WorldWideScience

Sample records for diii-d boundary plasma

  1. Particle transport in DIII-D plasmas

    Science.gov (United States)

    Kress, Peter; Mordijck, Saskia

    2017-10-01

    By analyzing the plasma opacity and density evolution during the ELM cycle in DIII-D H-mode plasmas in which the amount of gas fueling was altered, we find evidence for an inward particle pinch at the plasma edge which seems to become more pronounced at higher density. Furthermore, at the plasma edge we find a correlation between the pedestal density and opacity, which measures neutral penetration depth. The changes in edge opacity during an ELM cycle were calculated by using a detailed time history of measured plasma profiles. At the same time, the density evolution during an ELM cycle was investigated. We find that if the edge density increases through an increase in gas fueling, then opacity increases and neutral fueling penetration depth decreases. We also find that density at the top of the pedestal recovers faster following an ELM when the overall density level is higher, leading to a hollow profile inside of the pedestal top. All these results indicate that there must be an inward particle pinch in the pedestal which will be crucial in the fueling of future burning plasma devices. Supported by US DOE DE-SC0007880, DIII-D Grant Number DE-FC02-04ER54698.

  2. 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)

  3. 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

  4. 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

  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. DIII-D Edge Plasma, Disruptions, and Radiative Processes. Final Report

    International Nuclear Information System (INIS)

    Boedo, J. A.; Luckhardt, S.C.; Moyer, R. A.

    2001-01-01

    The scientific goal of the UCSD-DIII-D Collaboration during this period was to understand the coupling of the core plasma to the plasma-facing components through the plasma boundary (edge and scrape-off layer). To achieve this goal, UCSD scientists studied the transport of particles, momentum, energy, and radiation from the plasma core to the plasma-facing components under normal (e.g., L-mode, H-mode, and ELMs), and off-normal (e.g., disruptions) operating conditions

  7. DIII-D Edge Plasma, Disruptions, and Radiative Processes. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Boedo, J. A.; Luckhardt, S.C.; Moyer, R. A.

    2001-01-01

    The scientific goal of the UCSD-DIII-D Collaboration during this period was to understand the coupling of the core plasma to the plasma-facing components through the plasma boundary (edge and scrape-off layer). To achieve this goal, UCSD scientists studied the transport of particles, momentum, energy, and radiation from the plasma core to the plasma-facing components under normal (e.g., L-mode, H-mode, and ELMs), and off-normal (e.g., disruptions) operating conditions.

  8. 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

  9. 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.

  10. Enhanced scrape-off layer plasma in DIII-D double-null discharges

    International Nuclear Information System (INIS)

    Watkins, J.G.; Jong, R.A.; Moyer, R.A.

    1994-07-01

    In this paper, the authors examine a denser and broader scrape-off layer (SOL) plasma, first seen in VH mode, in the DIII-D tokamak. The enhanced SOL appears in many types of double-null (DN) discharges and is not a property of VH-mode only. The DN enhanced SOL density and temperature profiles exhibit a 5--6 cm broad profile outside the separatrix. For DN and single-null (SN) boundary geometry with similar core plasma conditions, the enhanced SOL is only observed in high triangularity discharges. The origin of the enhanced SOL is, however, not yet understood

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  12. 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

  13. 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

  14. Effects of Resonanat Magnetic Perturbations in the DIII-D Edge Plasma

    International Nuclear Information System (INIS)

    Boedo, J.A.; Rudakov, D.L.; McKee, G.R.; Joseph, I.; Reiser, D.; Evans, T.E.; Moyer, R.A.; Watkins, J.G.; Allen, S.L.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Holland, C.; Hollmann, E.M.; Lasnier, C.J.; Leonard, A.W.; Schaffer, M.J.; Tynan, G.R.; West, W.P.; Zeng, L.

    2007-01-01

    Resonant magnetic perturbations (RMPs) are applied to the boundary of DIII-D with a variety of global effects such as edge localized mode (ELM) suppression and global density increase/decrease. How the applied perturbations affect the transport and the plasma edge stability and thus suppress the ELMs, are among the fundamental questions to be answered because of the high heat load created by the ELMs on the plasma facing components. We present fast probe measurements of the effects of applying RMPs to: (1) low power (ohmic) and (2) H-mode DIII-D discharges. In the low power discharges, the effect of islands is clearly seen in the edge plasma as structures in the profiles and changes in the fluctuations as far as 4 cm inside the separatrix. These observations compare well to calculations using 3D field mapping codes, indicating that the island structures modulate the edge parameters and transport. On the high power discharges, measurements of probes and other diagnostics (such as BES) are made at various points in the edge and changes in the profiles and fluctuations are compared. We find that fluctuations can be affected (enhanced or reduced) in narrow (1-2 cm) regions in the pedestal and in the scrape-off layer. The changes in the profiles and fluctuations are dependent on the structure of the applied fields that can be varied in both intensity and mode number.

  15. 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

  16. 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)

  17. Simulations of a DIII-D plasma disruption with the NIMROD code

    International Nuclear Information System (INIS)

    Kruger, S.E.; Schnack, D.D.

    2005-01-01

    To investigate the dynamics of the disruption of DIII-D discharge number 87009, two types of initial-value simulations with the NIMROD code were performed to investigate different characteristics. In the first set of simulations, a conducting wall was placed on the last close flux surface, and an equilibrium that was close to the ideal-MHD marginal stability point was created. A heating source was added to the pressure equation to drive the plasma equilibrium through an ideal-MHD instability point. Excellent agreement with analytic theory was obtained. To investigate how the stored energy is deposited on the wall, free-boundary simulations were performed with an ideal-MHD unstable equilibria. The unstable modes grow until the magnetic islands overlap and the magnetic field is stochastic over a large part of the plasma domain. The rapid stochastization of the field allows the plasma to lose two thirds of its internal energy in approximately 200 microseconds in qualitative agreement with the experiment. The deposition of thermal energy on the wall is localized poloidally and toroidally on the wall due to helically-localized temperature gradients and the rapid parallel heat conduction which carries this heat flux to the wall. (author)

  18. 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

  19. 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)

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. DIII-D data for modeling the scrape-off-layer plasma

    Energy Technology Data Exchange (ETDEWEB)

    Hill, D.N.; Buchenauer, D.; Carlstrom, T.N.; Ferron, J.; Resink, M.

    1989-12-04

    We are in the process of assembling a database of edge and divertor plasma parameters suitable for use in benchmarking tious 2D models of the scrape-off- layer (SOL) plasma. Also, we are using the Braams B2 code to derive transport coefficients for the edge plssma. In parallel, work is starting on an upgrade to the B2 code that includes padlel current flow and EXB drifts. These efforts are directed at increasing the confidence level of models of the tokamak edge plasma so that we can predict the effect of planned upgrades to DIII-D (e.g., the Advanced Divertor Program) and the performance of next generation machines such as CIT or ITER, where initial design studies show that plasma conditions at the divertor targets can have a large impact on the lifetime and cost of the machine. This report summarizes our recent progress in characterizing the DIII-D SOL plasma and in modeling these data with the the B2 code. Section I contains a brief description of the diagnostics available for characterizing the SOL plasma. In Section II we present our measurements of the SOL parameters for H-mode plasmas. This includes data showing how the divertor plasma parameters (n{sub e}(r), T{sub e}(r), and Q(r)) vary from ohmic to L-mode to H-mode, and power balance for quasi-stationary H-mode plasmas. Section III covers divertor-target heat-flux asymmetries for double and single null operation with forward and reversed toroidal field. In Section IV we show the scaling of L-mode parameters with neutral beam power, and Section V concludes with a summary of the results obtained from the Braams B2 SOL simulation code.

  5. Survey and Cleaning of Metal Contamination in Graphite Plasma-Facing Tiles in DIII-D

    Science.gov (United States)

    Chrobak, C. P.; Chamberlain, F.; Lee, R. L.; Holtrop, K. L.; Taylor, P. L.; Jackson, G. L.; Wall, D.; Buchenauer, D. A.; Mills, B. E.

    2012-10-01

    During the DIII-D FY11 and FY12 campaigns, relatively high levels of high Z metallic core plasma impurities impeded high performance plasma operation. Observations made during a vessel entry revealed potential sources of the increased metals, including: copper and Inconel splatter from a probe head damaged by runaway electrons, partial melting of a neutral beam molybdenum shield plate, and exposed metals on the Fast Wave antenna Faraday shields. Portable beta-backscattering and x-ray fluorescence diagnostics were used to map the areal density of metals deposited on the graphite plasma-facing tiles around the vessel. Tile surfaces with deposits exceeding 7x10^16 metal atoms/cm^2 were sanded in place or grit blasted outside of the vessel to remove impurities. The distribution of metals before and after resurfacing and the effectiveness of the tile resurfacing techniques on subsequent plasmas will be presented.

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

    International Nuclear Information System (INIS)

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

    2004-03-01

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

  8. Correlation among plasma rotation, magnetic configurations and improved confinement regimes on the DIII-D Tokamak

    International Nuclear Information System (INIS)

    Alladio, F.; Micozzi, P.

    1995-12-01

    A correlation has been established between the improvement of the energy confinement time observed in some plasma regimes on the DIII-D tokamak (VH modes and shear reversed discharges) and a geometrical characteristic of the plasma column: the Pfirsch-Schluter-like factor, which multiplies the moment of inertia of the magnetic configuration. Such a quantity is generated by the compression that the flux tubes suffer going from the external to the internal part of the torus. Therefore the configurations in which the module of the total magnetic field is more constant upon the magnetic surfaces (near omnigeneous configurations) show a lower value of the moment of inertia. The geometric parameter of Pfirsch-Schluter determines the transient and steady state behaviour of the plasma rotation under the assumption that the anomalous parallel viscosity is greater that the neoclassical one. In this way, also the profile of the part of the radial electric field (and his absolute value) is influenced by the magnetic configuration. The radial electric field, or, at least, his radial derivative, is invoked by many authors as a principal factor in reducing the turbulence (and so the anomalous transport) in magnetically confined plasmas. In particular, DIII-D machine, the highly elongated and triangular plasma discharges that evolve toward the VH-mode show a lower value of the Pfirsch-Schluter quantity and a higher level of the radial electric field; also the shear reversed profiles tend to lower 1+2q 2 in the central region of the plasma column, driving towards very high values of the electric field within the reversal region

  9. 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

  10. C transport studies in L-mode divertor plasmas on DIII-D.

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, N. (General Atomics, San Diego, CA); Nagy A. (Princeton Plasma Physics Laboratory, Princeton NJ); McLean, A. G. (University of Toronto Institute for Aerospace Studies, Toronto, Canada); Whyte, D. G. (University of Wisconsin, Madison, WI); Rudakov, D. L. (University of California, San Diego, La Jolla, CA); Bozek, A. (General Atomics, San Diego, CA); Allen, S. L. (Lawrence Livermore National Laboratory, Livermore, CA); West, W.P. (General Atomics, San Diego, CA); Stangeby, P. C. (University of Toronto Institute for Aerospace Studies, Toronto, Canada); Wampler, William R.; Matthews, G. F. (Euratom/UKAEA Fusion Association, Culham Science Center, Abingdon, UK); Phillips, V. (FZJ Julich GmbH/Euratom Institut fur Plasmaphysik, Julich, Germany); Ellis, R. (Lawrence Livermore National Laboratory, Livermore, CA)

    2004-06-01

    {sup 13}CH{sub 4} was injected with a toroidally-symmetric gas system into 22 identical lower-single-null L-mode discharges on DIII-D. The injection level was adjusted so that it did not significantly perturb the core or divertor plasmas, with a duration of {approx}3 s on each shot, for a total of {approx}300 T L of injected particles. The plasma shape remained very constant; the divertor strike points were controlled to {approx}1 cm at the divertor plate. At the beginning of the subsequent machine vent, 29 carbon tiles were removed for nuclear reaction analysis of {sup 13}C content to determine regions of carbon deposition. It was found that only the tiles inboard of the inner strike point had appreciable {sup 13}C above background. Visible spectroscopy measurements of the carbon injection and comparisons with modeling are consistent with carbon transport by means of scrape-off layer flow.

  11. {sup 13}C transport studies in L-mode divertor plasmas on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L. [Lawrence Livermore National Laboratory, P.O. Box 808, 7000 East Avenue, L-637, Livermore, CA 94550 (United States)]. E-mail: allens@fusion.gat.com; Wampler, W.R. [Sandia National Laboratories, Albuquerque, NM 87185-1129 (United States); McLean, A.G. [University of Toronto Institute for Aerospace Studies, Toronto, Canada MH3 5T6 (Canada); Whyte, D.G. [University of Wisconsin, Madison, WI 53706 (United States); West, W.P. [General Atomics, San Diego, CA 92186-5608 (United States); Stangeby, P.C. [University of Toronto Institute for Aerospace Studies, Toronto, Canada MH3 5T6 (Canada); Brooks, N.H. [General Atomics, San Diego, CA 92186-5608 (United States); Rudakov, D.L. [University of California, San Diego, La Jolla, CA 92093-0417 (United States); Phillips, V. [FZJ Juelich GmbH/EURATOM Institut fuer Plasmaphysik, TEC, D-52425 Juelich (Germany); Rubel, M. [Alfven Laboratory, Royal Institute of Technology, Association EURATOM-VR, Stockholm (Sweden); Matthews, G.F. [EURATOM/UKAEA Fusion Association, Culham Science Center, 0X14 3DB Abingdon (United Kingdom); Nagy, A. [Princeton Plasma Physics Laboratory, Princeton NJ 08543-0451 (United States); Ellis, R. [Lawrence Livermore National Laboratory, P.O. Box 808, 7000 East Avenue, L-637, Livermore, CA 94550 (United States); Bozek, A.S. [General Atomics, San Diego, CA 92186-5608 (United States)

    2005-03-01

    {sup 13}CH{sub 4} was injected with a toroidally-symmetric gas system into 22 identical lower-single-null L-mode discharges on DIII-D. The injection level was adjusted so that it did not significantly perturb the core or divertor plasmas, with a duration of {approx}3 s on each shot, for a total of {approx}300 T L of injected particles. The plasma shape remained very constant; the divertor strike points were controlled to {approx}1 cm at the divertor plate. At the beginning of the subsequent machine vent, 29 carbon tiles were removed for nuclear reaction analysis of {sup 13}C content to determine regions of carbon deposition. It was found that only the tiles inboard of the inner strike point had appreciable {sup 13}C above background. Visible spectroscopy measurements of the carbon injection and comparisons with modeling are consistent with carbon transport by means of scrape-off layer flow.

  12. Co-toroidal plasma rotation with electron cyclotron power in DIII-D

    International Nuclear Information System (INIS)

    Grassie, J.S. de; Baker, D.R.; Luce, T.C.; Petty, C.C.; Prater, R.; Brennan, D.

    2001-01-01

    RF electron heating and current drive in DIII-D are observed to typically reduce the core toroidal rotation velocity and core ion temperature when added to target discharges with rotation established by neutral beam heating. Two cases are noted here in which electron cyclotron heating and current drive are observed to increase co-toroidal rotation in different discharge regimes. In the first case electron cyclotron current drive (ECCD) is used to stabilize a 3/2 neoclassical tearing mode (NTM) and the stabilization is accompanied by an increase in rotation, ion temperature and plasma beta. In the second case electron cyclotron heating (ECH) added to a nominally Ohmic target discharge results in an increase in the co-toroidal rotation

  13. Suppression of erosion in the DIII-D divertor with detached plasmas

    International Nuclear Information System (INIS)

    Wampler, William R.; Bastasz, Robert J.; Whyte, D.G.; Wong, C.P.C.; West, W.P.

    2000-01-01

    The ability to withstand disruptions makes carbon-based materials attractive for use as plasma-facing components in divertors. However, such materials suffer high erosion rates during attached plasma operation which, in high power long pulse machines, would give short component lifetimes and high tritium inventories. The authors present results from recent experiments in DIII-D, in which the Divertor Materials Evaluation System (DiMES) was used to examine erosion and deposition during short exposures to well defined plasma conditions. These studies show that during operation with detached plasmas, produced by gas injection, net erosion is suppressed everywhere in the divertor. Net deposition of carbon with deuterium was observed at the inner and outer strikepoints and in the private-flux region between strikepoints. For these low temperature plasmas (T e < 2eV), physical sputtering is eliminated. These results show that with detached plasmas, the location of carbon net erosion and the carbon impurity source, probably lies outside the divertor. Physical or chemical sputtering by charge-exchange neutrals or ions in the main plasma chamber is a probable source of carbon under these plasma conditions

  14. 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.

  15. 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

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

    NARCIS (Netherlands)

    Tobias, B.; Yu, L.; Domier, C.W.; N C Luhmann 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

  17. Fast wave current drive in H mode plasmas on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Grassie, J.S. de; Baity, F.W.

    1999-01-01

    Current driven by fast Alfven waves is measured in H mode and VH mode plasmas on the DIII-D tokamak for the first time. Analysis of the poloidal flux evolution shows that the fast wave current drive profile is centrally peaked but sometimes broader than theoretically expected. Although the measured current drive efficiency is in agreement with theory for plasmas with infrequent ELMs, the current drive efficiency is an order of magnitude too low for plasmas with rapid ELMs. Power modulation experiments show that the reduction in current drive with increasing ELM frequency is due to a reduction in the fraction of centrally absorbed fast wave power. The absorption and current drive are weakest when the electron density outside the plasma separatrix is raised above the fast wave cut-off density by the ELMs, possibly allowing an edge loss mechanism to dissipate the fast wave power since the cut-off density is a barrier for fast waves leaving the plasma. (author)

  18. 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.

  19. 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

  20. Advances in Understanding and Control of Plasma Rotation on DIII-D

    Science.gov (United States)

    Grierson, B. A.; Logan, N.; Haskey, S.; Ashourvan, A.; Ernst, D.; Chrystal, C.; Degrassie, J.; Boedo, J.; Tala, T.; Salmi, A.

    2017-10-01

    Momentum transport experiments on DIII-D have advanced our understanding of the origin of core and edge rotation by showing that (1) core rotation in low-torque electron-heated ITER-like plasmas displays hollowing driven by turbulence in the absence of MHD, (2) intrinsic rotation in torque-free electron-heated plasmas follows the favorable rho* and nu* scalings as previously found in intrinsic torque experiments using NBI, (3) the edge plasma rotation can be controlled through shaping of triangularity and X-point radius, and (4) rotation and density profiles have separate dependencies on the applied 3D field spectra. These advances inform strategies to avoid low torque disruptions by tailoring turbulent modes that minimize rotation hollowing, and provide confidence in dimensionless scaling of intrinsic torque and rotation to ITER. The triangularity and X-point position provide important new actuators on the rotation beyond neutral beam injection that are available for any diverted tokamak including ITER. The separate spectral dependencies of the momentum and density explain how quiescent braking as well as edge isolated ELM control are possible even in machines with limited toroidal harmonic EFC coils. Work supported by US DOE under DE-AC02-09CH11466,DE-FC02-04ER54698.

  1. Displaying DIII-D plasma data using DEC's X window system

    International Nuclear Information System (INIS)

    Greene, K.L.

    1992-01-01

    This paper reports on the DIII-D tokamak program funded by the Department of Energy, which carries out plasma physics and fusion energy research experiments. The machine began operation in February 1986; at that time, approximately 7 Mbytes of data was collected for each shot. Since that time, the shot size has steadily increased to over 50 Mbytes with the average shot size between 35 and 45 Mbytes. Shots are fired every 12 to 15 minutes and last approximately 5 to 10 seconds. Between 30 and 40 shots are fired each day when plasma experiments are scheduled. In 1990, both programs were converted from User Interface Services (UIS) routines, which are part of the MicroVMS workstation graphics software, to DEC's X Window System using the DECWindows window manager. These modifications were required because of a move by Digital Equipment Corporation (DEC) to support Xwindows and phase out UIS. Due to the nature and purpose of each program, MFITD needed only simple graphics conversion while MFITPLAY was completely rewritten. The DECWindows version of MFITPLAY offers a number of improvements, such as a more intuitive user interface

  2. 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

  3. 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)

  4. 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

  5. DIII-D Research Operations annual report to the US Department of Energy, October 1, 1990--September 30, 1991. Magnetic Fusion Research Program

    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.

  6. 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

  7. Plasma-Surface Interaction Studies on DIII-D and Their Implications for Next-Step Fusion Experiments

    International Nuclear Information System (INIS)

    Whyte, D.G.

    2005-01-01

    Unique diagnostic and access features of the DIII-D tokamak, including a sample exposure system, have been used to carry out controlled and well-diagnosed plasma-surface interactions (PSI) experiments. An important contribution of the experiments has been the ability to link a given plasma exposure condition to a measured response of the plasma-facing surface and to thus understand the interaction. This has allowed for benchmarking certain aspects of erosion models, particularly near-surface particle transport. DIII-D has empirically quantified some of the PSI effects that will limit the operation availability and lifetime of future fusion devices, namely, net erosion limiting divertor plate lifetime and hydrogenic fuel retention in deposit layers. Cold divertor plasmas obtained with detachment can suppress net carbon divertor erosion, but many low-temperature divertor PSI phenomena remain poorly understood: nondivertor erosion sources, long-range particle transport, global erosion/deposition patterns, the enhancement of carbon erosion with neon impurity seeding, the sputtered carbon velocity distribution, and the apparent suppression of carbon chemical erosion in detachment. Long-term particle and energy fluences have reduced the chemical erosion yield of lower-divertor tiles. Plasma-caused modification of a material's erosion properties, including material mixing, will occur quickly and be important in long-pulse fusion devices, making prediction of PSI difficult in future devices

  8. Analysis and correction of intrinsic non-axisymmetric magnetic fields in high-β DIII-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)

  9. Dependence of Helicon Antenna Loading on the Antenna/Plasma Gap and n|| in DIII-D Experiments

    Science.gov (United States)

    Pinsker, R. I.; Moeller, C. P.

    2017-10-01

    A comprehensive set of measurements of the plasma loading of a 12-element antenna array, designed to launch helicon waves (i.e., very-high-harmonic fast waves), were performed on DIII-D in 2016. The antenna, operated in the 466 - 486 MHz band, is prototypical of a wider array for a 1-MW-level experiment planned for 2018-9. The dependence of the antenna loading on antenna/plasma gap is of great practical significance, as the gap must be kept greater than a minimum distance to suppress deleterious plasma-material interactions, while the loading must be high enough to retain good efficiency of power transfer to the plasma. While the loading in all examined plasma regimes, including both limited and diverted L-mode discharges and H-mode discharges, decayed exponentially with increasing gap in agreement with simple theory, the characteristic decay length was in all cases larger than expected, motivating the development of a more realistic model. Furthermore, the characteristic decay length did not depend on the launched n||, though the absolute level of loading at a given gap increased as |n||| was decreased from 4 to 2. After the antenna was removed from DIII-D, measurements of the loading produced by a 100 Ω/sq resistive film were carried out on the bench. Both the antenna/film gap and n|| were scanned varied and the results compared with calculations done with the QuickWave FDTD electromagnetics solver. Very good agreement was found in this case. Work supported by the US DOE under DE-FC02-04ER54698.

  10. Comparison of wall/divertor deuterium retention and plasma fueling requirements on the DIII-D, TdeV, and ASDEX-upgrade tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R. [Oak Ridge Associated Universities, TN (United States); Terreault, B. [Inst. National de la Recherche Scientifique, Varennes, Quebec (Canada); Haas, G. [Max Planck Inst. fuer Plasmaphysik, Garching (Germany)] [and others

    1996-06-01

    The authors present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV, and ASDEX-Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without Helium Glow Discharge Cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by > 1,250 Torr-1 (equivalent to 150 {times} plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by < 35 Torr-{ell} (equivalent to 50{times} plasma particle content) plasma inventories at a maximum fueling efficiency {approximately} 1. Data from the ASDEX-Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D.

  11. 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

  12. 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

  13. 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

  14. 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.

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

    Science.gov (United States)

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

    2015-11-01

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

  16. 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

  17. 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.

  18. Implications of wall recycling and carbon source locations on core plasma fueling and impurity content in DIII-D

    International Nuclear Information System (INIS)

    Groth, M.; Porter, G.D.; Fenstermacher, M.E.; Lasnier, C.J.; Meyer, W.M.; Rensink, M.E.; Wolf, N.S.; Boedo, J.A.; Moyer, R.A.; Rudakov, D.L.; Brooks, N.H.; Groebner, R.J.; Petrie, T.W.; Owen, L.W.; Wang, G.; Zeng, L.; Watkins, J.G.

    2005-01-01

    Measurement and modeling of the 2-D poloidal D α intensity distribution in DIII-D low and medium density L-mode and ELMy H-mode plasmas indicate that hydrogen neutrals predominantly fuel the core from the divertor X-point region. The 2-D distribution of neutral deuterium and low-charge-state carbon were measured in the divertor and the high-field side midplane scrape-off layer (SOL) using tangentially viewing cameras. The emission in the high-field SOL at the equatorial plane was found to be three to four orders of magnitude lower than at the strike points in the divertor, suggesting a strong divertor particle source. Modeling using the UEDGE/DEGAS codes predicted the poloidal fueling distribution to be dependent on the direction of the ion Bx∇B drift. In plasmas with the Bx∇B drift into the divertor stronger fueling from the inner divertor than from the outer is predicted, due to a lower-temperature and higher-density plasma in the inner leg. UEDGE simulations with carbon produced by both physical and chemical sputtering at the divertor plates and walls only are in agreement with a large set of diagnostic data. The simulations indicate flow reversal in the inner divertor that augments the leakage of carbon ions from the divertor into the core. (author)

  19. CHANGES IN EDGE AND SCRAPE-OFF LAYER PLASMA BEHAVIOE DUE TO VAARIATION IN MAGNETIC BALANCE IN DIII-D

    International Nuclear Information System (INIS)

    PETRIE, T.W.; WATKINS, J.G.; BAYLOR, L.R.; BROOKS, N.H.; FENSTERMACHER, M.E.; HYATT, A.W.; JACKSON, G.L.; LASNIER, C.J.; LEONARD, A.W.; PIGAROV, A.YU.; RENSINK, M.E.; ROGNLIEN, T.D.; SCHAFFER, M.J.; WOLF, N.S.

    2002-01-01

    Changes in the divertor magnetic balance in DIII-D H-mode plasmas affects core, edge, and divertor plasma behavior. Both the pedestal density n e,PED and plasma stored energy W T were sensitive to changes in magnetic balance near the double-null (DN) configuration, e.g., both decreased 20%-30% when the DN shifted to a slightly unbalanced DN, where the B x (del)B drift direction pointed away from the main X-point. Recycling at each of the four divertor targets was sensitive to changes in magnetic balance and the B x (del)B drift direction. The poloidal distribution of the recycling in DN is in qualitative agreement with the predictions of UEDGE modeling with particle drifts included. The particle flux at the inner divertor target is shown to be much more sensitive to magnetic balance than the particle flux at the outer divertor target near the DN shape. These results suggest possible advantages and drawbacks for balanced DN operation

  20. Discovery of Stationary Operation of Quiescent H-mode Plasmas with Net-Zero NBI Torque and High Energy Confinement on DIII-D

    Science.gov (United States)

    Burrell, Keith

    2015-11-01

    Experiments this summer in DIII-D have used edge turbulence control to achieve stationary, high confinement operation without Edge Localized Mode (ELM) instabilities and with no 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 outstanding tokamak performance, well above the H98 international tokamak energy confinement scaling (H98 =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 ExB 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. 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 betan =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. Stationary operation with improved pedestal conditions is highly significant for future burning plasma devices, since operation without ELMs at low rotation and good confinement is key for fusion energy production. Supported by the US DOE under DE-FC02-04ER54698.

  1. Evaluation of an improved atomic data basis for carbon in UEDGE emission modeling for L-mode plasmas in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Muñoz Burgos, J.M., E-mail: munozj@fusion.gat.com [Oak Ridge Institute for Science and Education, Oak Ridge, TN 37831-0117 (United States); Leonard, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Loch, S.D.; Ballance, C.P. [Auburn University, Auburn, AL 36849 (United States)

    2013-07-15

    New scaled carbon atomic electron-impact excitation data is utilized to evaluate comparisons between experimental measurements and fluid emission modeling of detached plasmas at DIII-D. The C I and C II modeled emission lines for 909.8 and 514.7 nm were overestimated by a factor of 10–20 than observed experimentally for the inner leg, while the outer leg was within a factor of 2. Due to higher modeled emissions, a previous study using the UEDGE code predicted that a higher amount of carbon was required to achieve a detached outboard divertor plasma in L-mode at DIII-D. The line emission predicted by using the new scaled carbon data yields closer results when compared against experiment. We also compare modeling and measurements of D{sub α} emission from neutral deuterium against predictions from newly calculated R-Matrix with pseudostates data available at the ADAS database.

  2. 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)

  3. 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%)

  4. Carbon influx in He and D plasmas in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    West, W.P.; Brooks, N.H. [General Atomics, San Diego, CA (United States); Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States)] [and others

    1998-07-01

    Differences in the carbon behavior between He and D plasmas during VH-mode, L-mode and L-mode with excess gas puffing are reported and inferences on the importance of the various carbon sources during these modes of operation are discussed. During a VH-mode phase, VUV and visible charge exchange spectroscopy indicates that for both He and D operation the carbon behavior is very similar. In the edge plasma, carbon build up is quite rapid, and the carbon influx represents a large fraction of the total plasma density increase until the termination of the VH phase. During cold divertor operation induced by puffing the primary fueling gas, D and He discharges show a difference in the carbon behavior. The core carbon density is seen to be approximately constant during a D discharge as it transitions from an attached to a cold divertor. However in a He discharge, the core carbon density disappears soon after the cold divertor transition. Arguments are made that the primary carbon source in the ELM free H-mode period is physical sputtering by ion impact at the divertor strike point. In L-mode, both attached and cold divertor, the primary source is from the divertor region and two possibilities for this source are chemical sputtering or charge neutral sputtering. Existing data supports charge exchange neutrals as dominant.

  5. Effect of Divertor Shaping on Divertor Plasma Behavior on DIII-D

    Science.gov (United States)

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

    2012-10-01

    Recent experiments examined the dependence of divertor density (nTAR), temperature (TTAR), and heat flux at the outer divertor separatrix target on changes in the divertor separatrix geometry. The responses of nTAR and TTAR to changes in the parallel connection length in the scrape-off layer (SOL) (L||) are consistent with the predictions of the Two Point Model (TPM). However, nTAR and TTAR display a more complex response to changes in the radial location of the outer divertor strike point (RTAR) than expected based on the TPM. SOLPS transport analysis indicates that small differences in divertor geometry can change neutral trapping sufficient to explain differences between experiment and TPM predictions. The response of the core and divertor plasmas to changes in L|| and RTAR, under both radiating and non-radiating divertor conditions, will be shown.

  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. Data-driven robust control of the plasma rotational transform profile and normalized beta dynamics for advanced tokamak scenarios in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Shi, W.; Wehner, W.P.; Barton, J.E.; Boyer, M.D. [Mechanical Engineering and Mechanics, Lehigh University, Bethlehem, PA 18015 (United States); Schuster, E., E-mail: schuster@lehigh.edu [Mechanical Engineering and Mechanics, Lehigh University, Bethlehem, PA 18015 (United States); Moreau, D. [CEA, IRFM, F-13018 St Paul lez Durance (France); Walker, M.L.; Ferron, J.R.; Luce, T.C.; Humphreys, D.A.; Penaflor, B.G.; Johnson, R.D. [General Atomics, San Diego, CA 92121 (United States)

    2017-04-15

    A control-oriented, two-timescale, linear, dynamic, response model of the rotational transform ι profile and the normalized beta β{sub N} is proposed based on experimental data from the DIII-D tokamak. Dedicated system-identification experiments without feedback control have been carried out to generate data for the development of this model. The data-driven dynamic model, which is both device-specific and scenario-specific, represents the response of the ι profile and β{sub N} to the electric field due to induction as well as to the heating and current drive (H&CD) systems during the flat-top phase of an H-mode discharge in DIII-D. The control goal is to use both induction and the H&CD systems to locally regulate the plasma ι profile and β{sub N} around particular target values close to the reference state used for system identification. A singular value decomposition (SVD) of the plasma model at steady state is carried out to decouple the system and identify the most relevant control channels. A mixed-sensitivity robust control design problem is formulated based on the dynamic model to synthesize a stabilizing feedback controller without input constraints that minimizes the reference tracking error and rejects external disturbances with minimal control energy. The feedback controller is then augmented with an anti-windup compensator, which keeps the given controller well-behaved in the presence of magnitude constraints in the actuators and leaves the nominal closed-loop system unmodified when no saturation is present. The proposed controller represents one of the first feedback profile controllers integrating magnetic and kinetic variables ever implemented and experimentally tested in DIII-D. The preliminary experimental results presented in this work, although limited in number and constrained by actuator problems and design limitations, as it will be reported, show good progress towards routine current profile control in DIII-D and leave valuable lessons

  8. Impact of screening of resonant magnetic perturbations in three dimensional edge plasma transport simulations for DIII-D

    Czech Academy of Sciences Publication Activity Database

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

    2012-01-01

    Roč. 19, č. 5 (2012), 052507-052507 ISSN 1070-664X R&D Projects: GA ČR GAP205/11/2341 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * TEXTOR * divertors * plasma boundary layers * plasma density * plasma magnetohydrodynamics * plasma simulation * plasma temperature * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.376, year: 2012 http://pop.aip.org/resource/1/phpaen/v19/i5/p052507_s1

  9. Detection of plasma stability on DIII-D, using the experimentally extracted plasma transfer function based on 3D MHD spectroscopy

    Science.gov (United States)

    Wang, Zhirui; Logan, Nikolas; Park, Jongkyu; Menard, Jonathan; Nazikian, Raffi; Munaretto, Stefano; Liu, Yueqiang; Hanson, Jeremy

    2017-10-01

    Three-dimensional (3D) magnetohydrodynamic (MHD) spectroscopy is successfully applied to extract the plasma transfer function from DIII-D experiments. The method uses upper and lower internal coils to perform scans of frequency and poloidal mode spectrum, and measure the corresponding n =1 plasma response on 3D magnetic sensors. The transfer function is extracted, based on Padé approximation, by fitting the measured signals on different sensors simultaneously. The experimental transfer function not only points out the multi-mode plasma response but also shows the number of dominant modes and the contribution of each mode to the plasma response. The extracted damping rate of the least stable mode can be a new index indicating plasma stability quantitatively. This method has the potential to optimize ELM suppression and monitor the plasma stability in future fusion reactors. Results and analysis of 3D MHD spectroscopy experiments will be presented. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698, DE-AC02-09CH11466 and DE-FG02-04ER54761.

  10. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lohr, J. [ed.] [General Atomics, San Diego, CA (United States)

    1995-07-01

    The DIII-D tokamak research program is managed by General Atomics (GA) for the US Department of Energy (DOE). Major program participants include GA, Lawrence Livermore National Laboratory (LLNL), Oak Ridge National Laboratory (ORNL), and the University of California together with several other national laboratories and universities. The DIII-D is a moderate sized tokamak with great flexibility and extremely capable subsystems. The primary goal of the DIII-D tokamak research program is to provide data for development of a conceptual physics blueprint for a commercially attractive fusion power plant. In so doing, the DIII-D program provides physics and technology R&D output to aid the International Thermonuclear Experimental Reactor (ITER) and the Princeton Tokamak Physics Experiment (TPX) projects. Specific DIII-D objectives include the achievement of steady-state plasma current as well as the demonstration of techniques for radio frequency heating, divertor heat removal, particle exhaust and tokamak plasma control. The DIII-D program is addressing these objectives in an integrated fashion in plasmas with high beta and with high confinement. The long-range plan is organized with two principal elements, the development of an advanced divertor and the development of advanced tokamak concepts. These two elements have a common goal: an improved demonstration reactor (DEMO) with lower cost and smaller size than present DEMO concepts. In order to prepare for this long-range development, in FY94 the DIII-D research program concentrated on three major areas: Divertor and Boundary Physics, Advanced Tokamak studies, and Tokamak Physics.

  12. 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.

  13. 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.

  14. Scenario development for high β p low torque plasma with q min above 2 and large-radius internal transport barrier in DIII-D

    Science.gov (United States)

    Ding, S.; Xu, G. S.; Wang, Q.; Solomon, W. M.; Zhao, Y.; Gong, X.; Garofalo, A. M.; Holcomb, C. T.; McKee, G.; Yan, Z.; Wang, H. Q.; Qian, J.; Wan, B. N.

    2017-02-01

    A recent experiment on DIII-D, which was conducted by the joint research team from DIII-D and EAST, has extended the previous high {β\\text{p}} , high q min regime, which has been tested in the 2013 DIII-D/EAST joint experiment, to inductive operation at higher plasma current ({{I}\\text{p}}=0.8 MA) and significantly higher normalized fusion performance (G={{H}89}{β\\text{N}}/q952=0.16 ). The experiment aims at exploring high performance scenario with {{q}\\text{min}}>2 and reduced torque for long pulse operation, which can be potentially extrapolated to EAST. The effort was largely motivated by the interest in developing a feasible scenario for long-pulse high performance operation with low torque on EAST. Very high confinement, H 89  =  3.5 or {{H}98,\\text{y2}}=2.1 with {β\\text{N}}∼ 3.0 , has been achieved transiently in this experiment together with {{q}\\text{min}}>2 and reduced NBI torque (3∼ 5 N m). The excellent confinement is associated with the spontaneous formation of an internal transport barrier (ITB) in plasmas with {{I}\\text{p}}=0.8 MA at large minor radius (normalized ρ ∼ 0.7 ) in all channels (n e, T e, T i, {{V}φ} , especially strong in the T e channel). Fluctuation measurements show a significant reduction in the fluctuation levels, including AE modes and broadband turbulence, at the location where an ITB forms. Linear gyrokinetic simulations also support the decrease of the growth rate of the most unstable mode during strong ITB formation. The simulation implies that strong suppression of turbulence and a positive feedback loop may be active in this process and is responsible for the spontaneous formation of large-radius ITB. In an unstable ITB phase, an ELM crash is observed to have a positive effect on transient formation of large-radius ITB. The formation of this kind of ITB is found to have a shielding (protecting) effect on the core plasma while isolating the perturbation due to ELM crash.

