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DEVELOPMENT IN THE DIII-D TOKAMAK OF HYBRID OPERATION SCENARIOS FOR BURNING PLASMA EXPERIMENTS

Description: OAK-B135 The basic parameters of proposed burning plasma experiments such as ITER and FIRE have been chosen based on analysis of multi-machine databases of confinement, stability, and divertor operation. given these specifications, it is of interest to run discharges in present-day machines such as DIII-D to verify the design basis and evaluate the margin available to achieve the mission goals. it is especially important to operate discharges which are stationary with respect to the current relaxation time scale ({tau}{sub R}) since it is well-known that higher performance can be achieved transiently. Attention has been focused on validating the baseline scenario for diverted machines--ELMing H-mode discharges with q{sub 95} = 3 with sawteeth. However, there is also interest in the ITER program to assess the feasibility of operating the tokamak in a mode to maximize the neutron fluence for the purpose of testing the design of various components critical to the nuclear fuel cycle and energy conversion systems in a fusion power plant. It was originally envisioned that these discharges would be intermediate between an inductive burn (baseline) scenario and a fully noninductive (steady state) scenario; therefore, this type of discharge has become known as a hybrid scenario. In the course of investigating these hybrid scenarios in DIII-D, two key results have been obtained. First, stationary discharges with q{sub 95} > 4 have been obtained which project to Q{sub fus} {approx} 10 in ITER. The projected duration of these discharges in ITER when using the full inductive flux capability is > 4000 s. (The significant engineering issues of site heat capacity, activation, and tritium consumption are beyond the scope of this work). Second, utilizing the same plasma initiation techniques as developed for the hybrid scenario, discharges at q{sub 95} = 3.2 project to near ignition in ITER, even with reduced parameters. ...
Date: August 1, 2003
Creator: LUCE,TC; WADE,MR; FERRON,JR; HYATT,AW; POLITZER,PA & SIPS,ACC
Partner: UNT Libraries Government Documents Department

Recent H-mode density limit experiments on DIII-D

Description: A vast body of tokamak data is in good agreement with the empirical density limit scalings proposed by Hugill and Greenwald. These scalings have common puzzling features of showing no dependence on either impurity concentration or heating power, since the density limit is frequently correlated with a rapid rise of the edge radiation. Despite the resiliency of these scalings, several machines under restrictive conditions have operated at densities well above the predictions of these scalings, albeit with pellet injection and at varying degrees of confinement degradation. Furthermore, data from several machines display a weak dependence on heating power. These results cast doubt on the universal validity of both of these scalings. Nevertheless the fact remains that access to densities above Hugill-Greenwald scaling is very difficult. A number of theories based on radiative power balance in the plasma boundary have explained some but not all features of tokamak density limit behavior, and as ITER design studies recently brought to focus, a satisfactory understanding of this phenomenon is lacking. Motivated by a need for better understanding of effects of density and fueling on tokamak plasmas in general, the authors have conducted a series of experiments designed to identify and isolate physical effects suspected to be directly or indirectly responsible for the density limit. The physical effects being considered include: divertor power balance, MARFE, poloidally symmetric radiative instabilities, MHD instabilities, and transport. In this paper they first present a brief summary of the experimental results up to the writing of this paper. The remainder of the paper is devoted to a comparison of this data at the onset of the MARFE instability with predictions of theory and the implications of the results on access to densities beyond the Hugill-Greenwald limit.
Date: June 1997
Creator: Mahdavi, M.A.; Maingi, R. & Hyatt, A.W.
Partner: UNT Libraries Government Documents Department

