ELMs) as well as frequent temperature cycling from high to low extremes. Therefore we are motivated to understand the heat flux conditions that can cause not only enhanced erosion but also bulk thermo-mechanical damage to a tungsten divertor. For the metal ring campaign in DIII-D, tungsten-coated TZM tile inserts were installed making two toroidal arrays of metal tile inserts in the lower divertor. This study examines the deposited heat flux on these rings with embedded thermocouples (TCs) sampling at 10 kHz and compares them to Langmuir probe (LP) and infrared thermography (IRTV) heat flux measurements. We see agreement of the TC, LP, and IRTV data within 20% of the heat flux averaged over the entire discharge, and that all three diagnostics suggest parallel heat flux at the OSP location increases linearly with input heating power. The TC and LP heat flux time traces during the discharge trend together during large changes to the average heat flux. By subtracting the LP measured inter-ELM heat flux from TC data, using a rectangular ELM energy pulse shape, and taking the relative size and duration of each ELM from {{D}}α measurements, we extract the ELM heat fluxes from TC data. This over-estimates the IRTV measured ELM heat fluxes by a factor of 1.9, and could be due to the simplicity of the TC heat flux model and the assumed ELM energy pulse shape. ELM heat fluxes deposited on the inserts are used to model tungsten erosion in this campaign. These TC ELM heat flux estimates are used in addition to IRTV, especially in cases where the IRTV view to the metal ring is obstructed. We observe that some metal inserts were deformed due to exposed leading edges. The thermal conditions on these inserts are investigated with the thermal modeling code ABAQUS using our heat flux measurements when these edges

  3. Progress in high gain inertial confinement fusion

    International Nuclear Information System (INIS)

    Sun Jingwen

    2001-01-01

    The author reviews the progress in laboratory high gain inertial confinement fusion (ICF), including ICF capsule physics, high-energy-density science, inertial fusion energy, the National Ignition Facility (NIF) and its design of ignition targets and the peta watt laser breakthrough. High power laser, particle beam, and pulsed power facilities around the world have established the new laboratory field of high-energy- density plasma physics and have furthered development of inertial fusion. New capabilities such as those provided by high-brightness peta watt lasers have enabled the study of matter feasible in conditions previously unachievable on earth. Science and technology developed in inertial fusion research have found near-term commercial use and have enabled steady progress toward the goal of fusion ignition and high gain in the laboratory, and have opened up new fields of study for the 21 st century

  4. Predictions of the near edge transport shortfall in DIII-D L-mode plasmas using the trapped gyro-Landau-fluid model

    Energy Technology Data Exchange (ETDEWEB)

    Kinsey, J. E. [CompX, P.O. Box 2672, Del Mar, California 92014 (United States); Staebler, G. M.; Candy, J.; Petty, C. C.; Waltz, R. E. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Rhodes, T. L. [Physics Department and PSTI, University of California, Los Angeles, California 90095 (United States)

    2015-01-15

    Previous studies of DIII-D L-mode plasmas have shown that a transport shortfall exists in that our current models of turbulent transport can significantly underestimate the energy transport in the near edge region. In this paper, the Trapped Gyro-Landau-Fluid (TGLF) drift wave transport model is used to simulate the near edge transport in a DIII-D L-mode experiment designed to explore the impact of varying the safety factor on the shortfall. We find that the shortfall systematically increases with increasing safety factor and is more pronounced for the electrons than for the ions. Within the shortfall dataset, a single high current case has been found where no transport shortfall is predicted. Reduced neutral beam injection power has been identified as the key parameter separating this discharge from other discharges exhibiting a shortfall. Further analysis shows that the energy transport in the L-mode near edge region is not stiff according to TGLF. Unlike the H-mode core region, the predicted temperature profiles are relatively more responsive to changes in auxiliary heating power. In testing the fidelity of TGLF for the near edge region, we find that a recalibration of the collision model is warranted. A recalibration improves agreement between TGLF and nonlinear gyrokinetic simulations performed using the GYRO code with electron-ion collisions. The recalibration only slightly impacts the predicted shortfall.

  5. Development and validation of a critical gradient energetic particle driven Alfven eigenmode transport model for DIII-D tilted neutral beam experiments

    Science.gov (United States)

    Waltz, R. E.; Bass, E. M.; Heidbrink, W. W.; VanZeeland, M. A.