  15. Automated Fault Detection for DIII-D Tokamak Experiments

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

    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.

  16. Benchmarking UEDGE with DIII-D data

    Energy Technology Data Exchange (ETDEWEB)

    Porter, G.D.; Fenstermacher, M.; Rensink, M.E.; Rognlien, T.D. [Lawrence Livermore National Lab., CA (United States); Groebner, R.; Leonard, A. [General Atomics, La Jolla, CA (United States); Nguyen, Q. [Univ. of California, Berkeley, CA (United States)

    1993-09-01

    Comparisons between a 2-D fluid simulation of the scrape-off-layer (SOL) plasma of a diverted tokamak and experimental data from the DIII-D are shown. It is concluded that a simple diffusive model for perpendicular transport is consistent with the data. Discrepancies in the simulation suggest that impurity radiation may be playing a significant role in the experiment, and that further work is required to understand hydrogen recycling at the divertor.

  17. 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.

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

    Science.gov (United States)

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

    2016-12-01

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

  19. 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

  20. The investigation of structure, chemical composition, hydrogen isotope trapping and release processes in deposition layers on surfaces exposed to DIII-D divertor plasma

    International Nuclear Information System (INIS)

    Buzhinskij, O.I.; Opimach, I.V.; Barsuk, V.A.; Arkhipov, I.I.; Whyte, D.; Wampler, W.R.

    1998-05-01

    The exposure of ATG graphite sample to DIII-D divertor plasma was provided by the DiMES (Divertor Material Evaluation System) mechanism. The graphite sample arranged to receive the parallel heat flux on a small region of the surface was exposed to 600ms of outer strike point plasma. The sample was constructed to collect the eroded material directed downward into a trapping zone onto s Si disk collector. The average heat flux onto the graphite sample during the exposure was about 200W/cm 2 , and the parallel heat flux was about 10 KW/cm 2 . After the exposure the graphite sample and Si collector disk were analyzed using SEM, NRA, RBS, Auger spectroscopy. IR and Raman spectroscopy. The thermal desorption was studied also. The deposited coating on graphite sample is amorphous carbon layer. Just upstream of the high heat flux zone the redeposition layer has a globular structure. The deposition layer on Si disk is composed also from carbon but has a diamond-like structure. The areal density of C and D in the deposited layer on Si disk varied in poloidal and toroidal directions. The maximum D/C areal density ratio is about 0.23, maximum carbon density is about 3.8 x 10 18 cm -2 , maximum D area density is about 3 x 10 17 cm 2 . The thermal desorption spectrum had a peak at 1,250K

  1. 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

  2. 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

  3. Coupled two-dimensional edge-plasma and neutral gas modelling of the DIII-D scrape-off-layer

    International Nuclear Information System (INIS)

    Maingi, R.; Gilligan, J.; Hankins, O.; Rensink, M.; Owen, L.; Klepper, C.; Mioduszewski, P.

    1992-01-01

    This paper reports that in order to do consistent scrape-off-layer plasma and neutral transport calculations, the 2-D fluid code, B2 has been externally coupled to the neutral transport code, DEGAS, for Dlll-D. The coupling procedure is similar to recent simulations done for TFTR, Tore Supra, and ClT. An averaged source approach is utilized to allow convergence between the two codes. Initial comparison of plasma quantities between the coupled code set and the B2 code alone shows that a colder, denser plasma may exist at the divertor targets than predicted by the B2 code with its internal recycling model

  4. The fishbone instability in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Sager, G.

    1990-01-01

    Although most DIII-D plasmas are stable to the fishbone instability, fishbones are sometimes observed when β p approx. = 1.5 and n-bar e approx. = 5x10 13 cm -3 . These bursts are usually of minor significance operationally; however, under one condition, over 50% of the beam power was lost. The angle of beam injection has little effect on the virulence of the instability, suggesting that the fishbone instability in DIII-D is the ion diamagnetic branch of the internal kink. (author). 37 refs, 9 figs, 1 tab

  5. 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

  6. New Frontier Science Campaign on DIII-D launched in 2017

    Science.gov (United States)

    Koepke, M.; Buttery, R.; Carter, T.; Egedal, J.; Forest, C.; Fox, W.; Ji, H.; Howes, G.; Piovesan, P.; Sarff, J.; Skiff, F.; Spong, D.; DIII-D FSE Collaboration Collaboration

    2017-10-01

    The DIII-D Frontier Science Experiments initiative explores the potential to use the DIII-D tokamak facility to investigate questions of value beyond the usual fusion-energy science mission of DIII-D. The campaign is unique within DOE-SC-FES because the DIII-D tokamak supplied a multi-day-shot platform for non-fusion-energy-motivated research for the first time. All selected FSE campaign projects competed on the basis of potential intellectual impact and on the degree to which the ability to achieve success as a transformational advance relied on the capabilities of DIII-D. The motivation of the following FSE projects, as well as the selection process, will be summarized (1) Self-organization of Unstable Flux Ropes: Universal Structures in Space/Astrophysical Plasmas (2) Impact of Magnetic Perturbations on Turbulence: Zonal Flow Interactions and Saturation (3) Interaction of Alfvén/whistler fluctuations and Runaway Electrons (4) Self-consistent chaos in magnetic field dynamics These basic-plasma experiments, conducted in collaboration with the DIII-D team, were carried out during 5 shot days in FY2017. Additional days are earmarked in FY2018. Future studies with additional FSE-community members are envisioned. Opportunities exist to piggy back with DIII-D research A proper solicitation and peer review would be appropriate going forward if this activity on DIII-D continues Funding from U.S. DOE is gratefully acknowledged.

  7. Cross-calibrating Spatial Positions of Light-viewing Diagnostics using Plasma Edge Sweeps in DIII-D

    International Nuclear Information System (INIS)

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

    2003-01-01

    An experimental technique is presented that permits diagnostics viewing light from the plasma edge to be spatially calibrated relative to one another. By sweeping the plasma edge, each chord of each diagnostic sweeps out a portion of the light emission profile. A nonlinear least-squares fit to such data provides superior cross-calibration of diagnostics located at different toroidal locations compared with simple surveying. Another advantage of the technique is that it can be used to monitor the position of viewing chords during an experimental campaign to ensure that alignment does not change over time. Moreover, should such a change occur, the data can still be cross-calibrated and its usefulness retained

  8. Divertor erosion in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Whyte, D.G. [Univ. of California, San Diego, CA (United States); Bastasz, R.; Wampler, W.R. [Sandia National Labs., Albuquerque, NM (United States); Brooks, J.N. [Argonne National Lab., IL (United States); West, W.P.; Wong, C.P.C.; Buzhinskij, O.I. [General Atomics, San Diego, CA (United States); Opimach, I.V. [TRINITI Lab. (United States)

    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{sub e} > 40 eV) ELMing plasmas, and detached (T{sub e} < 2 eV) ELMing plasmas. For the attached cases, the erosion rates exceed 10 cm/exposure-year, even with incident heat flux < 1 MW/m{sup 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 {le} 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition ({approximately} 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 ({approximately} 50 MW/m{sup 2}) have very high net erosion rates at the OSP of an attached plasma ({approximately} 10 {micro}m/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.

  9. 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

  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. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    deGrassie, J.S.; Callis, R.; Lin-Liu, Y.R.; Moeller, C..; Petty, C.C.; Phelps, D.R.; Pinsker, R.I.; Remsen, D. (General Atomics, San Diego, CA (United States)); Baity, F.W.; Hoffman, D.J.; Taylor, D.J. (Oak Ridge National Lab., TN (United States)); Arnold, W.; Martin, S. (ANT-Nachrichtentechnik GmbH, Backnang (Germany))

    1992-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    deGrassie, J.S.; Callis, R.; Lin-Liu, Y.R.; Moeller, C..; Petty, C.C.; Phelps, D.R.; Pinsker, R.I.; Remsen, D. [General Atomics, San Diego, CA (United States); Baity, F.W.; Hoffman, D.J.; Taylor, D.J. [Oak Ridge National Lab., TN (United States); Arnold, W.; Martin, S. [ANT-Nachrichtentechnik GmbH, Backnang (Germany)

    1992-09-01

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

  14. Resistive Wall Mode Stabilization Studies at DIII-D

    International Nuclear Information System (INIS)

    Garofalo, A.M.

    2005-01-01

    The effort to understand the physics of the resistive wall mode (RWM) and develop methods to control this magnetohydrodynamic mode to allow achievement of higher pressure in advanced tokamak plasmas has been an example of successful multi-institutional collaboration at the DIII-D National Fusion Facility in San Diego, California. DIII-D research in this area has produced several advances and breakthroughs following a coordinated research plan involving a sequence of measurements, development of new analysis tools, and the installation of new diagnostic and feedback stabilization hardware: Suppression of the RWM by active magnetic feedback has been demonstrated using the DIII-D six-element error field correction coil, rotational stabilization of the RWM has been demonstrated and sustained for all values of the plasma pressure from the no-wall to the ideal-wall stability limits, improved RWM feedback stabilization has been shown using a new set of 12 internal control coils, and newly developed models of feedback have shown good agreement with the measurements. By so doing, the DIII-D work on RWM stabilization has become a cornerstone of the long-term advanced tokamak program and is having impact on the world fusion program. Presently both ITER and FIRE are including plans for RWM stabilization in their programs

  15. 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

  16. 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

  17. 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.

  18. Recent Progress in BOUT + + boundary plasma turbulence simulations

    Science.gov (United States)

    Xu, X. Q.; BOUT++ Team

    2017-10-01

    BOUT + + has been developed and applied for a range of problems that impact on boundary plasma turbulence and transport. A summary of simulation progress and results will be presented including, but not limited to: (1) Modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths; (2) Self-consistent calculation of the radial electric field with ion orbit loss mechanism; (3) Simulating the DIII-D and EAST grassy ELM regime; (4) Simulation comparison of EHO state and broadband MHD phase in near-zero torque QH-mode on DIII-D; (5) Simulation of the ELMs triggering by lithium pellet on EAST tokamak; (6) Ideal MHD Stability and Characteristics of Edge Localized Modes on CFETR Our latest transport module solves a set of transport equations with quasi-neutral constraint using vorticity formulation under the BOUT + + framework. This new capability enables BOUT + + team to simulate boundary plasma transport across the separatrix with self-consistent electric and magnetic drifts, ion orbit loss, and sheath boundary conditions in the scrape-off-layer. Preliminary results of the coupled turbulence and transport simulations will also presented. Prepared by LLNL under Contract DE-AC52-07NA27344.

  19. 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

  20. 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.

  1. 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

  2. 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.))

  3. 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

  4. 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

  5. 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

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

    Science.gov (United States)

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

    2008-11-01

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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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}.

  14. 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

  15. Development of inside launch reflectometer systems on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Doyle, E.J.; Rhodes, T.L.; Doane, J.L.; Peebles, W.A.

    1995-01-01

    Inside launch (high field side) reflectometry is necessary for routine core access on DIII-D and is a very attractive option for ITER and TPX. For high temperature reactor relevant plasmas, relativistic corrections for the left-hand cutoff (inside launched) are relatively small, whereas these substantially affect the location of the normal right hand cutoff. On DIII-D, core access using the right hand cutoff is restricted to low density plasmas, while O-mode polarization is incompatible with flat H-mode density profiles. Routine core access can, however, be obtained using the left-hand cutoff and inside launch. On ITER and TPX, outside launch X-mode systems need to operate at high frequency, ∼100--250 GHz, such that the left-hand cutoff, which is at much lower frequency, ∼0--75 GHz, is an attractive option for high density operation. In addition, inside launch reflectometry can be used to investigate plasma asymmetries; on DIII-D an inboard/outboard asymmetry in turbulence response is observed at the L-H transition. An upgraded inside launch reflectometer system was recently installed on DIII-D, specifically to investigate the use of the left hand X-mode cutoff and inside launch for profile and turbulence measurements in reactor relevant plasmas. This system will also provide routine core access for reflectometry on DIII-D

  16. Updated DIII-D experimental plan for FY-1989

    Energy Technology Data Exchange (ETDEWEB)

    Luxon, J.L. [ed.

    1989-08-01

    The program proposed here is designed to support and build toward the long-term plan put forward during 1987 for the DIII-D facility. This plan has as its ultimate goal developing sufficient understanding and predictive capability to enable the demonstration of a high beta plasma with non-inductively driven toroidal current. The early stages of this plan call for the optimization of the plasma configuration for good confinement at high beta while simultaneously developing the need rf power systems for current drive, profile control, and heating.

  17. 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

  18. 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

  19. DiMES divertor erosion experiments on DIII-D

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  20. VH-Mode discharges in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Greenfield, C.M.; Jackson, G.L.; Burrell, K.H.; DeBoo, J.C.; Lao, L.L.; Osborne, T.H.; Schissel, D.P.; Taylor, T.S. (General Atomics, San Diego, CA (United States)); Rettig, C. (California Univ., Los Angeles, CA (United States)); Winter, J. (Kernforschungsanlage Juelich GmbH (Germany))

    1992-05-01

    Introduction. A regime of very high confinement (VH-mode) has been observed in divertor discharges in DIII-D. The VH-mode, first seen following the initial boronization of the DIII-D vessel in 1991, exhibits total energy confinement a factor of 2.5 to 3.5 greater than that predicted by the ITER89-P L-mode scaling relation. Also, confinement of thermal energy alone is greater than 1.6 times that of the JET/DIII-D H-mode scaling and in many cases has exceeded twice that amount. VH-mode is observed during a long ({le}0.8 sec) ELM-free phase of the discharges. At the beginning of the ELM-free period, the plasma appears to be in H-mode, with confinement near that predicted by the JET/DIII-D scaling. In the usual H-mode, confinement is observed to decrease or remain constant over time. In the present discharges, confinement has been observed to remain nearly constant for up to hundreds of milliseconds, after which the behavior sharply deviates from H-mode as the confinement begins to increase over time. This increase in confinement continues until the occurrence of a beta- related ({beta}>2.8I/aB) global MHD event, which rapidly decreases the plasma stored energy with a temperature reduction across the entire profile. Magnetic measurements indicate that at least in some cases, this event includes both an internal n = 1 mode and a more localized high-n mode near the edge. After this event, the plasma relaxes into an ELMing H-mode phase. As a consequence of the boronization, the plasmas in these discharges are unusually clean, with very low radiated power. In previous H-mode discharges, the radiated power increased during the ELM-free, sometimes reaching levels comparable with the input power if the ELM-free period was long enough. Also, Z{sub eff}is constant or decreasing over the length of the discharge, with a central value of {approx}1. It is noted that most of the energy in these discharges is thermal energy, with {le}10% contained in fast ions.

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

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

  3. 3D Equilibrium Reconstructions in DIII-D

    Science.gov (United States)

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

    2013-10-01

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

  4. 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

  5. Motional stark effect upgrades on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Model-based current profile control at DIII-D

    International Nuclear Information System (INIS)

    Yongsheng Ou; Schuster, E.; Luce, T.; Ferron, J.; Walker, M.; Humphreys, D.

    2006-01-01

    There is consensus in the fusion community that control of the radial profiles of various plasma quantities (current, pressure, rotation, etc.) will be key to the optimization of burning plasma scenarios. It has been suggested, for instance, that global current profile control, eventually combined with pressure profile control, can be an effective mechanism for neoclassical tearing mode (NTM) control and avoidance. It has been also suggested that simultaneous real-time control of the current and pressure profiles could lead to the steady state sustainment of an internal transport barrier (ITB) and so to a stationary optimized plasma regime. A key goal in control of an advanced tokamak (AT) discharge is to maintain safety factor (q) and pressure profiles that are compatible with both MHD stability at high toroidal beta and a high fraction of the self-generated bootstrap current. This will enable high fusion gain and noninductive sustainment of 100% of the plasma current for steady-state operation. Active feedback control of the q profile evolution at DIII-D has been already demonstrated [J.R. Ferron, et al., '' Control of DIII-D Advanced Tokamak Discharges '', 32 nd EPS Conference on Plasma Physics, Tarragona, 27 June - 1 July 2005, ECA vol. 29C, p. 1,069 (2005)]. In this work we report progress towards enabling model-based active control of the current profile during both plasma current ramp-up and flattop phases. Initial results on modeling and simulation of the dynamic evolution of the poloidal flux profile are presented. Dynamic models will allow the exploitation of recent developments in the field of (nonlinear) control of distributed-parameter systems to solve present profile control problems in magnetic fusion energy. (author)

  7. 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)

  8. 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

  9. Gas Balance in Ohmic Discharges on DIII-D

    Science.gov (United States)

    West, W. P.; Brooks, N. H.; Leonard, A. W.; Whyte, D. G.; Lipschultz, B.; Watkins, J. G.; Groth, M.; Lasnier, C. J.; Fenstermacher, M. E.; Boedo, J. A.; Rudakov, D. L.; Unterberg, E. A.

    2008-11-01

    Wall retention of deuterium (D) fueling gas in ohmic discharges on DIII-D has been measured by operation in a closed system with no exhaust from the vacuum vessel. Vessel pressures after identical gas injection, with and without plasma operation, are compared. The ion flux to the divertor was measured with fixed Langmuir probes, and SOL plasma density and temperatures were measured with fast-stroke probes. Ten similar discharges with no in-vessel pumping were repeated, followed by three discharges with in-vesssel divetor cryopumps active then regenerated after each discharge. Preliminary analysis indicates the retained D in ohmic discharges is ˜90 (20)% of the injected gas in the un(pumped) discharges, whereas previous gas balance during cryopumped ELMing H-mode discharges indicated no retention. In both the pumped and unpumped ohmic cases, the retained D is ˜1% of the ion fluence to the wall.

  10. Kinetic simulations of scrape-off layer physics in the DIII-D tokamak

    OpenAIRE

    Churchill, R.M.; Canik, J.M.; Chang, C.S.; Hager, R.; Leonard, A.W.; Maingi, R.; Nazikian, R.; Stotler, D.P.

    2016-01-01

    Simulations using the fully kinetic code XGCa were undertaken to explore the impact of kinetic effects on scrape-off layer (SOL) physics in DIII-D H-mode plasmas. XGCa is a total-f, gyrokinetic code which self-consistently calculates the axisymmetric electrostatic potential and plasma dynamics, and includes modules for Monte Carlo neutral transport. Fluid simulations are normally used to simulate the SOL, due to its high collisionality. However, depending on plasma conditions, a number of dis...

  11. Engineering and design of a CO2 phase contrast interferometer system for DIII-D

    International Nuclear Information System (INIS)

    Phelps, R.D.; Coda, S.

    1994-11-01

    This report describes the development of a CO 2 laser interferometer system, the engineering, design and installation of the hardware, and the selection of materials specific to the requirements of a CO 2 laser diagnostic. A brief description of system operation is included. A phase contrast interferometer diagnostic has been designed and installed on the DIII-D tokamak to enhance studies of the physical characteristics of plasma turbulence, and specifically to analyze plasma density fluctuations in the boundary region of the plasma. A 20 watt CO 2 laser beam, operating at the 10.6 micron wavelength, is expanded to a diameter of 76 mm and directed through a series of mirrors which provide for entry of the beam into the vessel at a point 70 cm above the midplane at the 285 degree toroidal location. After being reflected from a mirror inside the vessel, the beam is directed downward so that it passes through the edge of the plasma immediately in front of a four-strap fast wave current drive rf antenna. The laser beam is then reflected by a second internal mirror and exits the vessel 70 cm below the midplane (also at 285 degrees) returning to an optical table through a final series of external steering mirrors

  12. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Porkolab, M. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Plasma Fusion Center

    1993-08-04

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

  14. 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

  15. Multi-field/-scale interactions of turbulence with neoclassical tearing modes and impact on plasma confinement in the DIII-D tokamak

    Science.gov (United States)

    Bardoczi, L.

    2016-10-01

    We present the first localized measurements of ITG scale temperature and density fluctuations and TEM scale density fluctuations modified by an m=2, n=1 magnetic island. These islands are formed by a Neoclassical Tearing Mode (NTM) deep in the core plasma at the q=2 surface. NTMs are important as they often degrade confinement and lead to disruption. This is the first experimental confirmation of a long-standing theory prediction of decreased local small-scale turbulence levels across large-scale magnetic islands. Our measurements capture a mean reduction of turbulence inside (and enhancement just outside) the island region during island evolution. Additionally, in the island saturated state, the fluctuations at the O-point are observed to be reduced compared to the X-point. These measurements allow the determination of the turbulence length scale at the island separatrix that is predicted to affect NTM stability. A novel, non-perturbative measurement technique finds reduced cross-field electron thermal diffusivity (by 1-2 orders of magnitude) at the O-point, consistent with the local turbulence reduction. Initial comparisons to the GENE non-linear gyrokinetic code are promising with GENE predicting the observed turbulence reduction inside the island and increase just outside the island and replicating the observed scaling with island size. These results allow the validation of gyrokinetic simulations modeling the interaction of multi-scale phenomena as well as have potential implications for improved NTM control. Supported by USDOE under DE-FG02-08ER54984, DE-FG02-08ER54999 and DE-FC02-04ER54698.

  16. 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.

  17. Improved confinement in highly powered high performance scenarios on DIII-D

    Science.gov (United States)

    Petrie, T. W.; Osborne, T.; Fenstermacher, M. E.; Ferron, J.; Groebner, R.; Grierson, B.; Holcomb, C.; Lasnier, C.; Leonard, A.; Luce, T.; Makowski, M.; Turco, F.; Solomon, W.; Victor, B.; Watkins, J.

    2017-08-01

    DIII-D has recently demonstrated improved energy confinement by injecting neutral deuterium gas into high performance near-double null divertor (DND) plasmas during high power operation. Representative parameters for these plasmas are: q 95  =  6, P IN up to 15 MW, H 98  =  1.4-1.8, and β N  =  2.5-4.0. The ion B   ×  \

  18. Fast wave current drive on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    deGrassie, J.S.; Petty, C.C.; Pinsker, R.I. [and others

    1995-07-01

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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  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. Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code

    International Nuclear Information System (INIS)

    Evans, T.E.; Mahdavi, M.A.; Sager, G.T.; West, W.P.; Fenstermacher, M.E.; Meyer, W.H.; Porter, G.D.

    1995-07-01

    A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DIII-D divertor is discussed. MCI simulation results are compared to experimental DIII-D carbon measurements

  10. 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

  11. Equilibrium reconstruction improvement via Kalman-filter-based vessel current estimation at DIII-D

    International Nuclear Information System (INIS)

    Yongsheng Ou; Schuster, E.; Walker, M.; Ferron, J.

    2006-01-01

    The efficient and safe operation of large fusion devices relies on accurate knowledge of many of the discharge parameters. Unfortunately, the values of several discharge parameters, such as plasma shape and current density distribution, are not directly measured. However, these values can be reconstructed from magnetic field and flux measurements. Equilibrium codes, such as EFIT, calculate the distributions of flux and toroidal current density over the plasma and surrounding vacuum region that best fit, in a least square sense, the external magnetic measurements, and that simultaneously satisfy the MHD equilibrium equation (Grad-Shafranov equation). Once the flux distribution is known, it is possible to reconstruct the plasma boundary for shape control purposes. The most general approach to the fitting problem treats all toroidal current sources as unknown values. Thus, in addition to the plasma toroidal current, the currents in the external poloidal field (PF) coils can be free parameters and, potentially, the induced currents in the vacuum vessel and support structures can be treated this way as well. There are direct measurements of the external PF coil currents, but these measurements have uncertainties that can be properly accounted for in the least squares fitting procedure by solving for the external currents using the measurements as constraints. A similar procedure could be followed for the vessel currents if they were measurable. Unfortunately, this is not usually the case and vessel currents are often neglected in the fitting procedure. The important effect of vessel or structure currents has been recognized in many plasma control applications. Kalman filtering theory is used in this work to optimally estimate the current in the tokamak vessel. With the ultimate goal of improving the equilibrium reconstruction for the DIII-D tokamak, the real-time version of the EFIT algorithm is modified to accept the estimated vessel currents. Furthermore, it will be

  12. Observation of suprathermal electrons during magnetic reconnection at the sawtooth instability in DIII-D TOKAMAK

    CERN Document Server

    Savrukhin, R V

    2002-01-01

    OAK A271 Observation of suprathermal electrons during magnetic reconnection at the sawtooth instability in DIII-D TOKAMAK. Intense bursts of x-ray and electron cyclotron emission are observed during sawtooth instabilities in high-temperature plasmas in the DIII-D tokamak. The bursts are initiated around the X-point of the m = 1, n = 1 magnetic island at the beginning of the sawtooth crash and are displaced to larger radii later during the temperature collapse. Reconstruction of the magnetic configuration using motional Stark effect (MSE) data and numerical simulations indicates that the bursts can be connected with suprathermal electrons (E sub r approx 30-40 keV) generated during reconnection of the magnetic field around the q = 1 surface.

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. Studies of the DIII-D disruption database using Machine Learning algorithms

    Science.gov (United States)

    Rea, Cristina; Granetz, Robert; Meneghini, Orso

    2017-10-01

    A Random Forests Machine Learning algorithm, trained on a large database of both disruptive and non-disruptive DIII-D discharges, predicts disruptive behavior in DIII-D with about 90% of accuracy. Several algorithms have been tested and Random Forests was found superior in performances for this particular task. Over 40 plasma parameters are included in the database, with data for each of the parameters taken from 500k time slices. We focused on a subset of non-dimensional plasma parameters, deemed to be good predictors based on physics considerations. Both binary (disruptive/non-disruptive) and multi-label (label based on the elapsed time before disruption) classification problems are investigated. The Random Forests algorithm provides insight on the available dataset by ranking the relative importance of the input features. It is found that q95 and Greenwald density fraction (n/nG) are the most relevant parameters for discriminating between DIII-D disruptive and non-disruptive discharges. A comparison with the Gradient Boosted Trees algorithm is shown and the first results coming from the application of regression algorithms are presented. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-SC0014264 and DE-FG02-95ER54309.

  19. 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

  20. 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

  1. Disruption mitigation using high-pressure noble gas injection on DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.

    2002-01-01

    High-pressure gas jet injection of neon and argon is shown to be a simple and robust method to mitigate the deleterious effects of disruptions on the DIII-D tokamak. The gas jet penetrates to the central plasma at its sonic velocity. The deposited species dissipates ∼100% of the plasma thermal energy by radiation and substantially reduces mechanical stresses on the vessel caused by poloidal halo currents. The gas jet species charge distribution can include >50% fraction neutral species which inhibits runaway electrons. The favorable scaling of this technique to burning fusion plasmas is discussed. (author)

  2. 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.

  3. 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.

  4. Tungsten erosion by unipolar arcing in DIII-D

    Science.gov (United States)

    Bykov, I.; Chrobak, C. P.; Abrams, T.; Rudakov, D. L.; Unterberg, E. A.; Wampler, W. R.; Hollmann, E. M.; Moyer, R. A.; Boedo, J. A.; Stahl, B.; Hinson, E. T.; Yu, J. H.; Lasnier, C. J.; Makowski, M.; McLean, A. G.

    2017-12-01

    Unipolar arcing was an important mechanism of metal surface erosion during the recently conducted Metal Rings Campaign in DIII-D when two toroidally continuous tile rings with 5 cm wide W-coated TZM inserts were installed in graphite tiles in the lower divertor, one on the floor and one on the shelf. Most of the arc damage occurred on the shelf ring. The total amount of W removed by arcing from the affected ˜4% of the shelf ring area was estimated ˜0.8 × 1021 at., about half of the total amount of W eroded and redeposited outside the inserts (1.8 ± 0.9)×1021 at. The rings were exposed for a total of ˜480 discharges, an equivalent of plasma time on W surfaces (with {{I}}{{p}}> 0.5 MA) ˜103 s. Arcing was monitored in situ with WI (400.9 nm) filtered camera and photomultipliers and showed that: (i) arcing only occurred during ELMs and disruptions, (ii) arcing rate was much lower on the floor than on the shelf ring, and (iii) arcing had a low cut off power flux density about 2 MW m-2. About half of arc tracks had large {10}\\circ pitch angle and probably were produced during disruptions. Such tracks were only found on the shelf. Moderate toroidal variation of the arc track density and W erosion with nearly n = 1 pattern has been measured.

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. 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.

  10. 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

  11. Developing Boundary/PMI Solutions for Next-Step Fusion Devices

    Science.gov (United States)

    Guo, H. Y.; Leonard, A. W.; Thomas, D. M.; Allen, S. L.; Hill, D. N.; Unterberg, Z.

    2014-10-01

    The path towards next-step fusion development requires increased emphasis on the boundary/plasma-material interface. The new DIII-D Boundary/Plasma-Material Interactions (PMI) Center has been established to address these critical issues on a timescale relevant to the design of FNSF, adopting the following transformational approaches: (1) Develop and test advanced divertor configurations on DIII-D compatible with core plasma high performance operational scenarios in FNSF; (2) Validate candidate reactor PFC materials at reactor-relevant temperatures in DIII-D high-performance plasmas, in collaboration with the broad material research/development community; (3) Integrate validated boundary-materials interface with high performance plasmas to provide viable boundary/PMI solutions for next-step fusion devices. This program leverages unique DIII-D capabilities, promotes synergistic programs within the broad PMI community, including linear material research facilities. It will also enable us to build a compelling bridge for the US research on long-pulse facilities. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344, DE-AC05-00OR2725.

  12. 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 target of DIII-D. The divertor baffle consists of two toroidally continuous Inconel 625 water-cooled rings and a toroidal array of discontinuous radiatively-cooled plates. The water-cooled rings are each comprised of four quadrants, mechanically formed, chem.-milled, and resistance and TIG welded Inconel 625 panels. The supports attaching the panels to the vessel wall are designed to accommodate the differential thermal expansion between the rings and vessel during bake and to react the electromagnetic loads induced during disruptions. They are made from either Inconel 625 or Inconel 718 depending on the stress levels predicted in Finite Element Analysis. Gas seals are designed to limit the leakage from the baffle chamber back to the core plasma to 2,500 ell/s and incorporate plasma sprayed alumina to minimize currents flowing through them. The bulk of the water-cooled ring fabrication was performed by a vendor, however, the final machining of penetrations in the conical ring for diagnostic access was performed in-house using a unique machining configuration. This configuration, and the machining of the diagnostic cutouts is described. Graphite tiles were machined from ATJ graphite to form a smooth plasma-facing surface. The installation of all divertor components required only four weeks

  13. ANOMALIES IN THE APPLIED MAGNETIC FIELDS ON DIII-D AND THEIR IMPLICATIONS FOR THE UNDERSTANDING OF STABILITY EXPERIMENTS

    International Nuclear Information System (INIS)

    LUXON, J.L; SCHAFFER, M.J; JACKSON, G.L; LEUER, J.A; NAGY, A; SCOVILLE, J.T; STRAIT, E.J

    2003-01-01

    Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments and resistive wall mode experiments done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made on DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported. Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all of the coils are one third of the values indicated from the stability experiments. These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks

  14. Anomalies in the applied magnetic fields in DIII-D and their implications for the understanding of stability experiments

    Energy Technology Data Exchange (ETDEWEB)

    Luxon, J. L.; Schaffer, M. J.; Jackson, G. L.; Leuer, J. A.; Nagy, A.; Scoville, J. T.; Strait, E. J.

    2003-12-01

    Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments and resistive wall mode experiments done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made on DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported. Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all of the coils are one third of the values indicated from the stability experiments. These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks.

  15. Design, Fabrication, and Installation of the Lower Divertor for DIII-D

    International Nuclear Information System (INIS)

    Anderson, P.M.; Murphy, C.J.; Reis, E.E.; Hu, Q.; Song, Y.; Yao, D.

    2006-01-01

    The geometry of the lower divertor of the DIII-D tokamak was modified to provide improved density control of the tokamak plasma during operation in a high-triangularity double null plasma. This divertor replaces the low triangularity Advanced Divertor in use since 1990. The design, analysis and fabrication were completed in 2005 and the installation was completed in March 2006. Plasma operations are planned for June 2006. The primary component of the lower divertor is a toroidally continuous flat cooling plate. Three rows of graphite tiles are mechanically attached to the plate to shield it from plasma impingement. The plate is water cooled for heat removal between shots and is heated to 350 o C with hot air and inductive current during vessel baking. The divertor ring is supported 100 mm from the vacuum vessel floor by two rows of 24 supports that must react the vertical loads due to halo currents. The space below the ring forms a pumping plenum between the floor strike point and the lower cryopump. The divertor plate was fabricated by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) in four 90 degree sectors from Type 316 stainless. Each sector consists of two plate halves with three machined coolant channels connected in parallel. Two plate halves are joined together by spot welds and perimeter TIG welds. During installation, the vacuum-tight 90 degree panel sectors were aligned and welded together inside the vessel forming a toroidally continuous ring. The water cooling/air bakeout lines connecting the 4 sectors into two 180 degree cooling circuits were then welded in place. The vacuum boundary for the cooling/air bakeout lines uses a reverse conflat design with the tubes welded into a modified, outward facing conflat flange. This design provides for copper gasket seal replacement without disturbing any welds. Plasma facing tile designs have been modified from previous designs to eliminate holes in high heat flux areas. Upgraded floor tiles

  16. Measurements of carbon and tungsten erosion/deposition in the DIII-D divertor

    International Nuclear Information System (INIS)

    Bastasz, R.; Wampler, W.R.; Cuthbertson, J.W.; Buchenauer, D.A.; Brooks, N.; Junge, R.; West, W.P.; Wong, C.P.C.

    1994-01-01

    Net erosion/deposition rates of carbon and tungsten were measured at the outer strike point of the divertor plasma on the floor of the DIII-D tokamak during deuterium H-mode operation at a peak power deposition of about 40 W/cm 2 . For carbon, net erosion rates of up to 4 nm/s were found. For a tungsten film, no appreciable erosion was detected. However, measurements of deposited tungsten on adjacent carbon surfaces indicated a net W erosion rate of 0.06 nm/s

  17. A 15 MeV proton dianostic for DIII-D

    International Nuclear Information System (INIS)

    Duong, Hau; Heidbrink, W.W.

    1990-10-01

    A 15 MeV proton diagnostic that is patterned after the ASDEX proton probe is presently being fabricated for the DIII-D tokamak. A bellows assembly inserts a silicon detector into the vacuum for plasma operation and retracts it for baking. The detector preamplifier is situated in a reentrant tube (at atmosphere) beside the detector; electrically, the whole assembly is referenced to vessel potential. Orbit calculations in realistic magnetic field geometries predict a proton detection efficiency of O(10 -7 ). The diagnostic will be used for burnup studies at high β and particle transport studies in the H-mode. 25 refs., 4 figs

  18. 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

  19. 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)

  20. 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.

  1. 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

  2. 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.

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

    Science.gov (United States)

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

    2017-10-01

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

  4. Experimental survey of the L-H transition conditions in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Carlstrom, T.N.; Gohil, P. [General Atomics, San Diego, CA (United States); Watkins, J.C. [Sandia National Labs., Albuquerque, NM (United States)] [and others

    1994-01-01

    We present the global analysis of a recent survey of the H-mode power threshold in DIII-D using D{sup o} {yields} D{sup +} NBI after boronization of the vacuum vessel. Single parameter scans of B{sub T}, I{sub 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{sub T} and n{sub e} but remains independent of I{sub p}. In double-null discharges, the power threshold is found to be approximately independent of both B{sub T} and n{sub 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.

  5. 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.

  6. 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

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

    Science.gov (United States)

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

    2017-10-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Zeng, L., E-mail: zeng@fusion.gat.com; Peebles, W. A.; Doyle, E. J.; Rhodes, T. L.; Crocker, N.; Nguyen, X.; Wannberg, C. W.; Wang, G. [Physics and Astronomy Department, University of California, Los Angeles, California 90095 (United States)

    2014-11-15

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. OEDGE modeling of DIII-D density scan discharges leading to detachment

    Science.gov (United States)

    Elder, J. D.; Stangeby, P. C.; Bray, B. D.; Brooks, N.; Leonard, A. W.; McLean, A. G.; Unterberg, E. A.; Watkins, J. G.

    2015-08-01

    The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for Dα, Dβ and Dγ for values of Te and ne characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to Te.

  11. OEDGE modeling of DIII-D density scan discharges leading to detachment

    Energy Technology Data Exchange (ETDEWEB)

    Elder, J. D. [Univ. of Toronto, ON (Canada); Stangeby, P. C. [Univ. of Toronto, ON (Canada); General Atomics, San Diego, CA (United States); Bray, B. D. [General Atomics, San Diego, CA (United States); Brooks, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Leonard, A. W. [General Atomics, San Diego, CA (United States); McLean, A. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Unterberg, Ezekial A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Watkins, J. G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-09-30

    Here, we study the OEDGE code that is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for Dα, Dβ and Dγ for values of Te and ne characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ~1 eV. Lastly, for the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to Te.

  12. OEDGE modeling of DIII-D density scan discharges leading to detachment

    Energy Technology Data Exchange (ETDEWEB)

    Elder, J.D., E-mail: david@starfire.utias.utoronto.ca [University of Toronto Institute for Aerospace Studies, Toronto M3H 5T6 (Canada); Stangeby, P.C. [University of Toronto Institute for Aerospace Studies, Toronto M3H 5T6 (Canada); General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Bray, B.D. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Brooks, N. [Lawrence Livermore National Laboratory, PO Box 808, Livermore, CA 94550 (United States); Leonard, A.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, PO Box 808, Livermore, CA 94550 (United States); Unterberg, E.A. [Oak Ridge National Laboratories, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Watkins, J.G. [Sandia National Laboratories, PO Box 5800, Albuquerque, NM 87185 (United States)

    2015-08-15

    The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for D{sub α}, D{sub β} and D{sub γ} for values of T{sub e} and n{sub e} characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to T{sub e}.

  13. Physics of electron cyclotron current drive on DIII-D

    CERN Document Server

    Petty, C C; Harvey, R W; Kinsey, J E; Lao, L L; Lohr, J; Luce, T C; Makowski, M A; Prater, R

    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. The narrow width of the measured ECCD profile is consistent with only low levels of radial transport for the current carrying electrons.

  14. 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

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

    Science.gov (United States)

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

    2012-09-01

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

  17. High Frequency ELM Pacing by Lithium Pellet Injection on DIII-D

    Science.gov (United States)

    Bortolon, A.; Maingi, R.; Mansfield, D. K.; Nagy, A.; Roquemore, A. L.; Lunsford, R.; Jackson, G. L.; Osborne, T. H.; Parks, P. B.