The electrical insulation of the DIII-D advanced divertor electrode

Description: The electrode for biasing experiments on the DIII-D tokamak was installed in the summer of 1990 and biasing experiments have shown positive results. For the electrode, electrical insulation had to provide voltage standoff in the DIII-D divertor environment of neutral pressures in the range of 10{sup {minus}8} to 5 {times} 10{sup {minus}2} torr, variable magnetic fields, and in the presence of ionizing radiation. The electrical insulation system was designed and tested in air and vacuum for voltages up to 3 kV. In this paper, we provide an update on our operating experience, problems encountered, and improvements to the system. Electrical breakdown of some components has occurred during tokamak operations and transient voltages, up to 5 kV, have been observed. The original concept for insulating the water and electrical feeds for the electrode, a thin layer of woven ceramic cloth insulation between the feeds and a ground plane to keep out stray plasma, was found to be prone to failure. A new scheme of rigid ceramic insulators surrounded by a ground plane was designed and is being implemented. Another problem was arcs from vessel potential surfaces to the electrode in several locations where vessel ground existed within 1 cm of the electrode. The arc traveled in a small crack between two insulators. Careful attention has been paid to closing this and other small gaps in the insulation. Coatings on the surface of plasma facing insulators have been found to be electrically conductive. Grooves are being machined into the insulators to give areas shadowed from the coating source. Tests are being done to demonstrate the design concepts in both vacuum and glow discharge environments. Plasma sprayed ceramic coatings were also tested to determine the voltage standoff capability in a glow plasma discharge. The results of these tests will be discussed. 2 refs., ...
Date: October 1, 1991
Creator: Smith, J.P.; Schaffer, M.J. & Hyatt, A.W.
Partner: UNT Libraries Government Documents Department

An edge density fluctuation diagnostic for DIII-D using lithium beams

Description: In order to investigate fluctuations in the edge regions of tokamaks and their influence on particle and energy transport, we are developing an {tilde n} diagnostic for DIII-D system based on an injected neutral lithium beam. Analysis of the 670.8 nm light emitted along the beam trajectory due to collisional excitation should yield the behavior of the electron density and its associated fluctuations in the edge region of DIII-D in both L- and H-mode plasma configurations. The planned system, design considerations, and expected performance levels will be presented. 22 refs., 2 figs.
Date: September 1, 1990
Creator: Thomas, D.M.; Hyatt, A.W. & Thomas, M.P.
Partner: UNT Libraries Government Documents Department

Investigation of the effect of large core changes in toroidal plasma rotation and radial electric field on confinement in H-mode discharges in the DIII-D tokamak

Description: The plasma toroidal rotation and the radial electric field in the core ({rho}{approx lt}0.9) of H-mode discharges in DIII-D are greatly altered by the drag produced by application of static, resonant magnetic field perturbations from an external coil ( the n = 1 coil''). Transport loss due to turbulent fluctuations can in theory be reduced by E{sub r} shear stabilization or suppression. This is tested experimentally in DIII-D by using the magnetic breaking'' of rotation (with concomitant change in E{sub r}) as an independent control. The magnetic braking produces reversal of the core radial electric field, E{sub r}, and E{sub r} shear. However, the plasma maintains a negative edge ({rho}{approx lt}0.95) E{sub r} and E{sub r} shear and remains in H-mode with insignificant changes in global confinement, density profile and temperature profiles. The H-mode confinement is remarkably robust to the increasing error fields and the slowed toroidal rotation up to the onset of a locked mode which induces a transition to L-mode, the virtual cessation of plasma rotation and large reconnected islands.
Date: May 1, 1992
Creator: LaHaye, R.J.; Groebner, R.J.; Hyatt, A.W. & Scoville, J.T.
Partner: UNT Libraries Government Documents Department

