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

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)

Plasma flow in the DIII-D divertor

International Nuclear Information System (INIS)

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

1998-07-01

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

Disruption Physics and Mitigation on DIII-D

International Nuclear Information System (INIS)

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

2005-01-01

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

Stability Limits of High-Beta Plasmas in DIII-D

International Nuclear Information System (INIS)

Strait, E.J.

2005-01-01

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

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)

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.

Stability in high gain plasmas in DIII-D

International Nuclear Information System (INIS)

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

1996-10-01

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

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

DIII-D edge physics database

International Nuclear Information System (INIS)

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

1992-03-01

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

Advanced divertor experiments on DIII-D

International Nuclear Information System (INIS)

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

1991-04-01

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

Dependence of β·τ on plasma shape in DIII-D

International Nuclear Information System (INIS)

Lazarus, E.A.

1993-05-01

In this paper we discuss the observed variation in plasma performance with plasma shape, in particular, we shall compare single and double null diverted plasmas. The product β·τ has been used as a figure-of-merit for comparing different toroidal magnetic configurations. Here we shall use it as the figure-of-merit for comparing differing configurations within the DIII-D tokamak

Status of DIII-D plasma control

International Nuclear Information System (INIS)

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

1995-10-01

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

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.

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

Science.gov (United States)

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

2017-10-01

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

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

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

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.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Effect of heating scheme on SOL width in DIII-D and EAST

Directory of Open Access Journals (Sweden)

L. Wang

2017-08-01

Full Text Available Joint DIII-D/EAST experiments in the radio-frequency (RF heated H-mode scheme with comparison to that of neutral beam (NB heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH H-mode plasmas.

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

International Nuclear Information System (INIS)

Simonen, T.C.; Baker, D.

1993-01-01

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

Dependence of {beta} {center_dot} {tau} on plasma shape in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Lazarus, E.A. [Oak Ridge National Lab., TN (United States)

1993-12-31

In this paper we discuss the observed variation in plasma performance with plasma shape, in particular, we shall compare single and double null diverted plasmas. The product {beta} {center_dot} {tau} has been used as a figure-of-merit for comparing different toroidal magnetic configurations. Here we shall use it as the figure-of-merit for comparing differing configurations within the DIII-D tokamak. (author) 5 refs., 5 figs.

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

Extending the capabilities of the DIII-D Plasma Control System for worldwide fusion research collaborations

International Nuclear Information System (INIS)

Penaflor, B.G.; Ferron, J.R.; Walker, M.L.; Humphreys, D.A.; Leuer, J.A.; Piglowski, D.A.; Johnson, R.D.; Xiao, B.J.; Hahn, S.H.; Gates, D.A.

2009-01-01

This paper will discuss the recent enhancements which have been made to the DIII-D Plasma Control System (PCS) in order to further extend its usefulness as a shared tool for worldwide fusion research. The PCS developed at General Atomics is currently being used in a number of fusion research experiments worldwide, including the DIII-D Tokamak Facility in San Diego, and most recently the KSTAR Tokamak in South Korea. A number of enhancements have been made to support the ongoing needs of the DIII-D Tokamak in addition to meeting the needs of other PCS users worldwide. Details of the present PCS hardware and software architecture along with descriptions of the latest enhancements will be given.

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

Automated Fault Detection for DIII-D Tokamak Experiments

International Nuclear Information System (INIS)

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

1999-01-01

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

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

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

Science.gov (United States)

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

2016-11-01

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

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

International Nuclear Information System (INIS)

Sayer, R.O.; Peng, Y.K.M.; Jardin, S.C.

1993-01-01

A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench and current quench. With a suitable halo surrounding the main plasma, the TSC predictions are in good agreement with experimental results for the plasma current decay, plasma trajectory, toroidal and poloidal vessel currents, and for the magnetic probe and flux loop values for the entire VDE. Simulations with no plasma halo yield much faster vertical motion and significantly worse agreement with the magnetics and flux loop data than do halo simulations. (author). 12 refs, 13 figs

Radiative divertor plasmas with convection in DIII-D

International Nuclear Information System (INIS)

Leonard, A.W.; Porter, G.D.; Wood, R.D.; Allen, S.L.; Boedo, J.; Brooks, N.H.; Evans, T.E.; Fenstermacher, M.E.; Hill, D.N.; Isler, R.C.; Lasnier, C.J.; Lehmer, R.D.; Mahdavi, M.A.; Maingi, R.; Moyer, R.A.; Petrie, T.W.; Schaffer, M.J.; Wade, M.R.; Watkins, J.G.; West, W.P.; Whyte, D.G.

1998-01-01

The radiation of divertor heat flux on DIII-D [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low-Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction-dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE [T. Rognlien, J. L. Milovich, M. E. Rensink, and G. D. Porter, J. Nucl. Mater. 196 endash 198, 347 (1992)] has reproduced many of the observed experimental features. copyright 1998 American Institute of Physics

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.

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.

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

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.

Wall stabilization of high beta plasmas in DIII-D

International Nuclear Information System (INIS)

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

1995-02-01

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

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

International Nuclear Information System (INIS)

1996-09-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Plasma rotation and rf heating in DIII-D

International Nuclear Information System (INIS)

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

1999-05-01

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

Plasma rotation and rf heating in DIII-D

International Nuclear Information System (INIS)

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

1999-01-01

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

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.

Dust Studies in DIII-D and TEXTOR

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-03-01

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

Recycling and particle control in DIII-D

International Nuclear Information System (INIS)

Jackson, G.L.

1991-11-01

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

ICRH coupling in DIII-D

International Nuclear Information System (INIS)

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

1986-01-01

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

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

Science.gov (United States)

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

2006-10-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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.

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Dust Studies in DIII-D and TEXTOR

International Nuclear Information System (INIS)

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

2009-01-01

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

Magnetic barriers and their q95 dependence at DIII-D

Science.gov (United States)

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

2012-05-01

It is well known that externally generated resonant magnetic perturbations (RMPs) can form islands in the plasma edge. In turn, large overlapping islands generate stochastic fields, which are believed to play a role in the avoidance and suppression of edge localized modes (ELMs) at DIII-D. However, large coalescing islands can also generate, in the middle of these stochastic regions, KAM surfaces effectively acting as ‘barriers’ against field-line dispersion and, indirectly, particle diffusion. It was predicted in Ali and Punjabi (2007 Plasma Phys. Control. Fusion 49 1565-82) that such magnetic barriers can form in piecewise analytic DIII-D plasma equilibria. In this work, the formation of magnetic barriers at DIII-D is corroborated by field-line tracing calculations using experimentally constrained EFIT (Lao et al 1985 Nucl. Fusion 25 1611) DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. According to these calculations, the occurrence and location of magnetic barriers depend on the edge safety factor q95. It was thus suggested that magnetic barriers might contribute to narrowing the edge stochastic layer and play an indirect role in the RMPs failing to control ELMs for certain values of q95. The analysis of DIII-D discharges where q95 was varied, however, does not show anti-correlation between barrier formation and ELM suppression.

Magnetic barriers and their q95 dependence at DIII-D

International Nuclear Information System (INIS)

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

2012-01-01

It is well known that externally generated resonant magnetic perturbations (RMPs) can form islands in the plasma edge. In turn, large overlapping islands generate stochastic fields, which are believed to play a role in the avoidance and suppression of edge localized modes (ELMs) at DIII-D. However, large coalescing islands can also generate, in the middle of these stochastic regions, KAM surfaces effectively acting as ‘barriers’ against field-line dispersion and, indirectly, particle diffusion. It was predicted in Ali and Punjabi (2007 Plasma Phys. Control. Fusion 49 1565–82) that such magnetic barriers can form in piecewise analytic DIII-D plasma equilibria. In this work, the formation of magnetic barriers at DIII-D is corroborated by field-line tracing calculations using experimentally constrained EFIT (Lao et al 1985 Nucl. Fusion 25 1611) DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. According to these calculations, the occurrence and location of magnetic barriers depend on the edge safety factor q 95 . It was thus suggested that magnetic barriers might contribute to narrowing the edge stochastic layer and play an indirect role in the RMPs failing to control ELMs for certain values of q 95 . The analysis of DIII-D discharges where q 95 was varied, however, does not show anti-correlation between barrier formation and ELM suppression. (paper)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

Design of DIII-D advanced divertor

International Nuclear Information System (INIS)

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

1989-01-01

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

Design of DIII-D Advanced Divertor

International Nuclear Information System (INIS)

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

1989-11-01

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

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

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.

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

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.

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

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

International Nuclear Information System (INIS)

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

1997-06-01

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

Simulation of DIII-D Flat q Discharges

International Nuclear Information System (INIS)

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

2004-01-01

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

Methane penetration in DIII-D ELMing H-mode plasmas

International Nuclear Information System (INIS)

West, W.P.; Lasnier, C.J.; Whyte, D.G.; Isler, R.C.; Evans, T.E.; Jackson, G.L.; Rudakov, D.; Wade, M.R.; Strachan, J.

2003-01-01

Carbon penetration into the core plasma during midplane and divertor methane puffing has been measured for DIII-D ELMing H-mode plasmas. The methane puffs are adjusted to a measurable signal, but global plasma parameters are only weakly affected (line average density, e > increases by E , drops by 6+ density profiles in the core measured as a function of time using charge exchange recombination spectroscopy. The methane penetration factor is defined as the difference in the core content with the puff on and puff off, divided by the carbon confinement time and the methane puffing rate. In ELMing H-mode discharges with ion ∇B drift direction into the X-point, increasing the line averaged density from 5 to 8x10 19 m -3 dropped the penetration factor from 6.6% to 4.6% for main chamber puffing. The penetration factor for divertor puffing was below the detection limit (<1%). Changing the ion ∇B drift to away from the X-point decreased the penetration factor by more than a factor of five for main chamber puffing

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).

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.

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

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

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

Comparison of calculated neutral beam shine through with measured shine-through in DIII-D

International Nuclear Information System (INIS)

Chiu, H.K.; Hong, R.

1997-11-01

A comparison of the calculated shine through of neutral particle beams in the DIII-D plasma to measured values inferred from the target temperature rise is reported. This provides an opportunity to verify the shine through calculations and makes them more reliable in those cases where the shine through can not be measured. The DIII-D centerpost neutral beam target tiles are safe-guarded against excessive beam shine-through by pyrometry and thermocouple (TC) arrays on the tiles. Shine-through beam power is calculated from the measured temperature changes reported by the target tile TC array. These measurements are performed at the beginning of each operational year at DIII-D. Theoretically, the beam energy deposited into the plasma can be expressed as a function of the change in beam density. Neutral beam energy deposition in plasma (of known density) is inferred by comparing the results of a series of shine-through measurements for the 1997 campaign at DIII-D to the expected shine-through given by theory

Upgrade of the DIII-D RF systems

International Nuclear Information System (INIS)

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

1995-10-01

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

Far SOL transport and main wall plasma interaction in DIII-D

International Nuclear Information System (INIS)

Rudakov, D.L.; Boedo, J.A.; Moyer, R.A.; Doerner, R.P.; Hollmann, E.M.; Krasheninnikov, S.I.; Pigarov, A.Yu.; Stangeby, P.C.; McLean, A.G.; Watkins, J.G.; Wampler, W.R.; Whyte, D.G.; McKee, G.R.; Zeng, L.; Wang, G.; Brooks, N.H.; Evans, T.E.; Leonard, A.W.; Mahdavi, M.A.; West, W.P.; Wong, C.P.C.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J.

2005-01-01

Far scrape-off layer (SOL) and near-wall plasma parameters in DIII-D depend strongly on the discharge parameters and confinement regime. In L-mode discharges cross-field transport increases with the average discharge density and flattens far SOL profiles, thus increasing plasma-wall contact. In H-mode between edge localized modes (ELMs), plasma-wall contact is generally weaker than in L-mode. During ELMs plasma fluxes to the wall increase to, or above the L-mode levels. Depending on the discharge conditions ELMs are responsible for 30-90% of the ion flux to the outboard chamber wall. Cross-field fluxes in far SOL are dominated by large amplitude intermittent transport events that may propagate all the way to the outer wall and cause sputtering. A Divertor Material Evaluation System (DiMES) probe containing samples of several ITER-relevant materials including carbon, beryllium and tungsten was exposed to a series of upper single null (USN) discharges as a proxy to measure the first wall erosion. (author)

Recent DIII-D results

International Nuclear Information System (INIS)

Petersen, P.I.

1994-07-01

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

A phase contrast interferometer on DIII-D

International Nuclear Information System (INIS)

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

1992-04-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Non-linear instability of DIII-D to error fields

International Nuclear Information System (INIS)

La Haye, R.J.; Scoville, J.T.

1991-10-01

Otherwise stable DIII-D discharges can become nonlinearly unstable to locked modes and disrupt when subjected to resonant m = 2, n = 1 error field caused by irregular poloidal field coils, i.e. intrinsic field errors. Instability is observed in DIII-D when the magnitude of the radial component of the m = 2, n = 1 error field with respect to the toroidal field is B r21 /B T of about 1.7 x 10 -4 . The locked modes triggered by an external error field are aligned with the static error field and the plasma fluid rotation ceases as a result of the growth of the mode. The triggered locked modes are the precursors of the subsequent plasma disruption. The use of an ''n = 1 coil'' to partially cancel intrinsic errors, or to increase them, results in a significantly expanded, or reduced, stable operating parameter space. Precise error field measurements have allowed the design of an improved correction coil for DIII-D, the ''C-coil'', which could further cancel error fields and help to avoid disruptive locked modes. 6 refs., 4 figs

110GHz ECH on DIII-D

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Science.gov (United States)

Maingi, R.

1996-11-01

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

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

Science.gov (United States)

Ali, Halima; Punjabi, Alkesh

2013-05-01

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

Scaling of H-mode pedestal characteristics in DIII-D and C-Mod

International Nuclear Information System (INIS)

Granetz, R.S.; Boivin, R.L.; Osborne, T.H.

1999-01-01

Since the H-mode edge pedestal effectively sets the boundary conditions for energy transport throughout the core, a better understanding of the pedestal region is necessary in order to fully predict H-mode performance. Pedestal characteristics in the DIII-D and Alcator C-Mod tokamaks are described, and scalings of the pedestal width with various plasma parameters are shown. The pedestal width in both tokamaks varies in an inverse sense with plasma current, and is independent of toroidal field. Other similarities, as well as differences, are discussed. It is also found that the pedestal widths of the various physical quantities involved (T e , T i , n e , n i ) may be different. (author)

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

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Fast wave current drive experiment on the DIII-D tokamak

International Nuclear Information System (INIS)

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

1992-06-01

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

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

Neutral beam current drive scaling in DIII-D

International Nuclear Information System (INIS)

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

1989-03-01

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

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.