    2015-11-01

    Full-shot, high-frequency pacing of edge localized modes (ELM) by lithium pellet injection has been demonstrated in DIII-D. A Lithium Granule Injector (LGI), recently installed on DIII-D to study pacing efficiency dependence on granule size and velocity, was tested in different ELMy scenarios (βN = 1.2-2.0) injecting granules of nominal diameter 0.3-0.9 mm, with injection speed 50-120 m/s and injection rates up to 500 Hz. Robust ELM pacing was documented on time windows up to 3.5 s, with triggering efficiency close to 100% obtained with 0.9 mm diameter granules, lower with smaller sizes and weakly dependent on granule velocity. Paced ELM frequencies up to 100 Hz were achieved, with a 2-5 fold increase over the natural ELM frequency and a consequent reduction of divertor peak heat flux. Overall, LGI high frequency pacing appeared to be compatible with high plasma performance, in terms of global confinement and pedestal characteristics. Work supported by the US Department of Energy under DE-AC02-09CH11466, DE-FC02-04ER54698.

  18. Erosion and deposition of metals and carbon in the DIII-D divertor

    International Nuclear Information System (INIS)

    Wampler, W.R.; Bastasz, R.; Buchenauer, D.

    1995-01-01

    Net erosion rates at the outer strike point of the DIII-D divertor plasma were measured for several materials during quiescent H-mode operation with deuterium plasmas. Materials examined include graphite, beryllium, tungsten, vanadium and molybdenum. For graphite, net erosion rates up to 4 nm/sec were found. Erosion rates for the metals were much smaller than for carbon. Ion fluxes from Langmuir probe measurements were used to predict gross erosion by sputtering. Measured net erosion was much smaller than predicted gross erosion. Transport of metal atoms by the plasma across the divertor surface was also examined. Light atoms were transported farther than heavy atoms as predicted by impurity transport models

  19. Origins and spatial distributions of core fueling in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Owen, L.W.; Colchin, R.J.; Maingi, R.; Fenstermacher, M.E.; Carlstrom, T.N.; Groebner, R.J.

    2001-01-01

    Analysis of DIII-D discharge data with fluid plasma and Monte Carlo neutrals transport codes reveals that core particle fueling stays relatively constant between the L-mode and the ELM-free H-mode phase immediately following the L-H transition. This indicates that in the ELM-free phase nearly all of the increase in plasma electron density comes from a decrease in the cross-field transport rate and an increase in the impurity influx. This result differs from conclusions of previous work in that the effects of the thinner H-mode scrape-off-layer do not appear to be as important in a plasma that is fueled primarily from divertor recycling as would be expected if the fueling from limiter recycling were dominant. In both L-mode and H-mode the calculated core particle confinement times are less than, but within 50% of, the corresponding energy confinement times

  20. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

  1. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs

  2. 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

  3. Current profile evolution during fast wave current drive on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Forest, C.B.; Baity, F.W.

    1995-06-01

    The effect of co and counter fast wave current drive (FWCD) on the plasma current profile has been measured for neutral beam heated plasmas with reversed magnetic shear on the DIII-D tokamak. Although the response of the loop voltage profile was consistent with the application of co and counter FWCD, little difference was observed between the current profiles for the opposite directions of FWCD. The evolution of the current profile was successfully modeled using the ONETWO transport code. The simulation showed that the small difference between the current profiles for co and counter FWCD was mainly due to an offsetting change in the o at sign c current proffie. In addition, the time scale for the loop voltage to reach equilibrium (i.e., flatten) was found to be much longer than the FWCD pulse, which limited the ability of the current profile to fully respond to co or counter FWCD

  4. A New Neutron Calibration Technique with Fast Scintillators on DIII-D Tokamak

    Science.gov (United States)

    Zhu, Y. B.; Heidbrink, W. W.; Taylor, P. L.; Carrig, W.

    2015-11-01

    Absolute calibrations are necessary for conventional neutron measurements based on proportional counters and fission chambers, at regular intervals. For the DIII-D tokamak, the wide span of fusion rates, approximately between 1.e9 - 1.e17 neutrons per second, from pure Ohmic to high power auxiliary heating plasmas requires careful cross-calibrations of a variety of neutron detectors with stepwise and overlapped sensitivities, with an intense isotope neutron source, e.g. californium-252 and real plasmas. Scintillators have been successfully utilized for fast time resolved neutron detection for decades. A new calibration approach with the help of scintillators is shown to be straightforward, simpler and trustworthy while the conventional approach is complicated, time consuming and costly. Details on the calibration setup and results will be presented. Supported by US DOE SC-G903402 and DE-FC02-04ER54698.

  5. Cooperative program on DIII-D

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1991-01-01

    The main contribution of the Berkeley group to data has concerned ion temperature profile data reduction and transport analysis using this data. In addition, our graduate students have worked on fundamental aspects of transport theory, under the guidance of the Principal Investigator, to prepare them for productive participation in the D3-D program. One of these students, Q. Nguyen, has written a paper with Drs. Stambaugh and Fowler on divertor design, a subject of increasing urgency for ITER and an area of increasing importance in the D3-D program. Finally, work has been completed on determining upper bounds on fluctuation levels and growth constants, relevant to core plasma transport calculations, using thermodynamic methods. This report contains a brief summary of this work, with emphasis on the accomplishments during the past year

  6. Relationship Between Locked Modes and Disruptions in the DIII-D Tokamak

    Science.gov (United States)

    Sweeney, Ryan

    This thesis is organized into three body chapters: (1) the first use of naturally rotating tearing modes to diagnose intrinsic error fields is presented with experimental results from the EXTRAP T2R reversed field pinch, (2) a large scale study of locked modes (LMs) with rotating precursors in the DIII-D tokamak is reported, and (3) an in depth study of LM induced thermal collapses on a few DIII-D discharges is presented. The amplitude of naturally rotating tearing modes (TMs) in EXTRAP T2R is modulated in the presence of a resonant field (given by the superposition of the resonant intrinsic error field, and, possibly, an applied, resonant magnetic perturbation (RMP)). By scanning the amplitude and phase of the RMP and observing the phase-dependent amplitude modulation of the resonant, naturally rotating TM, the corresponding resonant error field is diagnosed. A rotating TM can decelerate and lock in the laboratory frame, under the effect of an electromagnetic torque due to eddy currents induced in the wall. These locked modes often lead to a disruption, where energy and particles are lost from the equilibrium configuration on a timescale of a few to tens of milliseconds in the DIII-D tokamak. In fusion reactors, disruptions pose a problem for the longevity of the reactor. Thus, learning to predict and avoid them is important. A database was developed consisting of ˜ 2000 DIII-D discharges exhibiting TMs that lock. The database was used to study the evolution, the nonlinear effects on equilibria, and the disruptivity of locked and quasi-stationary modes with poloidal and toroidal mode numbers m = 2 and n = 1 at DIII-D. The analysis of 22,500 discharges shows that more than 18% of disruptions present signs of locked or quasi-stationary modes with rotating precursors. A parameter formulated by the plasma internal inductance li divided by the safety factor at 95% of the toroidal flux, q95, is found to exhibit predictive capability over whether a locked mode will

  7. 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

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

    Science.gov (United States)

    Lohr, John; Brambila, Rigo; Cengher, Mirela; Chen, Xi; Gorelov, Yuri; Grosnickle, William; Moeller, Charles; Ponce, Dan; Prater, Ron; Torrezan, Antonio; Austin, Max; Doyle, Edward; Hu, Xing; Dormier, Calvin

    2017-07-01

    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.

  9. 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.

  10. 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

  11. 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

  12. 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

  13. Design and Analysis of the Cryopump for the DIII-D Upper Divertor

    International Nuclear Information System (INIS)

    Reis, E.E.; Baxi, C.B.; Bozek, A.S.

    1999-01-01

    A cryocondensation pump for the upper inboard divertor on DIII-D is to be installed in the vacuum vessel in the fall of 1999. The cryopump removes neutral gas particles from the divertor and prevents recycling to the plasma. This pump is designed for a pumping speed of 18,000 ell/s at 0.4 mTorr. The cryopump is toroidally continuous to minimize inductive voltages and avoid electrical breakdown during disruptions. The cryopump consists of a 25 mm Inconel tube cooled by liquid helium and is surrounded by nitrogen cooled shields. A segmented ambient temperature radiation/particle shield protects the nitrogen shields. The pump is subjected to a steady state heat load of less than 10 W due to conduction and radiation heat transfer. The helium tube will be subjected to Joule heating of less than 300 J due to induced current and a particle load of less than 12 W during plasma operation. The thermal design of the cryopump requires that it be cooled by 5 g/s liquid helium at an inlet pressure of 115 kPa and a temperature of 4.35 K. Thermal analysis and tests show that the helium tube can absorb a transient heat load of up to 100 W for 10 s and still pump deuterium at 6.3 K. Disruptions induce toroidal currents in the helium line and nitrogen shields. These currents cross the rapidly changing magnetic fields, applying complex dynamic loads on the cryopump. The forces on the pump are extrapolated from magnetic measurements from DIII-D plasma disruptions and scaled to a 3 MA disruption. The supports for the nitrogen shield consist of a racetrack design, which are stiff for reacting the disruption loads, but are radially flexible to allow differential thermal displacements with the vacuum vessel. Static and dynamic finite element analyses of the cryopump show that the stresses and displacements over a range of disruption and thermal loadings are acceptable

  14. 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

  15. Search for a critical electron temperature gradient in DIII-D L-mode discharges

    International Nuclear Information System (INIS)

    DeBoo, J.C.; Cirant, S.; Luce, T.C.; Petty, C.C.; Baker, D.R.; Greenfield, C.M.; Staebler, G.M.; Manini, A.; Ryter, F.; Austin, M.E.; Gentle, K.W.; Kinsey, J.E.

    2005-01-01

    Two experiments on DIII-D have been performed with the purpose of searching for evidence of a critical electron temperature gradient or gradient scale length. Both experiments employed off-axis EC heating to vary the local value of ∇T e /T e while holding the total heating power and thus edge temperatures constant. No evidence of an inverse critical gradient scale length, k crit , was observed in these experiments, but the existence of one cannot be ruled out by the experimental results. If k crit exists, the experimental results indicate k crit -1 at ρ = 0.45 and k crit -1 at ρ = 0.29 corresponding to a critical gradient scale length larger than 43% and 65% of the plasma minor radius, respectively. Models other than one based on k crit are also consistent with the experimental observations. (author)

  16. Real-time protection of the Ohmic heating coil force limits in DIII-D

    International Nuclear Information System (INIS)

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

    1997-11-01

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

  17. Improved operation of the Michelson interferometer electron cyclotron emission diagnostic on DIII-D

    International Nuclear Information System (INIS)

    Austin, M.E.; Ellis, R.F.; Doane, J.L.; James, R.A.

    1997-01-01

    The measurement of accurate temperature profiles is critical for transport analysis and equilibrium reconstruction in the DIII-D tokamak. Recent refinements in the Michelson interferometer diagnostic have produced more precise electron temperature measurements from electron cyclotron emission and made them available for a wider range of discharge conditions. Replacement of a lens-relay with a low-loss corrugated waveguide transmission system resulted in an increase in throughput of 6 dB and a reduction of calibration error from 15% to 5%. The waveguide exhibits a small polarization scrambling fraction of 0.05 at the quarter-wavelength frequency and very stable transmission characteristics over time. Further reduction in error was realized through special signal processing of the calibration and plasma interferograms. copyright 1997 American Institute of Physics

  18. Improved operation of the Michelson interferometer ECE diagnostic on DIII-D

    International Nuclear Information System (INIS)

    Austin, M.E.; Ellis, R.F.; Doane, J.L.; James, R.A.

    1996-05-01

    The measurement of accurate temperature profiles is critical for transport analysis and equilibrium reconstruction in the DIII-D tokamak. Recent refinements in the Michelson interferometer diagnostic have produced more precise electron temperature measurements from electron cyclotron emission and made them available for a wider range of discharge conditions. Replacement of a lens-relay with a low-loss corrugated waveguide transmission system resulted in an increase in throughput of 6 dB and reduction of calibration error to around 5%. The waveguide exhibits a small polarization scrambling fraction of 0.05 at the quarter wavelength frequency and very stable transmission characteristics over time. Further reduction in error has been realized through special signal processing of the calibration and plasma interferograms

  19. 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

  20. 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

  1. 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

  2. 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

  3. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (n e and T e ) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  4. 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

  5. Plasma transport near material boundaries

    International Nuclear Information System (INIS)

    Singer, C.E.

    1985-06-01

    The fluid theory of two-dimensional (2-d) plasma transport in axisymmetric devices is reviewed. The forces which produce flow across the magnetic field in a collisional plasma are described. These flows may lead to up-down asymmetries in the poloidal rotation and radial fluxes. Emphasis is placed on understanding the conditions under which the known 2-d plasma fluid equations provide a valid description of these processes. Attempts to extend the fluid treatment to less collisional, turbulent plasmas are discussed. A reduction to the 1-d fluid equations used in many computer simulations is possible when sources or boundary conditions provide a large enough radial scale length. The complete 1-d fluid equations are given in the text, and 2-d fluid equations are given in the Appendix

  6. 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.

  7. 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

  8. 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

  9. Physics analysis database for the DIII-D tokamak

    International Nuclear Information System (INIS)

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

    1986-03-01

    Centralized databases for handling reduced data for physics analysis have been 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 dataset (a subset of the database) with quality assurance provided by a physicist. For the confinement database, these datasets will be used to create profile datasets which will be input to a transport code to create a transport dataset. The databases are implemented on a Digital Equipment Corporation VAX 8600 running VMS. The database management system is S1032 from Software House, Inc

  10. Thomson scattering diagnostic upgrade on DIII-D

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  11. Effect of Island Overlap on ELM Suppression by Resonant Magnetic Perturbations in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Fenstermacher, M E; Evans, T E; Osborne, T H; Schaffer, M J; Aldan, M P; deGrassie, J S; Gohil, P; Joseph, I; Moyer, R A; Snyder, P B; Groebner, R J; Jakubowski, M; Leonard, A W; Schmitz, O

    2007-11-08

    Recent DIII-D [J.L. Luxon, et al., Nucl. Fusion 43, 1813 (2003)] experiments show a correlation between the extent of overlap of magnetic islands induced in the edge plasma by perturbation coils and complete suppression of Type-I edge localized modes (ELMs) in plasmas with ITER-like electron pedestal collisionality {nu}*{sub e} {approx} 0.1, flux surface shape and low edge safety factor (q{sub 95} {approx} 3.6). With fixed n = 3 resonant magnetic perturbation (RMP) strength, ELM suppression is obtained only in a finite window in the edge safety factor (q{sub 95}) consistent with maximizing the resonant component of the applied helical field. ELM suppression is obtained over an increasing range of q{sub 95} by either increasing the n = 3 RMP strength, or by adding n = 1 perturbations to 'fill in' gaps between islands across the edge plasma. The suppression of Type-I ELMs correlates with a minimum width of the edge region having magnetic islands with Chirikov parameter >1.0, based on vacuum calculations of RMP mode components excluding the plasma response or rotational shielding. The fraction of vacuum magnetic field lines that are lost from the plasma, with connection length to the divertor targets comparable to an electron-ion collisional mean free path, increases throughout the island overlap region in the ELM suppressed case compared with the ELMing case.

  12. Progress on the multipulse Thomson Scattering diagnostic on DIII-D

    International Nuclear Information System (INIS)

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

    1994-05-01

    The DIII-D Thomson scattering diagnostic, operational since 1990, uses 8 Nd: YAG 20 Hz lasers to measure electron temperature and density profiles (40 spatial points) throughout the plasma discharge. Recent progress has enabled a new set of operating modes to better fullfill varying plasma physics requirements. Custom circuitry for laser control (programmable with los precision) has successfully replaced a previous scheme which used real-time 68030 software. Two new modes of operation have been demonstrated. Burst Mode is useful to study a transient plasma event: a series of laser pulses axe fired at a rate ≤10 kHz after an external asynchronous event trigger. Burst Mode is also useful to synchronize the Thomson lasers with other systems, such as an asynchronous Michelson ECE diagnostic scanning near 40 Hz. Group Mode allows a programmed set of lasers to fire simultaneously into the same (65 nanosecond) data acquisition gate. Improved signal/noise then yields smaller statistical errors in the profile results. This provides profile data for lower density plasmas, such as those anticipated during fast wave current drive experiments. Plans for a new CCD-based laser alignment system for position monitoring and feedback control will also be presented

  13. Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

    Science.gov (United States)

    Greenfield, C. M.; DIII-D Team

    2005-01-01

    Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available.

  14. Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

    International Nuclear Information System (INIS)

    Greenfield, C M

    2005-01-01

    Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available

  15. Simulation of experimentally achieved detached plasmas using the UEDGE code

    International Nuclear Information System (INIS)

    Porter, G.D.; Allen, S.; Fenstermacher, M.

    1995-01-01

    The introduction of a divertor Thomson scattering system in DIII-D has enabled accurate determination of the plasma properties in the divertor region. We identify two plasma regimes; detached and attached. The electron temperature in the detached regime is about 2 eV, much lower than 5 to 10 eV determined earlier. We show that fluid models of the DIII-D scrape-off layer plasma are able to reproduce many of the features of these two plasma regimes, including the boundaries for transition between them. Detailed comparison between the results obtained from the fluid models and experiment suggest the models underestimate the spatial extent of the low temperature region associated the detached plasma mode. We suggest that atomic physics processes at the low electron temperatures reported here may account for this discrepancy

  16. Statistical characterization of surface features from tungsten-coated divertor inserts in the DIII-D Metal Rings Campaign

    Science.gov (United States)

    Adams, Jacob; Unterberg, Ezekial; Chrobak, Christopher; Stahl, Brian; Abrams, Tyler

    2017-10-01

    Continuing analysis of tungsten-coated inserts from the recent DIII-D Metal Rings Campaign utilizes a statistical approach to study carbon migration and deposition on W surfaces and to characterize the pre- versus post-exposure surface morphology. A TZM base was coated with W using both CVD and PVD and allowed for comparison between the two coating methods. The W inserts were positioned in the lower DIII-D divertor in both the upper (shelf) region and lower (floor) region and subjected to multiple plasma shots, primarily in H-mode. Currently, the post-exposure W inserts are being characterized using SEM/EDX to qualify the surface morphology and to quantify the surface chemical composition. In addition, profilometry is being used to measure the surface roughness of the inserts both before and after plasma exposure. Preliminary results suggest a correlation between the pre-exposure surface roughness and the level of carbon deposited on the surface. Furthermore, ongoing in-depth analysis may reveal insights into the formation mechanism of nanoscale bumps found in the carbon-rich regions of the W surfaces that have not yet been explained. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  18. 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

  19. Development, installation, and initial operation of DIII-D graphite armor tiles

    International Nuclear Information System (INIS)

    Anderson, P.M.; Baxi, C.B.; Reis, E.E.; Smith, J.P.; Smith, P.D.

    1988-04-01

    An upgrade of the DIII-D vacuum vessel protection system has been completed. The ceiling, floor, and inner wall have been armored to enable operation of CIT-relevant doublenull diverted plasmas and to enable the use of the inner wall as a limiting surface. The all- graphite tiles replace the earlier partial coverage armor configuration which consisted of a combination of Inconel tiles and graphite brazed to Inconel tiles. A new all-graphite design concept was chosen for cost and reliability reasons. The 10 minute duration between plasma discharges required the tiles to be cooled by conduction to the water-cooled vessel wall. Using two and three- dimensional analyses, the tile design was optimized to minimize thermal stresses with uniform thermal loading on the plasma-facing surface. Minimizing the stresses around the tile hold-down feature and eliminating stress concentrators were emphasized in the design. The design of the tile fastener system resulted in sufficient hold-down forces for good thermal conductance to the vessel and for securing the tile against eddy current forces. The tiles are made of graphite, and a program to select a suitable grade of graphite was undertaken. Initially, graphites were compared based on published technical data. Graphite samples were then tested for thermal shock capacity in an electron beam test facility at the Sandia National Laboratory (SNLA) in Albuquerque, New Mexico, USA. 4 refs., 6 figs

  20. 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.

  1. Multichordal visible/near-UV spectroscopy on the DIII-D tokamak

    International Nuclear Information System (INIS)

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

    1988-01-01

    A pair of visible/near-UV spectrometers with eight viewing chords apiece have been installed on the DIII-D tokamak. Each system views a neutral heating beam and can acquire up to 250 complete spectra from each chord with 5--20-ms time resolution. Each viewing chord covers 60 A with 0.27-A spectral resolution, and the chords span about (2)/(3) of the plasma's full width. By viewing Doppler-broadened spectral lines from charge exchange recombination (CER) reactions between beam neutrals and plasma ions, ion temperatures up to 4 keV have been measured, and the bulk Doppler shift of these same lines has yielded plasma rotation velocities up to 200 km/s. The constancy of temperature on a magnetic flux surface and the rigid rotor model of a flux surface have been confirmed. These instruments have also been used to measure the neutral beam deposition profile, and preliminary experimental results agree with theoretical calculations of the beam deposition profile

  2. Multichordal visible/near uv spectroscopy in the DIII-D tokamak

    International Nuclear Information System (INIS)

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

    1988-02-01

    A pair of visible/near uv spectrometers with eight viewing chords apiece have been installed on the DIII-D tokamak. Each system views a neutral heating beam, and can acquire up to 250 complete spectra from each chord with 5-20 msec time resolution. Each viewing chord covers 60 A with 0.27 A spectral resolution, and the chords span about 2/3 of the plasma's full width. By viewing Doppler broadened spectral lines from charge exchange recombination (CER) reactions between beam neutrals and plasma ions, ion temperatures up to keV have been meassured, and the bulk Doppler shift of these same lines has yielded plasma rotation velocities up to 200 km/sec. The constancy of temperature on a magnetic flux surface and the rigid rotor model of a flux surface have been confirmed. These instruments have also been used to measure the neutral beam deposition profile, and preliminary experimental results agree with theoretical calculations of the beam deposition profile. 5 refs., 6 figs

  3. 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

  4. New measurements of coil-related magnetic field errors on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Luxon, J.L. E-mail: luxon@fusion.gat.com; Jackson, G.L.; Leuer, J.A.; Nagy, A.; Schaffer, M.J.; Scoville, J.T.; Strait, E.J

    2003-09-01

    Non-axisymmetric (error) fields in tokamaks lead to a number of instabilities including so-called locked modes [J.T. Scoville, R.J. La Haye, Nucl. Fusion 43 (4) (2003) 250-257] and resistive wall modes (RWM) [A.M. Garofab, R.J. La Haye, J.T. Scoville, Nucl. Fusion 42 (11) (2002) 1335-1339] and subsequent loss of confinement. They can also cause errors in magnetic measurements made by point probes near the plasma edge, error in measurements made by magnetic field sensitive diagnostics, and they violate the assumption of axisymmetry in the analysis of data. Most notably, the sources of these error fields include shifts and tilts in the coil positions from ideal, coil leads, and nearby ferromagnetic materials excited by the coils. New measurements have been made of the n=1 coil-related field errors in the DIII-D plasma chamber. These measurements indicate that the errors due to the plasma shaping coil system are smaller than previously reported and no additional sources of anomalous fields were identified. Thus they fail to support the suggestion of an additional significant error field suggested by locked mode and RWM experiments.

  5. 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

  6. Improvements, upgrades, and plans for Thomson scattering on DIII-D

    Science.gov (United States)

    Carlstrom, T. N.; Du, D.; Glass, F.; Liu, C.; Watkins, M.; McLean, A. G.

    2016-10-01

    The Thomson scattering diagnostic on DIII-D consists of 3 beam lines that probe vertically, horizontally, and in the divertor region of the tokamak, with 54 spatial locations, edge spatial resolution down to 5 mm, and 10 Nd:YAG lasers. In its 25-year history, the collection lens optics and interference filters degraded and have been replaced, restoring previous performance. In addition, improved calibrations and detector temperature control (+/- 0.1 C) have reduced systematic errors. Cross calibration with the CO2 interferometer and ECE cut-off have improved the density calibration. Improvements to the beam line and lasers have increased the laser energy delivered to the scattering volume in the plasma. Future plans include moving the divertor system to measure regions of high triangularity using in-vessel mirrors to redirect the laser beam; adding a wide angle lens to the horizontal system to view the entire plasma radius near the plasma mid plane; and reversing the direction of the laser beam on the horizontal system to reduce the scattering angle and compressing the spectrum in wavelength space so that higher central Te measurements (<5 KeV) can be made with improved accuracy. Work supported by the US DOE under DE-FC02-04ER54698 and by LLNL under DE-AC52-07NA27344.

  7. 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.

  8. X-ray Ross filter method for impurity transport studies on DIII-D (abstract)

    Science.gov (United States)

    Bogatu, I. N.; Kim, J. S.; Egdell, D. H.; Snider, R. T.; Brooks, N. H.; Wade, M. R.; West, W. P.

    2001-01-01

    The injection of Ar into the region of the DIII-D divertor is a promising technique for energy dissipation (through radiation and collisions) and consequently for reduction of the heat load on the plates. An important problem related to this technique, is the inherent poisoning of the core plasma by migrating Ar. The Ar core contamination seems also to improve the thermal transport in an advanced operating mode of the tokamak. It is therefore of great importance to measure the evolution of the impurity concentration profile within the core plasma. This goal could be achieved by using the Ross filter method in conjunction with the existing x-ray diagnostics on DIII-D. A basic Ross filter system consists of two identical detectors placed behind two different x-ray absorbing foils looking at the same plasma volume. The foils are made of different elements or compounds with adjacent or nearly adjacent atomic numbers. Their accurate thickness causes the x-ray transmission curves of the two foils to be effectively identical over the entire energy range except within the narrow region between their absorption edges. Since the transmission characteristics of the foils above and below their absorption edges are the same, any difference in the two detected signals is proportional to the total x-ray power of the emission spectrum between these two edge energies. An x-ray Ross filter with its energy pass band centered on the Ar XVII Kα line at 3.14 keV has been designed. This allows for the discrimination of the Ar Kα line only, regardless of Ar ionization state, against any background radiation with energies outside the energy pass band. The Ross filter was installed in front of two of the fan shaped poloidal x-ray arrays on DIII-D. The first measurements showed very good discrimination against Ne, another injected impurity. Emissivity profile evolution of the Kα lines and Ar enhanced continuum within the energy pass band of the Ross filter can be determined from the x

  9. 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

  10. Radiative and SOL experiments in open and baffled divertors on DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Brooks, N.H.; Bastasz, R.

    1998-11-01

    The authors present recent progress towards an understanding of the physical processes in the divertor and scrape-off-layer (SOL) plasmas in DIII-D. This has been made possible by a combination of new diagnostics, improved computational models, and changes in divertor geometry. They have focused primarily on ELMing H-mode discharges. The physics of Partially Detached Divertor (PDD) plasmas, with divertor heat flux reduction by divertor radiation enhancement using D 2 puffing, has been studied in 2-D, and a model of the heat and particle transport has been developed that includes conduction, convection, ionization, recombination, and flows. Plasma and impurity particle flows have been measured with Mach probes and spectroscopy and these flows have been compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation has been increased in the divertor and SOL with puff and pump techniques using SOL D 2 puffing, divertor cryopumping, and argon puffing. The important physical processes in plasma-wall interactions have been examined with a DiMES probe, plasma characterization near the divertor plate, and the REDEP code. Experiments comparing single-null (SN) plasma operation in baffled and open divertors have demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H-mode with pumping and baffling has resulted in reduction in H-mode core densities to n e /n gw ∼ 0.25. Divertor particle exhaust and heat flux has been studied as the plasma shape was varied from a lower SN, to a balanced double null (DN), and finally to an upper SN

  11. 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

  12. Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

    Science.gov (United States)

    Frerichs, H.; Schmitz, O.; Covele, B.; Feng, Y.; Guo, H. Y.; Hill, D.

    2018-05-01

    Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Small changes in the strike point location can be expected to have a large impact on divertor conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the divertor slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which 3D edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.

  13. 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

  14. 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

  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. 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

  17. 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

  18. 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

  19. 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

  20. Development of Surface Eroding Thermocouples in DIII-D

    Science.gov (United States)

    Ren, Jun; Donovan, David; Watkins, Jon; Wang, Huiqian; Rudakov, Dmitry; Murphy, Christopher; Unterberg, Ezekial; Thomas, Dan; Boivin, Rejean

    2017-10-01

    The Surface Eroding Thermocouple (SETC) is a specialized diagnostic for characterizing the surface temperature evolution with a high temporal resolution ( 1ms) which is especially useful in areas unobservable by line-of-sight diagnostics (e.g. IR cameras). Recently, SETCs were tested in DiMES and successfully acquired temperature signals during strike point sweeps on the lower divertor shelf. We observed that the SETCs have a sub-10 ms time response and is sufficient to resolve ELM heat pulses. Preliminary analysis shows heat fluxes measured by SETCs and IR camera agree within 20%. Comparison of SETCs, calorimeters and Langmuir probe also show good agreement. We plan to implement an array of SETCs embedded in the tiles forming the new DIII-D small angle slot (SAS) divertor. Strategies to improve the SNR of these SETCs through testing in DiMES before the final installation will be discussed. This work was supported by the US Department of Energy under DE-SC0016318 (UTK), DE-AC05-00OR22725 (ORNL), DE-FG02-07ER54917 (UCSD), DE-FC02-04ER54698 (GA), DE-AC04-94AL85000 (SNL).

  1. Overview of equilibrium reconstruction on DIII-D using new measurements from an expanded motional Stark effect diagnostic

    International Nuclear Information System (INIS)

    Holcomb, C; Makowski, M; Allen, S; Meyer, W; Van Zeeland, M

    2008-01-01

    Motional Stark effect (MSE) measurements constrain equilibrium reconstruction of DIII-D tokamak plasmas using the equilibrium code EFIT. In 2007, two new MSE arrays were brought online, bringing the system to three core arrays, two edge arrays, and 64 total channels. We present the first EFIT reconstructions using this expanded system. Safety factor and E R profiles produced by fitting to data from the two new arrays and one of the other three agree well with independent measurements. Comparison of the data from the three arrays that view the core shows that one of the older arrays is inconsistent with the other two unless the measured calibration factors for this array are adjusted. The required adjustments depend on toroidal field and plasma current direction, and on still other uncertain factors that change as the plasma evolves. We discuss possible sources of calibration error for this array

  2. 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

  3. 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

  4. Characterizing Low-Z erosion and deposition in the DIII-D divertor using aluminum

    Directory of Open Access Journals (Sweden)

    C.P. Chrobak

    2017-08-01

    Full Text Available We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth and realistic (rough surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms.

  5. Migration of Artificially Introduced Micron Size Carbon Dust in the DIII-D Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D; West, W; Wong, C; Brooks, N; Evans, T; Fenstermacher, M; Groth, M; Krasheninnikov, S; Lasnier, C; McLean, A; Pigarov, A Y; Solomon, W; Antar, G; Boedo, J; Doerner, R; Hollmann, E; Hyatt, A; Maingi, R; Moyer, R; Nagy, A; Nishino, N; Roquemore, L; Stangeby, P; Watkins, J

    2006-05-15

    Migration of pre-characterized carbon dust in a tokamak environment was studied by introducing about 30 milligrams of dust flakes 5-10 {micro}m in diameter in the lower divertor of DIII-D using the DiMES sample holder. The dust was exposed to high power ELMing Hmode discharges in lower-single-null magnetic configuration with the strike points swept across the divertor floor. When the outer strike point (OSP) passed over the dust holder exposing it to high particle and heat fluxes, part of the dust was injected into the plasma. In about 0.1 sec following the OSP pass over the dust, 1-2% of the total dust carbon content (2-4 x 10{sup 19} carbon atoms, equivalent to a few million dust particles) penetrated the core plasma, raising the core carbon density by a factor of 2-3. When the OSP was inboard of the dust holder, the dust injection continued at a lower rate. Individual dust particles were observed moving at velocities of 10-100 m/s, predominantly in the toroidal direction for deuteron flow to the outer divertor target, consistent with the ion drag force. The observed behavior of the dust is in qualitative agreement with modeling by the 3D DustT code.

  6. Role of turbulence in determining particle transport in DIII-D

    Science.gov (United States)

    Mordijck, Saskia; Zeng, Lei; Rhodes, Terry; Salmi, Antti; Tala, Tuomas

    2017-10-01

    Recent advances in DIII-D identify how changes in turbulence and ExB shear affect particle transport in H-mode plasmas. Using a combination of co- and counter- injected neutral beams to vary applied torque, the ExB shear is systematically scanned at fixed power and fueling. When the ExB shear is reduced below the linear gyro-kinetic growth rates inside the pedestal top (ρ =0.6-0.8), the particle confinement is strongly reduced by an increase in outward diffusion. Furthermore, a slow modulation in ECH power from 1 to 3 MW shows that the density reduction (``pump-out'') originates in the same region. Time-dependent analysis finds that the pump-out begins with a strong increase in density fluctuations measured by DBS at ρ =0.78, where the initial density reduction is largest, along with an increase in the linear growth rate of the Ion Temperature Gradient (ITG) mode. Turbulence modeling by TGLF shows that the plasma eventually transitions from an ITG mode to a Trapped Electron Mode (TEM) regime during high power ECH, but the TEM is not the initial cause of density pump out. For stationary density profiles, the frequency of the dominant unstable mode (i.e., ITG or TEM) correlates with the local density gradient, as predicated by theoretical simulations. Supported by US DOE DE-SC0007880, DE-FC02-04ER54698.

  7. Improved Confinement in Highly Powered Advanced Tokamak Scenarios on DIII-D

    Science.gov (United States)

    Petrie, T. W.; Leonard, A.; Luce, T.; Osborne, T.; Solomon, W.; Turco, F.; Fenstermacher, M. E.; Holcomb, C.; Lasnier, C.; Makowski, M.

    2016-10-01

    DIII-D has recently demonstrated improved energy confinement by injecting neutral gas into high performance Advanced Tokamak (AT) plasmas during high power operation. Representative parameters are: q95 = 6, PIN up to 15 MW, H98 = 1.4-1.8, and βN = 2.8-4.2. Unlike in lower and moderate powered AT plasmas, τE and βN increased (and νELM decreased) as density was increased by deuterium gas puffing. We discuss how the interplay between pedestal density and temperature with fueling can lead to higher ballooning stability and a peeling/kink current limit that increasers as the pressure gradient increases. Comparison of neon, nitrogen, and argon as ``seed'' impurities in high PIN ATs in terms of their effects on core dilution, τE, and heat flux (q⊥) reduction favors argon. In general, the puff-and-pump radiating divertor was not as effective in reducing q⊥ while maintaining density control at highest PIN than it was at lower PIN. Work supported by the US DOE under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC04-94AL85000, DE-AC52-07NA27344, and DE-FG02-07ER54917.

  8. 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

  9. Observation of poloidal current flow to the vacuum vessel wall during vertical instabilities in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Strait, E.J.; Lao, L.L.; Luxon, J.L.; Reis, E.E.

    1991-01-01

    An attached poloidal current, which flows in a circuit lying partly in the vacuum vessel wall and partly in the scrape-off layer of the plasma, is observed during vertical instabilities in the DIII-D tokamak. A direct measurement of the current, using Rogowski loops on several protective tiles at locations where the plasma contacts the wall, is in good agreement with the value determined from MHD equilibrium reconstructions using measured values of magnetic field and flux. This attached current, which can reach transient peaks of several hundred kilo-amperes, interacts with the toroidal magnetic field to create a large vertical force on the vacuum vessel. The predicted motion of the vessel resulting from the measured currents agrees well with the observed displacement of the vacuum vessel. (author). 14 refs, 5 figs

  10. Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

    International Nuclear Information System (INIS)

    Henline, P.A.

    1995-10-01

    The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DRI-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape control due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described

  11. Grating spectrometer installation for electron cyclotron emission measurements on the DIII-D tokamak using circular waveguide and synchronous detection

    International Nuclear Information System (INIS)

    Lohr, J.; Jahns, G.; Moeller, C.; Prater, R.

    1986-01-01

    The grating spectrometer installation on the DIII-D tokamak uses fundamental circular waveguide propagating the TE 11 lowest-order mode followed by oversized circular guide carrying the low-loss TE 01 mode. The short section of fundamental guide permits use of an electronic chopper operating at 100 kHz for both calibration and plasma operation. By using ac-coupled amplifiers tuned to the chopping frequency, the background signal generated in the indium antimonide detectors by neutrons and x rays is automatically subtracted and the system noise bandwidth is reduced. Compared with a quasi-optical system, the much smaller fundamental horn and front-end waveguide allow the waveguide system to be located outside a gate valve. With this configuration the entire waveguide run, including the actual horn and vacuum window used during plasma operations, can be included in the calibration setup

  12. 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.

  13. 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

  14. 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.; Paz-Soldan, C.; Hollmann, E. M.; Moyer, R. A.

    2016-10-01

    Effective runaway electron (RE) mitigation strategies are essential for protecting ITER from the potential damage of a first wall strike. In DIII-D, shattered pellet injection (SPI) with large Ne pellets demonstrates the dissipation of post-disruption RE plateaus by collisions with high-Z impurities, while equivalently sized D2 pellets lead to a reduction of the impurity content of the background plasma, reducing RE dissipation. Varying the relative quantities of Ne /D2 in mixed species pellets shows that the effect of D2 may be dominant in determining the RE/pellet interaction. Compared with injection of the same quantity of Ne by massive gas injection, SPI achieves a similar initial RE current decay rate, but residual RE current remains after SPI. This may be due to the effects of a small quantity of D2 (used as a ``shell'' for firing of the Ne pellets) displacing high-Z impurities. These results will help guide the optimization of injection schemes and pellet compositions for the RE mitigation system in ITER. Work supported by the U.S. DOE under DE-FC02-04ER54698.

  15. Analysis Tools for the Ion Cyclotron Emission Diagnostic on DIII-D

    Science.gov (United States)

    Del Castillo, C. A.; Thome, K. E.; Pinsker, R. I.; Meneghini, O.; Pace, D. C.