    2015-11-01

    Recent experiments with the DIII-D tilted neutral beam injection (NBI) varying the beam energetic particle (EP) source profiles have provided strong evidence that unstable Alfven eigenmodes (AE) drive stiff EP transport at a critical EP density gradient [Heidbrink et al 2013 Nucl. Fusion 53 093006]. Here the critical gradient is identified by the local AE growth rate being equal to the local ITG/TEM growth rate at the same low toroidal mode number. The growth rates are taken from the gyrokinetic code GYRO. Simulation show that the slowing down beam-like EP distribution has a slightly lower critical gradient than the Maxwellian. The ALPHA EP density transport code [Waltz and Bass 2014 Nucl. Fusion 54 104006], used to validate the model, combines the low-n stiff EP critical density gradient AE mid-core transport with the Angioni et al (2009 Nucl. Fusion 49 055013) energy independent high-n ITG/TEM density transport model controling the central core EP density profile. For the on-axis NBI heated DIII-D shot 146102, while the net loss to the edge is small, about half the birth fast ions are transported from the central core r/a  <  0.5 and the central density is about half the slowing down density. These results are in good agreement with experimental fast ion pressure profiles inferred from MSE constrained EFIT equilibria.

  6. Increased power delivery from the DIII-D neutral beam injection system

    International Nuclear Information System (INIS)

    Colleraine, A.P.; Callis, R.W.; Hong, R.M.; Kellman, D.H.; Kim, J.; Langhorn, A.R.; Lee, R.; Phillips, J.C.; Wight, J.J.

    1989-12-01

    The neutral beam system installed on the DIII-D tokamak employs eight 80 kV Long Pulse Sources (LPS) mounted on four beamlines and was originally designed to deliver a nominal 12 MW of H degree power to a plasma for pulses of up to 5 sec duration. Lawrence Berkeley Laboratory designed the LPS for the US Fusion Program to fill the requirements of both the DIII-D and the TFTR machines. Essentially all source components are of a common design; the DIII-D version is therefore conservative in its rated parameters. Recently a neutron shield has been constructed around the torus hall allowing D degree injection to become routine. Because deuterium beams have a better neutralization efficiency, the nominal power delivery per source has been measured to be approximately 2 MW (for a total of 16 MW) without any modifications. However, by reoptimizing the voltage gradients in the source, the perveance can be increased without degrading the optics. A change of gradient grid voltage from 0.83 V accel to 0.79 V accel raises the perveance from 2.5 to 3.0 μPerv with a corresponding gain in beam power of about 20%. The arc power required also must be increased to the range of 100 to 120 kW but this is well within the design limits of the LPS. Further studies of our systems are now underway to assess the possibilities of raising V accel above 80 kV. An additional gain in power is possible by this technique. 6 refs., 6 figs

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

    International Nuclear Information System (INIS)

    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. Upgrade of DIII-D toroidal magnetic field power supply controls

    International Nuclear Information System (INIS)

    Petrach, P.M.; Rouleau, A.R.; McNulty, R.D.; Patrick, D.B.; Walin, J.L.

    1993-11-01

    The toroidal magnetic field power supply for the DIII-D tokamak is of the 12 pulse line commutated variety. It consists of four individual modules and a main system control cabinet which are combined to deliver 127,000 A and 1000 V to the toroidal field coil. The modules are connected in a series-parallel configuration but can be run alone or two at a time as well. Normally on DIII-D experiments, the series-parallel configuration is required. The original design provided each individual module with its own voltage and current control loop and a main control loop. A problem with this design was that the individual control loops would cause a current sharing imbalance in the parallel modules if the calibrated loops drifted by the slightest amount. It was determined that individual control loops were not needed and a single phase lock firing circuit was employed in the system cabinet with fiber optic links to the modules for gate drive signals. Since all four modules have to be on line for DIII-D to operate, a problem in any of the five E ampersand I control loops resulted in the supply, and, therefore, the tokamak, being idled. By reducing the number of control loops to one, the sharing problem was eliminated, as well as 4 out of 5 potential control failures. The original supply employed relay logic for sequence control and fault monitoring. There were over 130 relays in each module plus an additional 100 in the system cabinet. The combination of the number of relays with the required interconnecting wiring, the age of the supply, the vibrations of the cabinets and the harsh environment, resulted in a continuously escalating number of phantom, and often intermittent, faults. The fault and sequence logic relays were replaced by a new Programmable Logic Controller (PLC). All existing interconnect wire was removed and replaced with multiconductor cables that connect directly from fault sensors and input devices to the PLC

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Fabrication and installation of the DIII-D radiative divertor structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.

    1997-11-01

    Phase 1A of the Radiative Divertor Program (RDP) is now installed in the DIII-D tokamak located at General Atomics. This hardware was added to enhance both the Divertor and Advanced Tokamak research elements of the DIII-D program. This installation consists of a divertor baffle enveloping a cryocondensation pump at the upper outer divertor ta