Experiments at high elongations in DIII-D

International Nuclear Information System (INIS)

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

1990-01-01

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

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

Science.gov (United States)

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

2010-11-01

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

Pedestal performance dependence upon plasma shape in DIII-D

International Nuclear Information System (INIS)

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

2007-01-01

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

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)

Simultaneous Feedback Control of Plasma Rotation and Stored Energy on the DIII-D Tokamak

International Nuclear Information System (INIS)

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

2006-01-01

One of the major modifications made to the DIII-D tokamak during the 2005 Long Torus Opening was the rotation of one of the four two-source neutral beam injection systems. Prior to this modification, all beams injected power with a component in the same direction as the usual plasma current ('' co-injection ''). Starting in early 2006, two of the seven beams inject with a component in the opposite direction ('' counter-injection ''). This new capability allows, for the first time, a partial decoupling of the injected energy and momentum during neutral beam heating experiments. An immediate advantage of mixed co- and counter-injection beams is the capability to control the plasma rotation velocity. High beta plasmas can now be studied over a wide range of the plasma rotation velocity. The stabilizing effect of rotation on the resistive wall mode (RWM), for example, can be directly compared to the stabilization achieved by external feedback coils. This is an advantage over previous techniques to control plasma rotation, such as magnetic braking, which have had only limited success. We describe development and implementation of a model-based control algorithm for simultaneous regulation of plasma rotation and beta. The model includes the two relevant plasma states (plasma rotation and stored energy), and describes the dynamic effects of the relevant actuators on those states. The actuators include the applied beam torque and beam power, which depend on the amount of co and counter-injected beams. Implementation of the model-based control within the plasma control system (PCS) [B.G. Penaflor, et al, '' Current Status of DIII-D Plasma Control System Computer Upgrades,'' Fusion Eng. and Design 71 (2004) 47] requires real-time measurements of the plasma rotation, obtained from the charge exchange recombination (CER) diagnostic, and stored energy calculated by the real-time EFIT equilibrium reconstruction. Details of this model and its development, and a comparison with

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

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

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.

Analysis of toroidal rotation data for the DIII-D tokamak

International Nuclear Information System (INIS)

John, H.St.; Burrell, K.H.; Groebner, R.; DeBoo, J.; Gohil, P.

1989-01-01

Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner et al. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Similar studies have been previously reported for Doublet III, ASDEX, TFTR, JET and other tokamaks. (author) 13 refs., 4 figs

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Effects of neutrals on plasma rotation in DIII-D

International Nuclear Information System (INIS)

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

1997-01-01

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

Currents in the DIII-D Tokamak

Science.gov (United States)

Azari, A.; Eidietis, N. W.

2012-10-01

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

Comparison of discharges with core transport barriers on DIII-D and JET

International Nuclear Information System (INIS)

Luce, T.C.; Alper, B.; Challis, C.D.

1997-07-01

The basic phenomenology of discharges with core transport barriers is the same for DIII-D and JET. The limitations on performance in both cases are well described by MHD stability calculations. Since the discharge behavior of the two machines is so similar, it seems reasonable to apply a simple parameterization of fusion performance which describes well the best performance discharges on DIII-D. The highest fusion performance shot on JET has Q DD = 3.1 10 -3 at 3.2 MA. Scaling from the highest Q DD DIII-D single-null discharge would predict Q DD = 4.2 10 -3 for JET. Raising the plasma current to 4.0 MA would increase the projection to 6.6 10 -3 . Realization of such performance would require significant effort to develop lower q plasmas with an H-mode edge. Because the performance is so closely tied to the current profile, this class of discharges also shows significant potential for steady state if current profile control can be demonstrated

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.)

Control of plasma poloidal shape and position in the DIII-D tokamak

International Nuclear Information System (INIS)

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

1997-11-01

Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks

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.

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)

An advanced plasma control system for the DIII-D tokamak

International Nuclear Information System (INIS)

Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J.; Lazarus, E.

1991-11-01

An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Development of this system is expected to lead to control system technology appropriate for use on future tokamaks such as ITER and BPX. The digital system will allow for increased precision in shape control through real time adjustment of the control algorithm to changes in the shape and discharge parameters such as β p , ell i and scrape-off layer current. The system will be used for research on real time optimization of discharge performance for disruption avoidance, current and pressure profile control, optimization of rf antenna loading, or feedback on heat deposition patterns through divertor strike point position control, for example. Shape control with this system is based on linearization near a target shape of the controlled parameters as a function of the magnetic diagnostic signals. This digital system is unique in that it is designed to have the speed necessary to control the unstable vertical motion of highly elongated tokamak discharges such as those produced in DIII-D and planned for BPX and ITER. a 40 MHz Intel i860 processor is interfaced to up to 112 channels of analog input signals. The commands to the poloidal field coils can be updated at 80 μs intervals for the control of vertical position with a delay between sampling of the analog signal and update of the command of less than 80 μs

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

Recent DIII-D high power heating and current drive experiments

International Nuclear Information System (INIS)

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

1994-02-01

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

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

International Nuclear Information System (INIS)

Politzer, P.A.

2005-01-01

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

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Recent DIII-D high power heating and current drive experiments

International Nuclear Information System (INIS)

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

1995-01-01

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

Recent DIII-D high power heating and current drive experiments

Energy Technology Data Exchange (ETDEWEB)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Analysis of toroidal rotation data for the DIII-D tokamak

International Nuclear Information System (INIS)

St John, H.; Stroth, U.; Burrell, K.H.; Groebner, R.J.; DeBoo, J.C.; Gohil, P.

1989-01-01

Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Our results are based on numerical inversions using the transport code ONETWO, modified to account for the radial diffusion of toroidal angular momentum. 13 refs., 4 figs

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)

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

UEDGE code comparisons with DIII-D bolometer data

Energy Technology Data Exchange (ETDEWEB)

Daniel, J.M.

1994-12-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

The Resistive Wall Mode Feedback Control System on DIII-D

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

This article is a continuation of previous work on determining the outgassing characteristics of materials used in the DIII-D magnetic fusion tokamak [K. L. Holtrop, J. Vac. Sci. Technol. A 17, 2064 (1999)]. Achievement of high performance plasma discharges in the DIII-D tokamak requires careful control of impurity levels. Among the techniques used to control impurities are routine bakes of the vacuum vessel to an average temperature of 350 deg. C. Materials used in DIII-D must release only very small amounts of impurities (below 2x10 -6 mole) at this temperature that could be transferred to the first wall materials and later contaminate plasma discharges. To better study the behavior of materials proposed for use in DIII-D at elevated temperatures, the initial outgassing test chamber was improved to include an independent heating control of the sample and a simple load lock chamber. The goal was to determine not only the total degassing rate of the material during baking, but to also determine the gas species composition and to obtain a quantitative estimate of the degassing rate of each species by the use of a residual gas analyzer. Initial results for aluminum anodized using three different processes, stainless steel plated with black oxide and black chrome, and a commercially available fiber optic feedthrough will be presented

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

International Nuclear Information System (INIS)

Carlstrom, T.N.; Gohil, P.; Watkins, J.C.

1994-01-01

We present the global analysis of a recent survey of the H-mode power threshold in DIII-D using D o → D + NBI after boronization of the vacuum vessel. Single parameter scans of B T , I p , density, and plasma shape have been carried out on the DIII-D tokamak for neutral beam heated single-null and double-null diverted plasmas. In single-null discharges, the power threshold is found to increase approximately linearly with B T and n e but remains independent of I p . In double-null discharges, the power threshold is found to be approximately independent of both B T and n e . Various shape parameters such as plasma-wall gaps had only a weak effect on the power threshold. Imbalancing the double null configuration resulted in a large increase in the threshold power

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

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

Using AORSA to simulate helicon waves in DIII-D

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-10

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

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

Comparing 1.5D ONETWO and 2D SOLPS analyses of inter-ELM H-mode plasma in DIII-D

International Nuclear Information System (INIS)

Owen, Larry W.; Canik, John; Groebner, R.; Callen, J.D.; Bonnin, X.; Osborne, T.H.

2010-01-01

A DIII-D inter-ELM H-mode plasma that is in approximate transport equilibrium is analysed with the 1.5D ONETWO core code and the 2D SOLPS code. In order to investigate the importance of core-edge coupling and 2D effects, including divertor fuelling across the X-point and poloidal asymmetries that are not explicitly included in ONETWO, the domain of SOLPS is extended to very near the magnetic axis. Two principal objectives are (1) to determine whether poloidal asymmetries in the plasma distributions are large enough to vitiate a core-type interpretive plasma transport analysis and (2) to determine whether the interpretive transport coefficients and neutral beam power and particle sources from ONETWO, when used in 2D SOLPS full plasma simulations, yield the same quality fits to the measured upstream density and temperature profiles as obtained with ONETWO. Results show that only a small increase in the separatrix value of the particle diffusion coefficient, and no change in the thermal diffusivities from ONETWO was needed to get excellent agreement of the upstream SOLPS density and temperature profiles and the Thomson scattering and CER data. Good agreement of the ONETWO and SOLPS flux surface averaged distributions of the core electron and D+ densities and temperatures are also obtained. Likewise the C6+ density, with a simple chemical sputtering model based on a constant fraction of the divertor D+ flux, the core heat and particle fluxes and the neutral density reveal no 2D effects in the core/pedestal region that would vitiate a 1.5D treatment of the inter-ELM H-mode plasma.

Advanced tokamak physics in DIII-D

Energy Technology Data Exchange (ETDEWEB)

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

2000-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-12-15

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

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

Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

Directory of Open Access Journals (Sweden)

T.D. Rognlien

2017-08-01

Full Text Available A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt, is provided by divertor Thomson scattering (DTS. Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te and density (ne across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

Performance, diagnostics, controls and plans for the gyrotron system on the DIII-D tokamak

Directory of Open Access Journals (Sweden)

Ponce D.M.

2012-09-01

Full Text Available The DIII-D ECH complex is being upgraded with three new depressed collector gyrotrons. The performance of the existing system has been very good. As more gyrotrons having higher power are added to the system, diagnostics of gyrotron operation, optimization of the performance and qualification of components for higher power become more important. A new FPGA-based gyrotron control system is being installed, additional capabilities for rapid real time variation of the rf injection angles by the DIII-D Plasma Control System are being tested and infrastructure enhancements are being completed. Longer term plans continue to include ECH as a major component in the DIII-D heating and current drive capabilities.

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

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

Particle control in DIII-D with helium glow discharge conditioning

International Nuclear Information System (INIS)

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

1990-01-01

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

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

International Nuclear Information System (INIS)

La Haye, R.J.

2005-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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.))

Development of a radiative divertor for DIII-D

Energy Technology Data Exchange (ETDEWEB)

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

1995-04-01

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

Thermal-stress analysis and testing of DIII-D armor tiles

International Nuclear Information System (INIS)

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

1987-10-01

It is planned to install about 1500 new armor tiles in the DIII-D tokamak. The armor tiles currently installed in DIII-D are made by brazing Poco AXF-5Q graphite onto Inconel X-750 stock. A small percentage of these have failed by breakage of graphite. These failures were believed to be related to significant residual stress remaining in graphite after brazing. Hence, an effort was undertaken to improve the design with all-graphite tiles. Three criteria must be satisfied by the armor tiles and the hardware used to attach the tiles to the vessel walls: tiles should not structurally fail, peak tile temperature must be less than 2500 K, and peak vessel stresses must be below acceptable levels. A number of alternate design concepts were first analyzed with the two-dimensional finite element codes TOPAZ2D and NIKE2D. Promising designs were optimized for best parameters such as thicknesses, etc. The two best designs were further analyzed for thermal stresses with the three-dimensional codes P/THERMAL and P/STRESS. Prototype tiles of a number of materials were fabricated by GA and tested at the Plasma Materials Test Facility of the Sandia National Laboratory at Albuquerque. The tests simulated the heat flux and cooling conditions in DIII-D. This paper describes the 2-D and 3-D thermal stress analyses, the test results and logic which led to the selected design of the DIII-D armor tiles. 5 refs., 7 figs., 3 tabs

Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

Science.gov (United States)

Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

2000-12-01

Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

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.

COMPLETE SUPPRESSION OF THE m=2/n-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D

International Nuclear Information System (INIS)

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

2003-01-01

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

Experiments at high elongations in DIII-D

International Nuclear Information System (INIS)

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

1990-06-01

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

Advanced tokamak research in DIII-D

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

1990-10-01

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

Automated Calculation of DIII-D Neutral Beam Availability

International Nuclear Information System (INIS)

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

1999-01-01

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

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.

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

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

Lack of dependence on resonant error field of locked mode island size in ohmic plasmas in DIII-D

Science.gov (United States)

La Haye, R. J.; Paz-Soldan, C.; Strait, E. J.

2015-02-01

DIII-D experiments show that fully penetrated resonant n = 1 error field locked modes in ohmic plasmas with safety factor q95 ≳ 3 grow to similar large disruptive size, independent of resonant error field correction. Relatively small resonant (m/n = 2/1) static error fields are shielded in ohmic plasmas by the natural rotation at the electron diamagnetic drift frequency. However, the drag from error fields can lower rotation such that a bifurcation results, from nearly complete shielding to full penetration, i.e., to a driven locked mode island that can induce disruption. Error field correction (EFC) is performed on DIII-D (in ITER relevant shape and safety factor q95 ≳ 3) with either the n = 1 C-coil (no handedness) or the n = 1 I-coil (with ‘dominantly’ resonant field pitch). Despite EFC, which allows significantly lower plasma density (a ‘figure of merit’) before penetration occurs, the resulting saturated islands have similar large size; they differ only in the phase of the locked mode after typically being pulled (by up to 30° toroidally) in the electron diamagnetic drift direction as they grow to saturation. Island amplification and phase shift are explained by a second change-of-state in which the classical tearing index changes from stable to marginal by the presence of the island, which changes the current density profile. The eventual island size is thus governed by the inherent stability and saturation mechanism rather than the driving error field.

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Lower-hybrid counter current drive for edge current density modification in DIII-D

International Nuclear Information System (INIS)

Fenstermacher, M.E.; Nevins, W.M.; Porkolab, M.; Bonoli, P.T.; Harvey, R.W.

1994-01-01

Each of the Advanced Tokamak operating modes in DIII-D is thought to have a distinctive current density profile. So far these modes have only been achieved transiently through experiments which ramp the plasma current and shape. Extension of these modes to steady state requires non-inductive current profile control, e.g., with lower hybrid current drive (LHCD). Calculations of LHCD have been done for DIII-D using the ACCOME and CQL3D codes, showing that counter driven current at the plasma edge can cancel some of the undesirable edge bootstrap current and potentially extend the VH-mode. Results will be presented for scenarios using 2.45 GHz LH waves launched from both the midplane and off-axis ports. The sensitivity of the results to injected power, n e and T e , and launched wave spectrum will also be shown

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

International Nuclear Information System (INIS)

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

1999-06-01

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-05-01

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

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

International Nuclear Information System (INIS)

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

1993-04-01

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

Dependence of helium transport on plasma current and ELM frequency in H-mode discharges in DIII-D

International Nuclear Information System (INIS)

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

1993-05-01

The removal of helium (He) ash from the plasma core with high efficiency to prevent dilution of the D-T fuel mixture is of utmost importance for future fusion devices, such as the International Thermonuclear Experimental Reactor (ITER). A variety of measurements in L-mode conditions have shown that the intrinsic level of helium transport from the core to the edge may be sufficient to prevent sufficient dilution (i.e., τ He /τ E < 5). Preliminary measurements in biased-induced, limited H-mode discharges in TEXTOR suggest that the intrinsic helium transport properties may not be as favorable. If this trend is shown also in diverted H-mode plasmas, then scenarios based on ELMing H-modes would be less desirable. To further establish the database on helium transport in H-mode conditions, recent studies on the DIII-D tokamak have focused on determining helium transport properties in H-mode conditions and the dependence of these properties on plasma current and ELM frequency

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

Science.gov (United States)

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

2017-10-01

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

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)

Demonstration in the DIII-D tokamak of an alternate baseline scenario for ITER and other burning plasma experiments

International Nuclear Information System (INIS)

Luce, T.C.; Ferron, J.R.; Wade, M.R.