    2017-10-01

    Ion cyclotron emission (ICE) waves are excited by suprathermal particles such as neutral beam particles and fusion products. An ICE diagnostic is in consideration for use at ITER, where it could provide important passive measurement of fast ions location and losses, which are otherwise difficult to determine. Simple ICE data analysis codes had previously been developed, but more sophisticated codes are required to facilitate data analysis. Several terabytes of ICE data were collected on DIII-D during the 2015-2017 campaign. The ICE diagnostic consists of antenna straps and dedicated magnetic probes that are both digitized at 200 MHz. A suite of Python spectral analysis tools within the OMFIT framework is under development to perform the memory-intensive analysis of this data. A fast and optimized analysis allows ready access to data visualizations as spectrograms and as plots of both frequency and time cuts of the data. A database of processed ICE data is being constructed to understand the relationship between the frequency and intensity of ICE and a variety of experimental parameters including neutral beam power and geometry, local and global plasma parameters, magnetic fields, and many others. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

  16. 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.

  17. ECE electron temperature mapping errors in high-performance DIII-D discharges

    International Nuclear Information System (INIS)

    Garstka, G.D.; Austin, M.E.; Ellis, R.F.

    2001-01-01

    The proper mapping of diagnostic profiles to equilibrium flux surfaces is an important step in the analysis of tokamak data. On DIII-D, the mapping of electron temperature profiles obtained from second- and third-harmonic electron cyclotron emission (ECE) occasionally conflicts with that of T e profiles obtained by Thomson scattering, especially in high-performance negative central shear (NCS) discharges. Particularly, the ECE diagnostics report higher electron temperatures than the Thomson scattering measurements on the low-field side, although the profiles coincide within the measurement uncertainties on the high-field side. It is believed that this is the result of errors in the mapping of B to ρ, the normalized minor radius. This problem can be characterized for a particular equilibrium by calculating the percent reduction in B required to force agreement between the ECE and Thomson profiles. Calculations taking relativistic line broadening into account show that the required reduction in B in the range 0.2<ρ<0.8 is less than 5% in almost all cases, and is typically closer to 2%. Since the vacuum B-field is known to within a fraction of a percent, a discrepancy in total B must represent errors involving finite beta or other plasma parameters not related to vacuum magnetic field. The required reduction in B versus various parameters is investigated for a database of shots. It is found that in many cases the overlap error is due to the presence of large magnetic islands associated with locked modes

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

    Science.gov (United States)

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

    2015-11-01

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

  19. Structure, stability and ELM dynamics of the H-mode pedestal in DIII-D

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Leonard, A.W.; Osborne, T.H.

    2005-01-01

    Experiments are described that have increased understanding of the transport and stability physics that set the H-mode edge pedestal width and height, determine the onset of Type-I edge localized modes (ELMs), and produce the nonlinear dynamics of the ELM perturbation in the pedestal and scrape-off layer (SOL). Predictive models now exist for the n e pedestal profile and the p e height at the onset of Type-I ELMs, and progress has been made toward predictive models of the T e pedestal width and nonlinear ELM evolution. Similarity experiments between DIII-D and JET suggested that neutral penetration physics dominates in the relationship between the width and height of the n e pedestal while plasma physics dominates in setting the T e pedestal width. Measured pedestal conditions including edge current at ELM onset agree with intermediate-n peeling-ballooning (P-B) stability predictions. Midplane ELM dynamics data show the predicted (P-B) structure at ELM onset, large rapid variations of the SOL parameters, and fast radial propagation in later phases, similar to features in nonlinear ELM simulations. (author)

  20. 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

  1. Integrated modeling of high βN steady state scenario on DIII-D

    Science.gov (United States)

    Park, J. M.; Ferron, J. R.; Holcomb, C. T.; Buttery, R. J.; Solomon, W. M.; Batchelor, D. B.; Elwasif, W.; Green, D. L.; Kim, K.; Meneghini, O.; Murakami, M.; Snyder, P. B.

    2018-01-01

    Theory-based integrated modeling validated against DIII-D experiments predicts that fully non-inductive DIII-D operation with βN > 4.5 is possible with certain upgrades. IPS-FASTRAN is a new iterative numerical procedure that integrates models of core transport, edge pedestal, equilibrium, stability, heating, and current drive self-consistently to find steady-state (d/dt = 0) solutions and reproduces most features of DIII-D high βN discharges with a stationary current profile. Projecting forward to scenarios possible on DIII-D with future upgrades, the high qmin > 2 scenario achieves stable operation at βN as high as 5 by using a very broad current density profile to improve the ideal-wall stabilization of low-n instabilities along with confinement enhancement from low magnetic shear. This modeling guides the necessary upgrades of the heating and current drive system to realize reactor-relevant high βN steady-state scenarios on DIII-D by simultaneous optimization of the current and pressure profiles.

  2. 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

  3. Suppression of Type-I ELMs with a Reduced I-coil Set in DIII-D

    Science.gov (United States)

    Orlov, D. M.

    2013-10-01

    Recent experiments in DIII-D have demonstrated that Edge Localized Modes (ELMs) in a tokamak can be controlled with a reduced number of magnetic perturbation coils, demonstrating an important role of spectral sidebands, and showing promise of the technique for future fusion devices, where ELMs risk potentially damaging heat loads. The ELMs can be controlled with external magnetic perturbation used to regulate pressure gradients and maintain stability. The new results show that the coil currents required for ELM suppression with the reduced coil sets are comparable to what is typically required for the full set. This counterintuitive result provides an important validation of recent modeling of the physical mechanisms involved. This modeling shows that the spectral sidebands introduced by deactivating individual coils can often increase the magnetic stochasticity within the plasma, thereby increasing transport and facilitating ELM suppression. Deactivating individual coils results not only in the reduction of the dominant n = 3 component of the perturbation field, but also in a significant increase in the amplitudes of n = 1 and n = 2 sidebands. These sidebands may also be amplified by the plasma response. Application to ITER finds that the ITER ELM coils may be able to tolerate a loss of up to five of its 27 coils, while leaving a sufficient margin of current in the remaining coils to still meet the DIII-D ELM suppression criterion. Further, the new experiments show that the presence of the spectral sidebands does not adversely affect the plasma rotation or confinement. Both vacuum and two-fluid modeling are used to interpret and understand these results. Supported by US DOE under DE-FG02-05ER54809 & DE-FC02-04ER54698.

  4. 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

  5. DIII-D radiation shielding procedures and experiences

    International Nuclear Information System (INIS)

    Taylor, P.L.

    1991-11-01

    The D3-D tokamak operates with a neutron radiation shield to allow enhanced plasma operations with increased neutron production while minimizing the site boundary dose level. Neutron rates as high as 4 x 10 15 neutrons/s and total neutron production as high as 4 x 10 15 neutrons per shot are obtained while maintaining the site dose below the DOE administrative level of 20 mrem per year; a much more restrictive level than the State of California radiation limits. The radiation shielding has increased by a factor of 300 over the preshield value and is in agreement with the design calculation. The maximum site neutron dose since installation of the shield has been less than 0.03 mrem for a shot and less than 0.4 mrem for a day. The site neutron and gamma dose are monitored continuously during operations by a PC-based computer system that provides the means of measuring the low dose levels that occur during a shot by including postshot background subtraction. The neutron and gamma dose are measured and archived by shot, hour, and day in a database. Activation of the machine after a run day and during vessel entries is monitored and the activated nuclides have been determined. A radiation monitoring program and procedures are used to control the exposures to facility personnel and the exposure at the site boundary

  6. 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

  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. 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)

  9. Fast wave and electron cyclotron current drive in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Petty, C.C.; Pinsker, R.I.; Austin, M.E.

    1995-01-01

    The non-inductive current drive from directional fast Alfven and electron cyclotron waves was measured in the DIII-D tokamak in order to demonstrate these forms of radiofrequency (RF) current drive and to compare the measured efficiencies with theoretical expectations. The fast wave frequency was 8 times the deuterium cyclotron frequency at the plasma centre, while the electron cyclotron wave was at twice the electron cyclotron frequency. 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 steady current discharges, an analysis of the loop voltage revealed up to 195 kA of a non-inductive current (out of 310 kA) during combined electron cyclotron and fast wave injection, with a maximum of 110 kA of FWCD and 80 kA of ECCD achieved (not simultaneously). The peakedness of the current profile increased with RF current drive, indicating that the driven current was centrally localized. The FWCD efficiency increased linearly with the central electron temperature as expected; however, the FWCD was severely degraded in low current discharges owing to incomplete fast wave absorption. The measured FWCD agreed with the predictions of a ray tracing code only when a parasitic loss of 4% per pass was included in the modelling along with multiple pass absorption. Enhancement of the second harmonic ECCD efficiency by the toroidal electric field was observed experimentally. The measured ECCD was in good agreement with Fokker-Planck code predictions. (author). 41 refs, 13 figs, 1 tab

  10. Development of a cross-polarization scattering system for the measurement of internal magnetic fluctuations in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, T. L., E-mail: trhodes@ucla.edu; Peebles, W. A.; Crocker, N. A.; Nguyen, X. [Physics and Astronomy Department, University of California, Los Angeles, California 90098 (United States)

    2014-11-15

    The design and performance of a new cross-polarization scattering (CPS) system for the localized measurement of internal magnetic fluctuations is presented. CPS is a process whereby magnetic fluctuations scatter incident electromagnetic radiation into a perpendicular polarization which is subsequently detected. A new CPS design that incorporates a unique scattering geometry was laboratory tested, optimized, and installed on the DIII-D tokamak. Plasma tests of signal-to-noise, polarization purity, and frequency response indicate proper functioning of the system. CPS data show interesting features related to internal MHD perturbations known as sawteeth that are not observed on density fluctuations.

  11. 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.

  12. 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)

  13. 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

  14. Analysis and testing of the DIII-D ohmic heating coil lead repair clamp

    International Nuclear Information System (INIS)

    Reis, E.E.; Anderson, P.M.; Chin, E.; Robinson, J.I.

    1997-11-01

    DIII-D has been operating for the last year with limited volt-second capabilities due to structural failure of a conductor lead to one of the ohmic heating (OH) solenoids. The conductor failure was due to poor epoxy impregnation of the overwrap of the lead pack, resulting in copper fatigue and a water leak. A number of structural analyses were performed to assist in determining the failure scenario and to evaluate various repair options. A fatigue stress analysis of the leads with a failed epoxy overwrap indicated crack initiation after 1,000 cycles at the maximum operating conditions. The failure occurred in a very inaccessible area which restricted design repair options to concepts which could be implemented remotely. Several design options were considered for repairing the lead so that it can sustain the loads for 7.5 Vs conditions at full toroidal field. A clamp, along with preloaded banding straps and shim bags, provides a system that guarantees that the stress at the crack location is always compressive and prevents further crack growth in the conductor. Due to the limited space available for the repair, it was necessary to design the clamp system to operate at the material yield stress. The primary components of the clamp system were verified by load tests prior to installation. The main body of the clamp contains a load cell and potentiometer for monitoring the load-deflection characteristics of the clamp and conductors during plasma operation. Strain gages provides redundant instrumentation. If required, the preload on the conductors can be increased remotely by a special wrench attached to the clamp assembly

  15. 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

  16. Boundary Plasma Turbulence Simulations for Tokamaks

    International Nuclear Information System (INIS)

    Xu, X.; Umansky, M.; Dudson, B.; Snyder, P.

    2008-05-01

    The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T e ; T i ) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics

  17. 2-D Laser-Calibrated Doppler Images of HeII and CIII Emission on DIII-D

    Science.gov (United States)

    Allen, S. L.; Samuell, Cameron; Meyer, W. H.

    2017-10-01

    Recent improvements to the DIII-D CIS system have reduced the error bars of the inferred Doppler velocity by over an order of magnitude, i.e. to 0.1 km/s. Coherence imaging of plasma emission superimposes an interferogram on the plasma image, and the interferometer phase is a sensitive measure of the central wavelength of the emission. A tuneable diode laser calibration image at 465 nm is automatically acquired between plasma shots and provides the rest wavelength in the lab frame; the wavelength is measured with a wavemeter to 0.01 pm. The interferometer is stabilized mechanically and thermally with a unique system so that the interferometer drift between calibrations is small. These improvements have enabled tomographically inverted images of main ion He II parallel flow in the divertor during He plasma operation. The parallel flow, as expected, is observed to depend on the direction of the B × ∇B drift, which is reversed by changing the direction of the toroidal field. For many conditions, the C III Doppler velocity is also in the same direction as the main ion. Work supported by the US DOE under DE-FC02-04ER54698 and DE-AC52-07NA27344. LLNL-ABS-88688.

  18. First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

    International Nuclear Information System (INIS)

    Litnovsky, A.; Rudakov, D.L.; De Temmerman, G.; Wienhold, P.; Philipps, V.; Samm, U.; McLean, A.G.; West, W.P.; Wong, C.P.C.; Brooks, N.H.; Watkins, J.G.; Wampler, W.R.; Stangeby, P.C.; Boedo, J.A.; Moyer, R.A.; Allen, S.L.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J.; Boivin, R.L.

    2008-01-01

    Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 o C neither carbon deposition nor degradation of optical properties was detected

  19. Hydrocarbon transport in a plasma boundary layer

    International Nuclear Information System (INIS)

    Langer, W.D.; Ehrhardt, A.B.

    1989-01-01

    The theory of carbon transport in a plasma boundary layers is important for understanding the impurity penetration, and carbon and hydrogen recycling, in tokamaks using carbon compounds as limiters and as wall coatings. Neutral carbon kinetics and transport at the edge of plasma devices where chemical release is a source of carbon are modeled. Plasma reactions with carbon and hydrocarbons are important for such modeling, and these collisional processes are summarized. Combining the reaction schemes and kinetics in the DEGAS code makes it possible to treat the neutral transport at the plasma boundary layer. Results of such modeling of the atomic carbon and methane distribution at the edge are presented for comparison with recent carbon probe experiments performed on the Divertor and Injection Tokamak Experiment (DITE)

  20. Physics of a fusion plasma boundary layer

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, B.K.

    1977-03-01

    A theoretical and computational study has been made of plasma phenomena occurring when a hot, dense plasma containing a transverse magnetic field is brought into sudden contact with a cold metal wall; thermal and magnetic boundary layers develop. The time evolution of the plasma temperature, pressure, the charged and neutral particle concentration, magnetic and electric field strengths, and the plasma current density in the neighborhood of the solid surface are investigated. The rate of energy transfer from the plasma to the wall is calculated, and the conditions under which wall surface melting occurs are estimated. The physical conditions previously studied experimentally by Feinberg, are calculated, and the predicted rate of energy transfer from the plasma to the wall is found to be in good agreement.

  1. Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, C. M., E-mail: coopercm@fusion.gat.com [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Pace, D. C.; Paz-Soldan, C.; Eidietis, N. W. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Commaux, N.; Shiraki, D. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California, San Diego, La Jolla, California 92093-0533 (United States)

    2016-11-15

    A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5–100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead “pinhole camera” mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20 000 pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.

  2. Confinement improvement in the high poloidal beta regime on DIII-D and application to steady-state H-mode on EAST

    Science.gov (United States)

    Ding, S.; Garofalo, A. M.; Qian, J.; Cui, L.; McClenaghan, J. T.; Pan, C.; Chen, J.; Zhai, X.; McKee, G.; Ren, Q.; Gong, X.; Holcomb, C. T.; Guo, W.; Lao, L.; Ferron, J.; Hyatt, A.; Staebler, G.; Solomon, W.; Du, H.; Zang, Q.; Huang, J.; Wan, B.

    2017-05-01

    Systematic experimental and modeling investigations on DIII-D show attractive transport properties of fully non-inductive high βp plasmas. Experiments on DIII-D show that the large-radius internal transport barrier (ITB), a key feature providing excellent confinement in the high βp regime, is maintained when the scenario is extended from q95 ˜ 12 to 7 and from rapid to near-zero toroidal rotation. The robustness of confinement versus rotation was predicted by gyrofluid modeling showing dominant neoclassical ion energy transport even without the E × B shear effect. The physics mechanism of turbulence suppression, we found, is the Shafranov shift, which is essential and sets a βp threshold for large-radius ITB formation in the high βp scenario on DIII-D. This is confirmed by two different parameter-scan experiments, one for a βN scan and the other for a q95 scan. They both give the same βp threshold at 1.9 in the experiment. The experimental trend of increasing thermal transport with decreasing βp is consistent with transport modeling. The progress toward the high βp scenario on Experimental Advanced Superconducting Tokamak (EAST) is reported. The very first step of extending the high βp scenario on DIII-D to long pulse on EAST is to establish a long pulse H-mode with ITB on EAST. This paper shows the first 61 s fully non-inductive H-mode with stationary ITB feature and actively cooled ITER-like tungsten divertor in the very recent EAST experiment. The successful use of lower hybrid wave as a key tool to optimize the current profile in the EAST experiment is also introduced. Results show that as the electron density is increased, the fully non-inductive current profile broadens on EAST. The improved understanding and modeling capability are also used to develop advanced scenarios for the China Fusion Engineering Test Reactor. Overall, these results provide encouragement that the high βp regime can be extended to a lower safety factor and very low rotation

  3. The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, Christopher [General Atomics, San Diego; Nygren, R. E. [Sandia National Laboratories (SNL); Chrobak, C P. [General Atomics, San Diego; Buchenauer, Dean [Sandia National Laboratories (SNL); Holtrop, Kurt [General Atomics, San Diego; Unterberg, Ezekial A. [ORNL; Zach, Mike P. [ORNL

    2017-08-01

    Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.

  4. Commissioning of the 110 GHz ECH System on DIII-D for Physics Applications

    Science.gov (United States)

    Lohr, J. M.; Callis, R. W.; Cary, W. P.; Delaware, S. W.; Gorelov, I.; Harris, T. E.; Legg, R. A.; Nerem, A.; Ponce, D.; Prater, R.; Pronko Baity, S. G., Jr.; Barber, G. C.

    2000-10-01

    A major upgrade to the gyrotron system for ECH/ ECCD on the DIII-D tokamak is in progress. Four gyrotrons in the MW class at 110 GHz, the second harmonic electron cyclotron resonance, are operational. The upgrade project has included construction of a major addition to the DIII-D building, installation of two Gycom gyrotrons and associated modulator/ regulators, construction of two new high voltage power supply systems, and acquisition of three CPI gyrotrons capable of generating 1.0 MW for 10 s long pulses. Two of the gyrotrons are connected to an articulating launcher system from PPPL featuring poloidal and toroidal scan capability. Launchers for four additional gyrotrons, two with oblique injection for current drive and two injecting perpendicularly for heating only, but with poloidal scan capability, are also installed. The waveguide lines are evacuated, windowless and up to 80 m in length. Full control of the polarization of the rf beam is provided.

  5. 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

  6. Applications of ECH on the DIII-D tokamak and projections for future ECH upgrades

    Directory of Open Access Journals (Sweden)

    Solomon W.M.

    2012-09-01

    Full Text Available Electron Cyclotron Heating and Current Drive plays an important role in the DIII-D program. In high performance discharges EC power contributes greatly to MHD stability, and this is particularly important for discharges with low rotational torque applied, as will be the case for ITER. Off-axis EC current drive also plays a key role in the actualization of steady-state scenarios by supporting the desired current profile. In order to carry out these applications at higher beta and higher field, an upgrade of the EC power to 15 MW is needed, and the best gyrotron frequency for the DIII-D program is 117.5 GHz.

  7. 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.; Ferron, J.; Mahdavi, M.A.; Osborne, T.H.; Petersen, P.I.; Seraydarian, R.; Strait, E.J.; Taylor, P.L.; Allen, S.L.; Hill, D.; Haas, G.; Nakamura, H.; Shimada, M.

    1989-01-01

    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 condition 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 auxiliary heating, with significant improvement in ohmic energy confinement. (orig.)

  8. 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

  9. 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.; Stambaugh, R.D.

    1996-04-01

    V-4Cr-4-Ti alloy has been recently selected for use in the manufacture of a portion of the DIII-D Radiative Divertor modification, as part of an overall DIII-D vanadium alloy deployment effort developed by General Atomics (GA) in conjunction with the Argonne and Oak Ridge National Laboratories (ANL or ORNL). The goal of this work is to produce a production-scale heat of the alloy and fabricate it into product forms for the manufacture of a portion of the Radiative Divertor (RD) for the DIII-D tokamak, to develop the fabrications technology for manufacture of the vanadium alloy radiative Divertor components, and to determine the effects of typical tokamak environments in the behavior of the vanadium alloy. The production of a {approx}1300-kg heat of V-4Cr-4Ti alloy is currently in progress at Teledyne Wah Chang of Albany, oregon (TWCA) to provide sufficient material for applicable product forms. Two unalloyed vanadium ingots for the alloy have already been produced by electron beam melting of raw processes vanadium. Chemical compositions of one ingot and a portion of the second were acceptable, and Charpy V-Notch (CVN) impact test performed on processed ingot samples indicated ductile behavior. Material from these ingots are currently being blended with chromium and titanium additions, and will be vacuum-arc remelted into a V-4Cr-4Ti alloy ingot and converted into product forms suitable for components of the DIII-D RD structure. Several joining methods selected for specific applications in fabrication of the RD components are being investigated, and preliminary trials have been successful in the joining of V-alloy to itself by both resistance and inertial welding processes and to Inconel 625 by inertial welding.

  10. Prototype testing of the ITER Toroidal Interferometer and Polarimeter (TIP) on DIII-D

    Science.gov (United States)

    Carlstrom, T. N.; van Zeeland, M. A.; Gattuso, A.; O'Neill, R.; Vasquez, J.; Finkenthal, D. K.; Colio, R. A.; Johnson, D.; Brower, D.; Chen, J.; Ding, W.>X.>

    2017-10-01

    A 10.6 micron CO2 laser based ITER TIP system has been designed and tested for density measurements on DIII-D. Features include vibration compensation using a 5.22 micron Quantum Cascade Laser, real-time measurements at 1 kHz with Cooperative Agreement DE-FC02-04ER54698 and Contract Number DE-AC-02-09CH11466.

  11. 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

  12. 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.

  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. Data Analysis Software Tools for Enhanced Collaboration at the DIII-D National Fusion Facility

    Energy Technology Data Exchange (ETDEWEB)

    Schachter, J.; Peng, Q.; Schissel, D.P.

    1999-07-01

    Data analysis at the DIII-D National Fusion Facility is simplified by the use of two software packages in analysis codes. The first is GAP1otObj, an IDL-based object-oriented library used in visualization tools for dynamic plotting. GAPlotObj gives users the ability to manipulate graphs directly through mouse and keyboard-driven commands. The second software package is MDSplus, which is used at DIED as a central repository for analyzed data. GAPlotObj and MDSplus reduce the effort required for a collaborator to become familiar with the DIII-D analysis environment by providing uniform interfaces for data display and retrieval. Two visualization tools at DIII-D that benefit from them are ReviewPlus and EFITviewer. ReviewPlus is capable of displaying interactive 2D and 3D graphs of raw, analyzed, and simulation code data. EFITviewer is used to display results from the EFIT analysis code together with kinetic profiles and machine geometry. Both bring new possibilities for data exploration to the user, and are able to plot data from any fusion research site with an MDSplus data server.

  16. 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.

  17. Modeling of impurity spectroscopy in the divertor and SOL of DIII-D using the 1D multifluid model NEWT1D

    International Nuclear Information System (INIS)

    West, W.P.; Evans, T.E.; Brooks, N.H.

    1996-10-01

    NEWT1D, a one dimensional multifluid model of the scrape-off layer and divertor plasma, has been used to model the plasma including the distribution of carbon ionization states in the SOL and divertor of ELMing H-mode at two injected power levels in DIII-D. Comparison of the code predictions to the measured divertor and scrape-off layer (SOL) plasma density and temperature shows good agreement. Comparison of the predicted line emissions to the spectroscopic data suggests that physically sputtered carbon from the strike point is not transported up the flux tube; a distributed source of carbon a few centimeters up the flux tube is required to achieve reasonable agreement

  18. 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.

  19. Self-regulation of turbulence in low rotation DIII-D QH-mode with an oscillating transport barrier

    Science.gov (United States)

    Barada, Kshitish; Rhodes, T. L.; Burrell, K. H.; Zeng, L.; Chen, Xi

    2016-10-01

    We present observations of turbulence and flow shear limit cycle oscillations (LCOs) in wide pedestal QH-mode DIII-D tokamak plasmas that are consistent with turbulence self-regulation. In this low input torque regime, both edge harmonic oscillations (EHOs) and ELMs are absent. LCOs of ExB velocity shear and ñ present predator-prey like behavior in these fully developed QH-mode plasmas. During these limit cycle oscillations, the ExB poloidal flows possess a long-range toroidal correlation consistent with turbulence generated zonal flow activity. Further, these limit cycle oscillations are observed in a broad range of edge parameters including ne, Te, floor Langmuir probe ion saturation current, and radial electric field Er. TRANSP calculations of transport indicate little change between the EHO and LCO wide pedestal phases. These observations are consistent with LCO driven transport that may play a role in maintaining the profiles below ELM threshold in the EHO-free steady state wide pedestal QH-mode regime. Work supported by the US DOE under DE-FG02-08ER54984 and DE-FC02-04ER54698.

  20. DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

    International Nuclear Information System (INIS)

    Rudakov, D; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W; Wong, C; Allen, S; Bastasz, R; Boedo, J; Brooks, N; Boivin, R; De Temmerman, G; Fenstermacher, M; Groth, M; Hollmann, E; Lasnier, C; McLean, A; Moyer, R; Stangeby, P; Wampler, W; Watkins, J; Wienhold, P; Whaley, J

    2006-01-01

    A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited virtually no carbon deposition

  1. OEDGE Modeling of Collector Probe measurements in L-mode from the DIII-D tungsten ring campaign

    Science.gov (United States)

    Elder, J. D.; Stangeby, P. C.; Unterberg, Z.; Donovan, D.; Wampler, W. R.; Watkins, J.; Abrams, T.; McLean, A. G.

    2017-10-01

    During the tungsten ring campaign on DIII-D, a collector probe system with multiple diameter, dual-facing collector rods was inserted into the far scrape off layer (SOL) near the outer midplane to measure the plasma tungsten content. For most probes more tungsten was observed on the side connected along field lines to the inner divertor, with the larger probes showing largest divertor-facing asymmetries The OEDGE code is used to model the tungsten erosion, transport and deposition. It has been enhanced with (i) a peripheral particle transport and deposition model to record the impurity content in the peripheral region outside the regular mesh, and (ii) a collector probe model. The OEDGE results approximately reproduce both the divertor-facing asymmetries and the radial decay of each collector probe profile. The effect of changing impurity transport assumptions and wall location are examined. The measured divertor-facing asymmetries imply a higher tungsten density in the plasma upstream of the probe; this was expected theoretically from the effect of the parallel ion temperature gradient force driving upstream transport of tungsten from the outer divertor and was also found in the code analysis. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-NA0003525, DE-AC05-00OR22725, and DE-AC52-07NA27344.

  2. Core and edge aspects of quiescent double barrier operation on DIII-D, with relevance to critical ITB physics issues

    International Nuclear Information System (INIS)

    Doyle, E.J.; Casper, T.A.; Burrell, K.H.

    2003-01-01

    Recent results from DIII-D address critical internal transport barrier (ITB) research issues relating to sustainability, impurity accumulation and ITB control, and have also demonstrated successful application of general profile control tools. In addition, substantial progress has been made in understanding the physics of the Quiescent Double Barrier (QDB) regime, increasing the demonstrated operating space for the regime and improving performance. Highlights include: (1) A clear demonstration of q-profile modification using electron cyclotron current drive (ECCD); (2) Successful use of localized profile control using electron cyclotron heating (ECH) or ECCD to reduce central high-Z impurity accumulation associated with density peaking; (3) Theory based modeling codes are now being used to design experiments; (4) The operating space for Quiescent H-mode (QH-mode) has been substantially broadened, in particular higher density operation has been achieved; (5) Both absolute (β≤ 3.8%, neutron rate S n ≤ 5.5x10 15 s -1 ) and relative (β N H 89 = 7 for 10τ E ) performance has been increased; (6) With regard to sustainment, QDB plasmas have been run for 3.8 s or 26 τ E . These results emphasize that it is possible to produce sustained high quality H-mode performance with an edge localized mode (ELM)-free edge, directly addressing a major issue in fusion research, of how to ameliorate or eliminate ELM induced pulsed divertor particle and heat loads. (author)

  3. A NEW SYSTEM TO MONITOR DATA ANALYSES AND RESULTS OF PHYSICS DATA VALIDATION BETWEEN PULSES AT DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    FLANAGAN,A; SCHACHTER,J.M; SCHISSEL,D.P

    2003-02-01

    A Data Analysis Monitoring (DAM) system has been developed to monitor between pulse physics analysis at the DIII-D National Fusion Facility (http://nssrv1.gat.com:8000/dam). The system allows for rapid detection of discrepancies in diagnostic measurements or the results from physics analysis codes. This enables problems to be detected and possibly fixed between pulses as opposed to after the experimental run has concluded thus increasing the efficiency of experimental time. An example of a consistency check is comparing the experimentally measured neutron rate and the expected neutron emission, RDD0D. A significant difference between these two values could indicate a problem with one or more diagnostics, or the presence of unanticipated phenomena in the plasma. This new system also tracks the progress of MDSplus dispatched data analysis software and the loading of analyzed data into MDSplus. DAM uses a Java Servlet to receive messages, CLIPS to implement expert system logic, and displays its results to multiple web clients via HTML. If an error is detected by DAM, users can view more detailed information so that steps can be taken to eliminate the error for the next pulse.

  4. Increased electron temperature turbulence during suppression of edge localized mode by resonant magnetic perturbations in the DIII-D tokamak

    Science.gov (United States)

    Sung, C.; Wang, G.; Rhodes, T. L.; Smith, S. P.; Osborne, T. H.; Ono, M.; McKee, G. R.; Yan, Z.; Groebner, R. J.; Davis, E. M.; Zeng, L.; Peebles, W. A.; Evans, T. E.

    2017-11-01

    The first observation of increased electron temperature turbulence during edge localized mode (ELM) suppression by resonant magnetic perturbations (RMPs) is presented. These are long wavelength fluctuations (kθρs ≤ 0.2, where kθ = poloidal wavenumber and ρs = ion sound gyroradius) observed during H-mode plasmas on the DIII-D. This increase occurs only after ELMs are suppressed and are not observed during the initial RMP application. The T˜ e/Te increases ( >60%) are coincident with changes in normalized density and electron temperature gradients in the region from the top of the pedestal outward to the upper portion of the steep edge gradient. Density turbulence (kθρs ≤ 0.4) in this location was also observed to increase only after ELM suppression. These results are significant since they indicate that increased gradient-driven turbulent transport is one possible mechanism to regulate and maintain ELM-free H-mode operation. Investigation of linear stability of drift wave instabilities using the CGYRO code [Candy et al., J. Comput. Phys. 324, 73 (2016)] shows that the dominant mode moves closer to the electron mode branch from the ion mode branch only after ELMs are suppressed, correlated with the increased turbulence. The increased turbulence during ELM suppression, rather than with the initial RMP application, indicates that the often observed RMP induced "density pump-out" cannot be attributed to long wavelength edge turbulence level changes.

  5. A study for the installation of the TEXT heavy-ion beam probe on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Edmonds, P.H.; Solano, E.R.; Bravenec, R.V.; Wootton, A.J. [Fusion Research Center, University of Texas, Austin, Texas 78712 (United States); Schoch, P.M.; Crowley, T.P.; Hickok, R.L. [Rensselaer Polytechnic Institute, Troy, New York 12180 (United States); West, W.P.; Leuer, J.; Anderson, P. [General Atomics, San Diego, California 92186 (United States)

    1997-01-01

    An assessment of the feasibility of installing the TEXT 2 MeV heavy-ion beam probe on the DIII-D tokamak has been completed. Detailed drawings of the machine cross section were imported into the CAD application AutoCAD. A set of programs written in AutoLisp were used to generate trajectories. Displays of the accessible cross section of the plasma, scan lines for the entire range of primary beam energy and injection angle ranges, and sample{endash}volume dimensions can be rapidly generated. Because of the large deflection between the primary input beam and the emergent secondary beam, either the analyzer needs to be tracked over a {plus_minus}20{degree} angle or secondary poloidal deflector plates need to be installed at the exit port. Toroidal deflector plates will be installed at both the injection and exit ports to compensate for toroidal displacements and deflections. The sample volumes generated by this procedure are within a few centimeters of the locations derived from a full three-dimensional calculation.{copyright} {ital 1997 American Institute of Physics.}

  6. A study for the installation of the TEXT heavy-ion beam probe on DIII-D

    Science.gov (United States)

    Edmonds, P. H.; Solano, E. R.; Bravenec, R. V.; Wootton, A. J.; Schoch, P. M.; Crowley, T. P.; Hickok, R. L.; West, W. P.; Leuer, J.; Anderson, P.

    1997-01-01

    An assessment of the feasibility of installing the TEXT 2 MeV heavy-ion beam probe on the DIII-D tokamak has been completed. Detailed drawings of the machine cross section were imported into the CAD application AutoCAD. A set of programs written in AutoLisp were used to generate trajectories. Displays of the accessible cross section of the plasma, scan lines for the entire range of primary beam energy and injection angle ranges, and sample-volume dimensions can be rapidly generated. Because of the large deflection between the primary input beam and the emergent secondary beam, either the analyzer needs to be tracked over a ±20° angle or secondary poloidal deflector plates need to be installed at the exit port. Toroidal deflector plates will be installed at both the injection and exit ports to compensate for toroidal displacements and deflections. The sample volumes generated by this procedure are within a few centimeters of the locations derived from a full three-dimensional calculation.

  7. A study for the installation of the TEXT heavy-ion beam probe on DIII-D

    International Nuclear Information System (INIS)

    Edmonds, P.H.; Solano, E.R.; Bravenec, R.V.; Wootton, A.J.; Schoch, P.M.; Crowley, T.P.; Hickok, R.L.; West, W.P.; Leuer, J.; Anderson, P.

    1997-01-01

    An assessment of the feasibility of installing the TEXT 2 MeV heavy-ion beam probe on the DIII-D tokamak has been completed. Detailed drawings of the machine cross section were imported into the CAD application AutoCAD. A set of programs written in AutoLisp were used to generate trajectories. Displays of the accessible cross section of the plasma, scan lines for the entire range of primary beam energy and injection angle ranges, and sample endash volume dimensions can be rapidly generated. Because of the large deflection between the primary input beam and the emergent secondary beam, either the analyzer needs to be tracked over a ±20 degree angle or secondary poloidal deflector plates need to be installed at the exit port. Toroidal deflector plates will be installed at both the injection and exit ports to compensate for toroidal displacements and deflections. The sample volumes generated by this procedure are within a few centimeters of the locations derived from a full three-dimensional calculation.copyright 1997 American Institute of Physics

  8. A NEW SYSTEM TO MONITOR DATA ANALYSES AND RESULTS OF PHYSICS DATA VALIDATION BETWEEN PULSES AT DIII-D

    International Nuclear Information System (INIS)

    FLANAGAN, A; SCHACHTER, J.M; SCHISSEL, D.P

    2003-01-01

    A Data Analysis Monitoring (DAM) system has been developed to monitor between pulse physics analysis at the DIII-D National Fusion Facility (http://nssrv1.gat.com:8000/dam). The system allows for rapid detection of discrepancies in diagnostic measurements or the results from physics analysis codes. This enables problems to be detected and possibly fixed between pulses as opposed to after the experimental run has concluded thus increasing the efficiency of experimental time. An example of a consistency check is comparing the experimentally measured neutron rate and the expected neutron emission, RDD0D. A significant difference between these two values could indicate a problem with one or more diagnostics, or the presence of unanticipated phenomena in the plasma. This new system also tracks the progress of MDSplus dispatched data analysis software and the loading of analyzed data into MDSplus. DAM uses a Java Servlet to receive messages, CLIPS to implement expert system logic, and displays its results to multiple web clients via HTML. If an error is detected by DAM, users can view more detailed information so that steps can be taken to eliminate the error for the next pulse

  9. Improved cross-calibration of Thomson scattering and electron cyclotron emission with ECH on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Brookman, M. W., E-mail: brookmanmw@fusion.gat.com [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States); Austin, M. E.; McLean, A. G. [Lawrence Livermore National Lab, Livermore, California 94500 (United States); Carlstrom, T. N.; Hyatt, A. W.; Lohr, J. [General Atomics, San Diego, California 92122 (United States)

    2016-11-15

    Thomson scattering produces n{sub e} profiles from measurement of scattered laser beam intensity. Rayleigh scattering provides a first calibration of the relation n{sub e} ∝ I{sub TS}, which depends on many factors (e.g., laser alignment and power, optics, and measurement systems). On DIII-D, the n{sub e} calibration is adjusted against an absolute n{sub e} from the density-driven cutoff of the 48 channel 2nd harmonic X-mode electron cyclotron emission system. This method has been used to calibrate Thomson n{sub e} from the edge to near the core (r/a > 0.15). Application of core electron cyclotron heating improves the quality of cutoff and depth of its penetration into the core, and also changes underlying MHD activity, minimizing crashes which confound calibration. Less fueling is needed as “ECH pump-out” generates a plasma ready to take up gas. On removal of gyrotron power, cutoff penetrates into the core as channels fall successively and smoothly into cutoff.

  10. CAMAC throughput of a new RISC-based data acquisition computer at the DIII-D tokamak

    International Nuclear Information System (INIS)

    VanderLaan, J.F.; Cummings, J.W.

    1993-10-01

    The amount of experimental data acquired per plasma discharge at DIII-D has continued to grow. The largest shot size in May 1991 was 49 Mbyte; in May 1992, 66 Mbyte; and in April 1993, 80 Mbyte. The increasing load has prompted the installation of a new Motorola 88100-based MODCOMP computer to supplement the existing core of three older MODCOMP data acquisition CPUs. New Kinetic Systems CAMAC serial highway driver hardware runs on the 88100 VME bus. The new operating system is MODCOMP REAL/IX version of AT ampersand T System V UNIX with real-time extensions and networking capabilities; future plans call for installation of additional computers of this type for tokamak and neutral beam control functions. Experiences with the CAMAC hardware and software will be chronicled, including observation of data throughput. The Enhanced Serial Highway crate controller is advertised as twice as fast as the previous crate controller, and computer I/O speeds are expected to also increase data rates

  11. First measurements of the ion energy distribution at the divertor strike point during DIII-D disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Parks, P.B.; Brooks, N.H.; West, W.P.; Wong, C.P.C. [General Atomics, San Diego, CA (United States); Bastasz, R.; Wampler, W.R. [Sandia National Labs., Albuquerque, NM (United States); Whyte, D. [Inst. National de la Recherche Scientifique, Varennes, Quebec (Canada)

    1996-03-01

    Plasma disruptions are a serious concern in tokamak design because of the high impulsive heat loads which can cause strong erosion of divertor materials due to enhanced sputtering, or melting/ablation in the most severe cases. Predictions of net erosion rates and hence component lifetimes are very difficult and are highly dependent on the plasma conditions over the divertor target. It is therefore necessary to characterize the properties of the scrape-off plasma near the divertor target plate under these special conditions. Here, plasma/wall interaction studies are being carried out using the Divertor Materials Exposure System (DiMES) on DIII-D. The objective of the experiment is to determine the kinetic energy and flux of deuterium ions reaching the divertor target during argon-induced radiative disruptions. The experiment utilizes a special slotted ion analyzer mounted over a Si sample to collect the fast charge-exchange (CX) deuterium neutrals emitted within the recycled cold neutral layer (CNL) which serves as a CX target for the incident ions. A theoretical interpretation of the experiment reveals a strong forward pitch-angle dependence in the approaching ion distribution function. The depth distribution of the trapped D in the Si sample was measured using low-energy direct recoil spectroscopy. Comparison with the TRIM code using monoenergetic ions indicated that the best fit to the data was obtained for an ion energy of 100 eV. An estimate of the CNL thickness {integral}nd{ell} indicates that during disruptions the CNL cushion is thick enough to reduce the local ion heat load by {approximately}30% due to CX refluxing.

  12. 4 MW upgrade to the DIII-D fast wave current drive system

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-10-01

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

  13. Environmental Assessment for the proposed modification and continued operation of the DIII-D facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The EA evaluates the proposed action of modifying the DIII-D fusion facility and conducting related research activities at the GA San Diego site over 1995-1999 under DOE contract number DE-ACO3-89ER51114. The proposed action is need to advance magnetic fusion research for future generation fusion devices such as ITER and TPX. It was determined that the proposed action is not a major action significantly affecting the quality of the human environment according to NEPA; therefore a finding of no significant impact is made and an environmental impact statement is not required.

  14. Environmental Assessment for the proposed modification and continued operation of the DIII-D facility

    International Nuclear Information System (INIS)

    1995-07-01

    The EA evaluates the proposed action of modifying the DIII-D fusion facility and conducting related research activities at the GA San Diego site over 1995-1999 under DOE contract number DE-ACO3-89ER51114. The proposed action is need to advance magnetic fusion research for future generation fusion devices such as ITER and TPX. It was determined that the proposed action is not a major action significantly affecting the quality of the human environment according to NEPA; therefore a finding of no significant impact is made and an environmental impact statement is not required

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

    International Nuclear Information System (INIS)

    McHarg, B.B. Jr.

    1993-10-01

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

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

    Science.gov (United States)

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

    2008-10-01

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

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

    Science.gov (United States)

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

    2008-10-01

    Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m(2) for 10 s exposures have been installed in the lower divertor target plates of the DIII-D tokamak. The 20 pyrolytic graphite probe tips have more than three times higher thermal conductivity and 16 times larger mass than the original DIII-D isotropic graphite probes. The probe tips have a fixed 12.5 degree surface angle to distribute the heat flux more uniformly than the previous 6 mm diameter domed collectors and a symmetric "rooftop" design to allow operation with reversed toroidal magnetic field. A large spring-loaded contact area improves heat conduction from each probe tip through a ceramic insulator into a cooled graphite divertor floor tile. The probe tips, brazed to molybdenum foil to ensure good electrical contact, are mounted in a ceramic tray for electrical isolation and reliable cable connections. The new probes are located 1.5 cm radially apart in a staggered arrangement near the entrance to the lower divertor pumping baffle and are linearly spaced 3 cm apart on the shelf above the in-vessel cryopump. Typical target plate profiles of J(sat), T(e), and V(f) with 4 mm spatial resolution are shown.

  18. 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

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

    International Nuclear Information System (INIS)

    Fenstermacher, M.E.; Meyer, W.H.; Wood, R.D.; Nilson, D.G.; Ellis, R.; Brooks, N.H.

    1997-01-01

    A video camera system has been installed on the DIII-D tokamak for two-dimensional spatial studies of line emission in the lower divertor region. The system views the divertor tangentially at approximately the height of the X point through an outer port. At the tangency plane, the entire divertor from the inner wall to outside the DIII-D bias ring is viewed with spatial resolution of ∼1 cm. The image contains information from ∼90 deg of toroidal angle. In a recent upgrade, remotely controllable filter changers were added which have produced images from nominally identical discharges using different spectral lines. Software was developed to calculate the response function matrix of the optical system using distributed computing techniques and assuming toroidal symmetry. Standard sparse matrix algorithms are then used to invert the three-dimensional images onto a poloidal plane. Spatial resolution of the inverted images is 2 cm; higher resolution simply increases the size of the response function matrix. Initial results from a series of experiments with multiple identical discharges show that the emission from CII and CIII, which appears along the inner scrape-off layer above and below the X point during ELMing H mode, moves outward and becomes localized near the X point in radiative divertor operation induced by deuterium injection. copyright 1997 American Institute of Physics

  20. Current drive with fast waves, electron cyclotron waves, and neutral injection in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Prater, R.; Petty, C.C.; Pinsker, R.I.

    1993-01-01

    Current drive experiments have been performed on the DIII-D tokamak using fast waves, electron cyclotron waves, and neutral injection. Fast wave experiments were performed using a 4-strap antenna with 1 MW of power at 60 MHz. These experiments showed effective heating of electrons, with a global heating efficiency equivalent to that of neutral injection even when the single pass damping was calculated to be as small as 5%. The damping was probably due to the effect of multiple passes of the wave through the plasma. Fast wave current drive experiments were performed with a toroidally directional phasing of the antenna straps. Currents driven by fast wave current drive (FWCD) in the direction of the main plasma current of up to 100 kA were found, not including a calculated 40 kA of bootstrap current. Experiments with FWCD in the counter current direction showed little current drive. In both cases, changes in the sawtooth behavior and the internal inductance qualitatively support the measurement of FWCD. Experiments on electron cyclotron current drive have shown that 100 kA of current can be driven by 1 MW of power at 60 GHz. Calculations with a Fokker-Planck code show that electron cyclotron current drive (ECCD) can be well predicted when the effects of electron trapping and of the residual electric field are included. Experiments on driving current with neutral injection showed that effective current drive could be obtained and discharges with full current drive were demonstrated. Interestingly, all of these methods of current drive had about the same efficiency. (Author)

  1. Modeling of noble gas injection into tokamak plasmas

    International Nuclear Information System (INIS)

    Morozov, D.Kh.; Yurchenko, E.I.; Lukash, V.E.; Baronova, E.O.; Rozhansky, V.A.; Senichenkov, I.Yu.; Veselova, I.Yu.; Schneider, R.

    2005-01-01

    Noble gas injection for mitigation of the disruption in DIII-D is simulated. The simulation of the first two stages is performed: of the neutral gas jet penetration through the background plasmas, and of the thermal quench. In order to simulate the first stage the 1.5-dimensional numerical code LLP with improved radiation model for noble gas is used. It is demonstrated that the jet remains mainly neutral and thus is able to penetrate to the central region of the tokamak in accordance with experimental observations. Plasma cooling at this stage is provided by the energy exchange with the jet. The radiation is relatively small, and the plasma thermal energy is spent mainly on the jet expansion. The magnetic surfaces in contact with the jet are cooled significantly. The cooling front propagates towards the plasma center. The simulations of the plasma column dynamics in the presence of moving jet is performed by means of the free boundary transport modeling DINA code. It has been shown that the cooling front is accompanied by strongly localized 'shark fin-like' perturbation in toroidal current density profile. After few milliseconds the jet (together with the current perturbation) achieves the region where safety factor is slightly higher than unity and a new type of the non-local kink mode develops. The unstable kink perturbation is non-resonant for any magnetic surface, both inside the plasma column, and in the vacuum space. The mode disturbs mainly the core region. The growth time of the 'shark fin-like' mode is higher than the Alfven time by a factor of 100 for DIII-D parameters. Hence, the simulation describes the DIII-D experimental results, at least, qualitatively. (author)

  2. 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

  3. Automatic determination of L/H transition times in DIII-D through a collaborative distributed environment

    International Nuclear Information System (INIS)

    Farias, G.; Vega, J.; González, S.; Pereira, A.; Lee, X.; Schissel, D.; Gohil, P.

    2012-01-01

    Highlights: ► An automatic predictor of L/H transition times has been implemented for the DIII-D tokamak. ► The system predicts the transition combining two techniques: a morphological pattern recognition algorithm and a support vector machines multi-layer model. ► The predictor is employed within a collaborative distributed computing environment. The system is trained remotely in the Ciemat computer cluster and operated on the DIII-D site. - Abstract: An automatic predictor of L/H transition times has been implemented for the DIII-D tokamak. The system predicts the transition combining two techniques: A morphological pattern recognition algorithm, which estimates the transition based on the waveform of a Dα emission signal, and a support vector machines multi-layer model, which predicts the L/H transition using a non-parametric model. The predictor is employed within a collaborative distributed computing environment. The system is trained remotely in the Ciemat computer cluster and operated on the DIII-D site.

  4. First measurements of electron temperature and density with divertor Thomson Scattering in radiative divertor discharges on DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Hill, D.N.; Carlstrom, T.N.; Nilson, D.G.

    1996-10-01

    We have obtained the first measurements of n e and T e in the DIII-D divertor region with a multi-pulse (20 Hz) Divertor Thomson Scattering (DTS) system. Eight measurement locations are distributed vertically up to 21 cm above the divertor plate. Two-dimensional distributions have been obtained by sweeping the divertor plasma across the DTS measurement location. Several operating modes have been studied, including ohmic, L-mode, Elming H-mode, and Radiative Divertor operation with puffing of D 2 and impurities. Mapping of the data to either the (L pol , φ) or (R, Z) planes with the EFIT equilibrium is used to analyze the 2D profiles. We find that in ELMing H-mode: n e , T e , and P e are relatively constant along field lines from the X-point to the divertor plate, especially near the separatrix field line. With D 2 puffing, the DTS profiles indicate that T e in a large part of divertor region below the X-point is dramatically reduced from ∼30-40 eV in ELMing H-mode to 1-2 eV. This results in a fairly uniform low-T e divertor, with an increased electron density in the range of 2 to 4 x 10 20 m -3 . Detailed comparisons of the spatial profiles of n e , T e , and electron pressure P e , are presented for several operating modes. In addition, these data are compared with initial calculations from the UEDGE fluid code

  5. Advances in the Understanding of ELM Suppression by Resonant Magnetic Perturbations (RMPs) in DIII-D and Implications for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Nazikian, R. [PPPL

    2014-09-01

    Experiments on DIII-D have expanding the operating window for RMP ELM suppression to higher q95 with dominant electron heating and fully non-inductive current drive relevant to advanced modes of ITER operation. Robust ELM suppression has also been obtained with a reduced coil set, mitigating the risk of coil failure in maintaining ELM suppression in ITER. These results significantly expand the operating space and reduce risk for obtaining RMP ELM suppression in ITER. Efforts have also been made to search for 3D cause of ELM suppression. No internal non-axisymmetric structure is detected at the top of the pedestal, indicating that the dominant effect of the RMP is to produce an n=0 transport modification of the profiles. Linear two fluid MHD simulations using M3D-C1 indicate resonant field penetration and significant magnetic stochasticity at the top of the pedestal, consistent with the absence of detectable 3D structure in that region. A profile database was developed to compare the scaling of the pedestal and global confinement with the applied 3D field strength in ELM suppressed and ELM mitigated plasmas. The EPED pedestal model accurately predicts the measured pedestal pressure at the threshold of ELM suppression, increasing confidence in theoretical projections to ITER pedestal conditions. Both the H-factor (H(sub)98y2) and thermal energy confinement time do not degrade substantially with applied RMP fields near the threshold of ELM suppression, enhancing confidence in the compatibility of ITER high performance operation with RMP ELM suppression.

  6. Kinetic simulations of scrape-off layer physics in the DIII-D tokamak

    Directory of Open Access Journals (Sweden)

    R.M. Churchill

    2017-08-01

    The XGCa simulation of the DIII-D tokamak in a nominally sheath-limited regime show many noteworthy features in the SOL. The density and ion temperature are higher at the low-field side, indicative of ion orbit loss. The SOL ion Mach flows are at experimentally relevant levels (Mi ∼ 0.5, with similar shapes and poloidal variation as observed in various tokamaks. Surprisingly, the ion Mach flows close to the sheath edge remain subsonic, in contrast to the typical fluid Bohm criterion requiring ion flows to be above sonic at the sheath edge. Related to this are the presence of elevated sheath potentials, eΔΦ/Te∼3−4, over most of the SOL, with regions in the near-SOL close to the separatrix having eΔΦ/Te > 4. These two results at the sheath edge are a consequence of non-Maxwellian features in the ions and electrons there.

  7. 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

  8. Hydrogen isotopes retention in divertor tiles of DIII-D tokamak

    International Nuclear Information System (INIS)

    Skorodumov, B.G.; Buzhinskij, O.I.; West, W.P.; Ulanov, V.G.

    1996-01-01

    The absolute concentration of hydrogen isotopes in graphite divertor tiles coated with boron carbide after the exposure in DIII-D during 16 operational weeks of the 1993 campaign was obtained using the 14 MeV neutron-induced recoil detection (NERD) method. It is shown that the absolute concentration of H in tile's surface layers correlates with thickness of the deposited layers. The graphite tile without boron carbide coating had a H concentration similar to that of the tile with the thickest deposited layer. Deuterium and tritium were not detected in any of the investigated tiles. The proposed method can be used for the determination of the thickness of coatings without sample destruction. Thus, the thickness of boron carbide coatings on the tiles obtained with this method varied from 80 to 115 μm, which corresponded well to electron microscope data. (orig.)

  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. 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)

  11. Optical boundary reconstruction of tokamak plasmas for feedback control of plasma position and shape

    NARCIS (Netherlands)

    Hommen, G.; de M. Baar,; Nuij, P.; McArdle, G.; Akers, R.; Steinbuch, M.

    2010-01-01

    A new diagnostic is developed to reconstruct the plasma boundary using visible wavelength images. Exploiting the plasma's edge localized and toroidally symmetric emission profile, a new coordinate transform is presented to reconstruct the plasma boundary from a poloidal view image. The plasma

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

    International Nuclear Information System (INIS)

    Lohr, J.; Calahan, P.; Callis, R.W.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-07-01

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

  14. Nonlinear interaction of the surface waves at a plasma boundary

    International Nuclear Information System (INIS)

    Dolgopolov, V.V.; El-Naggar, I.A.; Hussein, A.M.; Khalil, Sh.M.

    1976-01-01

    Amplitudes of electromagnetic waves with combination frequencies, radiating from the plasma boundary due to nonlinear interaction of the surface waves, have been found. Previous papers on this subject did not take into account that the tangential components of the electric field of waves with combination frequencies were discontinuous at the plasma boundary. (Auth.)

  15. Reduction of edge localized mode intensity on DIII-D by on-demand triggering with high frequency pellet injection and implications for ITERa)

    Science.gov (United States)

    Baylor, L. R.; Commaux, N.; Jernigan, T. C.; Meitner, S. J.; Combs, S. K.; Isler, R. C.; Unterberg, E. A.; Brooks, N. H.; Evans, T. E.; Leonard, A. W.; Osborne, T. H.; Parks, P. B.; Snyder, P. B.; Strait, E. J.; Fenstermacher, M. E.; Lasnier, C. J.; Moyer, R. A.; Loarte, A.; Huijsmans, G. T. A.; Futatani, S.

    2013-08-01

    The injection of small deuterium pellets at high repetition rates up to 12× the natural edge localized mode (ELM) frequency has been used to trigger high-frequency ELMs in otherwise low natural ELM frequency H-mode deuterium discharges in the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)]. The resulting pellet-triggered ELMs result in up to 12× lower energy and particle fluxes to the divertor than the natural ELMs. The plasma global energy confinement and density are not strongly affected by the pellet perturbations. The plasma core impurity density is strongly reduced with the application of the pellets. These experiments were performed with pellets injected from the low field side pellet in plasmas designed to match the ITER baseline configuration in shape and normalized β operation with input heating power just above the H-mode power threshold. Nonlinear MHD simulations of the injected pellets show that destabilization of ballooning modes by a local pressure perturbation is responsible for the pellet ELM triggering. This strongly reduced ELM intensity shows promise for exploitation in ITER to control ELM size while maintaining high plasma purity and performance.

  16. Experimentally-based ExB drifts in the DIII-D divertor and SOL calculated from integration of Ohm's law using Thomson scattering measurements of Te and ne

    Directory of Open Access Journals (Sweden)

    P.C. Stangeby

    2017-08-01

    Full Text Available The 2D spatial distributions of cross field drift velocities are calculated from 2D Thomson scattering measurements of Te and ne in the divertor and SOL of DIII-D. In contrast with the method that has been used on DIII-D where the 2D distribution of plasma potential Vplasma is obtained from measurements of the probe floating potential of reciprocating probes, the present method does not require insertion of a probe into the plasma and can therefore be used in high power discharges. The 2D spatial distribution of Vplasma is calculated from Ohm's Law for the parallel electric field E|| along each flux tube, E∥(s∥=−1.71dTe/ds∥−(Te/nedne/ds∥, where the Thomson scattering values of Te and ne are used. To within a constant of integration, Vplasma is obtained by integrating E|| along the flux-tubes (field lines; the constant is obtained for each flux tube using the sheath drop at the target calculated from the characteristic of Langmuir probes built into the divertor tiles. The 2D distributions of Eradial=−dVp/dsradial, Epoloidal=−dVp/dspoloidal, vExBpoloidal=Eradial/B and vExBradial=Epoloidal/B are then calculated as well as the particle drift flux densities ΓExBpoloidal=nvExBpoloidal and ΓExBradial=nvExBradial for electrons, fuel ions and impurity ions, using the appropriate values of particle density, n.

  17. Error field mode studies on JET, COMPASS-D and DIII-D, and implications for ITER

    International Nuclear Information System (INIS)

    Buttery, R.J.; Gates, D.A.; De Benedetti, M.

    1999-01-01

    New experiments on COMPASS-D, DIII-D and JET have identified the critical scalings of error field sensitivity and harmonic content effects, enabling predictions of the requirements for larger devices such as ITER. Thresholds are lowest at low density, a regime proposed for H mode access on ITER. Results suggest a moderate error field sensitivity (δB/B∼10 -4 ) for ITER, comparable with the size of its intrinsic error, although there are uncertainties in scaling behaviour. Other studies on COMPASS-D and DIII-D show that sideband harmonics to the (2,1) component play an important role. Thus a correction system for ITER will be important, with flexibility to correct sidebands desirable, possibly assisted by beam rotation. Such a system has been designed and is capable of reducing multiple harmonic error levels to ∼2x10 -5 . (author)

  18. Error field mode studies on JET, COMPASS-D and DIII-D, and implications for ITER

    International Nuclear Information System (INIS)

    Buttery, R.; Gates, D.; Benedetti, M. de

    2001-01-01

    New experiments on COMPASS-D, DIII-D and JET have identified the critical scalings of error field sensitivity and harmonic content effects, enabling predictions of the requirements for larger devices such as ITER. Thresholds are lowest at low density, a regime proposed for H mode access on ITER. Results suggest a moderate error field sensitivity (δB/B∼10 -4 ) for ITER, comparable with the size of its intrinsic error, although there are uncertainties in scaling behaviour. Other studies on COMPASS-D and DIII-D show that sideband harmonics to the (2,1) component play an important role. Thus a correction system for ITER will be important, with flexibility to correct sidebands desirable, possibly assisted by beam rotation. Such a system has been designed and is capable of reducing multiple harmonic error levels to ∼2x10 -5 . (author)

  19. Impact of environmental regulations on control of copper ion concentration in the DIII-D cooling water system

    International Nuclear Information System (INIS)

    Gootgeld, A.M.

    1993-10-01

    Tokamaks and industrial users are faced with the task of maintaining closed-loop, low conductivity, low impurity, cooling water systems. Operating these systems concentrates the impurities in the water requiring subsequent disposal. Environmental regulations are making this increasingly difficult. This paper will discuss the solution to the problem of removing and disposing of copper ions in the DIII-D low conductivity water system. Since the commissioning of the Doublet facility, the quality of the water in the 3000 gpm system that cools the DIII-D vacuum vessel coils, power supplies and auxiliary heating components has been controlled with mixed-bed ion exchangers. Low ion levels, particularly copper, are required to operate this equipment. In early 1992, the company that leases and regenerates DIII-D ion exchangers said they no longer can accept these resin beds for regeneration due to the level of copper ion on the resin. This change in policy, a change that has been adopted throughout their industry, was necessary to assure that the Metropolitan Sewerage System of the City of San Diego stays in compliance with State of California regulations and EPA-mandated national pretreatment standards and regulations. A cost effective solution was implemented which utilizes a reverse osmosis filtration system with the ion exchangers for make-up water. Levels of copper ion disposed to the sewer are in compliance with government standards. These measures have thus far proved effective in maintaining low conductivity and overall good quality cooling water. Specifically, this paper discusses DIII-D deionized cooling water quality requirements and an affective means to meet these requirements in order to be in compliance with government regulations for copper ion disposal. The problems discussed, the alternatives considered and the approach taken would be readily applicable to any deionized cooling water system containing copper where EPA standards and regulations are mandated

  20. Video digitizer (real time-frame grabber) with region of interest suitable for quantitative data analysis used on the infrared and H alpha cameras installed on the DIII-D experiment

    International Nuclear Information System (INIS)

    Ferguson, S.W.; Kevan, D.K.; Hill, D.N.; Allen, S.L.

    1987-01-01

    This paper describes a CAMAC based video digitizer with region of interest (ROI) capability that was designed for use with the infrared and H alpha cameras installed by Lawrence Livermore Laboratory on the DIII-D experiment at G.A. Technologies in San Diego, California. The video digitizer uses a custom built CAMAC video synchronizer module to clock data into a CAMAC transient recorder on a line-by-line basis starting at the beginning of a field. The number of fields that are recorded is limited only by the available transient recorder memory. In order to conserve memory, the CAMAC video synchronizer module provides for the alternative selection of a specific region of interest in each successive field to be recorded. Memory conservation can be optimized by specifying lines in the field, start time, stop time, and the number of data samples per line. This video frame grabber has proved versatile for capturing video in such diverse applications as recording video fields from a video tape recorder played in slow motion or recording video fields in real time during a DIII-D shot. In other cases, one or more lines of video are recorded per frame to give a cross sectional slice of the plasma. Since all the data in the digitizer memory is synchronized to video fields and lines, the data can be read directly into the control computer in the proper matrix format to facilitate rapid processing, display, and permanent storage

  1. Status of the Linux PC cluster for between-pulse data analyses at DIII-D

    International Nuclear Information System (INIS)

    Peng, Q.; Groebner, R.J.; Lao, L.L.; Schachter, J.; Schissel, D.P.; Wade, M.R.

    2002-01-01

    Some analyses that survey experimental data are carried out at a sparse sample rate between pulses during tokamak operation and/or completed as a batch job overnight because the complete analysis on a single fast workstation cannot fit in the narrow time window between two pulses. Scientists therefore miss the opportunity to use these results to guide experiments quickly. With a dedicated Beowulf type cluster at a cost less than that of a workstation, these analyses can be accomplished between pulses and the analyzed data made available for the research team during the tokamak operation. A Linux PC cluster comprised of 12 processors was installed at DIII-D National Fusion Facility in CY00 and expanded to 24 processors in CY01 to automatically perform between-pulse magnetic equilibrium reconstructions using the EFIT code written in FORTRAN, CER analyses using CERQUICK code written in interactive data language (IDL) and full profile fitting analyses (n e , T e , T i , V r , Z eff ) using IDL code ZIPFIT. This paper reports the current status of the system, the details of the between-pulse profile fitting analyses, and discusses some problems and concerns raised during the implementation and expansion of the system

  2. Use of open systems for control, analysis, and data acquisition of the DIII-D tokamak

    International Nuclear Information System (INIS)

    Henline, P.A.

    1993-10-01

    For the past several years, it has been evident that the very old MODCOMP 16-bit computers being used at DIII-D for control and data acquisition were no longer adequate to perform the services needed. In early 1992, the computer systems group began to look seriously into alternate systems to replace these aged MODCOMP systems. The decision was made to investigate open-quote OPEN close-quote system computers and also to maintain the compatibility with the large usage of CAMAC equipment as the real-time hardware interface. Information about the needs for real-time capabilities and open-quote OPEN close-quote systems ability to meet these needs is discussed. The needs include hardware requirements, operating system software which has known response rates, interconnectability and access of data from other workstations and computers. Some of the parameters and pitfalls of open systems are discussed as well as the advantages of OPEN systems for use in a real-time environment. The success at arriving at an OPEN systems solution is examined

  3. DISSOLVED OXYGEN REDUCTION IN THE DIII-D NEUTRAL BEAM ION SOURCE COOLING SYSTEM

    International Nuclear Information System (INIS)

    YIP, H.; BUSATH, J.; HARRISON, S.

    2004-03-01

    OAK-B135 Neutral beam ion sources (NBIS) are critical components for the neutral beam injection system supporting the DIII-D tokamak. The NBIS must be cooled with 3028 (ell)/m (800 gpm) of de-ionized and de-oxygenated water to protect the sources from overheating and failure. These ions sources are currently irreplaceable. Since the water cooled molybdenum components will oxidize in water almost instantaneously in the presence of dissolved oxygen (DO), de-oxygenation is extremely important in the NBIS water system. Under normal beam operation the DO level is kept below 5 ppb. However, during weeknights and weekends when neutral beam is not in operation, the average DO level is maintained below 10 ppb by periodic circulation with a 74.6 kW (100 hp) pump, which consumes significant power. Experimental data indicated evidence of continuous oxygen diffusion through non-metallic hoses in the proximity of the NBIS. Because of the intermittent flow of the cooling water, the DO concentration at the ion source(s) could be even higher than measured downstream, and hence the concern of significant localized oxidation/corrosion. A new 3.73 kW (5 hp) auxiliary system, installed in the summer of 2003, is designed to significantly reduce the peak and the time-average DO levels in the water system and to consume only a fraction of the power

  4. Neural Network Computed Bootstrap Current for Real Time Control in DIII-D

    Science.gov (United States)

    Tema Biwole, Arsene; Smith, Sterling P.; Meneghini, Orso; Belli, Emily; Candy, Jeff

    2017-10-01

    In an effort to provide a fast and accurate calculation of the bootstrap current density for use as a constraint in real-time equilibrium reconstructions, we have developed a neural network (NN) non-linear regression of the NEO code calculated bootstrap current jBS. A new formulation for jBS in NEO allows for a determination of the coefficients on the density and temperature scale lengths. The new formulation reduces the number of inputs to the NN, and the number of output coefficients is 2 times the number of species (including electrons). The NN can reproduce the NEO and Sauter coefficients to a high degree of accuracy (bootstrap current density calculated in NEO has been used as a constraint in an offline equilibrium reconstruction for comparison to the NN calculation. The computational time of this method (μs) makes it ideal for real time calculation in DIII-D. Work supported by US DOE under DE-FC02-04ER54698, DE-FG2-95ER-54309, DE-SC 0012656, DE-FC02-06ER54873.

  5. STATUS OF THE LINUX PC CLUSTER FOR BETWEEN-PULSE DATA ANALYSES AT DIII-D

    International Nuclear Information System (INIS)

    PENG, Q; GROEBNER, R.J; LAO, L.L; SCHACHTER, J.; SCHISSEL, D.P; WADE, M.R.

    2001-08-01

    OAK-B135 Some analyses that survey experimental data are carried out at a sparse sample rate between pulses during tokamak operation and/or completed as a batch job overnight because the complete analysis on a single fast workstation cannot fit in the narrow time window between two pulses. Scientists therefore miss the opportunity to use these results to guide experiments quickly. With a dedicated Beowulf type cluster at a cost less than that of a workstation, these analyses can be accomplished between pulses and the analyzed data made available for the research team during the tokamak operation. A Linux PC cluster comprises of 12 processors was installed at DIII-D National Fusion Facility in CY00 and expanded to 24 processors in CY01 to automatically perform between-pulse magnetic equilibrium reconstructions using the EFIT code written in Fortran, CER analyses using CERQUICK code written in IDL and full profile fitting analyses (n e , T e , T i , V r , Z eff ) using IDL code ZIPFIT. This paper reports the current status of the system and discusses some problems and concerns raised during the implementation and expansion of the system

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stoschus, H.; Hudson, B. [Oak Ridge Institute for Science and Education, Oak Ridge, Tennessee 37831-0117 (United States); Thomas, D. M.; Watkins, M.; Osborne, T. H. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Finkenthal, D. F. [Palomar College, 1140 West Mission Rd, San Marcos, California 92069-1487 (United States); Moyer, R. A. [University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States)

    2013-08-15

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

  8. Investigation of collisional effects within the bending magnet region of a DIII-D neutral beamline

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. SYSTEM DESIGN AND PERFORMANCE FOR THE RECENT DIII-D NEUTRAL BEAM COMPUTER UPGRADE

    International Nuclear Information System (INIS)

    PHILLIPS, J.C; PENAFLOR, B.G; PHAM, N.Q; PIGLOWSKI, D.A.

    2004-03-01

    OAK-B135 This operating year marks an upgrade to the computer system charged with control and data acquisition for neutral beam injection system's heating at the DIII-D National Fusion Facility, funded by the US Department of Energy and operated by General Atomics (GA). This upgrade represents the third and latest major revision to a system which has been in service over twenty years. The first control and data acquisition computers were four 16 bit mini computers running a proprietary operating system. Each of the four controlled two ion source over dedicated CAMAC highway. In a 1995 upgrade, the system evolved to be two 32 bit Motorola mini-computers running a version of UNIX. Each computer controlled four ion sources with two CAMAC highways per CPU. This latest upgrade builds on this same logical organization, but makes significant advances in cost, maintainability, and the degree to which the system is open to future modification. The new control and data acquisition system is formed of two 2 GHz Intel Pentium 4 based PC's, running the LINUX operating system. Each PC drives two CAMAC serial highways using a combination of Kinetic Systems PCI standard CAMAC Hardware Drivers and a low-level software driver written in-house expressly for this device. This paper discusses the overall system design and implementation detail, describing actual operating experience for the initial six months of operation

  10. A Handheld, Free Roaming, Data Display for DIII-D Diagnostic Data

    International Nuclear Information System (INIS)

    Broesch, J.D.; Phillips, J.C.; Petersen, P.I.; Hansink, M.J.

    1999-01-01

    Standard handheld test instruments such as voltmeters and portable oscilloscopes are useful for making basic measurements necessary for the operation and maintenance of large experiments such as the DIII-D magnetic fusion research facility. Some critical diagnostic information, however, is available only on system computers. Often this diagnostic information is located in computer databases and requires synthesis via computational algorithms to be of practical use to the technician. Unfortunately, this means the data is typically only available via computer screens located at fixed locations. One common way to provide mobile information is to have one operator sit at a console and read the data to the mobile technician via radio. This is inefficient in as much as it requires two people. Even more importantly the operator-to-technician voice link introduces significant delays and errors that may hinder response times. To address these concerns of personnel utilization and efficiency, we have developed a remote display based on an rf-data link that can be carried with a technician as he moves about the facility. This display can provide the technician with any information needed from the stationary database. This paper will discuss the overall architecture as well as the individual modules for the mobile data display. Lessons learned, as well as techniques for improving the usefulness of such systems, will be presented

  11. Electrostatic sheath at the boundary of a collisional dusty plasma

    Indian Academy of Sciences (India)

    Abstract. Considering the Boltzmann response of the ions and electrons in plasma dynamics and inertial dynamics of the dust charged grains in a highly collisional dusty plasma, the nature of the electrostatic potential near a boundary is investigated. Based on the fluid approximation, the forma- tion as well as the ...

  12. Analysis of time-dependent particle transport in the tokamak boundary plasma

    Science.gov (United States)

    Rognlien, T. D.; Groebner, R. J.; Nam, S. K.

    2010-11-01

    Plasma particle transport in the edge and scrape-off layer of tokamaks is not well understood but is important for core fueling, helium removal, and impurity intrusion. A simple 1D model is presented to clarify the time-dependent impact of different possible mechanisms including penetration and ionization of recycled or injected neutrals, and plasma diffusion and convection. More detail of edge profile dynamics between Edge-Localized-Modes (ELMs) corresponding to re-building of the pedestal region is studied with the 2D UEDGE transport code. The influence of the ion pinch associated with perpendicular ion viscosity is evaluated. It is assumed that during quasi-steady-state discharges with regularly-spaced bursts of Edge Localized Modes (ELMs), the net pumping of all walls and pumps averaged over an ELM cycle is just sufficient to remove the small neutral beam particle source. The simulation results are then compared with similar time-dependent data for DIII-D edge density profiles between ELMs.

  13. Validation of the model for ELM suppression with 3D magnetic fields using low torque ITER baseline scenario discharges in DIII-D

    Science.gov (United States)

    Moyer, R. A.; Paz-Soldan, C.; Nazikian, R.; Orlov, D. M.; Ferraro, N. M.; Grierson, B. A.; Knölker, M.; Lyons, B. C.; McKee, G. R.; Osborne, T. H.; Rhodes, T. L.; Meneghini, O.; Smith, S.; Evans, T. E.; Fenstermacher, M. E.; Groebner, R. J.; Hanson, J. M.; La Haye, R. J.; Luce, T. C.; Mordijck, S.; Solomon, W. M.; Turco, F.; Yan, Z.; Zeng, L.; DIII-D Team

    2017-10-01

    Experiments have been executed in the DIII-D tokamak to extend suppression of Edge Localized Modes (ELMs) with Resonant Magnetic Perturbations (RMPs) to ITER-relevant levels of beam torque. The results support the hypothesis for RMP ELM suppression based on transition from an ideal screened response to a tearing response at a resonant surface that prevents expansion of the pedestal to an unstable width [Snyder et al., Nucl. Fusion 51, 103016 (2011) and Wade et al., Nucl. Fusion 55, 023002 (2015)]. In ITER baseline plasmas with I/aB = 1.4 and pedestal ν * ˜ 0.15, ELMs are readily suppressed with co- I p neutral beam injection. However, reducing the beam torque from 5 Nm to ≤ 3.5 Nm results in loss of ELM suppression and a shift in the zero-crossing of the electron perpendicular rotation ω ⊥ e ˜ 0 deeper into the plasma. The change in radius of ω ⊥ e ˜ 0 is due primarily to changes to the electron diamagnetic rotation frequency ωe * . Linear plasma response modeling with the resistive MHD code m3d-c1 indicates that the tearing response location tracks the inward shift in ω ⊥ e ˜ 0. At pedestal ν * ˜ 1, ELM suppression is also lost when the beam torque is reduced, but the ω ⊥ e change is dominated by collapse of the toroidal rotation v T . The hypothesis predicts that it should be possible to obtain ELM suppression at reduced beam torque by also reducing the height and width of the ωe * profile. This prediction has been confirmed experimentally with RMP ELM suppression at 0 Nm of beam torque and plasma normalized pressure β N ˜ 0.7. This opens the possibility of accessing ELM suppression in low torque ITER baseline plasmas by establishing suppression at low beta and then increasing beta while relying on the strong RMP-island coupling to maintain suppression.

  14. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Beyer, P.

    1999-01-01

    Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

  15. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Beyer, P.

    2001-01-01

    Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

  16. The plasma-sheath boundary region

    International Nuclear Information System (INIS)

    Franklin, R N

    2003-01-01

    In this review an attempt is made to give a broad coverage of the problem of joining plasma and sheath over a wide range of physical conditions. We go back to the earliest works quoting them, where appropriate, to understand what those who introduced the various terms associated with the structure of the plasma-sheath had in mind. We try to bring out the essence of the insights that have been gained subsequently, by quoting from the literature selectively, indicating how misunderstandings have arisen. In order to make it accessible to the generality of those currently working in low temperature plasmas we have sought to avoid mathematical complexity but retain physical insight, quoting from published work where appropriate. Nevertheless, in clarifying my own ideas I have found it necessary to do additional original work in order to give a consistent picture. In this way I have sought to bring together work in the late 1920s, the 1960s, and now mindful of the commercial importance of plasma processing, work over the past 15 years that adds to the general understanding. (topical review)

  17. On nonlinear statistical thermodynamics of boundary plasma with postactions

    International Nuclear Information System (INIS)

    Temko, S.W.; Temko, K.W.; Kuz'min, S.K.

    1992-01-01

    The authors use the statistical thermodynamics of small systems proposed before their publications for boundary weakly ionized plasma with postaction. Boundary properties of the plasma is taken into account by two ways: (1) suppose that only small number of very quick particles are able to leave the cloud having done entrance into outer medium work; (2) take into account the interaction between particles and inner surface of the cloud. Interactions in the boundary plasma are described by corresponding potential functions. The potential functions are mathematical models of real interactions in boundary plasma. Choosing of potential functions, their numerical parameters, geometrical form and dimensions of the cloud is made by using the methods of optimal experiment planning, maximum likelihood and computer experiment. Free energy of the cloud is a likelihood function. State of boundary plasma with admixtures is described by vector-density of particles distribution. Term ''distribution'' is used here in Sobolev-Schwartc sense. The authors obtain the vector-density of particles distribution in cloud which gives the condition minimum of free energy for every time moment under quasistatistical equilibrium. The system of conditions for free energy conditional minimizing for every time moment includes integral equilibrium equations, ''non-hard normalization'' and additional conditions taken as a result of analyzing physical and physical-chemical nature of boundary plasma. To obtain conditional minimum of free energy it is necessary to solve the system of conditions. First of all they solve equilibrium problem by the authors method. They obtain vector-density of particles distribution in the cloud. Then using method of random walk with postaction between sets of random walk process they build distribution function of random vector-density

  18. Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

    Directory of Open Access Journals (Sweden)

    Yutthapong Pinanroj

    2014-04-01

    Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

  19. DIII-D electron cyclotron heating 2 MW upgrade project. Final report for the period FY89 through FY97

    International Nuclear Information System (INIS)

    Callis, R.W.