2003-01-01

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

Ion Bernstein wave antenna design for DIII-D

International Nuclear Information System (INIS)

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

1989-01-01

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

Performance of V-4Cr-4Ti material exposed to DIII-D tokamak environment

Energy Technology Data Exchange (ETDEWEB)

Tsai, H.; Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

1997-04-01

Test specimens made with the 832665 heat of V-4Cr-4Ti alloy were exposed in the DIII-D tokamak environment to support the installation of components made of a V-4Cr-4Ti alloy in the radiative divertor of the DIII-D. Some of the tests were conducted with the Divertor Materials Evaluation System (DiMES) to study the short-term effects of postvent bakeout, when concentrations of gaseous impurities in the DIII-D chamber are the highest. Other specimens were mounted next to the chamber wall behind the divertor baffle plate, to study the effects of longer-term exposures. By design, none of the specimens directly interacted with the plasma. Preliminary results from testing the exposed specimens indicate only minor degradation of mechanical properties. Additional testing and microstructural characterization are in progress.

Simulations of drift resistive ballooning L-mode turbulence in the edge plasma of the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Cohen, B. I.; Umansky, M. V.; Nevins, W. M.; Makowski, M. A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Boedo, J. A.; Rudakov, D. L. [University of California, San Diego, San Diego, California 92093 (United States); McKee, G. R.; Yan, Z. [University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Groebner, R. J. [General Atomics, P.O. Box 85608, San Diego, California 92186 (United States)

2013-05-15

Results from simulations of electromagnetic drift-resistive ballooning turbulence for tokamak edge turbulence in realistic single-null geometry are reported. The calculations are undertaken with the BOUT three-dimensional fluid code that solves Braginskii-based fluid equations [X. Q. Xu and R. H. Cohen, Contrib. Plasma Phys. 36, 158 (1998)]. The simulation setup models L-mode edge plasma parameters in the actual magnetic geometry of the DIII-D tokamak [J. L. Luxon et al., Fusion Sci. Technol. 48, 807 (2002)]. The computations track the development of drift-resistive ballooning turbulence in the edge region to saturation. Fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes are compared to experimental data near the outer midplane from Langmuir probe and beam-emission-spectroscopy for a few well-characterized L-mode discharges in DIII-D. The simulations are comprised of a suite of runs in which the physics model is varied to include more fluid fields and physics terms. The simulations yield results for fluctuation amplitudes, correlation lengths, particle and energy fluxes, and diffusivities that agree with measurements within an order of magnitude and within factors of 2 or better for some of the data. The agreement of the simulations with the experimental measurements varies with respect to including more physics in the model equations within the suite of models investigated. The simulations show stabilizing effects of sheared E × B poloidal rotation (imposed zonal flow) and of lower edge electron temperature and density.

Structural design of the DIII-D radiative divertor

International Nuclear Information System (INIS)

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

1996-10-01

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

Divertor particle exhaust and wall inventory on DIII-D

International Nuclear Information System (INIS)

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

1995-01-01

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

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Metastable beta limit in DIII-D

International Nuclear Information System (INIS)

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

1997-06-01

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

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

Active and passive spectroscopic imaging in the DIII-D tokamak

International Nuclear Information System (INIS)

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

2010-01-01

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

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)

Real-time mirror steering for improved closed loop neoclassical tearing mode suppression by electron cyclotron current drive in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Kolemen, E., E-mail: ekolemen@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 45, Princeton, NJ 08543-0451 (United States); Ellis, R. [Princeton Plasma Physics Laboratory, PO Box 45, Princeton, NJ 08543-0451 (United States); La Haye, R.J.; Humphreys, D.A.; Lohr, J.; Noraky, S.; Penaflor, B.G.; Welander, A.S. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States)

2013-11-15

Highlights: • We developed neoclassical tearing mode (NTM) control system for DIII-D, which uses six sets of real-time steerable mirrors in order to move the electron cyclotron current drive (ECCD) deposition location in plasma. • This algorithm accurately finds the NTM island location employing motional Stark effect EFIT MHD equilibrium reconstruction. • Successful NTM suppression and preemption has been achieved in DIII-D using this control system to automatically switches on and off gyrotrons when NTM is detected and rapidly align the NTM island and the ECCD deposition location. -- Abstract: The development and operation of the neoclassical tearing mode (NTM) avoidance and control system for DIII-D, which uses six sets of real-time steerable mirrors in order to move the electron cyclotron current drive (ECCD) deposition location in plasma, is described. The real-time DIII-D NTM control algorithm residing in the Plasma Control System (PCS) automatically detects an NTM by analysis of the Mirnov diagnostics, employs motional Stark effect (MSE) EFIT MHD equilibrium reconstruction to locate the rational q-surface where the NTM island can be found, then calculates the appropriate mirror position for alignment of the ECCD with the island using ray tracing. The control commands from PCS are sent to the electron cyclotron system to switch on and off or modulate the gyrotrons and to the steerable mirror system to move the steerable mirrors to the requested positions. Successful NTM suppression has been achieved in DIII-D using this control system to rapidly align the NTM island and the ECCD deposition location, and to actively maintain the alignment as plasma conditions change.

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}.

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

International Nuclear Information System (INIS)

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

1997-11-01

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

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

Recent results from DIII-D and future plans

International Nuclear Information System (INIS)

Simonen, T.

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1997-06-01

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

Interpretive modeling of simple-as-possible-plasma discharges on DIII-D using the OEDGE code

International Nuclear Information System (INIS)

Stangeby, P.C.; Elder, J.D.; Boedo, J.A.; Bray, B.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Isler, R.C.; Lao, L.L.; Lisgo, S.; Porter, G.D.; Reiter, D.; Rudakov, D.L.; Watkins, J.G.; West, W.P.; Whyte, D.G.

2003-01-01

Recently a number of major, unanticipated effects have been reported in tokamak edge research raising the question of whether we understand the controlling physics of the edge. This report is on the first part - here focused on the outer divertor - of a systematic study of the simplest possible edge plasma - no ELMs, no detachment, etc. - for a set of 10 repeat, highly diagnosed, single-null, divertor discharges in DIII-D. For almost the entire, extensive data set so far evaluated, the matches of experiment and model are so close as to imply that the controlling processes at the outer divertor for these simple plasma conditions have probably been correctly identified and quantitatively characterized in the model. The principal anomaly flagged so far relates to measurements of T e near the target, potentially pointing to a deficiency in our understanding of sheath physics in the tokamak environment

2-D tomography with bolometry in DIII-D

International Nuclear Information System (INIS)

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

1994-07-01

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

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.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

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

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

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

Directory of Open Access Journals (Sweden)

Wukitch S.J.

2017-01-01

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

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Alfven Eigenmode Control in DIII-D

Science.gov (United States)

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

2017-10-01

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

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

Active and passive spectroscopic imaging in the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2010-04-15

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

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

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

Science.gov (United States)

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

2004-06-11

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

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Scaling of divertor heat flux profile widths in DIII-D

International Nuclear Information System (INIS)

Lasnier, C.J.; Makowski, M.A.; Boedo, J.A.; Allen, S.L.; Brooks, N.H.; Hill, D.N.; Leonard, A.W.; Watkins, J.G.; West, W.P.

2011-01-01

New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

DIII-D Thomson Scattering Diagnostic Data Acquisition, Processing and Analysis Software

International Nuclear Information System (INIS)

Middaugh, K.R.; Bray, B.D.; Hsieh, C.L.; McHarg, B.B.Jr.; Penaflor, B.G.

1999-01-01

One of the diagnostic systems critical to the success of the DIII-D tokamak experiment is the Thomson scattering diagnostic. This diagnostic is unique in that it measures local electron temperature and density: (1) at multiple locations within the tokamak plasma; and (2) at different times throughout the plasma duration. Thomson ''raw'' data are digitized signals of scattered light, measured at different times and locations, from the laser beam paths fired into the plasma. Real-time acquisition of this data is performed by specialized hardware. Once obtained, the raw data are processed into meaningful temperature and density values which can be analyzed for measurement quality. This paper will provide an overview of the entire Thomson scattering diagnostic software and will focus on the data acquisition, processing, and analysis software implementation. The software falls into three general categories: (1) Set-up and Control: Initializes and controls all Thomson hardware and software, synchronizes with other DIII-D computers, and invokes other Thomson software as appropriate. (2) Data Acquisition and Processing: Obtains raw measured data from memory and processes it into temperature and density values. (3) Analysis: Provides a graphical user interface in which to perform analysis and sophisticated plotting of analysis parameters

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)

Study of particle pumping characteristics for different pumping geometries in JT-60U and DIII-D divertors

International Nuclear Information System (INIS)

Takenaga, H.; Sakasai, A.; Kubo, H.

2001-01-01

Particle pumping characteristics were compared between pumping from the inner side private flux region (IPP) and pumping from both sides of the private flux region (BPP) in the JT-60U W shaped divertor, and between JT-60U IPP and pumping in the DIII-D lower baffled divertor. The pumping flux for BPP is smaller than that for IPP by about a factor of 2 with weak in-out asymmetry of recycling neutral flux and by a factor of 3.5-6.5 with strong in-out asymmetry. The reduction of the pumping flux for BPP is consistent with Monte Carlo simulations, where backflow at the outer pumping slot is observed due to in-out recycling asymmetry. The pumping flux in DIII-D at I p =0.8 MA and B T =1.6 T is comparable to or smaller than that for JT-60U IPP at I p =1.0 MA, B T =3.8 T and I p =1.5 MA, B T =3.5 T in the same density regime. In the DIII-D divertor with pumping from the private flux region, the pumping flux decreases with increasing in-out asymmetry. The pumping flux normalized by the integrated D α emission over the whole plasma exhibits a similar dependence on the distance between the pumping slot and the strike point in JT-60U IPP and the DIII-D lower divertor with pumping through the outer divertor plasma region. (author)

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 D_α, D_β and D_γ for values of T_e and n_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. Lastly, for the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to T_e.

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.

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

International Nuclear Information System (INIS)

Fenstermacher, M.E.; Porkolab, M.

1993-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

Theoretical work, computation, and results from TCV [J. Decker “Effect of density fluctuations on ECCD in ITER and TCV,” EPJ Web of Conf. 32, 01016 (2012)] suggest that density fluctuations in the edge region of a tokamak plasma can cause broadening of the ECH deposition profile. In this paper, a GUI tool is presented which is used for analysis of ECH deposition as a first step towards looking for this broadening, which could explain effects seen in previous DIII-D ECH transport studies [K.W. Gentle “Electron energy transport inferences from modulated electron cyclotron heating in DIII-D,” Phys. Plasmas 13, 012311 (2006)]. By applying an FFT to the T e measurements from the University of Texas’s 40-channel ECE Radiometer, and using a simplified thermal transport equation, the flux surface extent of ECH deposition is determined. The Fourier method analysis is compared with a Break-In-Slope (BIS) analysis and predictions from the ray-tracing code TORAY. Examination of multiple Fourier harmonics and BIS fitting methods allow an estimation of modulated transport coefficients and thereby the true ECH deposition profile. Correlations between edge fluctuations and ECH deposition in legacy data are also explored as a step towards establishing a link between fluctuations and deposition broadening in DIII-D

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

Energy Technology Data Exchange (ETDEWEB)

Brookman, M. W., E-mail: brookmanmw@fusion.gat.com; Austin, M. E. [Institute for Fusion Studies, University of Texas at Austin, MS 13-505, 3483 Dunhill St, San Diego, CA 92121-1200 (United States); Petty, C. C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States)

2015-12-10

Theoretical work, computation, and results from TCV [J. Decker “Effect of density fluctuations on ECCD in ITER and TCV,” EPJ Web of Conf. 32, 01016 (2012)] suggest that density fluctuations in the edge region of a tokamak plasma can cause broadening of the ECH deposition profile. In this paper, a GUI tool is presented which is used for analysis of ECH deposition as a first step towards looking for this broadening, which could explain effects seen in previous DIII-D ECH transport studies [K.W. Gentle “Electron energy transport inferences from modulated electron cyclotron heating in DIII-D,” Phys. Plasmas 13, 012311 (2006)]. By applying an FFT to the T{sub e} measurements from the University of Texas’s 40-channel ECE Radiometer, and using a simplified thermal transport equation, the flux surface extent of ECH deposition is determined. The Fourier method analysis is compared with a Break-In-Slope (BIS) analysis and predictions from the ray-tracing code TORAY. Examination of multiple Fourier harmonics and BIS fitting methods allow an estimation of modulated transport coefficients and thereby the true ECH deposition profile. Correlations between edge fluctuations and ECH deposition in legacy data are also explored as a step towards establishing a link between fluctuations and deposition broadening in DIII-D.

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

Study of H-mode threshold conditions in DIII-D

International Nuclear Information System (INIS)

Groebner, R.J.; Carlstrom, T.N.; Burrell, K.H.

1996-10-01

Studies have been conducted in DIII-D to determine the dependence of the power threshold P lh for the transition to the H-mode regime and the threshold P hl for the transition from H-mode to L-mode as functions of external parameters. There is a value of the line-averaged density n e at which P lh has a minimum and P lh tends to increase for lower and higher values of n e . Experiments conducted to separate the effect of the neutral density n 0 from the plasma density n e give evidence of a strong coupling between n 0 and n e . The separate effect of neutrals on the transition has not been determined. Coordinated experiments with JET made in the ITER shape show that P lh increases approximately as S 0.5 where S is the plasma surface area. For these discharges, the power threshold in DIII-D was high by normal standards, thus suggesting that effects other than plasma size may have affected the experiment. Studies of H-L transitions have been initiated and hysteresis of order 40% has been observed. Studies have also been done of the dependence of the L-H transition on local edge parameters. Characterization of the edge within a few ms prior to the transition shows that the range of edge temperatures at which the transition has been observed is more restrictive than the range of densities at which it occurs. These results suggest that some temperature function is important for controlling the transition

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Motional stark effect upgrades on DIII-D

International Nuclear Information System (INIS)

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

1994-04-01

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

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

Science.gov (United States)

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

2015-11-01

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

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

2015-09-15

The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Effects of particle exhaust on neutral compression ratios in DIII-D

International Nuclear Information System (INIS)

Colchin, R.J.; Maingi, R.; Wade, M.R.; Allen, S.L.; Greenfield, C.M.

1998-08-01

In this paper, neutral particles in DIII-D are studied via their compression in the plenum and via particle exhaust. The compression of gas in the plena is examined in terms of the magnetic field configuration and wall conditions. DIII-D compression ratios are observed in the range from 1 to ≥ 1,000. Particle control ultimately depends on the exhaust of neutrals via plenum or wall pumping. Wall pumping or outgassing is calculated by means of a detailed particle balance throughout individual discharges, and its effect on particle control is discussed. It is demonstrated that particle control through wall conditioning leads to lower normalized densities. A two-region model shows that the gas compression ratio (C div = divertor plenum neutral pressure/torus neutral pressure) can be interpreted in relation to gas flows in the torus and divertor including the pumping speed of the plenum cryopumps, plasma pumping, and the pumping or outgassing of the walls

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

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

Science.gov (United States)

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

2002-12-23

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

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

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

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)

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

Free-Boundary 3D Equilibria and Resistive Wall Instabilities with Extended-MHD

Science.gov (United States)

Ferraro, N. M.

2015-11-01

The interaction of the plasma with external currents, either imposed or induced, is a critical element of a wide range of important tokamak phenomena, including resistive wall mode (RWM) stability and feedback control, island penetration and locking, and disruptions. A model of these currents may be included within the domain of extended-MHD codes in a way that preserves the self-consistency, scalability, and implicitness of their numerical methods. Such a model of the resistive wall and non-axisymmetric coils is demonstrated using the M3D-C1 code for a variety of applications, including RWMs, perturbed non-axisymmetric equilibria, and a vertical displacement event (VDE) disruption. The calculated free-boundary equilibria, which include Spitzer resistivity, rotation, and two-fluid effects, are compared to external magnetic and internal thermal measurements for several DIII-D discharges. In calculations of the perturbed equilibria in ELM suppressed discharges, the tearing response at the top of the pedestal is found to correlate with the onset of ELM suppression. Nonlinear VDE calculations, initialized using a vertically unstable DIII-D equilibrium, resolve in both space and time the currents induced in the wall and on the plasma surface, and also the currents flowing between the plasma and the wall. The relative magnitude of these contributions and the total impulse to the wall depend on the resistive wall time, although the maximum axisymmetric force on the wall over the course of the VDE is found to be essentially independent of the wall conductivity. This research was supported by US DOE contracts DE-FG02-95ER54309, DE-FC02-04ER54698 and DE-AC52-07NA27344.

Impurity penetration and transport during VH-mode on DIII-D

International Nuclear Information System (INIS)

Lippmann, S.I.; Evans, T.E.; Jackson, G.L.; West, W.P.

1992-05-01

A new modeling effort is made in order to understand the observed relatively low levels of impurity contamination during the VH-mode phase on DIII-D, as compared to those observed during the H-mode phase of selected discharges. The key element is the inclusion of the real 2-D flux surface geometry in the prediction of impurity penetration of sputtered atoms through the scrape-off layer into the core plasma. Of the elements which determine the impurity content in the plasma: sputtering yield, penetration, and core transport, the penetration through the scrape-off layer is found to be the most determinative factor. The low impurity content in VH-mode is attributed to the development of a scrape-off layer with higher density and temperature properties than those normally obtained in H-mode

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.

Increased power delivery from the DIII-D neutral beam injection system

International Nuclear Information System (INIS)

Colleraine, A.P.; Callis, R.W.; Hong, R.M.; Kellman, D.H.; Kim, J.; Langhorn, A.R.; Lee, R.; Phillips, J.C.; Wight, J.J.

1989-12-01

The neutral beam system installed on the DIII-D tokamak employs eight 80 kV Long Pulse Sources (LPS) mounted on four beamlines and was originally designed to deliver a nominal 12 MW of H degree power to a plasma for pulses of up to 5 sec duration. Lawrence Berkeley Laboratory designed the LPS for the US Fusion Program to fill the requirements of both the DIII-D and the TFTR machines. Essentially all source components are of a common design; the DIII-D version is therefore conservative in its rated parameters. Recently a neutron shield has been constructed around the torus hall allowing D degree injection to become routine. Because deuterium beams have a better neutralization efficiency, the nominal power delivery per source has been measured to be approximately 2 MW (for a total of 16 MW) without any modifications. However, by reoptimizing the voltage gradients in the source, the perveance can be increased without degrading the optics. A change of gradient grid voltage from 0.83 V accel to 0.79 V accel raises the perveance from 2.5 to 3.0 μPerv with a corresponding gain in beam power of about 20%. The arc power required also must be increased to the range of 100 to 120 kW but this is well within the design limits of the LPS. Further studies of our systems are now underway to assess the possibilities of raising V accel above 80 kV. An additional gain in power is possible by this technique. 6 refs., 6 figs

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

Improved edge charge exchange recombination spectroscopy in DIII-D.

Science.gov (United States)

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

2016-11-01

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

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

International Nuclear Information System (INIS)

Evans, T.E.; Taylor, P.L.; Whyte, D.G.

1998-12-01

Prompt runaway electron bursts, generated by rapidly cooling DIII-D plasmas with argon killer pellets, are used to test a recent knock-on avalanche theory describing the growth of multi-MeV runaway electron currents during disruptions in tokamaks. Runaway current amplitudes, observed during some but not all DIII-D current quenches, are consistent with growth rates predicted by the theory assuming a pre-current quench runaway electron density of approximately 10 15 m -3 . Argon killer pellet modeling yields runaway densities of between 10 15 --10 16 m -3 in these discharges. Although knock-on avalanching appears to agree rather well with the measurements, relatively small avalanche amplification factors combined with uncertainties in the spatial distribution of pellet mass and cooling rates make it difficult to unambiguously confirm the proposed theory with existing data

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

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

Beta-induced Alfven-acoustic Eigenmodes in NSTX and DIII-D Driven by Beam Ions

International Nuclear Information System (INIS)

Gorelenkov, N.N.; Van Zeeland, M.A.; Berk, H.L.; Crocker, N.A.; Darrow, D.; Fredrickson, E.; Fu, G.-Y.; Heidbrink, W.W.; Menard, J.; Nazikian, R.

2009-01-01

Kinetic theory and experimental observations of a special class of energetic particle driven instabilities called here Beta-induced Alfven-Acoustic Eigenmodes (BAAE) are reported confirming previous results [N.N. Gorelenkov H.L. Berk, N.A. Crocker et. al. Plasma Phys. Control. Fusion 49 B371 (2007)] The kinetic theory is based on the ballooning dispersion relation where the drift frequency effects are retained. BAAE gaps are recovered in kinetic theory. It is shown that the observed certain low-frequency instabilities on DIII-D [J.L. Luxon, Nucl. Fusion 42 614 (2002)] and National Spherical Torus Experiment [M. Ono, S.M. Kaye, Y.-K M. Peng et. al., Nucl. Fusion 40 3Y 557 (2000)] are consistent with their identification as BAAEs. BAAEs deteriorated the fast ion confinement in DIII-D and can have a similar effect in next-step fusion plasmas, especially if excited together with multiple global Toroidicity-induced shear Alfven Eigenmode (TAE) instabilities. BAAEs can also be used to diagnose safety factor profiles, a technique known as magnetohydrodynamic spectroscopy

Beta-induced Alfven-acousti Eigenmodes in NSTX and DIII-D Driven by Beam Ions

Energy Technology Data Exchange (ETDEWEB)

Gorelenkov, N. N.; Van Zeeland, M. A.; Berk, H. L.; Crocker, N. A.; Darrow, D.; Fredrickson, E.; Fu, G. Y.; Heidbrink, W. W.; Menard, J.; Nazikian, R.

2009-03-06

Kinetic theory and experimental observations of a special class of energetic particle driven instabilities called here Beta-induced Alfven-Acoustic Eigenmodes (BAAE) are reported confirming previous results [N.N. Gorelenkov H.L. Berk, N.A. Crocker et. al. Plasma Phys. Control. Fusion 49 B371 (2007)] The kinetic theory is based on the ballooning dispersion relation where the drift frequency effects are retained. BAAE gaps are recovered in kinetic theory. It is shown that the observed certain low-frequency instabilities on DIII-D [J.L. Luxon, Nucl. Fusion 42 614 (2002)] and National Spherical Torus Experiment [M. Ono, S.M. Kaye, Y.-K M. Peng et. al., Nucl. Fusion 40 3Y 557 (2000)] are consistent with their identification as BAAEs. BAAEs deteriorated the fast ion confinement in DIII-D and can have a similar effect in next-step fusion plasmas, especially if excited together with multiple global Toroidicity-induced shear Alfven Eigenmode (TAE) instabilities. BAAEs can also be used to diagnose safety factor profiles, a technique known as magnetohydrodynamic spectroscopy.

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

DIII-D physics analysis database

International Nuclear Information System (INIS)

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

1990-10-01

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

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

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�

The ground-fault detection system for DIII-D

International Nuclear Information System (INIS)

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

1987-10-01

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

STATIONARY HIGH-PERFORMANCE DISCHARGES IN THE DIII-D TOKAMAK

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Science.gov (United States)

Zhao, Michael; Punjabi, Alkesh; Ali, Halima

2009-11-01

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

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

International Nuclear Information System (INIS)

Strait, E.J.; Chu, M.S.; Ferron, J.R.

1996-12-01

Discharges with negative central magnetic shear (NCS) hold the promise of enhanced fusion performance in advanced tokamaks. However, stability to long wavelength magnetohydrodynamic modes is needed to take advantage of the improved confinement found in NCS discharges. The stability limits seen in DIII-D experiments depend on the pressure and current density profiles and are in good agreement with stability calculations. Discharges with a strongly peaked pressure profile reach a disruptive limit at low beta, β N = β (I/aB) -1 ≤ 2.5 (% m T/MA), caused by an n = 1 ideal internal kink mode or a global resistive instability close to the ideal stability limit. Discharges with a broad pressure profile reach a soft beta limit at significantly higher beta, β N = 4 to 5, usually caused by instabilities with n > 1 and usually driven near the edge of the plasma. With broad pressure profiles, the experimental stability limit is independent of the magnitude of negative shear but improves with the internal inductance, corresponding to lower current density near the edge of the plasma. Understanding of the stability limits in NCS discharges has led to record DIII-D fusion performance in discharges with a broad pressure profile and low edge current density

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.

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Design of the divertor Thomson scattering system on DIII-D

International Nuclear Information System (INIS)

Carlstrom, T.N.; Foote, J.H.; Nilson, D.G.; Rice, B.W.

1994-05-01

Local measurements of n e and T e in the divertor region are necessary for a more complete understanding of divertor physics. We have designed an extension to the existing multipulse Thomson scattering system to measure n e in the range 5 x 10 18 to 5 x 10 20 m -3 and T e 5--500 eV, with 1 cm resolution from 1--21 cm above the floor of the DIII-D vessel, in the region of the X-point for lower single-null diverted plasmas. One of the existing 8, 20 Hz, ND:YAG lasers will be redirected to a separate vertical port, and viewed radially with a specially designed, f/6.8 lens. Fiber optics carry the light to additional polychromators whose interference filters have been optimized for low T e measurements. Other aspect of the system, including the beam path to the vessel, polychromator design, real time data acquisition, laser control, calibration facility, and DIII-D timing and data acquisition interface will be shared with the existing multipulse Thomson system. An in-situ laser alignment monitor will provide alignment information for each laser pulse

An overview of the DIII-D program

International Nuclear Information System (INIS)

Luxon, J.L.

1996-10-01

The DIII-D program focuses on developing fusion physics in an integrated program of tokamak concept improvement. The intent is both to support the present ITER physics R and D and to develop more efficient concepts for the later phases of ITER and eventual power plants. Progress in this effort can be best summarized by recent results for a diverted deuterium discharge with negative central shear which reached a performance level of Q DT = 0.32. The ongoing development of the tools needed to carry out this program of understanding and optimization continues to be crucial to its success. Control of the plasma cross-sectional shape and the internal distributions of plasma current, density, and rotation has been essential to optimizing plasma performance. Advanced divertor concepts provide edge power and particle control for future devices such as ITER and provide techniques to help manage the edge power and particle flows for advanced tokamak concepts. New divertor diagnostics and improved modeling are developing excellent divertor understanding. Many of the plasma physics issues being posed by ITER are being addressed. Scrapeoff layer power flow is being characterized to provide an accurate basis for the design of reactor devices. Ongoing studies of the density limit focus on identifying ways in which ITER can achieve the required densities in excess of the Greenwald limit. Better understanding of disruptions is crucial to the design of future reactors

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

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.

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

The effect of plasma collisionality on pedestal current density formation in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Thomas, D M; Leonard, A W; Osborne, T H; Groebner, R J; West, W P; Burrell, K H [General Atomics, PO Box 85608, San Diego, California 92186-5608 (United States)

2006-05-15

The evolution and performance limits for the pedestal in H-mode are dependent on the two main drive terms for instability: namely the edge pressure gradient and the edge current density. These terms are naturally coupled though neoclassical (Pfirsch-Schluter and bootstrap) effects. On DIII-D, local measurements of the edge current density are made using an injected lithium beam in conjunction with Zeeman polarimetry and compared with pressure profile measurements made with other diagnostics. These measurements have confirmed the close spatial and temporal correlation that exists between the measured current density and the edge pressure in H- and QH-mode pedestals, where substantial pressure gradients exist. In the present work we examine the changes in the measured edge current for DIII-D pedestals which have a range of values for the ion and electron collisionalities {l_brace}{upsilon}{sub i}*,{upsilon}{sub e}*{r_brace} due to fuelling effects. Such changes in the collisionality in the edge are expected to significantly alter the level of the bootstrap current from the value predicted from the collisionless limit and therefore should correspondingly alter the pedestal stability limits. We find a clear decrease in measured current as {nu} increases, even for discharges having similar edge pressure gradients.

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.

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

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

Edge density fluctuation diagnostic for DIII-D using lithium beams: 1992 annual report

International Nuclear Information System (INIS)

Thomas, D.M.

1994-01-01

During the past several months the Lithium beam diagnostic was commissioned of DIII-D and began yielding useful information. The author developed the remote control and monitoring of the ion source operation and beam formation and focussing, and integrated the control system and data acquisition into the DIII-D operating system. Several detector types were fabricated, and fluorescence data were collected using several differing detector arrangements. Beam-gas measurements were conducted to analyze the intrinsic beam fluctuations and stability. Fluorescence data was then obtained on a number of Tokamak discharges under varying discharge conditions. Analysis of this initial data is proceeding but has already yielded some interesting features. These include changes in the edge plasma density behavior during the l- to h-transition, disruptions, and edge localized modes (ELMs). Based on the quality of data obtained the author proceeded with the design and construction of the full 16-channel detection system which will be completed and tested shortly

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

Closed-loop feedback of MHD instabilities on DIII-D

Energy Technology Data Exchange (ETDEWEB)

Fredrickson, E.D.; Johnson, L.C.; Manickam, J.; Okabayashi, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Bialek, J.; Garofalo, A.M.; Navratil, G.A. [Columbia University, New York, NY 10027 (United States); La Haye, R.J.; Scoville, J.T.; Strait, E.J. [General Atomics, La Jolla, CA 92186 (United States); Lazarus, E.A. [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

2001-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

An overview of the DIII-D long pulse neutral beam system

International Nuclear Information System (INIS)

Callis, R.W.; Colleraine, A.P.; Hong, R.M.; Langhorn, A.R.; Lee, R.L.; Kim, J.; Phillips, J.C.; Wight, J.J.