    1997-08-01

    The 2 MW, 110 GHz ECH system was based on the General Atomics Proposal to the Department of Energy: DIII-D Fusion Research Program Vol. I Technical, and Vol. II Cost (GACP-72-166, July 1987 and revised). This proposal was reviewed in August 1987 by a senior technical review committee, who recommended to vigorously pursue increasing the ECH power to 6 MW. The realization of the higher frequency and power ECH on DIII-D was recognized by the committee to be important, not only for the DIII-D program, but also for future devices and the whole ECH area. Subsequently, an engineering cost and schedule review was conducted by DOE-OAK which confirmed the GA costs and schedules and recommended proceeding directly to 10 MW. However, because of budgetary constraints, in the April 1988 Field Task Proposal submission, GA proposed a phased ECH approach, Phase I being 2 MW and Phase II increasing the power to 10 MW. After review, DOE instructed GA to initiate the prototype 2 MW, 110 GHz program. The contract to procure four 500 kW, 110 GHz, 10 s gyrotrons from Varian Associates was initiated in April 1989 with final delivery by November 1990. Because of difficulties in spreading the energy of the electron beam over the collector area, the testing of the first gyrotron delayed its delivery until February 1991. The second gyrotron was able to operate for 1 s at 500 kW and 2 s at 300 kW, but failed when the cavity suffered thermal damage

  20. 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)

  1. The free-boundary equilibrium problem for helically symmetric plasmas

    International Nuclear Information System (INIS)

    Gardner, H.J.; Dewar, R.L.; Sy, W.N-C.

    1987-05-01

    An iterative technique for solving the ideal MHD equilibrium equations for a helically symmetric plasma with a free boundary is described. The method involves an application of Green's theorem and has been formulated for the geometry of a heliac. It is used to determine a stability diagram for the SHEILA heliac as a function of the plasma pressure and the current in one of the external coils

  2. Electromagnetic pulses at the boundary of a nonlinear plasma

    International Nuclear Information System (INIS)

    Satorius, E.H.

    1975-01-01

    An investigation was made of the behavior of strong electromagnetic pulses at the boundary of a nonlinear, cold, collisionless, and uniform plasma. The nonlinearity considered here is due to the nonlinear terms in the fluid equation which is used to describe the plasma. Two cases are studied. First, the case where there is a voltage pulse applied across the plane boundary of a semi-infinite, nonlinear plasma. Two different voltage pulses are considered, i.e., a delta function pulse and a suddenly turned-on sinusoidal pulse. The resulting electromagnetic fields propagating in the nonlinear plasma are found in this case. In the second case, the reflection of incident E-polarized and H-polarized, electromagnetic pulses at various angles of incidence from a nonlinear, semi-infinite plasma are considered. Again, two forms of incident pulses are considered: a delta function pulse and a suddenly turned-on sinusoidal pulse. In case two, the reflected electromagnetic fields are found. In both cases, the method used for finding the fields is to first solve the fluid equation (which describes the plasma) for the nonlinear conduction current in terms of the electric field using a perturbation method (since the nonlinear effects are assumed to be small). Next, this current is substituted into Maxwell's equations, and finally the electromagnetic fields which satisfy the boundary conditions are found. (U.S.)

  3. 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

  4. Near Real-Time Gyrotron Data Streaming and Data Acquisition with ns Resolution on the DIII-D ECH System

    Science.gov (United States)

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

    2014-10-01

    As part of the expansion and upgrade of the electron cyclotron heating (ECH) systen on DIII-D, a new data acquisition setup has been implemented to acquire and display waveform data from all gyrotrons in near real time with high time resolution. The data acquisition for each gyrotron system is based on a fast digitizer with 8 channels running at 2 MS/s/channel and a resolution of 14 bits. This enables the operator to monitor all gyrotron-relevant variables as well as fast diagnostic signals such as window arcs. The data are transferred to a local computer through a 132 MB/s PCI bus, and then are streamed to the ECH operator and to a local network attached storage using 1 GB Ethernet links. The data are displayed to the ECH operator by means of a graphical user interface developed in LabVIEW, replacing physical scopes. Acquired gyrotron data are accessible at DIII-D through a local database (PTDATA) connected to the ECH data acquisition system by an Ethernet line, a configuration that eliminates the need for legacy CAMAC hardware in the data link. Work supported by the US Department of Energy under DE-FC02-04ER54698.

  5. Production and fabrication of vanadium alloys for the radiative divertor program of DIII-D - Annual report input for 1996

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

    V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor (RD) upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy has been completed at Teledyne Wah Chang of Albany, Oregon (TWCA) to provide {approximately}800-kg of applicable product forms, and two billets have been extruded from the ingot. Chemical compositions of the ingot and both extruded billets were acceptable. Material from these billets will be converted into product forms suitable for components of the DIII-D Radiative Divertor structure. 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.

  6. Coherent structures in the boundary plasma of EAST Tokamak

    DEFF Research Database (Denmark)

    Yan, Ning

    -facing material, leading to intensive transient heat load and particle load on the local areas of both the divertor target plates and the first wall, which damages the material and causes enhanced recycling and impurity generation, then further pollutes the core plasma. In this project, we carried out experiment...... in the boundary plasma using multi-pin Langmuir probe in L-mode discharge. It was found that the coherent structures (Blobs and Holes) are created in the edge shear layer of poloidal flows where the plasma shows steep pressure gradient. Simulations have been performed using the ESEL code, which is a 2D fluid...... turbulence-simulation code based on the interchange instability as the main drive for the turbulence and structure motion in the scrape-off layer (SOL) plasma, with the input parameters from the EAST experiments. The simulations successfully reproduce the statistical characteristics of the SOL turbulence...

  7. Plasma Transport at the Magnetospheric Flank Boundary. Final report

    International Nuclear Information System (INIS)

    Otto, Antonius

    2012-01-01

    Progress is highlighted in these areas: 1. Model of magnetic reconnection induced by three-dimensional Kelvin Helmholtz (KH) modes at the magnetospheric flank boundary; 2. Quantitative evaluation of mass transport from the magnetosheath onto closed geomagnetic field for northward IMF; 3. Comparison of mass transfer by cusp reconnection and Flank Kelvin Helmholtz modes; 4. Entropy constraint and plasma transport in the magnetotail - a new mechanism for current sheet thinning; 5. Test particle model for mass transport onto closed geomagnetic field for northward IMF; 6. Influence of density asymmetry and magnetic shear on (a) the linear and nonlinear growth of 3D Kelvin Helmholtz (KH) modes, and (b) three-dimensional KH mediated mass transport; 7. Examination of entropy and plasma transport in the magnetotail; 8. Entropy change and plasma transport by KH mediated reconnection - mixing and heating of plasma; 9. Entropy and plasma transport in the magnetotail - tail reconnection; and, 10. Wave coupling at the magnetospheric boundary and generation of kinetic Alfven waves

  8. Volterra's series solutions of free boundary plasma equilibria

    International Nuclear Information System (INIS)

    De Menna, L.; Miano, G.; Rubinacci, G.

    1987-01-01

    In this paper the nonlinear problem of the free boundary equilibrium of a plasma inside a conducting circular shell has been solved analytically in the high beta limit. The unknown function describing the plasma shape has been expanded in a Volterra functional series, the functional analogous to the Taylor series. The hierarchy of the linear integral equations obtained from the expansion is, at least in principle, analytically solvable, so that the solution of each equation can be given in a closed form. The analytical computations have been carried out up to the fourth order and the results compared with numerical computations

  9. Electron temperature gradient driven instability in the tokamak boundary plasma

    International Nuclear Information System (INIS)

    Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.

    1992-01-01

    A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t -1/2 e γmt

  10. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    1999-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  11. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    2001-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  12. 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.

  13. 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

  14. 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

  15. On the nature of the plasma sheet boundary layer

    Energy Technology Data Exchange (ETDEWEB)

    Hones, E.W. Jr. (Mission Research Corp., Los Alamos, NM (USA) Los Alamos National Lab., NM (USA))

    1990-01-01

    The regions of the plasma sheet adjacent to the north and south lobes of the magnetotail have been described by many experimenters as locations of beams of energetic ions and fast-moving plasma directed primarily earthward and tailward along magnetic field lines. Measurements taken as satellites passed through one or the other of these boundary layers have frequently revealed near-earth mirroring of ions and a vertical segregation of velocities of both earthward-moving and mirroring ions with the fastest ions being found nearest the lobe-plasma sheet interface. These are features expected for particles from a distant tail source {bar E} {times} {bar B} drifting in a dawn-to-dusk electric field and are consistent with the source being a magnetic reconnection region. The plasma sheet boundary layers are thus understood as separatrix layers, bounded at their lobeward surfaces by the separatrices from the distant neutral line. This paper will review the observations that support this interpretation. 10 refs., 7 figs.

  16. Validating the energy transport modeling of the DIII-D and EAST ramp up experiments using TSC

    Science.gov (United States)

    Liu, Li; Guo, Yong; Chan, Vincent; Mao, Shifeng; Wang, Yifeng; Pan, Chengkang; Luo, Zhengping; Zhao, Hailin; Ye, Minyou

    2017-06-01

    The confidence in ramp up scenario design of the China fusion engineering test reactor (CFETR) can be significantly enhanced using validated transport models to predict the current profile and temperature profile. In the tokamak simulation code (TSC), two semi-empirical energy transport models (the Coppi-Tang (CT) and BGB model) and three theory-based models (the GLF23, MMM95 and CDBM model) are investigated on the CFETR relevant ramp up discharges, including three DIII-D ITER-like ramp up discharges and one EAST ohmic discharge. For the DIII-D discharges, all the transport models yield dynamic {{\\ell}\\text{i}} within +/- 0.15 deviations except for some time points where the experimental fluctuation is very strong. All the models agree with the experimental {β\\text{p}} except that the CT model strongly overestimates {β\\text{p}} in the first half of ramp up phase. When applying the CT, CDBM and GLF23 model to estimate the internal flux, they show maximum deviations of more than 10% because of inaccuracies in the temperature profile predictions, while the BGB model performs best on the internal flux. Although all the models fall short in reproducing the dynamic {{\\ell}\\text{i}} evolution for the EAST tokamak, the result of the BGB model is the closest to the experimental {{\\ell}\\text{i}} . Based on these comparisons, we conclude that the BGB model is the most consistent among these models for simulating CFETR ohmic ramp-up. The CT model with improvement for better simulation of the temperature profiles in the first half of ramp up phase will also be attractive. For the MMM95, GLF23 and CDBM model, better prediction of the edge temperature will improve the confidence for CFETR L-mode simulation. Conclusive validation of any transport model will require extensive future investigation covering a larger variety discharges.

  17. Physics of a fusion plasma boundary layer. Report No. 67. [Hot plasma interactions with cold wall

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, B.

    1976-01-01

    This thesis is a theoretical and computational study of plasma phenomena occurring when a hot, dense plasma containing a transverse magnetic field is brought into sudden contact with a cold metal wall. Thermal and magnetic boundary layers develop. This study includes the effects caused by thermal conduction, plasma convection, magnetohydrodynamic phenomena, atomic recombination and ionization, diffusion of neutral atoms which are formed near the wall, and radiation. The time evolution of the plasma temperature, pressure, the charged and neutral particle concentrations, magnetic and electric field strengths, and the plasma current density in the neighborhood of the solid surface are investigated. The rate of energy transfer from the plasma to the wall is calculated, and the conditions under which wall surface melting occurs are estimated.

  18. Fixed boundary toroidal plasma equilibria with toroidal flows

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Yanqiang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Yemin; Xiang, Nong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-04-15

    The fixed boundary toroidal plasma equilibria with toroidal flows are investigated by solving the modified Grad-Shafranov equation numerically in the cylindrical coordinate system. For normal equilibrium configurations with geometry and profiles similar to usual tokamaks with no flow, it is found that the effect of flow is to lead to an outward shift of the magnetic flux surfaces, together with the profiles of pressure, and mass and current densities. The shifts could become significant when the toroidal flow Mach number exceeds 0.5. For non-conventional current profiles, even for the usual tokamak geometry, novel current reversal equilibrium configurations may result, sometimes with changed topology in the poloidal flux function. This change in the topology of plasma equilibrium can be attributed to the large toroidal flow. The computed results may correspond to situations of intense tangential injection during the low toroidal current phase in expected experimental situations.

  19. Electron temperature gradient driven instability in the tokamak boundary plasma

    Energy Technology Data Exchange (ETDEWEB)

    Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.

    1992-12-15

    A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t{sup {minus}1/2}e{sup {gamma}mt}.

  20. Non-perturbative measurement of cross-field thermal diffusivity reduction at the O-point of 2/1 neoclassical tearing mode islands in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bardóczi, L.; Rhodes, T. L.; Carter, T. A.; Crocker, N. A.; Peebles, W. A. [University of California Los Angeles, Los Angeles, California 90095 (United States); Grierson, B. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2016-05-15

    Neoclassical tearing modes (NTMs) often lead to the decrease of plasma performance and can lead to disruptions, which makes them a major impediment in the development of operating scenarios in present toroidal fusion devices. Recent gyrokinetic simulations predict a decrease of plasma turbulence and cross-field transport at the O-point of the islands, which in turn affects the NTM dynamics. In this paper, a heat transport model of magnetic islands employing spatially non-uniform cross-field thermal diffusivity (χ{sub ⊥}) is presented. This model is used to derive χ{sub ⊥} at the O-point from electron temperature data measured across 2/1 NTM islands in DIII-D. It was found that χ{sub ⊥} at the O-point is 1 to 2 orders of magnitude smaller than the background plasma transport, in qualitative agreement with gyrokinetic predictions. As the anomalously large values of χ{sub ⊥} are often attributed to turbulence driven transport, the reduction of the O-point χ{sub ⊥} is consistent with turbulence reduction found in recent experiments. Finally, the implication of reduced χ{sub ⊥} at the O-point on NTM dynamics was investigated using the modified Rutherford equation that predicts a significant effect of reduced χ{sub ⊥} at the O-point on NTM saturation.

  1. 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.; Petersen, P.I.; Winter, J.

    1991-09-01

    A system has been added to the D3-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 of the boron film is to reduce the levels of impurity atoms in the D3-D plasmas. Experiments following the application of the boron film in D3-D have led to significant reductions in plasma impurity levels and the observation of a new, very high confinement regime. 9 refs., 1 fig

  2. Engineering study of the neutral beam and rf heating systems for DIII-D, MFTF-B, JET, JT-60 and TFTR

    International Nuclear Information System (INIS)

    Lindquist, W.B.; Staten, S.H.

    1987-01-01

    An engineering study was performed on the rf and neutral beam heating systems implemented for DIII-D, MFTF-B, JET, JT-60 and TFTR. Areas covered include: methodology used to implement the systems, technology, cost, schedule, performance, problems encountered and lessons learned. Systems are compared and contrasted in the areas studied. Summary statements were made on common problems and lessons learned. 3 refs., 6 tabs

  3. Theory and Fluid Simulations of Boundary Plasma Fluctuations

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, R H; LaBombard, B; LoDestro, L L; Rognlien, T D; Ryutov, D D; Terry, J L; Umansky, M V; Xu, X Q; Zweben, S

    2007-01-09

    Theoretical and computational investigations are presented of boundary plasma microturbulence that take into account important effects of the geometry of diverted tokamaks--in particular, the effect of x-point magnetic shear and the termination of field lines on divertor plates. We first generalize our previous 'heuristic boundary condition' which describes, in a lumped model, the closure of currents in the vicinity of the x-point region to encompass three current-closure mechanisms. We then use this boundary condition to derive the dispersion relation for low-beta flute-like modes in the divertor-leg region under the combined drives of curvature, sheath impedance, and divertor tilt effects. The results indicate the possibility of strongly growing instabilities, driven by sheath boundary conditions, and localized in either the private or common flux region of the divertor leg depending on the radial tilt of divertor plates. We re-visit the issue of x-point effects on blobs, examining the transition from blobs terminated by x-point shear to blobs that extend over both the main SOL and divertor legs. We find that, for a main-SOL blob, this transition occurs without a free-acceleration period as previously thought, with x-point termination conditions applying until the blob has expanded to reach the divertor plate. We also derive propagation speeds for divertor-leg blobs. Finally, we present fluid simulations of the C-Mod tokamak from the BOUT edge fluid turbulence code, which show main-SOL blob structures with similar spatial characteristics to those observed in the experiment, and also simulations which illustrate the possibility of fluctuations confined to divertor legs.

  4. New D3-D tokamak plasma control system

    Science.gov (United States)

    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 divertor 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 U.S. 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.

  5. Sensitivity of the Boundary Plasma to the Plasma-Material Interface

    International Nuclear Information System (INIS)

    Canik, John M.; Tang, X.-Z.

    2017-01-01

    While the sensitivity of the scrape-off layer and divertor plasma to the highly uncertain cross-field transport assumptions is widely recognized, the plasma is also sensitive to the details of the plasma-material interface (PMI) models used as part of comprehensive predictive simulations. Here in this paper, these PMI sensitivities are studied by varying the relevant sub-models within the SOLPS plasma transport code. Two aspects are explored: the sheath model used as a boundary condition in SOLPS, and fast particle reflection rates for ions impinging on a material surface. Both of these have been the study of recent high-fidelity simulation efforts aimed at improving the understanding and prediction of these phenomena. It is found that in both cases quantitative changes to the plasma solution result from modification of the PMI model, with a larger impact in the case of the reflection coefficient variation. Finally, this indicates the necessity to better quantify the uncertainties within the PMI models themselves, and perform thorough sensitivity analysis to propagate these throughout the boundary model; this is especially important for validation against experiment, where the error in the simulation is a critical and less-studied piece of the code-experiment comparison.

  6. 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.

  7. Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

    Science.gov (United States)

    Frerichs, Heinke; Schmitz, Oliver; Covele, Brent; Guo, Houyang; Hill, David; Feng, Yuhe

    2017-10-01

    In the Small Angle Slot (SAS) divertor in DIII-D, the combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field causes the strike point to vary radially along the divertor slot and even leave it at some toroidal locations. This effect essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade performance of the slot divertor. This effect has been approximated by a finite gap in the divertor baffle. Simulations with EMC3-EIRENE show that a toroidally localized loss of divertor closure can result in non-axisymmetric divertor densities and temperatures. This introduces a density window of 10-15% on top of the nominal threshold separatrix density during which a non-axisymmetric onset of local detachment occurs, initially leaving the gap and up to 60 deg beyond that still attached. Conversely, the impact of such toroidally localized divertor perturbations on the toroidal symmetry of midplane separatrix conditions is small. This work has been funded by the U.S. Department of Energy under Early Career Award Grant DE-SC0013911, and Grant DE-FC02-04ER54698.

  8. DIVIMP Modeling of the Toroidally-Symmetrical Injection of 13CH4 into the Upper SOL of DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    McLean, A G; Elder, J D; Stangeby, P C; Allen, S L; Brooks, N H; Fenstermacher, M E; Groth, M; Lisgo, S; Nagy, A; Wampler, W R; Watkins, J G; West, W P; Whyte, D G

    2004-12-03

    As part of a study of carbon-tritium co-deposition, we carried out an experiment on DIII-D involving a toroidally symmetric injection of {sup 13}CH{sub 4} at the top of a LSN discharge. A Monte Carlo code, DIVIMP-HC, which includes molecular breakup of hydrocarbons, was used to model the region near the puff. The interpretive analysis indicates a parallel flow in the SOL of M{sub l} {approx} 0.4 directed toward the inner divertor. The CH{sub 4} is ionized in the periphery of the SOL and so the particle confinement time, {tau}{sub c}, is not high, only {approx}5 ms, and about 4X lower than if the CH{sub 4} were ionized at the separatrix. For such a wall injection location, however, most of the CH{sub 4} gets ionized to C{sup +}, C{sup ++}, etc., and is efficiently transported along the SOL to the inner divertor, trapping hydrogen by co-deposition there.

  9. DIVIMP modeling of the toroidally-symmetrical injection of 13 CH4 into the upper SOL of DIII-D.

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, N. (General Atomics); Nagy A. (General Atomics); McLean, A. G. (University of Toronto Institute for Aerospace Studies); Groth, M. (Lawrence Livermore National Laboratory); Elder, J. D. (University of Toronto Institute for Aerospace Studies); Fenstermacher, M. E. (Lawrence Livermore National Laboratory); Whyte, D. G. (University of Wisconsin - Madison); Lisgo, S. (University of Toronto Institute for Aerospace Studies); Allen, S. L. (Lawrence Livermore National Laboratory); West, W.P. (General Atomics); Stangeby, P C (University of Toronto Institute for Aerospace Studies); Watkins, Jonathan G.; Wampler, William R.

    2004-05-01

    As part of a study of carbon-tritium co-deposition, we carried out an experiment on DIII-D involving a toroidally symmetric injection of {sup 13}CH{sub 4} at the top of a LSN discharge. A Monte Carlo code, DIVIMP-HC, which includes molecular breakup of hydrocarbons, was used to model the region near the puff. The interpretive analysis indicates a parallel flow in the SOL of M {parallel} {approx} 0.4 directed toward the inner divertor. The CH{sub 4} is ionized in the periphery of the SOL and so the particle confinement time, T{sub C}, is not high, only {approx} 5 ms, and about 4X lower than if the CH{sub 4} were ionized at the separatrix. For such a wall injection location, however, approximately 60-75% of the CH{sub 4} gets ionized to C{sup +}, C{sup 2+}, etc., and is efficiently transported along the SOL to the inner divertor, trapping hydrogen by co-deposition there.

  10. Improvements and new features for the diagnostics and control system of the ECH system on DIII-D

    Science.gov (United States)

    Torrezan, A. C.; Ponce, D.; Gorelov, Y. A.; Cengher, M.; Lohr, J.; Ellis, R. A.

    2016-10-01

    In this work we discuss improvements and new features for the diagnostics and controls of the electron cyclotron heating (ECH) system on the DIII-D tokamak. As for diagnostics, a new mapping system to measure the power density on a gyrotron collector, which should not exceed 500 W/cm2, has been designed and assembled. The new mapping system was designed to enable the test of new depressed collector gyrotrons that requires more RTD channels and for easier servicing and expansion compared with a previous system. First results from this diagnostic will be presented. As for controls, obsolete timing generators and auxiliary circuitry are being replaced by a more flexible approach using a FPGA. Besides being a simple replacement, the FPGA design will also add new features to the ECH control system such as an attempt to restart RF generation after RF loss or the recovery of selected interlocks. Upgrades made to the ECH launchers and issues found in the last experimental campaign as well as planned improvements will also be described. Work supported by US DOE under DE-FC02-04ER54698 and DE-AC02-09CH11466.

  11. Modelling and control of a tokamak plasma

    International Nuclear Information System (INIS)

    Bremond, S.

    1995-01-01

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes

  12. Depletion of solar wind plasma near a planetary boundary

    International Nuclear Information System (INIS)

    Zwan, B.J.; Wolf, R.A.

    1976-01-01

    A mathematical model is presented that describes the squeezing of solar wind plasma out along interplanetary magnetic field lines in the region between the bow shock and the effective planetary boundary (in the case of the earth, the magnetopause). In the absence of local magnetic merging the squeezing process should create a 'depletion layer,' a region of very low plasma density just outside the magnetopause. Numerical solutions are obtained for the dimensionless magnetohydrodynamic equations describing this depletion process for the case where the solar wind magnetic field is perpendicular to the solar wind flow direction. For the case of the earth with a magnetopause standoff distance of 10 R/subE/, the theory predicts that the density should be reduced by a factor > or =2 in a layer about 700--1300 km thick if M/subA/, the Alfven Mach number in the solar wind, is equal to 8. The layer thickness should vary as M/subA/ -2 and should be approximately uniform for a large area of the magnetopause around the subsolar point. Computed layer thicknesses are somewhat smaller than those derived from Lees' axisymmetric model. Depletion layers should develop fully only where magnetic merging is locally unimportant. Scaling of the model calculations to Venus and Mars suggest layer thicknesses about 1/10 and 1/15 those of the earth, respectively, neglecting diffusion and ionospheric effects

  13. 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.

    1991-01-01

    Hydrogen recycling and impurity influx are important issues in obtaining high confinement discharges in the D3-D tokamak. To reduce metallic impurities in D3-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 side of the tiles which form the small gaps separating each of the tiles in the divertor. The highest areal densities of both deuterium and metals were found on the plasma-facing surface near the inner strike point region of each set of divertor tiles. Significant deposits, extending as fast a 1 cm from the plasma-facing and containing up to forty percent of the total divertor deposition, were also observed on the gap-forming surfaces of the tiles

  14. Modeling and measurement of the motion of the DIII-D vacuum vessel during vertical instabilities

    International Nuclear Information System (INIS)

    Reis, E.; Blevins, R.D.; Jensen, T.H.; Luxon, J.L.; Petersen, P.I.; Strait, E.J.

    1991-11-01

    The motions of the D3-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which were input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hertz that are sufficiently loud to be felt as well as heard by the D3-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds. The correlation of the theoretical and measured vessel currents, the dynamic measurements and analysis, and the acoustic measurements and analysis show that: (1) The physics model can predict vessel forces for selected values of plasma resistivity. The model also predicts poloidal and toroidal wall currents which agree with measured values; (2) The force-time history from the above model, used in conjunction with an axisymmetric structural model of the vessel, predicts vessel motions which agree well with measured values; (3) The above results, input to a simple acoustic model predicts the magnitude of sounds emitted from the vessel during disruptions which agree with acoustic measurements; (4) Correlation of measured vessel motions with structural analysis shows that a maximum vertical motion of the vessel up to 0.24 in will not overstress the vessel or its supports. 11 refs., 10 figs., 1 tab

  15. Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S.

    1995-10-18

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.

  16. Feedforward and feedback control of locked mode phase and rotation in DIII-D with application to modulated ECCD experiments

    Science.gov (United States)

    Choi, W.; La Haye, R. J.; Lanctot, M. J.; Olofsson, K. E. J.; Strait, E. J.; Sweeney, R.; Volpe, F. A.; The DIII-D Team

    2018-03-01

    The toroidal phase and rotation of otherwise locked magnetic islands of toroidal mode number n  =  1 are controlled in the DIII-D tokamak by means of applied magnetic perturbations of n  =  1. Pre-emptive perturbations were applied in feedforward to ‘catch’ the mode as it slowed down and entrain it to the rotating field before complete locking, thus avoiding the associated major confinement degradation. Additionally, for the first time, the phase of the perturbation was optimized in real-time, in feedback with magnetic measurements, in order for the mode’s phase to closely match a prescribed phase, as a function of time. Experimental results confirm the capability to hold the mode in a given fixed-phase or to rotate it at up to 20 Hz with good uniformity. The control-coil currents utilized in the experiments agree with the requirements estimated by an electromechanical model. Moreover, controlled rotation at 20 Hz was combined with electron cyclotron current drive (ECCD) modulated at the same frequency. This is simpler than regulating the ECCD modulation in feedback with spontaneous mode rotation, and enables repetitive, reproducible ECCD deposition at or near the island O-point, X-point and locations in between, for careful studies of how this affects the island stability. Current drive was found to be radially misaligned relative to the island, and resulting growth and shrinkage of islands matched expectations of the modified Rutherford equation for some discharges presented here. Finally, simulations predict the as designed ITER 3D coils can entrain a small island at sub-10 Hz frequencies.

  17. A local expansion method applied to fast plasma boundary reconstruction for EAST

    Science.gov (United States)

    Guo, Yong; Xiao, Bingjia; Luo, Zhengping

    2011-10-01

    A fast plasma boundary reconstruction technique based on a local expansion method is designed for EAST. It represents the poloidal flux distribution in the vacuum region by a limited number of expansions. The plasma boundary reconstructed by the local expansion method is consistent with EFIT/RT-EFIT results for an arbitrary plasma configuration. On a Linux server with Intel (R) Xeon (TM) CPU 3.2 GHz, the method completes one plasma boundary reconstruction in about 150 µs. This technique is sufficiently reliable and fast for real-time shape control.

  18. Microwave Imaging Reflectometry for the study of Edge Harmonic Oscillations on DIII-D

    Science.gov (United States)

    Ren, X.; Chen, M.; Chen, X.; Domier, C. W.; Ferraro, N. M.; Kramer, G. J.; Luhmann, N. C., Jr.; Muscatello, C. M.; Nazikian, R.; Shi, L.; Tobias, B. J.; Valeo, E.

    2015-10-01

    Quiescent H-mode (QH-mode) is an ELM free mode of operation in which edge-localized harmonic oscillations (EHOs) are believed to enhance particle transport, thereby stabilizing ELMs and preventing damage to the divertor and plasma facing components. Microwave Imaging Reflectometer (MIR) enabling direct comparison between the measured and simulated 2D images of density fluctuations near the edge can determine the 2D structure of density oscillation, which can help to explain the physics behind EHO modes. MIR data sometimes indicate a counter-propagation between dominant (n=1) and higher harmonic modes of coherent EHOs in the steep gradient regions of the pedestal. To preclude diagnostic artifacts, we have performed forward modeling that includes possible optical mis-alignments to show that offsets between transmitting and receiving antennas do not account for this feature. We have also simulated the non-linear structure of the EHO modes, which induces multiple harmonics that are properly charaterized in the synthetic diagnostic. By excluding mis-alignments of optics as well as patially eliminating non-linearity of EHO mode structure as possible explanation for the data, counter-propagation observed in MIR data, which is not corroborated by external Mirnov coil array measurements, may be due to subtleties of the eigenmode structure, such as an inversion radius consistent with a magnetic island. Similar effects are observed in analysis of internal ECE-Imaging and BES data. The identification of a non-ideal structure motivates further exploration of nonlinear models of this instability. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

  19. Relationship between locked modes and thermal quenches in DIII-D

    Science.gov (United States)

    Sweeney, R.; Choi, W.; Austin, M.; Brookman, M.; Izzo, V.; Knolker, M.; La Haye, R. J.; Leonard, A.; Strait, E.; Volpe, F. A.; The DIII-D Team

    2018-05-01

    Locked modes are known to be one of the major causes of disruptions, but the physical mechanisms by which locking leads to disruptions are not well understood. Here we analyze the evolution of the temperature profile in the presence of multiple coexisting locked modes during partial and full thermal quenches. Partial quenches are often observed to be an initial, distinct stage in the full thermal quench. Near the onset of partial quenches, locked island O-points are observed to align with each other on the midplane, and their widths are sufficient to overlap each other, as indicated by the Chirikov parameter. Energy conservation analysis of one partial thermal quench shows that the energy lost is both radiated in the divertor region, and conducted or convected to the divertor. Nonlinear resistive magnetohydrodynamic simulations support the interpretation of stochastic fields causing a partial axisymmetric collapse, though the simulated temperature profile exhibits less degradation than the experimental profiles. In discharges with minimum values of the safety factor above  ∼1.2, locked modes are observed to self-stabilize by inducing, possibly via double tearing modes, a minor disruption that removes their neoclassical drive. These high q min discharges often exhibit relatively low ratios of the plasma internal inductance to the safety factor at 95% of the poloidal flux, which might imply classical stability, in agreement with the decay of the mode when the neoclassical drive is removed.

  20. A fast CCD detector for charge exchange recombination spectroscopy on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Thomas, D.M.; Burrell, K.H.; Groebner, R.J.; Gohil, P.

    1996-05-01

    Charge Exchange Recombination (CER) spectroscopy has become a standard diagnostic for tokamaks. CER measurements have been used to determine spatially and temporally resolved ion temperature, toroidal and poloidal ion rotation speed, impurity density and radial electric field. Knowledge of the spatial profile and temporal evolution of the electric field shear in the plasma edge is crucial to understanding the physics of the L to H transition. High speed CER measurements are also valuable for Edge Localized Mode (ELM) studies. Since the 0.52 ms minimum time resolution of our present system is barely adequate to study the time evolution of these phenomena, we have developed a new CCD detector system with about a factor of two better time resolution. In addition, our existing system detects sufficient photons to utilize the shortest time resolution only under exceptional conditions. The new CCD detector has a quantum efficiency of about 0.65, which is a factor of 7 better than our previous image intensifier-silicon photodiode detector systems. We have also equipped the new system with spectrometers of lower f/number. This combination should allow more routine operation at the minimum integration time, as well as improving data quality for measurements in the divertor-relevant region outside of the separatrix. Construction details, benchmark data and initial tokamak measurements for the new system will be presented

  1. Experiments on steady state particle control in Tore Supra and DIII-D

    Science.gov (United States)

    Mioduszewski, P. K.; Hogan, J. T.; Owen, L. W.; Maingi, R.; Lee, D. K.; Hillis, D. L.; Klepper, C. C.; Menon, M. M.; Thomas, C. E.; Uckan, T.; Wade, M. R.; Chatelier, M.; Grisolia, C.; Ghendrih, Ph.; Grosman, A.; Hutter, T.; Loarer, T.; Pégourié, B.; Mahdavi, M. A.; Schaffer, M.

    1995-04-01

    Particle control is playing an increasingly important role in tokamak plasma performance. The present paper discusses particle control of hydrogen/deuterium by wall pumping on graphite or carbonized surfaces, as well as by external exhaust with pumped limiters and pumped divertors. Wall pumping is ultimately a transient effect and by itself not suitable for steady state particle exhaust. Therefore, external exhaust techniques with pumped divertors and limiters are being developed. How wall pumping phenomena interact and correlate with these inherently steady state, external exhaust techniques, is not well known to date. In the present paper, the processes involved in wall pumping and in external pumping are investigated in an attempt to evaluate the effect of external exhaust on wall pumping. Some of the key elements of this analysis are: (1) charge-exchange fluxes to the wall play a crucial role in the core-wall particle dynamics, (2) the recycling fluxes of thermal molecules have a high probability of ionization in the scrape-off layer, (3) thermal particles originating from the wall, which are ionized within the scrape-off layer, can be directly exhausted, thus providing a direct path between wall and exhaust which can be used to control the wall inventory. This way, the wall can be kept in a continuous pumping state in the sense that it continuously absorbs energetic particles and releases thermal molecules which are then removed by the external exhaust mechanism. While most of the ingredients of this analysis have been observed individually before, the present evaluation is an attempt to correlate effects of wall recycling and external exhaust.

  2. STRUCTURAL RESPONSE OF THE DIII-D TOROIDAL FIELD COIL TO INCREASED LATERAL LOADS

    International Nuclear Information System (INIS)

    REIS, E.E; CHIN, E.

    2004-03-01

    OAK-B135 Recent calibration shots in which full toroidal field (TF) coil current interacted with the maximum poloidal field coils have produced increased lateral loads on the outer sections of the TF-coil. The increased lateral loads have resulted in deflections that have been sufficient to cause the TF-coil to contact adjacent equipment and produce a transient short to ground within the coil. The six outer turns of each TF-coil bundle are clamped together by insulated preloaded studs to provide increased bending stiffness. These sections of the outer bundles depend on friction to react the lateral loads as a bundle rather than six individual turns. A major concern is that the increased loads will produce slip between turns resulting in excessive lateral deflections and possible damage to the insulating sleeve on the preloaded studs. A finite element structural model of the TF-coil was developed for the calculation of deflections and the shear load distribution throughout the coil for the applied lateral loads from a full current calibration shot. The purpose of the updated structural model is to correlate the applied lateral loads to the total shear force between the unbonded sections of the outer turns. An allowable integrated lateral load applied to the outer turns is established based on the maximum shear force that can be reacted by friction. A program that calculates the magnetic fields and integrated lateral load along the outer turns can be incorporated into the plasma control system. The integrated load can then be compared to the calculated allowable value prior to execution of calibration shots. Calibration shots with a calculated total lateral load greater than the allowable value will be prevented

  3. Fusion programs in Applied Plasma Physics

    International Nuclear Information System (INIS)

    1992-07-01

    The Applied Plasma Physics (APP) program at General Atomics (GA) described here includes four major elements: (a) Applied Plasma Physics Theory Program, (b) Alpha Particle Diagnostic, (c) Edge and Current Density Diagnostic, and (d) Fusion User Service Center (USC). The objective of the APP theoretical plasma physics research at GA is to support the DIII-D and other tokamak experiments and to significantly advance our ability to design a commercially-attractive fusion reactor. We categorize our efforts in three areas: magnetohydrodynamic (MHD) equilibria and stability; plasma transport with emphasis on H-mode, divertor, and boundary physics; and radio frequency (rf). The objective of the APP alpha particle diagnostic is to develop diagnostics of fast confined alpha particles using the interactions with the ablation cloud surrounding injected pellets and to develop diagnostic systems for reacting and ignited plasmas. The objective of the APP edge and current density diagnostic is to first develop a lithium beam diagnostic system for edge fluctuation studies on the Texas Experimental Tokamak (TEXT). The objective of the Fusion USC is to continue to provide maintenance and programming support to computer users in the GA fusion community. The detailed progress of each separate program covered in this report period is described in the following sections

  4. Collisional boundary layer analysis for neoclassical toroidal plasma viscosity in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Shaing, K.C.; Cahyna, Pavel; Bécoulet, M.; Park, J.-K.; Sabbagh, S.A.; Chu, M.S.

    2008-01-01

    Roč. 15, č. 8 (2008), 082506-1-7 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2969434

  5. Toothbrush probe for instantaneous measurement of radial profile in tokamak boundary plasma

    Energy Technology Data Exchange (ETDEWEB)

    Uehara, Kazuya; Sengoku, Seio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Amemiya, Hiroshi

    1997-04-01

    A new probe for the instantaneous measurement of radial profiles of the boundary scrape-off layer (SOL) plasma has been developed in a tokamak. Five asymmetric double-probe chips are aligned in parallel to a strong magnetic field in the boundary plasma in a tokamak. This probe is named the `toothbrush probe` and can measure the ion temperature as well as the electron temperature and the plasma density in the SOL plasma within only one tokamak plasma shot. First, only one asymmetric probe is mounted on the divertor plate and it is tried to determine the ion temperature. Then, a manufactured toothbrush probe is mounted in the SOL plasma and the radial plasma profiles are simultaneously obtained. Data on the e-folding length of the plasma profile obtained by the toothbrush probe can determine the information on the transport properties such as the diffusion coefficient and the thermal conductivity of electrons and ions. (author)

  6. Initial tests and operation of a 110 GHz, 1 MW gyrotron with evacuated waveguide system on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lohr, J.; Ponce, D.; Tooker, J.F. [and others

    1996-08-01

    A gyrotron producing nominally 1 MW at 110 GHz has been installed at the DIII-D tokamak and operated in a program of initial tests with a windowless evacuated transmission line. The alignment and first test operation were performed in an air environment at atmospheric pressure. Under these conditions, the tube produced rf output in excess of 800 kW for pulse lengths greater than 10 msec and power near 500 kW for pulse lengths of about 100 msec into a free space dummy load. The gyrotron was operated into evacuated corrugated waveguide in the full power parameter regime for pulse lengths of up to 500 msec injecting greater than 0.5 MW into DIII-D for a preliminary series of experiments. Generated powers greater than 900 kW were achieved. A parasitic oscillation at various frequencies between 20 and 100 MHz, which was generated during the pulsing of the gyrotron electron beam, was suppressed somewhat by a capacitive filter attached to the gyrotron itself. Addition of a magnetic shield intended to alter the magnetic field geometry below the cathode eliminated internal tube sparks. Rework of the external power and interlock circuitry to improve the immunity to electromagnetic interference was also done in parallel so that the fast interlock circuitry could be used. The latest results of the test program, the design of the free space load and other test hardware, and the transmission line will be presented.

  7. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which 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 iteration of the design-test loop to optimize performance.

  8. Destiny of earthward streaming plasma in the plasmasheet boundary layer

    Science.gov (United States)

    Green, J. L.; Horwitz, J. L.

    1986-01-01

    The dynamics of the earth's magnetotail have been investigated, and it has become clear that the plasmasheet boundary layer field lines map into the Region I Field-Aligned Currents (FAC) of the auroral zone. It is pointed out that the role of earthward streaming ions in the plasmasheet boundary layer may be of fundamental importance in the understanding of magnetotail dynamics, auroral zone physics, and especially for ionospheric-magnetospheric interactions. The present paper has the objective to evaluate propagation characteristics for the earthward streaming ions observed in the plasmasheet boundary layer. An investigation is conducted of the propagation characteristics of protons in the plasmasheet boundary layer using independent single particle dynamics, and conclusions are discussed. The density of earthward streaming ions found in the plasmasheet boundary layer should include the ring current as well as the auroral zone precipitaiton and inner plasmasheet regions of the magnetosphere.

  9. Measurements of ion energies in the tokamak plasma boundary

    Czech Academy of Sciences Publication Activity Database

    Kocan, M.; Gunn, J. P.; Carpentier-Chouchana, S.; Herrmann, A.; Kirk, A.; Komm, Michael; Müller, H.W.; Pascal, J.Y.; Pitts, R.A.; Rohde, V.; Tamain, P.

    2011-01-01

    Roč. 415, č. 1 (2011), S1133-S1138 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI)/19./. San Diego, CA, 24.05.2010-28.05.2010] Institutional support: RVO:61389021 Keywords : plasma * tokamak * scrape-off layer Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.052, year: 2011 http://www.sciencedirect.com/science/article/pii/S0022311510004277#

  10. Experimental and theoretical research in applied plasma physics

    International Nuclear Information System (INIS)

    Porkolab, M.

    1992-01-01

    This report discusses research in the following areas: fusion theory and computations; theory of thermonuclear plasmas; user service center; high poloidal beta studies on PBX-M; fast ECE fluctuation diagnostic for balloning mode studies; x-ray imaging diagnostic; millimeter/submillimeter-wave fusion ion diagnostics; small scale turbulence and nonlinear dynamics in plasmas; plasma turbulence and transport; phase contrast interferometer diagnostic for long wavelength fluctuations in DIII-D; and charged and neutral fusion production for fusio plasmas

  11. Experimental and theoretical research in applied plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    Porkolab, M.

    1992-01-01

    This report discusses research in the following areas: fusion theory and computations; theory of thermonuclear plasmas; user service center; high poloidal beta studies on PBX-M; fast ECE fluctuation diagnostic for balloning mode studies; x-ray imaging diagnostic; millimeter/submillimeter-wave fusion ion diagnostics; small scale turbulence and nonlinear dynamics in plasmas; plasma turbulence and transport; phase contrast interferometer diagnostic for long wavelength fluctuations in DIII-D; and charged and neutral fusion production for fusio plasmas.

  12. Electrostatic sheath at the boundary of a collisional dusty plasma

    Indian Academy of Sciences (India)

    Department of Physics, Cotton College, Guwahati 781 001, India. Abstract. Considering the Boltzmann response of the ions ... respect to normal electronic charge (q ~105. –106e). The mass of the dust grains can have very high value too, up to ... degrees of plasma dynamics. Thus, the theoretical modeling of a dusty plasma ...

  13. The separatrix tentacle effect of ion acceleration to the plasma sheet boundary

    Science.gov (United States)

    Buechner, Joerg; Zelenyi, Lev M.

    1990-02-01

    The effect of a continuous ion acceleration in the earth's magnetotail due to chaotic particle scattering caused by separatrix traversals in the velocity space is described. This effect operates almost everywhere in the plasma sheet, but outside and independent of neutral lines. As a distributed inner source, it supports fast ion streams at the boundary of the plasma sheet.

  14. The separatrix tentacle effect of ion acceleration to the plasma sheet boundary

    Energy Technology Data Exchange (ETDEWEB)

    Buechner, J. (Central Institute for Astrophysics, Potsdam (German Democratic Republic)); Zelenyi, L.M. (Space Research Institute, Moscow (USSR))

    1990-02-01

    The authors describe the effect of a continuous ion-acceleration in the Earth's magnetotail due to chaotic particle scattering caused by separatrix traversals in the velocity space. This effect operates almost everywhere in the plasma sheet, but outside and independent of neutral lines. As a distributed inner source it supports fast ion streams at the boundary of the plasma sheet.

  15. COMPARISON OF SENSORS FOR RESISTIVE WALL MODE FEEDBACK CONTROL. MILESTONE No.145 ''CONTAINING PLASMA INSTABILITIES WITH METAL WALLS''

    International Nuclear Information System (INIS)

    STRAIT, E.J.; CHU, M.S.; GAROFALO, A.M.; LAHAYE, R.J.; OKABAYASHI, M.; REIMERDES, H.; SCOVILLE, J.T.; TURNBULL, A.D.

    2002-01-01

    OAK A271 COMPARISON OF SENSORS FOR RESISTIVE WALL MODE FEEDBACK CONTROL MILESTONE No.145 CONTAINING PLASMA INSTABILITIES WITH METAL WALLS. The most serious instabilities in the tokamak are those described by ideal magneto-hydrodynamic theory. These modes limit the stable operating space of the tokamak. The ideal MHD calculations predict the stable operating space of the tokamak may be approximately doubled when a perfectly conducting metal wall is placed near the plasma boundary, compared to the case with no wall (free boundary). The unstable mode distortions of the plasma column cannot bulge out through a perfectly conducting wall. However, real walls have finite conductivity and when plasmas are operated in the regime between the free boundary stability limit and the perfectly conducting wall limit, the unstable mode encountered in that case the resistive wall mode, can leak out through the metal wall, allowing the mode to keep slowly growing. The slow growth affords the possibility of feedback stabilizing this mode with external coils. DIII-D is making good progress in such feedback stabilization research and in 2002 will use an improved set of mode sensors inside the vacuum vessel and closer to the plasma surface which are expected theoretically to improve the ability to stabilize the resistive wall mode

  16. On hot tenuous plasmas, fireballs, and boundary layers in the earth's magnetotail

    International Nuclear Information System (INIS)

    Frank, L.A.; Ackerson, K.L.; Lepping, R.P.

    1976-01-01

    Intensive correlative studies of magnetic fields and plasmas within the earth's magnetotail at geocentric radial distances of approx. 23--46 R/sub E/ during March--October 1974 revealed striking new features. The hot tenuous plasmas within the plasma sheet were found to be in a state of almost continual flow and were threaded with northward, or closed, geomagnetic field lines. Proton bulk speeds were in the range 50--500 km s -1 . The magnetic fields are directed northward. These observations demand a strong persistent source of magnetic flux and hot plasmas for the plasma sheet. No characteristic proton bulk flows were evident during crossings of the neutral sheet. Occasionally, the satellite encountered the region of acceleration in the magnetotail, the 'fireball.' This spectacular phenomenon exhibits strong jetting of plasmas in exces of 1000 km s -1 , proton temperatures of approx. 10 7 degreeK (kT approx. 1 keV), disordered magnetic fields, southward magnetic fields during tailward jetting of the plasmas. Earthward plasma flows within the fireball are threaded with closed geomagnetic field lines, and open magnetic field lines are embedded in the tailward jetting plasmas. The magnetosheathlike plasmas within the boundary layers which are positioned contiguous to the plasma sheet display striking evidences of plasma heating, great changes in bulk flow velocities and acceleration of energetic electrons with E > 45 keV. Persistent zones of southward magnetic fields are detected, which are often positioned adjacent to the plasma sheet and within the boundary layer plasmas. Rotations of the magnetic fields from southward to northward, or vice versa, in these boundary layers are accompanied by large enhancements of energetic electron intensities, substantial heating of the low-energy electron distributions, and strong perturbations of the proton velocity distribution functions

  17. The 8.4 MW Modulator/Regulator Power Systems for the Electron Cyclotron Heating Facility Upgrade at DIII-D

    International Nuclear Information System (INIS)

    Pronko, S.G.E.; Baggest, D.S.

    1999-01-01

    Over the next three years the DIII-D National Fusion Facility at General Atomics will upgrade its electron cyclotron heating (ECH) capability from the present 3 MW at 110 GHz to 10 MW of injected microwave power. There will be ten gyrotron tubes supplied by five 8.4 MW modulator/regulator (M/R) power systems. The project has gained considerable leverage from the acquisition of surplus hardware from the MFTF program that was conducted at LLNL in the early 1980s. One of these systems had been refurbished and converted for use as an ECH power supply earlier. The experience gained and the lessons learned from operating that system have proved valuable in guiding the engineering of the new systems. This paper provides an overview of the power system design and a report on the present status of the project

  18. Solutions to remove a boundary image sticking in an ac plasma display panel

    International Nuclear Information System (INIS)

    Park, Choon-Sang; Tae, Heung-Sik

    2009-01-01

    When displaying a square-type image with peak luminance for approximately 500 h in a 42 in. plasma display panel TV with high Xe (15%) content, halo-type boundary image sticking was observed in the nondischarge region adjacent to the discharge region. The halo-type boundary image sticking phenomenon is due mainly to the redeposit of the Mg species on the MgO layer in the nondischarge region adjacent to the discharge region, which is verified by measuring the redeposited Mg species in the boundary image sticking region using a cross-sectional scanning electron microscope. Based on this result, three kinds of solution to remove the boundary image sticking of an ac plasma display panel are introduced. First, we completely recover the boundary image sticking cells by using a full-white aging process. Second, we prohibit the inherent production of boundary image sticking by sealing the plasma display panel under vacuum. The final solution is to prohibit the inherent production of boundary image sticking by use of lower gas pressure.

  19. Statistical characterization of turbulence in the boundary plasma of EAST

    DEFF Research Database (Denmark)

    Yan, Ning; Nielsen, Anders Henry; Xu, G.S.

    2013-01-01

    In Ohmic heated low confinement mode (L-mode) discharges, the intermittent statistical characteristics of turbulent fluctuations have been investigated in the edge and the scrape-off layer (SOL) plasma on EAST (the experimental advanced superconducting tokamak) by fast reciprocating Langmuir probe...

  20. The influence of beam boundaries and velocity reduction on Pierce instability in laboratory plasmas

    International Nuclear Information System (INIS)

    Jovanovic, D.

    1982-01-01

    The influences of the beam-plasma boundary and of weak nonlinearities on the Pierce instability are investigated. It is shown that the finite width of the beam has negligible influence on both the stability of the system and growth rate. In the nonlinear regime the wavelength decreases and enhancement of the wave potential close to the beam inlet boundary is observed. The relationship between this effect and the formation of double layers is discussed. (Auth.)

  1. Role of boundary plasma in lower-hybrid-frequency heating of a tokamak

    International Nuclear Information System (INIS)

    Uehara, Kazuya; Yamamoto, Takumi; Fujii, Tsuneyuki

    1982-01-01

    Boundary plasma of a circular tokamak has been investigated by means of electrostatic probes during lower-hybrid heating. The reflection coefficient is affected by the density gradient in front of the launcher. An effective ion heating is performed in the main plasma region when the boundary electron temperature is relatively high enough to suppress the parametric decay instabilities. The simultaneous injection of neutral beams as well as the lower-hybrid wave brings the suppression of instabilities with increase of the electron temperature coming from the neutral beam heating. (author)

  2. Adverse consequences of a moving vacuum-plasma boundary on axisymmetric ac helicity injection

    International Nuclear Information System (INIS)

    Bellan, P.M.

    1986-01-01

    The recent prediction of Liewer, Gould, and Bellan that a moving plasma-vacuum boundary significantly lowers the effectiveness of ac helicity injection is generalized by resolution of the apparent discrepancy between the helicity-conservation equations of Jensen and Chu and of Moffatt. It is shown that, if there are axisymmetric circular flux surfaces and a moving vacuum-plasma boundary, the helicity injected by oscillating fields (if net injection occurs) is simply consumed by an increase in helicity dissipation due to the same oscillating fields

  3. Resonant Excitation of Boundary Layer Instability of DC Arc Plasma Jet by Current Modulation

    Czech Academy of Sciences Publication Activity Database

    Kopecký, Vladimír; Hrabovský, Milan

    2011-01-01

    Roč. 31, č. 6 (2011), s. 827-838 ISSN 0272-4324 R&D Projects: GA ČR GAP205/11/2070 Institutional research plan: CEZ:AV0Z20430508 Keywords : dc arc jet * plasma jet oscillations * boundary layer instability * frequency spectra Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2011 http://www.springerlink.com/content/v160841757161758/

  4. Turbulence and intermittent transport at the boundary of magnetized plasmas

    DEFF Research Database (Denmark)

    Garcia, O.E.; Naulin, V.; Nielsen, A.H.

    2005-01-01

    fluctuation wave forms and transport statistics are also in a good agreement with those derived from the experiments. Associated with the turbulence bursts are relaxation oscillations in the particle and heat confinements as well as in the kinetic energy of the sheared poloidal flows. The formation of blob...... a forcing region with spatially localized sources of particles and heat outside which losses due to the motion along open magnetic-field lines dominate, corresponding to the edge region and the scrape-off layer, respectively. Turbulent states reveal intermittent eruptions of hot plasma from the edge region...

  5. Modification of boundary plasma behavior by Ion Bernstein Wave heating on the HT-7 tokamak

    International Nuclear Information System (INIS)

    Xu, G.S.; Wan, B.N.; Song, M.; Ling, B.L.; Li, C.F.; Li, J.

    2003-01-01

    The boundary plasma behavior during Ion Bernstein Wave heating was investigated using Langmuir probe arrays on the HT-7 tokamak. A distinct weak turbulence regime was reproducibly observed in the 30 MHz IBW heated plasmas with RF power larger than 120 kW, which resulted in a particle confinement improvement of a factor of 2. The strong suppression and decorrelation effect of fluctuations resulted in the turbulent particle flux dropping by more than an order of magnitude in the plasma boundary region. An additional inward radial electric field and associated poloidal ExB flows were produced, which could account for the additional poloidal velocity in the electron diamagnetic direction at some radial locations of the boundary plasma. The electrostatic fluctuations were nearly completely decorrelated in the high frequency region and only low frequency fluctuations remained. The poloidal correlation was considerably reduced in the high poloidal wave number region and only the fluctuations with long poloidal wavelength remained. Three-wave nonlinear phase coupling between the whole frequency domain and the very low frequency region increased significantly in both the plasma edge and the SOL. Quite low frequency fluctuations (about 5 kHz) were generated, which dominated the boundary turbulence during IBW heating. Detailed analyses suggested that, when an IBW with a frequency of 30 MHz was launched into a plasma with the toroidal magnetic field between 1.75 T and 2.0 T, the ion cyclotron resonant layer of 5/2.D was located in the plasma edge region. The poloidal ExB sheared flows generated by IBW near this layer due to a ponderomotive interaction were found to be the mechanism underlying these phenomena. (author)

  6. On tenuous plasmas, fireballs, and boundary layers in the earth's magnetotail

    Science.gov (United States)

    Frank, L. A.; Ackerson, K. L.; Lepping, R. P.

    1976-01-01

    The plasma instrumentation (the Lepedea) and the magnetometer aboard IMP 8 performed correlative measurements of magnetic fields and plasmas within the geomagnetic tail at geocentric radial distances of about 23-46 R-E during March-October 1974. The hot tenuous plasmas within the plasma sheet were found to be in a state of almost continuous flow and were threaded with northward, or closed geomagnetic lines. The satellite encountered a region of acceleration in the magnetotail, the 'fireball' which exhibits strong jetting of plasmas in excess of 1000 km/s, proton temperatures of about 10 to the 7th K, disordered magnetic fields, southward magnetic fields during tailward jetting of plasmas, and northward magnetic fields for fast plasma flows toward earth. In addition, the magnetosheath plasmas within the boundary layers which are contiguous to the plasma sheet display evidence of plasma heating, great changes in bulk flow velocities, and acceleration of energetic electrons with an energy of greater than 45 keV.

  7. Realization and classification of symmetric stellarator configurations through plasma boundary modulations

    International Nuclear Information System (INIS)

    Yokoyama, M.; Nakajima, N.; Okamoto, M.

    1997-12-01

    The basic roles of several principle modulations of plasma boundary shape on magnetohydrodynamic (MHD) equilibria are investigated. The appropriate combination of principle helical modulations for elimination of bumpy field component to realize the quasi-axisymmetric (QAS) and quasi-helically symmetric (QHS) configurations is explained by considering the variation of area of magnetic surface cross sections. The triangular modulation is effectively utilized to form the magnetic well by shifting the magnetic axis outward compared to the center of mass of magnetic surface cross section. The spatialization of the magnetic axis or the bumpy modulations of plasma boundary is rather important to reduce the toroidicity in the magnetic field, which can lead to QHS configurations. Some stellarator magnetic configurations under design activity or in the existing experimental device are mentioned from the point of views of plasma boundary modulations. Based on these principle understandings of plasma boundary modulations, examples of essential approach to QAS and QHS configurations are demonstrated step by step. The possibility of quasi-bumpy (or poloidally) symmetric (QBS) configuration is also mentioned. (author)

  8. Structure of boundary layers of plasmas confined by magnetic fields

    International Nuclear Information System (INIS)

    Laval, G.; Pellat, R.

    1963-01-01

    The structure of a transition sheath between a magnetically confined plasma and vacuum is investigated. A one-dimensional, static solution of the collisionless Vlasov equation is sought. The two kinds of particles are assumed to have unequal masses, M- and M+, and temperatures, T- and T+; and the parameter λ 2 = (M-/M+) (T-/T+) = (R-/R+) (R-, R+ are Larmor radii) is introduced. For non-relativistic particles, the result is obtained that no physical solution can exist except for λ = 1 without trapped particles. The problem for Maxwellian distribution functions, no charge separation field λ = 1 + α with α small, and with trapped particles, is investigated. For negative α, only trapped ions are needed, and the distribution function of trapped particles that provides the thinnest sheath can be calculated. A complete solution of the problem with unequal masses, charge separation field, and trapped electrons is given. The sheath thickness is about four Larmor radii, a good physical picture if stable. (authors) [fr

  9. Nonlinear fluid simulation of particle and heat fluxes during burst of ELMs on DIII-D with BOUT++  code

    Science.gov (United States)

    Xia, T. Y.; Xu, X. Q.

    2015-09-01

    In order to study the distribution and evolution of the transient particle and heat fluxes during edge-localized mode (ELM) bursts, a BOUT++  six-field two-fluid model based on the Braginskii equations with non-ideal physics effects is used to simulate pedestal collapse in divertor geometry. The profiles from the DIII-D H-mode discharge #144382 with fast target heat flux measurements are used as the initial conditions for the simulations. A flux-limited parallel thermal conduction is used with three values of the flux-limiting coefficient {αj} , free streaming model with {αj}=1 , sheath-limit with {αj}=0.05 , and one value in between. The studies show that a 20 times increase in {αj} leads to  ∼6 times increase in the heat flux amplitude to both the inner and outer targets, and the widths of the fluxes are also expanded. The sheath-limit model of flux-limiting coefficient is found to be the most appropriate one, which shows ELM sizes close to the measurements. The evolution of the density profile during the burst of ELMs of DIII-D discharge #144382 is simulated, and the collapse in width and depth of {{n}\\text{e}} are reproduced at different time steps. The growing process of the profiles for the heat flux at divertor targets during the burst of ELMs measured by IRTV (infrared television) is also reproduced by this model. The widths of heat fluxes towards targets are a little narrower, and the peak amplitudes are twice the measurements possibly due to the lack of a model of divertor radiation which can effectively reduce the heat fluxes. The magnetic flutter combined with parallel thermal conduction is found to be able to increase the total heat loss by around 33% since the magnetic flutter terms provide the additional conductive heat transport in the radial direction. The heat flux profile at both the inner and outer targets is obviously broadened by magnetic flutter. The lobe structures near the X-point at LFS are both broadened and elongated due

  10. Plasmas in Multiphase Media: Bubble Enhanced Discharges in Liquids and Plasma/Liquid Phase Boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Kushner, Mark Jay [University of Michigan

    2014-07-10

    In this research project, the interaction of atmospheric pressure plasmas with multi-phase media was computationally investigated. Multi-phase media includes liquids, particles, complex materials and porous surfaces. Although this investigation addressed fundamental plasma transport and chemical processes, the outcomes directly and beneficially affected applications including biotechnology, medicine and environmental remediation (e.g., water purification). During this project, we made advances in our understanding of the interaction of atmospheric pressure plasmas in the form of dielectric barrier discharges and plasma jets with organic materials and liquids. We also made advances in our ability to use computer modeling to represent these complex processes. We determined the method that atmospheric pressure plasmas flow along solid and liquid surfaces, and through endoscopic like tubes, deliver optical and high energy ion activation energy to organic and liquid surfaces, and produce reactivity in thin liquid layers, as might cover a wound. We determined the mechanisms whereby plasmas can deliver activation energy to the inside of liquids by sustaining plasmas in bubbles. These findings are important to the advancement of new technology areas such as plasma medicine

  11. Experimental and theoretical research in applied plasma physics. Technical progress report, October 15, 1990--October 14, 1993

    Energy Technology Data Exchange (ETDEWEB)

    Porkolab, M.

    1992-06-01

    This report discusses research in the following areas: fusion theory and computations; theory of thermonuclear plasmas; user service center; high poloidal beta studies on PBX-M; fast ECE fluctuation diagnostic for balloning mode studies; x-ray imaging diagnostic; millimeter/submillimeter-wave fusion ion diagnostics; small scale turbulence and nonlinear dynamics in plasmas; plasma turbulence and transport; phase contrast interferometer diagnostic for long wavelength fluctuations in DIII-D; and charged and neutral fusion production for fusio plasmas.

  12. The role of the boundary plasma in defining the viability of a magnetic fusion reactor: A review

    Science.gov (United States)

    Whyte, Dennis

    2012-10-01

    The boundary of magnetic confinement devices, from the pedestal through to the surrounding surfaces, encompasses an enormous range of plasma and material physics, and their integrated coupling. It is becoming clear that due to fundamental limits of plasma stability and material response the boundary will largely define the viability of an MFE reactor. However we face an enormous knowledge deficit in stepping from present devices and ITER towards a demonstration power plant. We review the boundary and plasma-material interaction (PMI) research required to address this deficit as well as related theoretical/scaling methods for extending present results to future devices. The research activities and gaps are reviewed and organized to three major axes of challenges: power density, plasma duration, and material temperature. The boundary can also be considered a multi-scale system of coupled plasma and material science regulated through the non-linear interface of the sheath. Measurement, theory and modeling across these scales are reviewed. Dimensionless parameters, often used to organized core plasma transport on similarity arguments, can be extended to the boundary plasma, plasma-surface interactions and material response. The scaling methodology suggests intriguing ways forward to prescribe and understand the boundary issues of an eventual reactor in intermediate size devices. Finally, proposed technology and science innovations towards solving the extreme PMI/boundary challenges of magnetic fusion energy will be reviewed.

  13. Pulsed Plasma with Synchronous Boundary Voltage for Rapid Atomic Layer Etching

    Energy Technology Data Exchange (ETDEWEB)

    Economou, Demetre J.; Donnelly, Vincent M.

    2014-05-13

    Atomic Layer ETching (ALET) of a solid with monolayer precision is a critical requirement for advancing nanoscience and nanotechnology. Current plasma etching techniques do not have the level of control or damage-free nature that is needed for patterning delicate sub-20 nm structures. In addition, conventional ALET, based on pulsed gases with long reactant adsorption and purging steps, is very slow. In this work, novel pulsed plasma methods with synchronous substrate and/or “boundary electrode” bias were developed for highly selective, rapid ALET. Pulsed plasma and tailored bias voltage waveforms provided controlled ion energy and narrow energy spread, which are critical for highly selective and damage-free etching. The broad goal of the project was to investigate the plasma science and engineering that will lead to rapid ALET with monolayer precision. A combined experimental-simulation study was employed to achieve this goal.

  14. Demonstration of ECCD Stabilization of m/n =2/1 NTMs in the Equivalent Low-Torque ITER Baseline Scenario in DIII-D

    Science.gov (United States)

    La Haye, Robert; Strait, Edward; Olofsson, Kej; Welander, Anders; Hanson, Jeremy; Sauter, Olivier

    2017-10-01

    Experiments in DIII-D are studying how best to minimize the average Electron Cyclotron Current Drive power directed at q =2 for stabilization of neoclassical tearing modes in discharges with the ITER shape and equivalent low-torque, low q95 3.1 and low betaN 1.8. ITER relies on localized ECCD to stabilize NTMs that would otherwise wall-lock and lead to disruption. The work contrasts the control strategies of pre-emption by continuous ECCD at the rational surface (``Active Tracking'') vs. suppression by a pulse of ECCD whenever a growing mode is detected (``Catch & Subdue''). The large rho 0.75 for q =2 and concomitant low Te make the EC current drive relatively weak per MW so that the EC power from 4 5 well-aligned gyrotrons of 2.5 2.8 MW, is just marginal for stabilization at about 70% of the neutral beam injection power. The low-torque makes early mode detection and good initial alignment imperative for prompt suppression before wall-locking. Requirements for stabilization will be presented. Work supported by the US DOE under DE-FC02-04ER54698.

  15. Multi-phase simulation of fast ion profile flattening due to Alfvén eigenmodes in a DIII-D experiment

    International Nuclear Information System (INIS)

    Todo, Y.; Van Zeeland, M.A.; Bierwage, A.; Heidbrink, W.W.

    2014-01-01

    A multi-phase simulation that is a combination of classical simulation and hybrid simulation for energetic particles interacting with a magnetohydrodynamic (MHD) fluid is developed to simulate the nonlinear dynamics on the slowing down time scale of the energetic particles. The hybrid simulation code is extended with realistic beam deposition profile, collisions and losses, and is used for both the classical and hybrid phases. The code is run without MHD perturbations in the classical phase, while the interaction between the energetic particles and the MHD fluid is simulated in the hybrid phase. In a multi-phase simulation of DIII-D discharge #142111, the stored beam ion energy is saturated due to Alfvén eigenmodes (AE modes) at a level lower than in the classical simulation. After the stored fast ion energy is saturated, the hybrid simulation is run continuously. It is demonstrated that the fast ion spatial profile is significantly flattened due to the interaction with the multiple AE modes with amplitude v/v A  ∼ δB/B ∼ O(10 −4 ). The dominant AE modes are toroidal Alfvén eigenmodes (TAE modes), which is consistent with the experimental observation at the simulated moment. The amplitude of the temperature fluctuations brought about by the TAE modes is of the order of 1% of the equilibrium temperature. This is also comparable with electron cyclotron emission measurements in the experiment. (paper)

  16. DIVIMP modeling of the toroidally symmetrical injection of {sup 13}CH{sub 4} into the upper SOL of DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    McLean, A.G. [Institute for Aerospace Studies, University of Toronto, 4925 Dufferin Street, Toronto, Ont., M3H 5T6 (Canada)]. E-mail: adam.mclean@utoronto.ca; Elder, J.D. [Institute for Aerospace Studies, University of Toronto, 4925 Dufferin Street, Toronto, Ont., M3H 5T6 (Canada); Stangeby, P.C. [Institute for Aerospace Studies, University of Toronto, 4925 Dufferin Street, Toronto, Ont., M3H 5T6 (Canada); Allen, S.L. [University of California, San Diego, La Jolla, CA 92093-0417 (United States); Boedo, J.A. [University of Wisconsin, Madison, WI 53706 (United States); Brooks, N.H. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551 (United States); Fenstermacher, M.E. [University of California, San Diego, La Jolla, CA 92093-0417 (United States); Groth, M. [University of California, San Diego, La Jolla, CA 92093-0417 (United States); Lisgo, S. [Institute for Aerospace Studies, University of Toronto, 4925 Dufferin Street, Toronto, Ont., M3H 5T6 (Canada); Nagy, A. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Rudakov, D.L. [University of Wisconsin, Madison, WI 53706 (United States); Wampler, W.R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Watkins, J.G. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); West, W.P. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551 (United States); Whyte, D.G. [Sandia National Laboratory, P.O. Box 5800, Albuquerque, NM 87185 (United States)

    2005-03-01

    As part of a study of carbon-tritium co-deposition, we carried out an experiment on DIII-D involving a toroidally symmetric injection of {sup 13}CH{sub 4} at the top of a LSN discharge. A Monte Carlo code, DIVIMP-HC, which includes molecular breakup of hydrocarbons, was used to model the region near the puff. The interpretive analysis indicates a parallel flow in the SOL of M {sub parallel} {approx} 0.4 directed toward the inner divertor. The CH{sub 4} is ionized in the periphery of the SOL and so the particle confinement time, {tau} {sub c}, is not high, only {approx}5 ms, and about 4X lower than if the CH{sub 4} were ionized at the separatrix. For such a wall injection location, however, approximately 60-75% of the CH{sub 4} gets ionized to C{sup +}, C{sup 2+}, etc., and is efficiently transported along the SOL to the inner divertor, trapping hydrogen by co-deposition there.

  17. Nonlocal control of electron temperature in short-discharge plasma with active boundaries

    Science.gov (United States)

    Demidov, V. I.; Adams, S. F.; Bogdanov, E.; Koepke, M. E.; Kudryavtsev, A. A.

    2012-10-01

    It is known that boundaries are very important in formation of nonlocal plasma properties [1]. This study combines experimental and modeling demonstration of controlling electron temperature in a plasma with active boundaries. To demonstrate that, a short dc discharge with cold cathode and application of different voltages to the conducting discharge wall for argon plasma at 1 Torr pressure has been used in experiments and modeling. It is demonstrated in the model for this discharge that spatial distributions of electron density and temperature and argon metastable atom density depend on the dc voltage applied to different conducting parts of the wall. Applied voltage can trap within the device volume energetic electrons arising from atomic and molecular processes in the plasma. This leads to a modification in the heating of slow electrons by energetic electrons and as a result modifies the electron temperature. Conducted experiments also demonstrate that the measured electron temperature is a function of potential applied to the wall and it is possible to see increasing the electron temperature with increasing absolute value of the applied negative potential.[4pt] [1] E. Bogdanov, S. Adams, V. Demidov, A. Kudryavtsev, J. M. Williamson, Phys. Plasmas 17, 103502 (2010)

  18. Scattering of magnetized electrons at the boundary of low temperature plasmas

    Science.gov (United States)

    Krüger, Dennis; Trieschmann, Jan; Brinkmann, Ralf Peter

    2018-02-01

    Magnetized technological plasmas with magnetic fields of 10–200 mT, plasma densities of 1017‑1019 m‑3, gas pressures of less than 1 Pa, and electron energies from a few to (at most) a few hundred electron volts are characterized by electron Larmor radii r L, that are small compared to all other length scales of the system, including the spatial scale L of the magnetic field and the collisional mean free path λ. In this regime, the classical drift approximation applies. In the boundary sheath of these discharges, however, that approximation breaks down: The sheath penetration depth of electrons (a few to some ten Debye length λ D; depending on the kinetic energy; typically much smaller than the sheath thickness of tens/hundreds of λ D) is even smaller than r L. For a model description of the electron dynamics, an appropriate boundary condition for the plasma/sheath interface is required. To develop such, the interaction of magnetized electrons with the boundary sheath is investigated using a 3D kinetic single electron model that sets the larger scales L and λ to infinity, i.e. neglects magnetic field gradients, the electric field in the bulk, and collisions. A detailed comparison of the interaction for a Bohm sheath (which assumes a finite Debye length) and a hard wall model (representing the limit {λ }{{D}}\\to 0; also called the specular reflection model) is conducted. Both models are found to be in remarkable agreement with respect to the sheath-induced drift. It is concluded that the assumption of specular reflection can be used as a valid boundary condition for more realistic kinetic models of magnetized technological plasmas.

  19. Plasma-based actuators for turbulent boundary layer control in transonic flow

    Science.gov (United States)

    Budovsky, A. D.; Polivanov, P. A.; Vishnyakov, O. I.; Sidorenko, A. A.

    2017-10-01

    The study is devoted to development of methods for active control of flow structure typical for the aircraft wings in transonic flow with turbulent boundary layer. The control strategy accepted in the study was based on using of the effects of plasma discharges interaction with miniature geometrical obstacles of various shapes. The conceptions were studied computationally using 3D RANS, URANS approaches. The results of the computations have shown that energy deposition can significantly change the flow pattern over the obstacles increasing their influence on the flow in boundary layer region. Namely, one of the most interesting and promising data were obtained for actuators basing on combination of vertical wedge with asymmetrical plasma discharge. The wedge considered is aligned with the local streamlines and protruding in the flow by 0.4-0.8 of local boundary layer thickness. The actuator produces negligible distortion of the flow at the absence of energy deposition. Energy deposition along the one side of the wedge results in longitudinal vortex formation in the wake of the actuator providing momentum exchange in the boundary layer. The actuator was manufactured and tested in wind tunnel experiments at Mach number 1.5 using the model of flat plate. The experimental data obtained by PIV proved the availability of the actuator.

  20. Observation of Abrupt- and Fast-rising SOL Current during Trigger Phase of ELMs in DIII-D Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    H. Takahashi; E.D. Fredrickson; M.J. Schaffer; M.E. Austin; N.H. Brooks; T.E. Evans; G.L. Jackson; L.L. Lao; J.G. Watkins

    2005-06-27

    Extensive studies to date of edge localized modes (ELMs) have sought their origin inside the separatrix, i.e., MHD instability from steep gradients in the plasma edge, and examined their consequences outside the separatrix, i.e., transport of heat and particles in the scrape-off-layer (SOL) and divertors. Recent measurement by a high-speed scrape-off-layer current (SOLC) diagnostic may indicate that the ELM trigger process lies, in part, in the SOL. Thermoelectrically driven SOLC precedes, or co-evolves with, other parameters of the ELM process, and thus can potentially play a causal role: error field generated by non-axisymmetric SOLC, flowing in the immediate vicinity (approximately 1 cm) of the plasma edge, may contribute toward destabilizing MHD modes. The SOLC, observed concurrently with MHD activity, including ELMs, has been reported elsewhere.

  1. Modification of boundary plasma behavior by Ion Bernstein Wave heating on HT-7 tokamak

    International Nuclear Information System (INIS)

    Xu Guoshen

    2002-01-01

    Cooperated with Fusion Research Center, the University of Texas at Austin, U.S.A. The boundary plasma behavior during Ion Bernstein Wave (IBW) heating was investigated using Langmuir probe arrays on HT-7 tokamak. The particle confinement improvement of over a factor of 2 was observed in 30 MHz IBW heated plasma with RF power > 120 kW. The strong de-correlation effect of fluctuations resulted in that the turbulent particle flux dropped more than an order of magnitude. In IBW heated plasma, an additional inward E r and associated poloidal ExB flows were produced, which could account for the additional poloidal velocity in the electron diamagnetic direction in the scrape-of layer (SOL). Three-wave nonlinear phase coupling increased evidently and low frequency fluctuations (about 5 kHz) were generated, which dominated the boundary turbulence during IBW heating. The 5/2-D resonant layer was located in the plasma edge region, which is found to be the mechanism underlying these phenomena. (author)

  2. Effect of remote field electromagnetic boundary conditions on microwave-induced plasma torches

    Science.gov (United States)

    Jimenez-Diaz, M.; van Dijk, J.; van der Mullen, J. J. A. M.

    2011-04-01

    A flexible versatile electromagnetic model constructed with the PLASIMO platform is employed to explore electromagnetic features of microwave-induced plasma torches. The bases, formed by a full-vector formulation of the Maxwell equations, provide the possibility to formulate the boundary conditions in a natural way. Together with the use of a direct matrix solver this gives a convergence speed-up of more than a factor of 100 when compared with a scalar formulation on an azimuthal magnetic field that uses an iterative solver. As a result, this electromagnetic model is ready to act in future studies as part of the self-consistent description of plasma-electromagnetic coupling. With the electromagnetic model three types of configuration were studied: the closed, semi-open and open configurations, all three based on the same simplified model plasmas. It is found that the closed configuration, acting as a cavity for which (de)tuning is extremely sensitive for the plasma conditions, is less suitable for applications in which changes in plasma compositions can be expected. The semi-open configuration can be seen as a model for the practice often used in laboratories to place microwave-induced plasma torches in a grid that aims at protecting the environment against microwave electromagnetic radiation. Calculations show that this is good practice provided the radius of this cylindrical grid is in the order of 90 mm. For the most often used configuration, the open version, we found that the power balance as expressed by the coefficients of absorption, transmission and reflection depends on the electron density of the plasma. The reason is that the plasma acts as an antenna, which converts the electromagnetic waves from the coaxial structure to that of the expansion region, and that this antenna function depends on the electron density. The influence of various other antenna elements is investigated as well.