1988-09-01

The four beamlines on the DIII-D tokamak have been upgraded to long pulse operation with the addition of eight 80 kV, 80 A, 5 sec long pulse sources. The eight sources have proven to be very reliable and have performed well. Up to 12 MW of H 0 has been injected into a plasma. Inertially cooled beam absorbers have proven capable of handling multi-second pulses. General performance characteristics and some recent long-pulse physics results are presented. 12 refs., 7 figs

An overview of the DIII-D long pulse neutral beam system

International Nuclear Information System (INIS)

Callis, R.W.; Colleraine, A.P.; Hong, R.-M.; Langhorn, A.R.; Lee, R.L.; Kim, J.; Phillips, J.C.; Wight, J.J.

1989-01-01

The four beamlines on the DIII-D tokamak have been upgraded to long pulse operation with the addition of eight 80 kV, 80 A, 5 sec long pulse sources. The eight sources have proven to be very reliable and have performed well. Up to 12 MW of H 0 has been injected into a plasma. Inertially cooled beam absorbers have proven capable of handling multi-second pulses. General performance characteristics and some recent long-pulse physics results are presented. (author). 12 refs.; 7 figs.; 1 tab

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

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.

Design and analysis of the DIII-D radiative divertor water-cooled structures

International Nuclear Information System (INIS)

Hollerbach, M.A.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Phelps, R.D.; Redler, K.M.; Reis, E.E.

1995-01-01

The Radiative Divertor is a major modification to the divertor of DIII-D and is being designed and fabricated for installation in late 1996. The Radiative Divertor Program (RDP) will enhance the dissipative processes in the edge and divertor plasmas to reduce the heat flux and plasma erosion at the divertor target. This approach will have major implications for the heat removal methods used in future devices. The divertor is of slot-type configuration designed to minimize the flow of sputtered and injected impurities back to the core plasma. The new divertor will be composed of toroidally continuous, Inconel 625 water-cooled rings of sandwich construction with an internal water channel, incorporating seam welding to provide the water-to-vacuum seal as well as structural integrity. The divertor structure is designed to withstand electro-magnetic loads as a result of halo currents and induced toroidal currents. It also accommodates the thermal differences experienced during the 400 C bake used on DIII-D. A low Z plasma-facing surface is provided by mechanically attached graphite tiles. Water flow through the rings will inertially cool these tiles which will be subjected to 38 MW, 10 second pulses. Current schedules call for detailed design in 1996 with installation completed in March 1997. A full size prototype, one-quarter of one ring, is being built to validate manufacturing techniques, machining, roll-forming, and seam welding. The experience and knowledge gained through the fabrication of the prototype is discussed. The design of the electrically isolated (5 kV) vacuum-to-air water feedthroughs supplying the water-cooled rings is also discussed

Comprehensive energy transport scalings derived from DIII-D similarity experiments

International Nuclear Information System (INIS)

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

1998-12-01

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

Comprehensive energy transport scalings derived from DIII-D similarity experiments

International Nuclear Information System (INIS)

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

1999-01-01

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

Comprehensive energy transport scalings derived from DIII-D similarity experiments

International Nuclear Information System (INIS)

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

2001-01-01

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

Software upgrade for the DIII-D neutral beam control systems

International Nuclear Information System (INIS)

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

1991-11-01

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

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

Science.gov (United States)

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

2012-10-01

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

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

Magnetic Transport Barriers in the DIII-D Tokamak

Science.gov (United States)

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

2009-11-01

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

Gamma ray imager on the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.; Van Zeeland, M. A.; Watkins, M. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Cooper, C. M. [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California-San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Riso, V. [State University of New York-Buffalo, 12 Capen Hall, Buffalo, New York 14260-1660 (United States)

2016-04-15

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

Measurement of deuterium density profiles in the H-mode steep gradient region using charge exchange recombination spectroscopy on DIII-D.

Science.gov (United States)

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

2016-11-01

Recent completion of a thirty two channel main-ion (deuterium) charge exchange recombination spectroscopy (CER) diagnostic on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] enables detailed comparisons between impurity and main-ion temperature, density, and toroidal rotation. In a H-mode DIII-D discharge, these new measurement capabilities are used to provide the deuterium density profile, demonstrate the importance of profile alignment between Thomson scattering and CER diagnostics, and aid in determining the electron temperature at the separatrix. Sixteen sightlines cover the core of the plasma and another sixteen are densely packed towards the plasma edge, providing high resolution measurements across the pedestal and steep gradient region in H-mode plasmas. Extracting useful physical quantities such as deuterium density is challenging due to multiple photoemission processes. These challenges are overcome using a detailed fitting model and by forward modeling the photoemission using the FIDASIM code, which implements a comprehensive collisional radiative model.

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

Science.gov (United States)

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

2018-03-01

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

Preliminary oscillating fluxes current drive experiment in DIII-D tokamak

International Nuclear Information System (INIS)

Yamaguchi, S.; Schaffer, M.; Kondoh, Y.

1995-01-01

A preliminary oscillating flux helicity injection experiment was done on DIII-D tokamak. The toroidal flux was modulated by programming the plasma elongation. Instead of programming the surface voltage directly, the plasma current was programmed with a periodic modulation at some phase shift. The theoretical basis of this modulation is discussed in terms of the helicity injection and also introduced by cross-field motion of the modulated plasma. Because the primary winding is well coupled with the plasma current and the power supply is strong, the plasma current behaves as programmed. However, as the plasma shape is not coupled strongly with the shaping and equilibrium coils, the elongation amplitude and phase are affected by the change of plasma current and do not behave as programmed. Because of this, the voltage induced by the helicity injection is low, and the experiment did not test the principle of helicity injection. The injection powers of helicity and energy, and the electric field intensity of the helicity injection model and the cross-field motion of plasma are compared with each other experimentally. The improvement necessary to do the experiment is also proposed. ((orig.))

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

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

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

Overview of Recent DIII-D Experimental Results

Science.gov (United States)

Fenstermacher, Max

2015-11-01

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

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

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

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

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

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

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

Science.gov (United States)

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

2017-10-01

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

Argon Kα measurement on DIII endash D by Ross filters technique (abstract)

International Nuclear Information System (INIS)

Snider, R.T.; Bogatu, I.N.; Brooks, N.H.; Wade, M.R.

1999-01-01

Techniques to reduce the heat flux to the divertor plates in tokamak power plants and the consequent erosion of, and subsequent damage to the divertor target plates include the injection of impurities such as argon, that can dissipate the energy (through radiative or collisional processes) before it reaches the target plates. An important issue with this type of scheme is poisoning of the plasma core by the impurities introduced in the divertor region. Subsequently, there is a desire to measure the profiles of the injected impurities in the core. X-ray Ross filters with an effective narrow band pass centered on the argon Kα line at 3.2 keV, have been installed on two of the existing x-ray arrays on DIII endash D in order to help determine the argon concentration profiles. Emissivity profiles of the Kα lines and the emissivity profiles for the argon enhanced continuum can be inferred from the inverted filtered x-ray brightness signals if T e , n e , and Ar 18+ profiles are known. The MIST code is used to couple the filtered x-ray signals to the time dependent measurements of T e and n e . Further, the Ar 16+ profiles measured by charge transfer spectroscopy, are used as a constraint on the MIST code runs to calculate Ar 18+ profiles and unfold the argon emissivity profiles. A discussion of the Ross filters, the DIII endash D argon data, and the data analysis scheme for inferring argon emissivity profiles will be discussed. Estimates of the total argon concentration in the core determined from this technique in DIII endash D argon puff experiments will be presented. copyright 1999 American Institute of Physics

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.

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

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

International Nuclear Information System (INIS)

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

1995-02-01

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

Measurements of the internal magnetic field using the B-Stark motional Stark effect diagnostic on DIII-D (inivited)

Energy Technology Data Exchange (ETDEWEB)

Pablant, N. A. [University of California-San Diego, La Jolla, California 92093 (United States); Burrell, K. H.; Groebner, R. J.; Kaplan, D. H. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Holcomb, C. T. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

2010-10-15

Results are presented from the B-Stark diagnostic installed on the DIII-D tokamak. This diagnostic provides measurements of the magnitude and direction of the internal magnetic field. The B-Stark system is a version of a motional Stark effect (MSE) diagnostic based on the relative line intensities and spacing of the Stark split D{sub {alpha}} emission from injected neutral beams. This technique may have advantages over MSE polarimetry based diagnostics in future devices, such as the ITER. The B-Stark diagnostic technique and calibration procedures are discussed. The system is shown to provide accurate measurements of B{sub {theta}}/B{sub T} and |B| over a range of plasma conditions. Measurements have been made with toroidal fields in the range of 1.2-2.1 T, plasma currents in the range 0.5-2.0 MA, densities between 1.7 and 9.0x10{sup 19} m{sup -3}, and neutral beam voltages between 50 and 81 keV. The viewing direction and polarization dependent transmission properties of the collection optics are found using an in situ beam into gas calibration. These results are compared to values found from plasma equilibrium reconstructions and the MSE polarimetry system on DIII-D.

Measurements of the internal magnetic field using the B-Stark motional Stark effect diagnostic on DIII-D (inivited).

Science.gov (United States)

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

2010-10-01

Results are presented from the B-Stark diagnostic installed on the DIII-D tokamak. This diagnostic provides measurements of the magnitude and direction of the internal magnetic field. The B-Stark system is a version of a motional Stark effect (MSE) diagnostic based on the relative line intensities and spacing of the Stark split D(α) emission from injected neutral beams. This technique may have advantages over MSE polarimetry based diagnostics in future devices, such as the ITER. The B-Stark diagnostic technique and calibration procedures are discussed. The system is shown to provide accurate measurements of B(θ)/B(T) and ∣B∣ over a range of plasma conditions. Measurements have been made with toroidal fields in the range of 1.2-2.1 T, plasma currents in the range 0.5-2.0 MA, densities between 1.7 and 9.0×10(19) m(-3), and neutral beam voltages between 50 and 81 keV. The viewing direction and polarization dependent transmission properties of the collection optics are found using an in situ beam into gas calibration. These results are compared to values found from plasma equilibrium reconstructions and the MSE polarimetry system on DIII-D.

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

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.

Measurements of flows in the DIII-D divertor by Mach probes

International Nuclear Information System (INIS)

Boedo, J.A.; Lehmer, R.; Moyer, R.A.; Watkins, J.G.; Porter, G.D.; Evans, T.E.; Leonard, A.W.; Schaffer, M.J.

1998-06-01

First measurements of Mach number of background plasma in the DIII-D divertor are presented in conjunction with temperature T e and density n e using a fast scanning probe array. To validate the probe measurements, the authors compared the T e , n e and J sat data to Thomson scattering data and find good overall agreement in attached discharges and some discrepancy for T e and n e in detached discharges. The discrepancy is mostly due to the effect of large fluctuations present during detached plasmas on the probe characteristic; the particle flux is accurately measured in every case. A composite 2-D map of measured flows is presented for an ELMing H-mode discharge and they focus on some of the details. They have also documented the temperature, density and Mach number in the private flux region of the divertor and the vicinity of the X-point, which are important transition regions that have been little studied or modeled. Background parallel plasma flows and electric fields in the divertor region show a complex structure

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

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

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

ICRF [ion cyclotron range of frequencies] coupling on DIII-D and the implications on ICRF technology development

International Nuclear Information System (INIS)

Hoffman, D.J.; Baity, F.W.; Mayberry, M.J.; Swain, D.W.

1987-01-01

Low-power coupling tests have been carried out with a prototype ion cyclotron range of frequencies (ICRF) compact loop antenna on the DIII-D tokamak. Plasma load resistance values higher than originally calculated are measured in ohmic and L-mode, beam-heated plasmas. Load resistance decreases by a factor of ∼2 in H-mode operation. When edge localized modes (ELMs) occur, the antenna loading increases transiently to several ohms. Results indicate that fast-wave ICRF antenna coupling characteristics are highly sensitive to changes in the edge plasma profiles associated with the H-mode regime

Comparison between the electron cyclotron current drive experiments on DIII-D and predictions for T-10

International Nuclear Information System (INIS)

Lohr, J.; Harvey, R.W.; Luce, T.C.; Matsuda, Kyoko; Moeller, C.P.; Petty, C.C.; Prater, R.; James, R.A.; Giruzzi, G.; Gorelov, Y.; DeHaas, J.

1990-11-01

Electron cyclotron current drive has been demonstrated on the DIII-D tokamak in an experiment in which ∼1 MW of microwave power generated ∼50 kA of non-inductive current. The rf-generated portion was about 15% of the total current. On the T-10 tokamak, more than 3 MW of microwave power will be available for current generation, providing the possibility that all the plasma current could be maintained by this method. Fokker-Planck calculations using the code CQL3D and ray tracing calculations using TORAY have been performed to model both experiments. For DIII-D the agreement between the calculations and measurements is good, producing confidence in the validity of the computational models. The same calculations using the T-10 geometry predict that for n e (0) ∼ 1.8 x 10 13 cm -3 , and T e (0) ∼ 7 keV, 1.2 MW, that is, the power available from only three gyrotrons, could generate as much as 150 kA of non-inductive current. Parameter space scans in which temperature, density and resonance location were varied have been performed to indicate the current drive expected under different experimental conditions. The residual dc electric field was considered in the DIII-D analysis because of its nonlinear effect on the electron distribution, which complicates the interpretation of the results. A 110 GHz ECH system is being installed on DIII-D. Initial operations, planned for late 1991, will use four gyrotrons with 500 kW each and 10 second output pulses. Injection will be from the low field side from launchers which can be steered to heat at the desired location. These launchers, two of which are presently installed, are set at 20 degrees to the radial and rf current drive studies are planned for the initial operation. 8 refs., 10 figs

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.

Edge Pedestal Control in Quiescent H-Mode Discharges in DIII-D Using Co Plus Counter Neutral Beam Injection

International Nuclear Information System (INIS)

Burrell, K.H.; Osborne, T.H.; Snyder, P.B.; West, W.P.; Chu, M.S.; Fenstermacher, M.E.; Gohil, P.; Solomon, W.M.

2008-01-01

We have made two significant discoveries in our recent studies of quiescent H-mode (QH-mode) plasmas in DIII-D. First, we have found that we can control the edge pedestal density and pressure by altering the edge particle transport through changes in the edge toroidal rotation. This allows us to adjust the edge operating point to be close to, but below the ELM stability boundary, maintaining the ELM-free state while allowing up to a factor of two increase in edge pressure. The ELM boundary is significantly higher in more strongly shaped plasmas, which broadens the operating space available for QH-mode and leads to improved core performance. Second, for the first time on any tokamak, we have created QH-mode plasmas with strong edge co-rotation; previous QH-modes in all tokamaks had edge counter rotation. This result demonstrates that counter NBI and edge counter rotation are not essential conditions for QH-mode. Both these investigations benefited from the edge stability predictions based on peeling-ballooning mode theory. The broadening of the ELM-stable region with plasma shaping is predicted by that theory. The theory has also been extended to provide a model for the edge harmonic oscillation (EHO) that regulates edge transport in the QH-mode. Many of the features of that theory agree with the experimental results reported either previously or in the present paper. One notable example is the prediction that co-rotating QH-mode is possible provided sufficient shear in the edge rotation can be created

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

Energy Technology Data Exchange (ETDEWEB)

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

1991-01-01

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

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

Disruption studies in DIII-D

International Nuclear Information System (INIS)

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

1996-09-01

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

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

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.