  3. Fatigue strength of inconel 625 plate and weldments used in the DIII-D configuration vacuum vessel

    Science.gov (United States)

    Trester, P. W.; Kaae, J. L.; Gallix, R.

    1985-08-01

    The Doublet III-D vacuum vessel will enable fusion experiments with large dee-configuration plasmas in the tokamak device at GA Technologies Inc. The vessel is a welded structure of the Ni-Cr-Mo-Nb alloy Inconel 625 in the form of mill-annealed plate and sheet. An emphasis was placed on specifying material with small grain size to enhance fatigue resistance. Fatigue strength was determined by tension-compression axial strain cycling of specimens excised from stock plate of 25 mm thickness. In addition tests were conducted on gas-tungsten-arc weldments. The strain ranges and the resulting alternating stresses, which produced fatigue lives between 10 3 and 10 6 cycles, were determined in the test program. Comparisons are presented with published data obtained using the rotating-beam test method. A discussion is also presented on the effect of the grain size on fatigue strength.

  4. Observation of SOL Current Correlated with MHD Activity in NBI-heated DIII-D Tokamak Discharges

    International Nuclear Information System (INIS)

    Takahashi, H.; Fredrickson, E.D.; Schaffer, M.J.; Austin, M.E.; Evans, T.E.; Lao, L.L.; Watkins, J.G.

    2004-01-01

    This work investigates the potential roles played by the scrape-off-layer current (SOLC) in MHD activity of tokamak plasmas, including effects on stability. SOLCs are found during MHD activity that are: (1) slowly growing after a mode-locking-like event, (2) oscillating in the several kHz range and phase-locked with magnetic and electron temperature oscillations, (3) rapidly growing with a sub-ms time scale during a thermal collapse and a current quench, and (4) spiky in temporal behavior and correlated with spiky features in Da signals commonly identified with the edge localized mode (ELM). These SOLCs are found to be an integral part of the MHD activity, with a propensity to flow in a toroidally non-axisymmetric pattern and with magnitude potentially large enough to play a role in the MHD stability. Candidate mechanisms that can drive these SOLCs are identified: (a) toroidally non-axisymmetric thermoelectric potential, (b) electromotive force (EMF) from MHD activity, and (c) flux swing, both toroidal and poloidal, of the plasma column. An effect is found, stemming from the shear in the field line pitch angle, that mitigates the efficacy of a toroidally non-axisymmetric SOLC to generate a toroidally non-axisymmetric error field. Other potential magnetic consequences of the SOLC are identified: (i) its error field can introduce complications in feedback control schemes for stabilizing MHD activity and (ii) its toroidally non-axisymmetric field can be falsely identified as an axisymmetric field by the tokamak control logic and in equilibrium reconstruction. The radial profile of a SOLC observed during a quiescent discharge period is determined, and found to possess polarity reversals as a function of radial distance

  5. Observation of SOL Current Correlated with MHD Activity in NBI-heated DIII-D Tokamak Discharges

    Energy Technology Data Exchange (ETDEWEB)

    H. Takahashi; E.D. Fredrickson; M.J. Schaffer; M.E. Austin; T.E. Evans; L.L. Lao; J.G. Watkins

    2004-03-26

    This work investigates the potential roles played by the scrape-off-layer current (SOLC) in MHD activity of tokamak plasmas, including effects on stability. SOLCs are found during MHD activity that are: (1) slowly growing after a mode-locking-like event, (2) oscillating in the several kHz range and phase-locked with magnetic and electron temperature oscillations, (3) rapidly growing with a sub-ms time scale during a thermal collapse and a current quench, and (4) spiky in temporal behavior and correlated with spiky features in Da signals commonly identified with the edge localized mode (ELM). These SOLCs are found to be an integral part of the MHD activity, with a propensity to flow in a toroidally non-axisymmetric pattern and with magnitude potentially large enough to play a role in the MHD stability. Candidate mechanisms that can drive these SOLCs are identified: (a) toroidally non-axisymmetric thermoelectric potential, (b) electromotive force (EMF) from MHD activity, and (c) flux swing, both toroidal and poloidal, of the plasma column. An effect is found, stemming from the shear in the field line pitch angle, that mitigates the efficacy of a toroidally non-axisymmetric SOLC to generate a toroidally non-axisymmetric error field. Other potential magnetic consequences of the SOLC are identified: (i) its error field can introduce complications in feedback control schemes for stabilizing MHD activity and (ii) its toroidally non-axisymmetric field can be falsely identified as an axisymmetric field by the tokamak control logic and in equilibrium reconstruction. The radial profile of a SOLC observed during a quiescent discharge period is determined, and found to possess polarity reversals as a function of radial distance.

  6. Edge stability and performance of the ELM-free quiescent H-mode and the quiescent double barrier mode on DIII-D

    International Nuclear Information System (INIS)

    West, W.P.; Burrell, K.H.; Snyder, P.B.; Gohil, P.; Lao, L.L.; Leonard, A.W.; Osborne, T.H.; Thomas, D.M.; Casper, T.A.; Lasnier, C.J.; Doyle, E.J.; Wang, G.; Zeng, L.; Nave, M.F.F.

    2005-01-01

    The quiescent H (QH) mode, an edge localized mode (ELM)-free, high-confinement mode, combines well with an internal transport barrier to form quiescent double barrier (QDB) stationary state, high performance plasmas. The QH-mode edge pedestal pressure is similar to that seen in ELMing phases of the same discharge, with similar global energy confinement. The pedestal density in early ELMing phases of strongly pumped counter injection discharges drops and a transition to QH-mode occurs, leading to lower calculated edge bootstrap current. Plasmas current ramp experiment and ELITE code modeling of edge stability suggest that QHmodes lie near an edge current stabilty boundary. At high triangularity, QH-mode discharges operate at higher pedestal density and pressure, and have achieved ITER level values of β PED and ν*. The QDB achieves performance of β N H 89 ∼ 7 in quasi-stationary conditions for a duration of 10 τ E , limited by hardware. Recently we demonstrated stationary state QDB discharges with little change in kinetic and q profiles (q 0 > 1) for 2 s, comparable to ELMing 'hybrid scenarios', yet without the debilitating effects of ELMs. Plasma profile control tools, including electron cyclotron heating and current drive and neutral beam heating, have been demonstrated to control simultaneously the q profile development, the density peaking, impurity accumulation and plasma beta. (author)

  7. Applying Boundary Conditions Using a Time-Dependent Lagrangian for Modeling Laser-Plasma Interactions

    Science.gov (United States)

    Reyes, Jonathan; Shadwick, B. A.

    2016-10-01

    Modeling the evolution of a short, intense laser pulse propagating through an underdense plasma is of particular interest in the physics of laser-plasma interactions. Numerical models are typically created by first discretizing the equations of motion and then imposing boundary conditions. Using the variational principle of Chen and Sudan, we spatially discretize the Lagrangian density to obtain discrete equations of motion and a discrete energy conservation law which is exactly satisfied regardless of the spatial grid resolution. Modifying the derived equations of motion (e.g., enforcing boundary conditions) generally ruins energy conservation. However, time-dependent terms can be added to the Lagrangian which force the equations of motion to have the desired boundary conditions. Although some foresight is needed to choose these time-dependent terms, this approach provides a mechanism for energy to exit the closed system while allowing the conservation law to account for the loss. An appropriate time discretization scheme is selected based on stability analysis and resolution requirements. We present results using this variational approach in a co-moving coordinate system and compare such results to those using traditional second-order methods. This work was supported by the U. S. Department of Energy under Contract No. DE-SC0008382 and by the National Science Foundation under Contract No. PHY- 1104683.

  8. Ion beam generation at the plasma sheet boundary layer by kinetic Alfven waves

    International Nuclear Information System (INIS)

    Moghaddam-Taaheri, E.; Goertz, C.K.; Smith, R.A.

    1989-01-01

    The kinetic Alfven wave, an Alfven wave with a perpendicular wavelength comparable to the ion gyroradius, can diffuse ions both in velocity and coordinate spaces with comparable transport rates. This may lead to the generation of ion beams in the plasma sheet boundary layer (PSBL). To investigate the ion beam generation process numerically, a two-dimensional quasi-linear code was constructed. Assuming that the plasma β (the ratio of plasma pressure to the magnetic pressure) varies from β = 1 to β << 1 across the magnetic field, the dynamics of the ion beam generation in the PSBL was studied. It was found that if your start with an ion distribution function which monotonically decreases with velocity along the magnetic field and a density gradient across the magnetic field, ions diffuse in velocity-coordinate space until nearly a plateau is established along the diffusion path. Depending on the topology of the magnetic field at the lobe side of the simulation system, i.e., open or closed field lines, the ion distribution function may or may not reach a steady state. If the field lines are open there, i.e., if the diffusion extends into the lobe, the double diffusion process may provide a mechanism for continuously transferring the ions from the central plasma sheet to the lobe. The authors comment on the effect of the particle loss on the establishment of the pressure balance in the plasma sheet

  9. Effects of the current boundary conditions at the plasma-gun gap on density in SSPX

    Science.gov (United States)

    Kolesnikov, Roman; Lodestro, L. L.; Meyer, W. H.

    2012-10-01

    The Sustained Spheromak Physics Experiment (SSPX) was a toroidal magnetic-confinement device without toroidal magnetic-field coils or a central transformer but which generated core-plasma currents by dynamo processes driven by coaxial plasma-gun injection into a flux-conserving vessel. Record electron temperatures in a spheromak (Te˜500eV) were achieved, and final results of the SSPX program were reported in [1]. Plasma density, which depended strongly on wall conditions, was an important parameter in SSPX. It was observed that density rises with Igun and that confinement improved as the density was lowered. Shortly after the last experiments, a new feature was added to the Corsica code's solver used to reconstruct SSPX equilibria. Motivated by n=0 fields observed in NIMROD simulations of SSPX, an insulating boundary condition was implemented at the plasma-gun gap. Using this option we will perform new reconstructions of SSPX equilibria and look for correlations between the location of the separatrix (which moves up the gun wall and onto the insulating gap as Igun increases) and plasma density and magnetic-flux amplification [2].[4pt] [1] H. S. McLean, APS, DPP, Dallas, TX, 2008.[0pt] [2] E. B. Hooper et al., Nucl. Fusion 47, 1064 (2007).

  10. Plasma boundary shape control and real-time equilibrium reconstruction on NSTX-U

    Science.gov (United States)

    Boyer, M. D.; Battaglia, D. J.; Mueller, D.; Eidietis, N.; Erickson, K.; Ferron, J.; Gates, D. A.; Gerhardt, S.; Johnson, R.; Kolemen, E.; Menard, J.; Myers, C. E.; Sabbagh, S. A.; Scotti, F.; Vail, P.

    2018-03-01

    The upgrade to the National Spherical Torus eXperiment (NSTX-U) included two main improvements: a larger center-stack, enabling higher toroidal field and longer pulse duration, and the addition of three new tangentially aimed neutral beam sources, which increase available heating and current drive, and allow for flexibility in shaping power, torque, current, and particle deposition profiles. To best use these new capabilities and meet the high-performance operational goals of NSTX-U, major upgrades to the NSTX-U control system (NCS) hardware and software have been made. Several control algorithms, including those used for real-time equilibrium reconstruction and shape control, have been upgraded to improve and extend plasma control capabilities. As part of the commissioning phase of first plasma operations, the shape control system was tuned to control the boundary in both inner-wall limited and diverted discharges. It has been used to accurately track the requested evolution of the boundary (including the size of the inner gap between the plasma and central solenoid, which is a challenge for the ST configuration), X-point locations, and strike point locations, enabling repeatable discharge evolutions for scenario development and diagnostic commissioning.

  11. Elemental and topographical imaging of microscopic variations in deposition on NSTX-U and DIII-D samples

    Science.gov (United States)

    Skinner, C. H.; Kaita, R.; Koel, B. E.; Chrobak, C. P.; Wampler, W. R.

    2017-10-01

    Tokamak plasma facing components (PFCs) have surface roughness that can cause microscopic spatial variations in erosion and deposition and hence influence material migration. Previous RBS measurements showed indirect evidence for this but the spatial (0.5mm) resolution was insufficient for direct imaging. We will present elemental images at sub-micron resolution of deposition on NSTX-U and DiMES samples that show strong microscopic variations and correlate this with 3D topographical maps of surface irregularities. The elemental imaging is performed with a Scanning Auger Microprobe (SAM) that measures element-specific Auger electrons excited by an SEM electron beam. 3D topographical maps of the samples are performed with a Leica DCM 3D confocal light microscope and compared to the elemental deposition pattern. The initial results appear consistent with erosion at the downstream edges of the surface pores exposed to the incident ion flux, whereas the deeper regions are shadowed and serve as deposition traps. Support was provided through DOE Contract Numbers DE-AC02-09CH11466, DE-FC02-04ER54698 and DE-NA0003525.

  12. Dust appearance rates during neutral beam injection and after oxygen bake in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Yu, J.H.; Smirnov, R.D.; Rudakov, D.L.

    2011-01-01

    A simple model to quantify source and sink terms of dust observed in tokamaks using fast visible imaging is presented. During neutral beam injection (NBI), dust appearance rates increase in front of the neutral beam port by up to a factor of 5. The images show dust streaming from the port box as previously settled dust becomes mobilized during beam injection. Following an oxygen bake and vent, the dust observation rate is a factor of 2 lower than that after a vessel entry vent with no oxygen bake. Detected dust levels decay on a shot-to-shot basis in a roughly exponential fashion, with a decay time of approximately 20 s of plasma exposure. Appearance rates of dust mass are estimated using assumed lognormal and power law functional forms for the dust size distribution. The two dust size distributions differ significantly on the amount the dust material carried by the largest particles, highlighting the need for further dust studies in order to make accurate forecasts to ITER.

  13. Transport studies in boundary and divertor plasmas of JT-60U

    International Nuclear Information System (INIS)

    Kumagai, Akira

    1999-03-01

    This thesis describes an investigation on transport of plasma, neutral particle and impurity in the boundary and divertor of the JT-60U tokamak to provide a better understanding of plasma-surface interactions and divertor physics. The asymmetry between the inboard and outboard divertor on plasma parameters (in-out asymmetry) are usually observed in tokamaks with the divertor. In this study, the in-out asymmetry was investigated under various plasma conditions and discharge parameters. The observed results were discussed with several mechanisms that can produce the in-out asymmetry. It was confirmed experimentally that the importance of each mechanism depends on the plasma parameters and discharge conditions. The current flowing in the scrape-off layer (SOL) due to the in-out asymmetry was observed. The SOL currents in the high density plasma with the occurrence of the plasma detachment were investigated for the first time in this study. The ion temperature in the divertor region is one of the most important factors for both generation and transport of impurity. However, the background ion temperature in the divertor region has not been measured in any tokamak so far. The ion temperature in the divertor region has been measured for the first time with the Doppler broading of the C 3+ ion emission line. The measured temperature was analyzed by an impurity particle transport code. The code calculation showed that the measured temperature reflects the low temperature at the outside of the separatrix in the inboard region. The spectral profile of Balmer-α (D α ) line emitted from the deuterium atoms reflects the velocity distribution of neutral particles by the Doppler effect and is effective for investigating the detailed neutral behavior and recycling process. The spatial variation of the D α line spectral profile in the divertor region has been measured for the first time in this study. The observed results were compared with the calculated one by a neutral

  14. Mathematical and numerical study of nonlinear boundary problems related to plasma physics

    International Nuclear Information System (INIS)

    Sermange, M.

    1982-06-01

    After the study of some equations based on the Hodgkin-Huxley model, the work presented here is concerned with nonlinear boundary problems in MHD. They are gathered in two subjects: equilibrium equations and stability equations. The axisymmetric MHD equilibrium equations with free boundary have been studied by different authors, particularly the existence, regularity, unicity and non-unicity. Here, bifurcation, convergence of calculation methods existence of solutions in a discontinuous frame are studied. MHD stability can be determined by the principle of Bernstein et al; the mathematical work concerned here bears on the equivalence, in the case of two-dimensional or axisymmetric stability, between this model and a scalar eigenvalue problem which is introduced. At last, modules for computing MHD equilibrium for the simulation of plasma confinement in a tokamak are described [fr

  15. Design and construction of Keda Space Plasma Experiment (KSPEX) for the investigation of the boundary layer processes of ionospheric depletions.

    Science.gov (United States)

    Liu, Yu; Zhang, Zhongkai; Lei, Jiuhou; Cao, Jinxiang; Yu, Pengcheng; Zhang, Xiao; Xu, Liang; Zhao, Yaodong

    2016-09-01

    In this work, the design and construction of the Keda Space Plasma EXperiment (KSPEX), which aims to study the boundary layer processes of ionospheric depletions, are described in detail. The device is composed of three stainless-steel sections: two source chambers at both ends and an experimental chamber in the center. KSPEX is a steady state experimental device, in which hot filament arrays are used to produce plasmas in the two sources. A Macor-mesh design is adopted to adjust the plasma density and potential difference between the two plasmas, which creates a boundary layer with a controllable electron density gradient and inhomogeneous radial electric field. In addition, attachment chemicals can be released into the plasmas through a tailor-made needle valve which leads to the generation of negative ions plasmas. Ionospheric depletions can be modeled and simulated using KSPEX, and many micro-physical processes of the formation and evolution of an ionospheric depletion can be experimentally studied.

  16. Stabilisation of a three-dimensional boundary layer by base-flow manipulation using plasma actuators

    International Nuclear Information System (INIS)

    Dörr, P C; Kloker, M J

    2015-01-01

    The applicability of dielectric barrier discharge plasma actuators for controlling the crossflow-vortex-induced laminar breakdown in a three-dimensional swept-wing-type boundary-layer flow is investigated using direct numerical simulation. Similar to the classical application of suction at the wall the aim is to modify the quasi two-dimensional base flow and to weaken primary crossflow (CF) instability, mainly due to a reduction of the basic CF. Not only localised volumetric forcing by plasma actuators but also CF counter-blowing and spots with a moving wall are investigated to identify effective fundamental mechanisms. It is found that counter blowing always results in partial blockage of the flow and eventually increased CF velocity, whereas moving-wall spots can slightly reduce the CF and the amplitude of crossflow vortices. Using discrete volumetric forcing a significant attenuation even of finite-amplitude crossflow vortices and thus a distinct transition delay is achieved. (paper)

  17. Finite element fluid modeling of axisymmetric magnetized boundary plasma with recycling neutrals

    International Nuclear Information System (INIS)

    Zanino, R.

    1992-01-01

    Finite elements should provide a natural and flexible method for fluid modeling of the tokamak SOL, in particular when the SOL geometry is complex, and/or the poloidal magnetic field is very inclined to the limiter/divertor target. Here we present a Galerkin finite element code, FELS, for transport modeling of a 2-fluid magnetized boundary plasma in an axisymmetry domain, in the presence of recycling neutrals. The classical collisional plasma dynamics along magnetic field lines is taken into account, and a simple diffusive Ansatz is used for the fluxes across magnetic surfaces; electric currents and diamagnetic flows are neglected for the time being. An analytical fluid model is used for the recycling neutrals. Results are shown and discussed for the case of a simple geometry. (orig.)

  18. A review on application of MHD theory to plasma boundary problems in tokamaks

    International Nuclear Information System (INIS)

    Itoh, Kimitaka.

    1992-08-01

    A survey is made on the problems of the edge plasmas, to which the analyses based on the MHD theory have been successfully applied. Also discussed are the efforts to extend the model equation to more general (and important as well) problems such as H-mode physics. An overview is first made on the advantages of the MHD picture, and the necessary supplementary physics are examined. Next, one- and two-dimensional models of the spatial structure of the edge plasma is discussed. The results on the stationary structure, both analytical and numerical, are reviewed: Typical example as well as the scaling law are shown. The instabilities associated with edge plasma is next reviewed. The surface kink mode, ballooning mode, interchange mode, resistive interchange mode and thermal instability are discussed. Role of the geometry such as the location of the X-point is studied. Influences of the atomic processes, and those of the radial electric field are also discussed. The analysis of the H-mode transition physics is finally discussed. The boundary plasma is a nonlinear media which possesses the possibility for bifurcation in which the radial electric field plays a key role. The model of the ion viscosity is also studied. Transition physics is developed. Analysis on the self-generating oscillation is shown and the relation with ELMs is discussed. After reviewing these problems, several comments are made to what directions the study can be deepened. (author) 53 refs

  19. Control of a shock wave-boundary layer interaction using localized arc filament plasma actuators

    Science.gov (United States)

    Webb, Nathan Joseph

    Supersonic flight is currently possible, but expensive. Inexpensive supersonic travel will require increased efficiency of high-speed air entrainment, an integral part of air-breathing propulsion systems. Although mixed compression inlet geometry can significantly improve entrainment efficiency, numerous Shock Wave-Boundary Layer Interactions (SWBLIs) are generated in this configuration. The boundary layer must therefore develop through multiple regions of adverse pressure gradient, causing it to thicken, and, in severe cases, separate. The associated increase in unsteadiness can have adverse effects on downstream engine hardware. The most severe consequence of these interactions is the increased aerodynamic blockage generated by the thickened boundary layer. If the increase is sufficient, it can choke the flow, causing inlet unstart, and resulting in a loss of thrust and high transient forces on the engine, airframe, and aircraft occupants. The potentially severe consequences associated with SWBLIs require flow control to ensure proper operation. Traditionally, boundary layer bleed has been used to control the interaction. Although this method is effective, it has inherent efficiency penalties. Localized Arc Filament Plasma Actuators (LAFPAs) are designed to generate perturbations for flow control. Natural flow instabilities act to amplify certain perturbations, allowing the LAFPAs to control the flow with minimal power input. LAFPAs also have the flexibility to maintain control over a variety of operating conditions. This work seeks to examine the effectiveness of LAFPAs as a separation control method for an oblique, impinging SWBLI. The low frequency unsteadiness in the reflected shock was thought to be the natural manifestation of a Kelvin-Helmholtz instability in the shear layer above the separation region. The LAFPAs were therefore placed upstream of the interaction to allow their perturbations to convect to the receptivity region (near the shear layer origin

  20. Plasma structure near the low-latitude boundary layer: A rebuttal

    International Nuclear Information System (INIS)

    Sckopke, N.

    1991-01-01

    A recent reanalysis of a well-documented interval of plasma and magnetic field data led its authors to offer a new model for the structure of the outer magnetosphere and the magnetosheath on the northern dawnside. On November 6, 1977, ISEE 1 and 2 had observed a series of quasi-periodic pulses of magnetosheath-like plasma on northward oriented magnetic field lines which were originally interpreted as repeated encounters of a pulsed low-latitude boundary layer inside a smooth magnetopause followed by a single outward crossing of the magnetopause. D.G. Sibeck and coworkers reinterpreted the ISEE observations as being due to quasi-periodic magnetopause motion causing the satellites to repeatedly exit the magnetosphere and to observe draped northward magnetosheath magnetic field lines in the plasma depletion layer. Their model is based on qualitative arguments concerning the amount of field line draping in the magnetosheath as well as the behavior of energetic electrons near the magnetopause. It is shown in this paper that both arguments are not in accordance with the available evidence

  1. Time-dependent free boundary equilibrium and resistive diffusion in a tokamak plasma

    International Nuclear Information System (INIS)

    Selig, G.

    2012-12-01

    In a Tokamak, in order to create the necessary conditions for nuclear fusion to occur, a plasma is maintained by applying magnetic fields. Under the hypothesis of an axial symmetry of the tokamak, the study of the magnetic configuration at equilibrium is done in two dimensions, and is deduced from the poloidal flux function. This function is solution of a non linear partial differential equation system, known as equilibrium problem. This thesis presents the time dependent free boundary equilibrium problem, where the circuit equations in the tokamak coils and passive conductors are solved together with the Grad-Shafranov equation to produce a dynamic simulation of the plasma. In this framework, the Finite Element equilibrium code CEDRES has been improved in order to solve the aforementioned dynamic problem. Consistency tests and comparisons with the DINA-CH code on an ITER vertical instability case have validated the results. Then, the resistive diffusion of the plasma current density has been simulated using a coupling between CEDRES and the averaged one-dimensional diffusion equation, and it has been successfully compared with the integrated modeling code CRONOS. (author)

  2. Observations of temporal and spatial behaviour of plasmas in relation to the interchange stability boundary scaling in GAMMA 10

    International Nuclear Information System (INIS)

    Minami, R.; Cho, T.; Kohagura, J.

    2002-01-01

    Observations of internal core plasma structural behaviour during the magnetohydrodynamic (MHD) destabilization of the central cell plasmas are carried out by the use of our developed semiconductor x-ray detector arrays installed in both central cell and anchor regions of the GAMMA 10 tandem mirror. In the present paper, it is found from the developed x-ray diagnostics that the bulk plasmas rotate without a change in its shape and structure with an ExB velocity during the destabilization. The onset of the off-axis rotation is identified to be closely related to a scaling of the MHD stability boundary (i.e. the anchor beta requirements for stabilizing central cell hot ion plasmas). These data confirm pressure driven interchange instability in tandem mirror plasmas, and reveal the rigid rotational bulk plasma structure as the first demonstrated interior plasma property during the destabilization. (author)

  3. Transport studies in boundary and divertor plasmas of JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Akira [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1999-03-01

    This thesis describes an investigation on transport of plasma, neutral particle and impurity in the boundary and divertor of the JT-60U tokamak to provide a better understanding of plasma-surface interactions and divertor physics. The asymmetry between the inboard and outboard divertor on plasma parameters (in-out asymmetry) are usually observed in tokamaks with the divertor. In this study, the in-out asymmetry was investigated under various plasma conditions and discharge parameters. The observed results were discussed with several mechanisms that can produce the in-out asymmetry. It was confirmed experimentally that the importance of each mechanism depends on the plasma parameters and discharge conditions. The current flowing in the scrape-off layer (SOL) due to the in-out asymmetry was observed. The SOL currents in the high density plasma with the occurrence of the plasma detachment were investigated for the first time in this study. The ion temperature in the divertor region is one of the most important factors for both generation and transport of impurity. However, the background ion temperature in the divertor region has not been measured in any tokamak so far. The ion temperature in the divertor region has been measured for the first time with the Doppler broading of the C{sup 3+} ion emission line. The measured temperature was analyzed by an impurity particle transport code. The code calculation showed that the measured temperature reflects the low temperature at the outside of the separatrix in the inboard region. The spectral profile of Balmer-{alpha} (D{sub {alpha}}) line emitted from the deuterium atoms reflects the velocity distribution of neutral particles by the Doppler effect and is effective for investigating the detailed neutral behavior and recycling process. The spatial variation of the D{sub {alpha}} line spectral profile in the divertor region has been measured for the first time in this study. The observed results were compared with the

  4. Design and first plasma measurements of the ITER-ECE prototype radiometer

    Energy Technology Data Exchange (ETDEWEB)

    Austin, M. E.; Brookman, M. W.; Rowan, W. L. [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States); Danani, S. [ITER-India/Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Bryerton, E. W.; Dougherty, P. [Virginia Diodes, Inc., Charlottesville, Virginia 22902 (United States)

    2016-11-15

    On ITER, second harmonic optically thick electron cyclotron emission (ECE) in the range of 220-340 GHz will supply the electron temperature (T{sub e}). To investigate the requirements and capabilities prescribed for the ITER system, a prototype radiometer covering this frequency range has been developed by Virginia Diodes, Inc. The first plasma measurements with this instrument have been carried out on the DIII-D tokamak, with lab bench tests and measurements of third through fifth harmonic ECE from high T{sub e} plasmas. At DIII-D the instrument shares the transmission line of the Michelson interferometer and can simultaneously acquire data. Comparison of the ECE radiation temperature from the absolutely calibrated Michelson and the prototype receiver shows that the ITER radiometer provides accurate measurements of the millimeter radiation across the instrument band.

  5. Researchers build a secure plasma prison

    International Nuclear Information System (INIS)

    Glanz, J.

    1995-01-01

    Research groups at Princeton University's Tokamak Fusion Test Reactor and at the DIII-D tokamak at General Atomics in San Diego have made a major breakthrough. By tailoring the magnetic fields with unprecedented finesse, they appear to have tamed the plasma instabilities that rattle and tear the fragile magnetic cage, allowing particles to leak out and limiting a tokamak's performance. In the process they increased the central density of the plasma by as much as threefold and reduced the particle leakage by a factor of 50

  6. Magnetic field structure near the plasma boundary in helical systems and divertor tokamaks

    International Nuclear Information System (INIS)

    Nagasaki, Kazunobu; Itoh, Kimitaka

    1990-02-01

    Magnetic field structure of the scrape off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall is calculated. In helical systems, the connection length, L, has a logarithmic dependence on the distance from the outermost magnetic surface or that from the residual magnetic islands. The effect of axisymmetric fields on the field structure is also determined. In divertor tokamaks, the connection length also has logarithmic properties near the separatrix. Even when the perturbations, which resonate to rational surfaces near the plasma boundary, are added, logarithmic properties still remain. We compare the connection length of torsatron/helical-heliotron systems with that of divertor tokamaks. It is found that the former is shorter than the latter by one order magnitude with similar aspect ratio. (author)

  7. Worldwide collaborative efforts in plasma control software development

    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.

    2008-01-01

    This presentation will describe the DIII-D collaborations with various tokamak experiments throughout the world which have adapted custom versions of the DIII-D plasma control system (PCS) software for their own use. Originally developed by General Atomics for use on the DIII-D tokamak, the PCS has been successfully installed and used for the NSTX experiment in Princeton, the MAST experiment in Culham UK, the EAST experiment in China, and the Pegasus experiment in the University of Wisconsin. In addition to these sites, a version of the PCS is currently being developed for use by the KSTAR tokamak in Korea. A well-defined and robust PCS software infrastructure has been developed to provide a common foundation for implementing the real-time data acquisition and feedback control codes. The PCS infrastructure provides a flexible framework that has allowed the PCS to be easily adapted to fulfill the unique needs of each site. The software has also demonstrated great flexibility in allowing for different computing, data acquisition and real-time networking hardware to be used. A description of the current PCS software architecture will be given along with experiences in developing and supporting the various PCS installations throughout the world

  8. Time-resolved measurements of hydrogen and deuterium fluxes in the ASDEX plasma boundary

    International Nuclear Information System (INIS)

    Roth, J.; Varga, P.; Martinelli, A.P.; Scherzer, B.M.U.; Chen, C.K.; Wampler, W.R.; Taglauer, E.

    1982-01-01

    Hydrogen and deuterium fluxes parallel to the toroidal magnetic field were measured in the plasma boundary of ASDEX using graphite collector probes. Time resolution of the order of 100 ms can be obtained by rotating the cylindrical probes behind an aperture during the discharge. The trapped amount of hydrogen was determined by subsequent thermal desorption; in the analyses of deuterium the D( 3 He,p) 4 He nuclear reaction was used. Both methods yield quantitative results. Measurements were done for limiter and divertor discharges in the range of 4 to 20 cm outside the limiter or separatrix. The time distributions show a maximum flux at the beginning and the end of the discharge. The relatively lower flux during the plateau phase of the discharge is in the range 10 15 to 2 x 10 17 cm - 2 sec - 1 , depending on the radial probe position; the maximum values are higher by a factor of 5 to 50. During neutral hydrogen injection, an additional maximum can be observed. The radial l/e-decay length is about 0.9 cm in front and 0.4 cm behind the fixed limiter. The results are compared with independent measurements in ASDEX and other plasma machines

  9. Material migration in tokamak plasmas with a three-dimensional boundary

    International Nuclear Information System (INIS)

    Laengner, Ruth

    2013-01-01

    In this work, the influence of a 3D boundary induced by resonant magnetic perturbations (RMPs) on the material migration, i.e. the erosion from wall material due to the plasma surface interaction and the transport of these impurities is investigated for the first time. With applied RMPs two new domains occur in the magnetic field structure. Three dimensional SOL flux tubes with predominantly transport parallel to short magnetic field lines and a region of longer stochastic field lines with diffusive gradient driven radial transport. The plasma wall interaction and the material transport in these domains were investigated. A globally higher radial electric field E r as well as local changes in the magnetic field structure such as pressure driven sonic flows or locally higher E r fields can potentially influence the material transport with applied RMPs. The experiments were performed at the tokamak TEXTOR, the RMPs were induced by the dynamic ergodic divertor (DED). The plasma discharges and DED application was chosen to have a spatially separated 3D structure to be able to investigate the underlying physics. Two spherical carbon test limiters were positioned in different poloidal and toroidal positions which allowed to analyse the material migration in a 3D SOL flux tube and a stochastic region at the same time. Methane doped with 13 C was injected through the test limiters during three different plasma scenarios, without RMPs, with static RMPs and an RMP sweep. The test limiters and the injected methane were monitored in situ with different cameras and spectrometers. The deposition of the injected particles was measured post mortem by colourimetry, nuclear reaction analysis and secondary ion mass spectrometry. The most profound change from no RMP to the RMP cases is a 90 re-direction of the low ionised carbon C + and C 2+ into the E r x B-drift direction. From a comparison of the experiments and numerical field line tracing it was found that this is a global effect

  10. Cluster view of the plasma sheet boundary layer and bursty bulk flow connection

    Directory of Open Access Journals (Sweden)

    O. W. Lennartsson

    2009-04-01

    Full Text Available The high-latitude boundaries of the plasma sheet (PSBL are dynamic latitude zones of recurring and transient (minutes to tens of minutes earthward and magnetic field-aligned bursts of plasma, each being more or less confined in longitude as well, whose ionic component is dominated by protons with flux, energies and density that are consistent with a central plasma sheet (CPS source at varying distance (varying rates of energy time dispersion, sometimes as close as the ~19 RE Cluster apogees, or closer still. The arguably most plausible source consists of so called "bursty bulk flows" (BBFs, i.e. proton bulk flow events with large, positive and bursty GSE vx. Known mainly from CPS observations made at GSE x>−30 RE, the BBF type events probably take place much further downtail as well. What makes the BBFs an especially plausible source are (1 their earthward bulk flow, which helps explain the lack of distinctive latitudinal PSBL energy dispersion, and (2 their association with a transient strong increase of the local tail Bz component ("local dipolarization". The enhanced Bz provides intermittent access to higher latitudes for the CPS plasma, resulting in local density reductions in the tail midplane, as illustrated here by proton data from the Cluster CIS CODIF instruments. Another sign of kinship between the PSBL bursts and the BBFs is their similar spatial fine structure. The PSBL bursts have prominent filaments aligned along the magnetic field with transverse flux gradients that are often characterized by local ~10 keV proton gyroradii scale size (or even smaller, as evidenced by Cluster measurements. The same kind of fine structure is also found during Cluster near-apogee traversals of the tail midplane, as illustrated here and implied by recently published statistics on BBFs obtained with Cluster multipoint observations at varying satellite separations. Altogether, the Cluster observations described here mesh rather well with theories

  11. Effect of magnetic boundary on edge plasma profiles studied using probe measurements in EXTRAP T2R

    OpenAIRE

    Moustaphawi, Hawra

    2012-01-01

    In this Master’s thesis project, several experiments are conducted under three different conditions in order to study their effect on the edge plasma profiles. In the first case, the standard case, there is no external interference and the plasma is studied under normal lab environments. In the second case, the plasma position inside the EXTRAP T2R device is changed by a few millimeters and in the third case a magnetic boundary is inserted into the experiment. For each set of the experiment, ...

  12. Erosion/redeposition analysis : status of modeling and code validation for semi-detached tokamak edge plasmas

    International Nuclear Information System (INIS)

    Brooks, J. N.

    1999-01-01

    We are analyzing erosion and tritium codeposition for ITER, DIII-D, and other devices with a focus on carbon divertor and metallic wall sputtering, for detached and semi-detached edge plasmas. Carbon chemical-sputtering hydrocarbon-transport is computed in detail using upgraded models for sputtering yields, species, and atomic and molecular processes. For the DIII-D analysis this includes proton impact and dissociative recombination for the full methane and higher hydrocarbon chains. Several mixed material (Si-C doping and Be/C) effects on erosion are examined. A semi-detached reactor plasma regime yields peak net wall erosion rates of ∼1.0 (Be), ∼0.3 (Fe), and ∼0.01 (W) cm/burn-yr, and ∼50 cm/burn-yr for a carbon divertor. Net carbon erosion is dominated by chemical sputtering in the ∼1-3 eV detached plasma zone. Tritium codeposition in divertor-sputtered redeposited carbon is high (∼10-20 g-T/1000 s ). Silicon and beryllium mixing tends to reduce carbon erosion. Initial hydrocarbon transport calculations for the DIII-D DiMES-73 detached plasma experiment show a broad spectrum of redeposited molecules with ∼90% redeposition fraction

  13. Final Report DOE Grant No. DE-FG03-01ER54617 Computer Modeling of Microturbulence and Macrostability Properties of Magnetically Confined Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Jean-Noel Leboeuf

    2004-03-04

    OAK-B135 We have made significant progress during the past grant period in several key areas of the UCLA and national Fusion Theory Program. This impressive body of work includes both fundamental and applied contributions to MHD and turbulence in DIII-D and Electric Tokamak plasmas, and also to Z-pinches, particularly with respect to the effect of flows on these phenomena. We have successfully carried out interpretive and predictive global gyrokinetic particle-in-cell calculations of DIII-D discharges. We have cemented our participation in the gyrokinetic PIC effort of the SciDAC Plasma Microturbulence Project through working membership in the Summit Gyrokinetic PIC Team. We have continued to teach advanced courses at UCLA pertaining to computational plasma physics and to foster interaction with students and junior researchers. We have in fact graduated 2 Ph. D. students during the past grant period. The research carried out during that time has resulted in many publications in the premier plasma physics and fusion energy sciences journals and in several invited oral communications at major conferences such as Sherwood, Transport Task Force (TTF), the annual meetings of the Division of Plasma Physics of the American Physical Society, of the European Physical Society, and the 2002 IAEA Fusion Energy Conference, FEC 2002. Many of these have been authored and co-authored with experimentalists at DIII-D.

  14. Final Report DOE Grant No. DE-FG03-01ER54617 Computer Modeling of Microturbulence and Macrostability Properties of Magnetically Confined Plasmas