Development of improved methods for remote access of DIII-D data and data analysis

International Nuclear Information System (INIS)

Greene, K.L.; McHarg, B.B. Jr.

1997-11-01

The DIII-D tokamak is a national fusion research facility. There is an increasing need to access data from remote sites in order to facilitate data analysis by collaborative researchers at remote locations, both nationally and internationally. In the past, this has usually been done by remotely logging into computers at the DIII-D site. With the advent of faster networking and powerful computers at remote sites, it is becoming possible to access and analyze data from anywhere in the world as if the remote user were actually at the DIII-D site. The general mechanism for accessing DIII-D data has always been via the PTDATA subroutine. Substantial enhancements are being made to that routine to make it more useful in a non-local environment. In particular, a caching mechanism is being built into PTDATA to make network data access more efficient. Studies are also being made of using Distributed File System (DFS) disk storage in a Distributed Computing Environment (DCE). A data server has been created that will migrate, on request, shot data from the DIII-D environment into the DFS environment

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

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

Experimental/theoretical comparisons of the turbulence in the scrape-off-layers of Alcator C-Mod, DIII-D, and NSTX

International Nuclear Information System (INIS)

Terry, J.L. . E-mail : terry@psfc.mit.edu; Zweben, S.J.; Rudakov, D.L.

2003-01-01

The intermittent turbulent transport in the scrape-off-layers of Alcator C-Mod, DIII-D, and NSTX is studied experimentally. On DIII-D the fluctuations of both density and temperature have strongly non-Gaussian statistics, and events with amplitudes above 10 times the mean level are responsible for large fractions of the net particle and heat transport, indicating the importance of turbulence on the transport. In C-Mod and NSTX the turbulence is imaged with a very high density of spatial measurements. The 2-D structure and dynamics of emission from a localized gas puff are observed, and intermittent features (also sometimes called 'blobs') are typically seen. On DIII-D the turbulence is imaged using BES and similar intermittent features are seen. The dynamics of these intermittent features are discussed. The experimental observations are compared with numerical simulations of edge turbulence. The electromagnetic turbulence in a 3-D geometry is computed using non-linear plasma fluid equations. The wavenumber spectra in the poloidal dimension of the simulations are in reasonable agreement with those of the C-Mod experimental images once the response of the optical system is accounted for. The resistive ballooning mode is the dominant linear instability in the simulations. (author)

Overview of the DIII-D program computer systems

International Nuclear Information System (INIS)

McHarg, B.B. Jr.

1997-11-01

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

Helium Exhaust Studies in H-Mode Discharges in the DIII-D Tokamak Using an Argon-Frosted Divertor Cryopump

International Nuclear Information System (INIS)

Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Mahdavi, M.A.; Maingi, R.; West, W.P.; Brooks, N.H.; Burrell, K.H.; Groebner, R.J.; Jackson, G.L.; Klepper, C.C.; Laughon, G.; Menon, M.M.; Mioduszewski, P.K.

1995-01-01

The first experiments demonstrating exhaust of thermal helium in a diverted, H-mode deuterium plasma have been performed on the DIII-D tokamak. The helium, introduced via gas puffing, is observed to reach the plasma core, and then is readily removed from the plasma with a time constant of ∼10--20 energy-confinement times by an in-vessel cryopump conditioned with argon frosting. Detailed analysis of the helium profile evolution suggests that the exhaust rate is limited by the exhaust efficiency of the pump (∼5%) and not by the intrinsic helium-transport properties of the plasma

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

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}.

Evaluation of Particle Pinch and Diffusion Coefficients in the Edge Pedestal of DIII-D H-mode Discharges

Science.gov (United States)

Stacey, W. M.; Groebner, R. J.

2009-11-01

Momentum balance requires that the radial particle flux satisfy a pinch-diffusion relationship. The pinch can be evaluated in terms of measurable quantities (rotation velocities, Er, etc.) by the use of momentum and particle balance [1,2], the radial particle flux can be determined by momentum balance, and then the diffusion coefficient can be evaluated from the pinch diffusion relation using the measured density gradient. Applications to several DIII-D H-mode plasmas are presented. 6pt [1] W.M. Stacey, Contr. Plasma Phys. 48, 94 (2008). [2] W.M. Stacey and R.J. Groebner, Phys. Plasmas 15, 012503 (2008).

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

Science.gov (United States)

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

2017-05-01

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

Plasma boundaries at Mars: a 3-D simulation study

Directory of Open Access Journals (Sweden)

A. Bößwetter

2004-12-01

Full Text Available The interaction of the solar wind with the ionosphere of planet Mars is studied using a three-dimensional hybrid model. Mars has only a weak intrinsic magnetic field, and consequently its ionosphere is directly affected by the solar wind. The gyroradii of the solar wind protons are in the range of several hundred kilometers and therefore comparable with the characteristic scales of the interaction region. Different boundaries emerge from the interaction of the solar wind with the continuously produced ionospheric heavy-ion plasma, which could be identified as a bow shock (BS, ion composition boundary (ICB and magnetic pile up boundary (MPB, where the latter both turn out to coincide. The simulation results regarding the shape and position of these boundaries are in good agreement with the measurements made by Phobos-2 and MGS spacecraft. It is shown that the positions of these boundaries depend essentially on the ionospheric production rate, the solar wind ram pressure, and the often unconsidered electron temperature of the ionospheric heavy ion plasma. Other consequences are rays of planetary plasma in the tail and heavy ion plasma clouds, which are stripped off from the dayside ICB region by some instability.

Key words. Magnetospheric physics (solar wind interactions with unmagnetized bodies – Space plasma physics (discontinuities; numerical simulation studies

A fast scanning probe for DIII--D

International Nuclear Information System (INIS)

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

1992-01-01

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

Critical β of a D-shaped toroidal plasma with sharp boundary

International Nuclear Information System (INIS)

Kito, M.

1981-01-01

Under the cap-cyclide coordinate system, the marginal stability of a sharp boundary toroidal plasma whose cross section is D shaped is examined using the energy principle. The analytical results are valid for arbitrary b, arbitrary aspect ratio, and arbitrary elongation of the plasma. There exists a maximum elongation for a given aspect ratio. The critical beta b/sub c/ is computed for several aspect ratios and elongations. For given aspect ratio, b/sub c/ increases as long as the elongation increases

Edge radial electric field structure in quiescent H-mode plasmas in the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); West, W P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Doyle, E J [University of California, Los Angeles, CA 90095-1597 (United States); Austin, M E [University of Texas at Austin, Austin, TX 78712 (United States); DeGrassie, J S [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Gohil, P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Greenfield, C M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Jayakumar, R [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); Kaplan, D H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lao, L L [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Leonard, A W [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M A [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); McKee, G R [University of Wisconsin, Madison, WI 53706-1687 (United States); Solomon, W M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Thomas, D M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Rhodes, T L [University of California, Los Angeles, CA 90095-1597 (United States); Wade, M R [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wang, G [University of California, Los Angeles, CA 90095-1597 (United States); Watkins, J G [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Zeng, L [University of California, Los Angeles, CA 90095-1597 (United States)

2004-05-01

H-mode operation is the choice for next step tokamak devices based on either conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the {beta} limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D over the past four years have demonstrated a new operating regime, the quiescent H-mode (QH-mode) regime, that solves these problems. QH-mode plasmas have now been run for over 4 s (>30 energy confinement times). Utilizing the steady-state nature of the QH-mode edge allows us to obtain unprecedented spatial resolution of the edge ion profiles and the edge radial electric field, E{sub r}, by sweeping the edge plasma slowly past the view points of the charge exchange spectroscopy system. We have investigated the effects of direct edge ion orbit loss on the creation and sustainment of the QH-mode. Direct loss of ions injected into the velocity-space loss cone at the plasma edge is not necessary for creation or sustainment of the QH-mode. The direct ion orbit loss has little effect on the edge E{sub r} well. The E{sub r} at the bottom of the well in these cases is about -100 kV m{sup -1} compared with -20 to -30 kV m{sup -1} in the standard H-mode. The well is about 1 cm wide, which is close to the diameter of the deuteron gyro-orbit. We also have investigated the effect of changing edge triangularity by changing the plasma shape from upwardly biased single null to magnetically balanced double null. We have now achieved the QH-mode in these double-null plasmas. The increased triangularity allows us to increase pedestal density in QH-mode plasmas by a factor of about 2.5 and overall pedestal pressure by a factor of 2. Pedestal {beta} and {nu}{sup *} values matching the values desired for ITER have been achieved. In

Data analysis software tools for enhanced collaboration at the DIII-D National Fusion Facility

International Nuclear Information System (INIS)

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

2000-01-01

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

Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Pinsker, R. I.; Jackson, G. L.; Luce, T. C.; Politzer, P. A. [General Atomics, PO Box 85608, San Diego, California 92186-5608 (United States); Austin, M. E. [University of Texas at Austin, Austin, Texas 78712 (United States); Diem, S. J.; Kaufman, M. C.; Ryan, P. M. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Doyle, E. J.; Zeng, L. [University of California Los Angeles, Los Angeles, California 90095 (United States); Grierson, B. A.; Hosea, J. C.; Nagy, A.; Perkins, R.; Solomon, W. M.; Taylor, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Maggiora, R.; Milanesio, D. [Politecnico di Torino, Dipartimento di Elettronica, Torino (Italy); Porkolab, M. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Turco, F. [Columbia University, New York, New York 10027 (United States)

2014-02-12

Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ∼2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedly strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. The AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.

DEMONSTRATION OF THE ITER IGNITION FIGURE OF MERIT AT q95>4 IN STATIONARY PLASMAS IN DIII-D

International Nuclear Information System (INIS)

WADE, M.R.; LUCE, T.C.; POLITZER, P.A.; FERRON, J.R.; HYATT, A.W.; SCOVILLE, J.T.; La HAYE, R.J.; KINSEY, J.E.; LASNIER, C.J.; MURAKAMI, M.; PETY, C.C.

2002-01-01

In order to maximize the probability of achieving ignition, the present International Thermonuclear Experimental Reactor (ITER) [1] design (as well as many of its predecessors) is based on operation at high plasma current. This constraint poses many significant engineering challenges, primarily related to the possibility of a sudden termination of the plasma current. Currents induced in the vessel and associated systems in such an event can lead to large forces, and runaway electrons may cause damage to the interior of the vacuum vessel. Present design methods (including those used for ITER) assume that the probability of experiencing such a major disruption increases with plasma current at fixed magnetic field and size. Because fusion performance is assumed to scale in a similar manner, reactor designs tend to seek a compromise between increased fusion performance and reduced susceptibility to disruptions, generally resulting in a design with q 95 ∼ 3.0. Discharges recently developed in the DIII-D tokamak offer a way to obtain equivalent fusion performance with more margin against disruption consequences, having obtained an ignition figure of merit comparable to the ITER baseline scenario with q 95 = 4.5. These discharges have been shown to be stationary on the thermal, resistive, and wall time scales and involve feedback control only of global quantities rather than profiles

Systematic Characterization of Component Failures for the DIII-D Tokamak

International Nuclear Information System (INIS)

Petersen, P.I.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1998-07-01

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

Dynamic divertor control using resonant mixed toroidal harmonic magnetic fields during ELM suppression in DIII-D

Science.gov (United States)

Jia, M.; Sun, Y.; Paz-Soldan, C.; Nazikian, R.; Gu, S.; Liu, Y. Q.; Abrams, T.; Bykov, I.; Cui, L.; Evans, T.; Garofalo, A.; Guo, W.; Gong, X.; Lasnier, C.; Logan, N. C.; Makowski, M.; Orlov, D.; Wang, H. H.

2018-05-01

Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness of the RMP ELM suppression technique for the simultaneous control of divertor heat and particle load required in ITER.

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)

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

International Nuclear Information System (INIS)

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

2004-03-01

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

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

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.

COMPLETE SUPPRESSION OF THE M/N = 2/1 NEOCLASSICAL TEARING MODE USING RADIALLY LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D AND THE REQUIREMENTS FOR ITER

International Nuclear Information System (INIS)

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

2003-01-01

A271 COMPLETE SUPPRESSION OF THE M/N = 2/1 NEOCLASSICAL TEARING MODE USING RADIALLY LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D AND THE REQUIREMENTS FOR ITER. DIII-D experiments demonstrate the first real-time feedback control of the relative location of a narrow beam of microwaves to completely suppress and eliminate a growing tearing mode at the q = 2 surface. long wavelength tearing modes such as the m/n = 2/1 instability are particularly deleterious to tokamak operation. Confinement is seriously degraded by the island, plasma rotation can cease (mode-lock) and disruption can occur. The neoclassical tearing mode (NTM) becomes unstable due to the presence of a helically-perturbed bootstrap current and can be stabilized by replacing the missing bootstrap current in the island O-point by precisely located co-electron cyclotron current drive (ECCD). The optimum position is found when the DIII-D plasma control system (PCS) is put into a search and suppress mode that makes small radial shifts (in about 1 cm steps) in the ECCD location based on minimizing the Mirnov amplitude. Requirements for ITER are addressed

Particle exhaust schemes in the DIII-D advanced divertor configuration

International Nuclear Information System (INIS)

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

1989-01-01

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

Thermal deposition analysis during disruptions on DIII-D using infrared scanners

International Nuclear Information System (INIS)

Lee, R.L.; Hyatt, A.W.; Kellman, A.G.; Taylor, P.L.; Lasnier, C.J.

1995-12-01

The DIII-D tokamak generates plasma discharges with currents up to 3 MA and auxiliary input power up to 20 MW from neutral beams and 4 MW from radio frequency systems. In a disruption, a rapid loss of the plasma current and internal thermal energy occurs and the energy is deposited onto the torus graphite wall. Quantifying the spatial and temporal characteristics of the heat deposition is important for engineering and physics-related issues, particularly for designing future machines such as ITER. Using infrared scanners with a time resolution of 120 micros, measurements of the heat deposition onto the all-graphite walls of DIII-D during two types of disruptions have been made. Each scanner contains a single point detector sensitive to 8--12 microm radiation, allowing surface temperatures from 20 C to 2,000 C to be measured. A zinc selenide window that transmits in the infrared is used as the vacuum window. Views of the upper and lower divertor regions and the centerpost provide good coverage of the first wall for single and double null divertor discharges. During disruptions, the thermal energy is not deposited evenly onto the inner surface of the tokamak, but is deposited primarily in the divertor region when operating diverted discharges. Analysis of the heat deposition during a radiative collapse disruption of a 1.5 MA discharge revealed power densities of 300--350 MW/m 2 in the divertor region. During the thermal quench of the disruption, the energy deposited onto the divertor region was more than 70% of the stored thermal energy in the discharge prior to the disruption. The spatial distribution and temporal behavior of power deposition during high β disruptions will also be presented

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.

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

Dynamic ELM and divertor control using resonant toroidal multi-mode magnetic fields in DIII-D and EAST

Science.gov (United States)

Sun, Youwen

2017-10-01

A rotating n = 2 Resonant Magnetic Perturbation (RMP) field combined with a stationary n = 3 RMP field has validated predictions that access to ELM suppression can be improved, while divertor heat and particle flux can also be dynamically controlled in DIII-D. Recent observations in the EAST tokamak indicate that edge magnetic topology changes, due to nonlinear plasma response to magnetic perturbations, play a critical role in accessing ELM suppression. MARS-F code MHD simulations, which include the plasma response to the RMP, indicate the nonlinear transition to ELM suppression is optimized by configuring the RMP coils to drive maximal edge stochasticity. Consequently, mixed toroidal multi-mode RMP fields, which produce more densely packed islands over a range of additional rational surfaces, improve access to ELM suppression, and further spread heat loading on the divertor. Beneficial effects of this multi-harmonic spectrum on ELM suppression have been validated in DIII-D. Here, the threshold current required for ELM suppression with a mixed n spectrum, where part of the n = 3 RMP field is replaced by an n = 2 field, is smaller than the case with pure n = 3 field. An important further benefit of this multi-mode approach is that significant changes of 3D particle flux footprint profiles on the divertor are found in the experiment during the application of a rotating n = 2 RMP field superimposed on a static n = 3 RMP field. This result was predicted by modeling studies of the edge magnetic field structure using the TOP2D code which takes into account plasma response from MARS-F code. These results expand physics understanding and potential effectiveness of the technique for reliably controlling ELMs and divertor power/particle loading distributions in future burning plasma devices such as ITER. Work supported by USDOE under DE-FC02-04ER54698 and NNSF of China under 11475224.

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.

A decade of DIII-D research. Final report for the period of work, October 1, 1989--September 30, 1998

International Nuclear Information System (INIS)

1999-03-01

During the ten-year DIII-D tokamak operating period of 1989 through 1998, major scientific advances and discoveries were made and facility upgrades and improvements were implemented. Each year, annual reports as well as journal and international conference proceedings document the year-by-year advances (summarized in Section 7). This final contract report, provides a summary of these historical accomplishments. Section 2 encapsulates the 1998 status of DIII-D Fusion Science research. Section 3 summarizes the DIII-D facility operations. Section 4 describes the major upgrades to the DIII-D facility during this period. During the ten-year period, DIII-D has grown from predominantly a General Atomics program to a national center for fusion science with participants from over 50 collaborating institutions and 300 users who spend more than one week annually at DIII-D to carry out experiments or data analysis. In varying degrees, these collaborators participate in formulating the research program directions. The major collaborating institution programs are described in Section 6

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

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

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)

RF power diagnostics and control on the DIII-D, 4 MW 30--120 MHz fast wave current drive system (FWCD)

International Nuclear Information System (INIS)

Ferguson, S.W.; Allen, J.C.; Callis, R.W.; Cary, W.P.; Harris, T.E.

1995-10-01

The Fast Wave Current Drive System uses three 2 MW transmitters to drive three antennas inside the DIII-D vacuum vessel. This paper describes the diagnostics for this system. The diagnostics associated with the General Atomics Fast Wave Current Drive System allow the system tuning to be analyzed and modified on a between shot basis. The transmitters can be exactly tuned to match the plasma with only one tuning shot into the plasma. This facilitates maximum rf power utilization

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.

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

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.

Progress towards a predictive model for pedestal height in DIII-D

International Nuclear Information System (INIS)

Groebner, R.J.; Leonard, A.W.; Snyder, P.B.; Osborne, T.H.; Petty, C.C.; Maggi, C.F.; Fenstermacher, M.E.; Owen, L.W.

2009-01-01

Recent DIII-D pedestal studies provide improved characterization of pedestal scaling for comparison with models. A new pedestal model accurately predicts the maximum achieved pedestal width and height in type I ELMing discharges over a large range of DIII-D operational space, including ITER demonstration discharges. The model is a combination of the peeling-ballooning theory for the MHD stability limits on the pedestal with a simple pedestal width scaling in which the width is proportional to the square root of the pedestal poloidal beta. Width scalings based on the ion toroidal or poloidal gyroradius are much poorer descriptions of DIII-D data. A mass scaling experiment in H and D provides support for a poloidal beta scaling and is not consistent with an ion poloidal gyroradius scaling. Studies of pedestal evolution during the inter-ELM cycle provide evidence that both the pedestal width and height increase during pedestal buildup. Model studies with a 1D kinetic neutrals calculation show that the temporal increase in density width cannot be explained in terms of increased neutral penetration depth. These studies show a correlation of pedestal width with both the square root of the pedestal poloidal beta and the square root of the pedestal ion temperature during the pedestal buildup.

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.

1995-01-01

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 fulfill varying plasma physics requirements. Custom circuitry for laser control (programmable with 1 μs 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 are 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 ns) 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

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

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

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

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

International Nuclear Information System (INIS)

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

1990-09-01

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

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

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

Electron cyclotron current drive experiments on DIII-D

International Nuclear Information System (INIS)

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

1990-05-01

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

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.

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

UEDGE simulations of He transport in DIII-D progress report FY 1996

International Nuclear Information System (INIS)

Fenstermacher, M. E.; Hill, D.N.

1997-01-01

In this report we present the status of numerical simulations of helium exhaust efficiency in the DIII-D tokamak. These computations are intended to serve eventually as a benchmark for simulations carried out for the ITER divertor geometry. Helium ash removal is an important issue for ITER since the helium ash can dilute the central fuel concentration and reduce the fusion power. Present experiments have shown that helium transport in the core plasma is sufficiently rapid to limit the ash buildup to acceptable levels if sufficient helium pumping can be maintained in the divertor. The question of pumping helium gas from the divertor has also been addressed in tokamak experiments, where it was found that the helium concentration in the divertor was about 5-10x lower than in the core plasma (deenrichment). Even so, the exhaust rate was adequate to meet the ITER requirements for central helium concentration. However, the experiments did not reproduce the anticipated ITER divertor geometry or operating conditions. Therefore, the predicted helium exhaust for ITER is still based on numerical simulation. In order to increase the confidence level in the simulations of helium exhaust in ITER, we decided to test the ability of the UEDGE code to simulate the measured enrichment of divertor helium in the DIII-D pumping plenum. Section II presents a description of the experimental discharge used for comparison with the present UEDGE simulations. The UEDGE runs which most closely match the data are presented in Section III including simulations with and without carbon impurity. Section IV presents UEDGE simulations of helium transport and comparison with the helium measurements for these discharges. Conclusions and plans for future work, to complete the detailed benchmarking of UEDGE helium transport models, are given in Section V. 6 refs., 26 figs., 1 tab

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Modification of adhered dust on plasma-facing surfaces due to exposure to ELMy H-mode plasma in DIII-D

Directory of Open Access Journals (Sweden)

I. Bykov

2017-08-01

Full Text Available Transient heat load tests have been conducted in the lower divertor of DIII-D using DiMES manipulator in order to study the behavior of dust on tungsten Plasma Facing Components (PFCs during ELMy H-mode discharges. Samples with pre-adhered, pre-characterized dust have been exposed at the outer strike point (OSP in a series of discharges with varied intra-(inter- ELM heat fluxes. We used C dust because of its high sublimation temperature and non-metal properties. Al dust as a surrogate for Be and W dust were employed as relevant to that in the ITER divertor. The poor initial thermal contact between the substrate and the particles led to overheating, sublimation and shrinking of the carbon dust, and wetting induced coagulation of Al dust. Little modification of the W dust was observed. An enhanced surface adhesion and improvement of the thermal contact of C and Al dust were the result of exposure. A post mortem “adhesive tape” sampling showed that 70% of Al, <5% of W and C particles could not be removed from the surface owing to the improved adhesion. Al and C but not W particles that could be lifted had W inclusions indicating damage to the substrate. This suggests that non destructive methods may be inefficient for removal of dust in ITER.

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.

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

Science.gov (United States)

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

2018-04-01

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

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

Science.gov (United States)

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

2018-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Shen, Wei; Sheng, Zheng-Mao [Department of Physics, Institute for Fusion Theory and Simulation, Zhejiang University, Hangzhou 310027 (China); Fu, G. Y.; Tobias, Benjamin [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Zeeland, Michael Van [General Atomics, San Diego, California 92186-5608 (United States); Wang, Feng [School of Physics and Optoelectronic Engineering, Dalian University of Technology, Dalian 116024 (China)

2015-04-15

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

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

International Nuclear Information System (INIS)

Cooper, C. M.; Pace, D. C.; Paz-Soldan, C.; Eidietis, N. W.; Commaux, N.; Shiraki, D.; Hollmann, E. M.

2016-01-01

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.

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.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Nonrigid, Linear Plasma Response Model Based on Perturbed Equilibria for Axisymmetric Tokamak Control Design

International Nuclear Information System (INIS)

Welander, A.S.; Deranian, R.D.; Humphreys, D.A.; Leuer, J.A.; Walker, M.L.

2005-01-01

Tokamak control design relies on an accurate linear model of the plasma response, which can often dominate the local field variations in regions under active feedback control. For example, when fluxes at selected points on the plasma boundary are regulated in DIII-D, the plasma response to a change in a coil current gives rise to a flux change which can be larger than and opposite to the flux change caused by the coil alone.In the past, rigid plasma models have been used for linear stability and shape control design. In a rigid model, the plasma current profile is considered fixed and moves rigidly in response to control coils to maintain radial and vertical force balance. In a nonrigid model, however, changes in the plasma shape and current profile are taken into account. Such models are expected to be important for future advanced tokamak control design. The present work describes development of a nonrigid plasma response model for high-accuracy multivariable control design and provides comparisons of model predictions against DIII-D experimental data. The linear perturbed plasma response model is calculated rapidly from an existing equilibrium solution

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.

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.

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.

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Canik, J.M. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Leonard, A.W.; Mahdavi, M.A. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Watkins, J.G. [Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Ferron, J.R.; Groebner, R.J. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Hill, D.N. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Hyatt, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Luce, T.C. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Moyer, R.A. [University of California–San Diego, La Jolla, CA 92093-0417 (United States); Stangeby, P.C. [University of Toronto Institute of Aerospace Studies, Toronto (Canada)

2013-07-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-12-15

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

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

Practical aspects of a 2-D edge-plasma model

International Nuclear Information System (INIS)

Rensink, M.E.; Hill, D.N.; Porter, G.D.; Braams, B.J.; Princeton Univ., NJ

1989-07-01

The poloidal divertor configuration is considered the most promising solution to the particle and energy exhaust problem for a tokamak reactor. The scrape-off layer plasma surrounding the core and the high-recycling plasma near the divertor plates can be modelled by fluid equations for particle, momentum and energy transport. A numerical code (B2) based on a two-dimensional multi-fluid model has been developed for the study of edge plasmas in tokamaks. In this report we identify some key features of this model as applied to the DIII-D tokamak. 2 refs., 1 fig

Scrape-off layer plasma modeling for the DIII-D tokamak

International Nuclear Information System (INIS)

Porter, G.D.; Rognlien, T.D.; Allen, S.L.

1994-09-01

The behavior of the scrape-off layer (SOL) region in tokamaks is believed to play an important role determining the overall device performance. In addition, control of the exhaust power has become one of the most important issues in the design of future devices such as ITER and TPX. This paper presents the results of application of 2-D fluid models to the DII-D tokamak, and research into the importance of processes which are inadequately treated in the fluid models. Comparison of measured and simulated profiles of SOL plasma parameters suggest the physics model contained in the UEDGE code is sufficient to simulate plasmas which are attached to the divertor plates. Experimental evidence suggests the presence of enhanced plasma recombination and momentum removal leading to the existence of detached plasma states. UEDGE simulation of these plasmas obtains a bifurcation to a low temperature plasma at the divertor, but the plasma remains attached. Understanding the physics of this detachment is important for the design of future devices. Analytic studies of the behavior of SOL plasmas enhance our understanding beyond that achieved with fluid modeling. Analysis of the effect of drifts on sheath structure suggest these drifts may play a role in the detachment process. Analysis of the turbulent-transport equations indicate a bifurcation which is qualitatively similar to the experimentally different behavior of the L- and H-mode SOL. Electrostatic simulations of conducting wall modes suggest possible control of the SOL width by biasing

Characterization of wall conditions in DIII-D

International Nuclear Information System (INIS)

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

1996-10-01

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

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.

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

Science.gov (United States)

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

2018-06-01

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

Global Alfven Eigenmodes in DIII-D

International Nuclear Information System (INIS)

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

1992-06-01

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

Experimental tests of linear and nonlinear three-dimensional equilibrium models in DIII-D

Energy Technology Data Exchange (ETDEWEB)

King, J. D., E-mail: kingjd@fusion.gat.com [Oak Ridge Institute for Science Education, Oak Ridge, Tennessee 37830-8050 (United States); General Atomics, P.O. Box 85608, San Diego, California 92816-5608 (United States); Strait, E. J.; Ferraro, N. M.; Lanctot, M. J.; Paz-Soldan, C.; Turnbull, A. D. [General Atomics, P.O. Box 85608, San Diego, California 92816-5608 (United States); Lazerson, S. A.; Logan, N. C.; Park, J.-K.; Nazikian, R.; Okabayashi, M. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States); Haskey, S. R. [Plasma Research Laboratory, Research School of Physical Sciences and Engineering, The Australia National University, Canberra, Australian Capital Territory 0200 (Australia); Hanson, J. M. [Columbia University, 2960 Broadway, New York, New York 10027 (United States); Liu, Yueqiang [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Shiraki, D. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, Tennessee 37831 (United States)

2015-07-15

DIII-D experiments using new detailed magnetic diagnostics show that linear, ideal magnetohydrodynamics (MHD) theory quantitatively describes the magnetic structure (as measured externally) of three-dimensional (3D) equilibria resulting from applied fields with toroidal mode number n = 1, while a nonlinear solution to ideal MHD force balance, using the VMEC code, requires the inclusion of n ≥ 1 to achieve similar agreement. These tests are carried out near ITER baseline parameters, providing a validated basis on which to exploit 3D fields for plasma control development. Scans of the applied poloidal spectrum and edge safety factor confirm that low-pressure, n = 1 non-axisymmetric tokamak equilibria are determined by a single, dominant, stable eigenmode. However, at higher beta, near the ideal kink mode stability limit in the absence of a conducting wall, the qualitative features of the 3D structure are observed to vary in a way that is not captured by ideal MHD.

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

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

International Nuclear Information System (INIS)

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

1997-01-01

V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

V-4Cr-4Ti alloy has been selected for use in the manufacture of a portion of the DIII-D Radiative Divertor upgrade. The production of a 1200-kg ingot of V-4Cr-4Ti alloy, and processing into final sheet and rod product forms suitable for components of the DIII-D Radiative Divertor structure, has been completed at Wah Chang (formerly Teledyne Wah Chang) of Albany, Oregon (WCA). Joining of V-4Cr-4Ti alloy has been identified as the most critical fabrication issue for its use in the RD Program, and research into several joining methods for fabrication of the RD components, including resistance seam, friction, and electron beam welding, is continuing. Preliminary trials have been successful in the joining of V-alloy to itself by electron beam, resistance, and friction welding processes, and to Inconel 625 by friction welding. An effort to investigate the explosive bonding of V-4Cr-4Ti alloy to Inconel 625 has also been initiated, and results have been encouraging. In addition, preliminary tests have been completed to evaluate the susceptibility of V-4Cr-4Ti alloy to stress corrosion cracking in DIII-D cooling water, and the effects of exposure to DIII-D bakeout conditions on the tensile and fracture behavior of V-4Cr-4Ti alloy.

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

Plasma Shape Control on the National Spherical Torus Experiment using Real-time Equilibrium Reconstruction

International Nuclear Information System (INIS)

Gates, D.A.; Ferron, J.R.; Bell, M.; Gibney, T.; Johnson, R.; Marsala, R.J.; Mastrovito, D.; Menard, J.E.; Mueller, D.; Penaflor, B.; Sabbagh, S.A.; Stevenson, T.

2005-01-01

Plasma shape control using real-time equilibrium reconstruction has been implemented on the National Spherical Torus Experiment (NSTX). The rtEFIT code originally developed for use on DIII-D was adapted for use on NSTX. The real-time equilibria provide calculations of the flux at points on the plasma boundary, which is used as input to a shape control algorithm known as isoflux control. The flux at the desired boundary location is compared to a reference flux value, and this flux error is used as the basic feedback quantity for the poloidal-field coils on NSTX. The hardware that comprises the control system is described, as well as the software infrastructure. Examples of precise boundary control are also presented

Experiences with delta compression of data produced by DIII

International Nuclear Information System (INIS)

Henline, P.A.

1986-01-01

The amount of data collected for each tokamak experimental shot is rapidly increasing. This is caused by many factors, including more diagnostic experiments, reduced cost of electronics hardware (especially memory), and longer plasma duration. The design goal for the DIII-D tokamak is 25 Mbytes of data per shot. In order to store the shot data as one logical unit, the delta compression algorithm, as it was implemented at ORNL by E. Blair, is being used. Statistics on compression factors, times, and general usage will be presented for actual DIII data. Data for 8-, 10-, and 12-bit digitizers will be highlighted since this hardware is common to most sites

The charge exchange recombination diagnostic system on the DIII-D tokamak

International Nuclear Information System (INIS)

Gohil, P.; Burrell, K.H.; Groebner, R.J.; Kim, J.; Martin, W.C.; McKee, E.L.; Seraydarian, R.P.

1991-11-01

The charge exchange recombination (CER) diagnostic system on the DIII-D tokamak is used to make spatially and temporally resolved measurements of the ion temperature and toroidal and poloidal rotation velocities. This is performed through visible spectroscopic measurements of the Doppler broadened and Doppler shifted HE II 468.6 nm, the CVI 529.1 nm, and the BV 494.5 nm spectral lines which have been excited by charge exchange recombination interactions between the fully stripped ions and the neutral atoms from the heating beams. The plasma viewing optics comprises 32 viewing chords spanning a typical plasma minor radius of 63 cm across the midplane, of which 15 spatial chords span 4.2 cm at the plasma edge just within the separatrix and provide a chord-to-chord spatial resolution of 0.3 cm. Fast camera readout electronics can provide a temporal resolution of 260 μs per time slice, but the effective minimum integration time, at present, is 1 ms which is limited by the detected photon flux from the plasma and the decay times of the phosphors used on the multichannel plate image intensifiers. Significant changes in the edge plasma radial electric field at the L-H transition have been observed, as determined from the CER measurements, and these results are being extensively compared to theories which consider the effects of sheared electric fields on plasma turbulence. 13 refs., 10 figs

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

International Nuclear Information System (INIS)

Phelps, R.D.

1994-01-01

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

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

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.

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

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

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.

The importance of matched poloidal spectra to error field correction in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Paz-Soldan, C., E-mail: paz-soldan@fusion.gat.com; Lanctot, M. J.; Buttery, R. J.; La Haye, R. J.; Strait, E. J. [General Atomics, P.O. Box 85608, San Diego, California 92121 (United States); Logan, N. C.; Park, J.-K.; Solomon, W. M. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Shiraki, D.; Hanson, J. M. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

2014-07-15

Optimal error field correction (EFC) is thought to be achieved when coupling to the least-stable “dominant” mode of the plasma is nulled at each toroidal mode number (n). The limit of this picture is tested in the DIII-D tokamak by applying superpositions of in- and ex-vessel coil set n = 1 fields calculated to be fully orthogonal to the n = 1 dominant mode. In co-rotating H-mode and low-density Ohmic scenarios, the plasma is found to be, respectively, 7× and 20× less sensitive to the orthogonal field as compared to the in-vessel coil set field. For the scenarios investigated, any geometry of EFC coil can thus recover a strong majority of the detrimental effect introduced by the n = 1 error field. Despite low sensitivity to the orthogonal field, its optimization in H-mode is shown to be consistent with minimizing the neoclassical toroidal viscosity torque and not the higher-order n = 1 mode coupling.

Characterisation of the ELM synchronized H-mode edge pedestal in ASDEX upgrade and DIII-D

Energy Technology Data Exchange (ETDEWEB)

Schneider, Philip A.; Wolfrum, Elisabeth; Guenter, Sibylle; Kurzan, Bernd; Zohm, Hartmut [Max Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Groebner, Rich; Osborne, Tom H.; Ferron, John; Snyder, Philip B. [General Atomics, San Diego, CA (United States); Dunne, Mike G. [Department of Physics, University College Cork, Association Euratom-DCU, Cork (Ireland); Collaboration: ASDEX Upgrade Team; DIII-D Team

2011-07-01

The results of a large database of edge pedestal data from type-I ELMy H-mode discharges from ASDEX Upgrade and DIII-D are presented. The data from high resolution edge diagnostics of both devices is analysed with the same analysis code in order to avoid systematic differences. Furthermore, sophisticated equilibrium reconstructions are used to asses uncertainties which arise during mapping from 2D real space coordinates to 1D flux coordinates. ELM synchronization allows the study of the pedestal structure at the ELM stability boundary. The pedestal is characterized by its top value, the gradient and the width. A large parameter range is covered by the two devices. Over this parameter range the profile shape of edge electron density differs from that of the temperature, irrespective of the device. However, the resulting electron pressure profile shape remains similar for all analysed H-Mode discharges.

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

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

1.5D quasilinear model and its application on beams interacting with Alfven eigenmodes in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Ghantous, K.; Gorelenkov, N. N. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, New Jersey 08543-0451 (United States); Berk, H. L. [Institute for Fusion Studies, University of Texas, 2100 San Jacinto Blvd., Austin, Texas 78712-1047 (United States); Heidbrink, W. W. [Department of Physics and Astronomy, University of California Irvine, Irvine, California 92697 (United States); Van Zeeland, M. A. [General Atomics, PO Box 85608, San Diego, California 92186-560 (United States)

2012-09-15

We propose a model, denoted here by 1.5D, to study energetic particle (EP) interaction with toroidal Alfvenic eigenmodes (TAE) in the case where the local EP drive for TAE exceeds the stability limit. Based on quasilinear theory, the proposed 1.5D model assumes that the particles diffuse in phase space, flattening the pressure profile until its gradient reaches a critical value where the modes stabilize. Using local theories and NOVA-K simulations of TAE damping and growth rates, the 1.5D model calculates the critical gradient and reconstructs the relaxed EP pressure profile. Local theory is improved from previous study by including more sophisticated damping and drive mechanisms such as the numerical computation of the effect of the EP finite orbit width on the growth rate. The 1.5D model is applied on the well-diagnosed DIII-D discharges no. 142111 [M. A. Van Zeeland et al., Phys. Plasmas 18, 135001 (2011)] and no. 127112 [W. W. Heidbrink et al., Nucl. Fusion. 48, 084001 (2008)]. We achieved a very satisfactory agreement with the experimental results on the EP pressure profiles redistribution and measured losses. This agreement of the 1.5D model with experimental results allows the use of this code as a guide for ITER plasma operation where it is desired to have no more than 5% loss of fusion alpha particles as limited by the design.

Three-dimensional plasma transport in open chaotic magnetic fields. A computational assessment for tokamak edge layers

International Nuclear Information System (INIS)

Frerichs, Heinke Gerd

2010-04-01

The development of nuclear fusion as an alternative energy source requires the research on magnetically confined, high temperature plasmas. In particular, the quantification of plasma flows in the domain near exposed material surfaces of the plasma container by computer simulations is of key importance, both for guiding interpretation of present fusion experiments and for aiding the ongoing design activities for large future devices such as ITER, W7-X or the DEMO reactor. There is a large number of computational issues related to the physics of hot, fully ionized and magnetized plasmas near surfaces of the vacuum chamber. This thesis is dedicated to one particular such challenge, namely the numerical quantification of self-consistent kinetic neutral gas and plasma fluid flows in very complex 3D (partially chaotic) magnetic fields, in the absence of any common symmetries for plasma and neutral gas dynamics. Such magnetic field configurations are e.g. generated by externally applied magnetic perturbations at the plasma edge, and are of great interest for the control of particle and energy exhausts. In the present thesis the 3D edge plasma and neutral particle transport code EMC3-EIRENE is applied to two distinct configurations of open chaotic magnetic system: at the TEXTOR and DIII-D tokamaks. Improvements of the edge transport model and extensions of the transport code are presented, which have allowed such simulations for the first time for 3D scenarios at DIII-D with ITER similar plasmas. A strong 3D effect of the chaotic magnetic field on the DIII-D edge plasma is found and analyzed in detail. It is found that a pronounced striation pattern of target particle and heat fluxes at DIII-D can only be obtained up to a certain upper limiting level of anomalous cross-field transport. Hence, in comparison to experimental data, these findings allow to narrow down the range of this model parameter. One particular interest at TEXTOR is the achievement of a regime with

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

Science.gov (United States)

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

2015-11-01

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

ELMs IN DIII-D HIGH PERFORMANCE DISCHARGES

International Nuclear Information System (INIS)

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

2003-01-01

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

Turbulence imaging and applications using beam emission spectroscopy on DIII-D (invited)

Science.gov (United States)

McKee, G. R.; Fenzi, C.; Fonck, R. J.; Jakubowski, M.

2003-03-01

Two-dimensional measurements of density fluctuations are obtained in the radial and poloidal plane of the DIII-D tokamak with the Beam Emission Spectroscopy (BES) diagnostic system. The goals are to visualize the spatial structure and time evolution of turbulent eddies, as well as to obtain the 2D statistical properties of turbulence. The measurements are obtained with an array of localized BES spatial channels configured to image a midplane region of the plasma. 32 channels have been deployed, each with a spatial resolution of about 1 cm in the radial and poloidal directions, thus providing measurements of turbulence in the wave number range 0movies have broad application to a wide variety of fundamental turbulence studies: imaging of the highly complex, nonlinear turbulent eddy interactions, measurement of the 2D correlation function, and S(kr,kθ) wave number spectra, and direct measurement of the equilibrium and time-dependent turbulence flow field. The time-dependent, two-dimensional turbulence velocity flow-field is obtained with time-delay-estimation techniques.

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

International Nuclear Information System (INIS)

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

2011-01-01

Fast beam-ion losses were studied in DIII-D in the presence of a scaled mockup of two Test Blanket Modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam-ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot which is predicted to be different among the various codes.

DIII-D dust particulate characterization (June 1998 Vent)

International Nuclear Information System (INIS)

Carmack, W.J.

1999-01-01

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

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.

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

International Nuclear Information System (INIS)

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

1997-06-01

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

Observations of ELM stabilization during neutral beam injection in DIII-D

Science.gov (United States)

Bortolon, Alessandro; Kramer, Gerrit; Diallo, Ahmed; Knolker, Matthias; Maingi, Rajesh; Nazikian, Raffi; Degrassie, John; Osborne, Thomas

2017-10-01

Edge localized modes (ELMs) are generally interpreted as peeling-ballooning instabilities, driven by the pedestal current and pressure gradient, with other subdominant effects possibly relevant close to marginal stability. We report observations of transient stabilization of type-I ELMs during neutral beam injection (NBI), emerging from a combined dataset of DIII-D ELMy H-mode plasmas with moderate heating obtained through pulsed NBI waveforms. Statistical analysis of ELM onset times indicates that, in the selected dataset, the likelihood of onset of an ELM lowers significantly during NBI modulation pulses, with the stronger correlation found with counter-current NBI. The effect is also found in rf-heated H-modes, where ELMs appear inhibited when isolated diagnostic beam pulses are applied. Coherent average analysis is used to determine how plasma density, temperature, rotation as well as beam ion quantities evolve during a NB modulation cycle, finding relatively small changes ( 3%) of pedestal Te and ne and toroidal and poloidal rotation variations up to 5 km/s. The effect of these changes on pedestal stability will be discussed. Work supported by US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466.

1.5D quasilinear model and its application on beams interacting with Alfvén eigenmodes in DIII-D

Science.gov (United States)

Ghantous, K.; Gorelenkov, N. N.; Berk, H. L.; Heidbrink, W. W.; Van Zeeland, M. A.

2012-09-01

We propose a model, denoted here by 1.5D, to study energetic particle (EP) interaction with toroidal Alfvenic eigenmodes (TAE) in the case where the local EP drive for TAE exceeds the stability limit. Based on quasilinear theory, the proposed 1.5D model assumes that the particles diffuse in phase space, flattening the pressure profile until its gradient reaches a critical value where the modes stabilize. Using local theories and NOVA-K simulations of TAE damping and growth rates, the 1.5D model calculates the critical gradient and reconstructs the relaxed EP pressure profile. Local theory is improved from previous study by including more sophisticated damping and drive mechanisms such as the numerical computation of the effect of the EP finite orbit width on the growth rate. The 1.5D model is applied on the well-diagnosed DIII-D discharges #142111 [M. A. Van Zeeland et al., Phys. Plasmas 18, 135001 (2011)] and #127112 [W. W. Heidbrink et al., Nucl. Fusion. 48, 084001 (2008)]. We achieved a very satisfactory agreement with the experimental results on the EP pressure profiles redistribution and measured losses. This agreement of the 1.5D model with experimental results allows the use of this code as a guide for ITER plasma operation where it is desired to have no more than 5% loss of fusion alpha particles as limited by the design.

State transitions, hysteresis, and control parameters on DIII-D

International Nuclear Information System (INIS)

Thomas, D.M.; Groebner, R.J.; Carlstrom, T.N.; Osborne, T.H.; Petrie, T.W.

1998-07-01

The theory of turbulence decorrelation by ExB velocity shear is the leading candidate to explain the changes in turbulence and transport that are seen at the plasma edge at the L to H transition. Based on this, a key question is: What are the conditions or control parameters needed to begin the formation of the E r shear layer and thus trigger the L to H transition? On the DIII-D tokamak, the authors are attacking this question both through direct tests of the various theories and by trying to gain insight into the fundamental physics by investigating the control parameters which have a major effect on the power threshold. In this paper the authors describe results of studies on oscillating discharges where the plasma transitions continuously between L and H states. By following the dynamics of the plasma state through the forward and back transitions, they can represent the evolution of various control parameter candidates as a trajectory in various parametric spaces. The shape of these control curves can illustrate the specific nonlinearities governing the L-H transition problem, and under the proper conditions may be interpreted in the context of various phase-transition based models. In particular, the hysteresis exhibited in the various curves may help to clarify causality (what are the critical parameters) and may serve as tests of the models, given sufficient experimental accuracy. At present they are looking at T e , E r and ballooning/diamagnetic parameters as possible control parameter candidates

Performance characteristics of the DIII-D advanced divertor cryopump

International Nuclear Information System (INIS)

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

1993-10-01

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

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)

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

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

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

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