Bias-sustained shield plasma

Description: Divertor biasing may provide a method for density and impurity control by enhancing the shielding efficiency of the scrape-off layer. The idea is to make the scrape-off plasma denser and thicker by heating it with a bias-driven current, and by inducing a radial E [times] B drift. If the bias is applied to flux surfaces at the outer edge of the usual scrape-off layer, a new layer of plasma can be added which is sustained by the bias-supplied power. A simple theoretical model will be presented which shows that there is a threshold condition which must be satisfied in order for the bias-heated plasma to be self-sustaining. The bias-sustained plasma must also be opaque enough to neutrals in order for it to be fueled by a gas puff, which means that it win serve as a shield to the core plasma against neutral impurities and hydrogen. Experiments performed on DIII-D have demonstrated both a modification of the central nickel impurity concentration and an increase in the ionization of hydrogen within the scrape-off layer due to biasing.
Date: September 1, 1992
Creator: Staebler, G.M.; Hyatt, A.W.; Schaffer, M.J. & Mahdavi, M.A.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: July 1, 2003
Creator: LAHAYE,RJ; LUCE,TC; PETTY,CC; HUMPHREYS,DA; HYATT,AW; PERKINS,FW et al.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: November 1, 1997
Creator: Broesch, J.D.; Scoville, J.T.; Hyatt, A.W. & Coon, R.M.
Partner: UNT Libraries Government Documents Department

Automated Fault Detection for DIII-D Tokamak Experiments

Description: 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.
Date: November 1, 1999
Creator: Walker, M.L.; Scoville, J.T.; Johnson, R.D.; Hyatt, A.W. & Lee, J.
Partner: UNT Libraries Government Documents Department

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

Description: A271 COMPLETE SUPPRESSION OF THE M=2/N-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D. The first suppression of the important and deleterious m=2/n=1 neoclassical tearing mode (NTM) is reported using electron cyclotron current drive (ECCD) to replace the ''missing'' bootstrap current in the island O-point. Experiments on the DIII-D tokamak verify the maximum shrinkage of the m=2/n=1 island occurs when the ECCD location coincides with the q = 2 surface. The DIII-D plasma control system is put into search and suppress mode to make small changes in the toroidal field to find and lock onto the optimum position, based on real time measurements of dB{sub {theta}}/dt, for complete m=2/n=1 NTM suppression by ECCD. The requirements on the ECCD for complete island suppression are well modeled by the modified Rutherford equation for the DIII-D plasma conditions.
Date: March 1, 2003
Creator: PETTY,CC; LAHAYE,LA; LUCE,TC; HUMPHREYS,DA; HYATT,AW; PRATER,R et al.
Partner: UNT Libraries Government Documents Department

Magnetic and thermal energy flow during disruptions in DIII-D

Description: The authors present results from disruption experiments where they measure magnetic energy flow across a closed surface surrounding the plasma using a Poynting flux analysis to measure the electromagnetic power, bolometers to measure radiation power and IR scanners to measure radiation and particle heat conduction to the divertor. The initial and final stored energies within the volume are found using the full equilibrium reconstruction code EFIT. From this analysis they calculate an energy balance and find that they can account for all energy deposited on the first wall and the divertor to within about 10%.
Date: July 1, 1996
Creator: Hyatt, A.W.; Lee, R.L.; Humphreys, D.A.; Kellman, A.G.; Taylor, P.L.; Cuthbertson, J.W. et al.
Partner: UNT Libraries Government Documents Department

DISRUPTION MITIGATION WITH HIGH-PRESSURE NOBLE GAS INJECTION

Description: OAK A271 DISRUPTION MITIGATION WITH HIGH-PRESSURE NOBLE GAS INJECTION. High-pressure gas jets of neon and argon are used to mitigate the three principal damaging effects of tokamak disruptions: thermal loading of the divertor surfaces, vessel stress from poloidal halo currents and the buildup and loss of relativistic electrons to the wall. The gas jet penetrates as a neutral species through to the central plasma at its sonic velocity. The injected gas atoms increase up to 500 times the total electron inventory in the plasma volume, resulting in a relatively benign radiative dissipation of >95% of the plasma stored energy. The rapid cooling and the slow movement of the plasma to the wall reduce poloidal halo currents during the current decay. The thermally collapsed plasma is very cold ({approx} 1-2 eV) and the impurity charge distribution can include > 50% fraction neutral species. If a sufficient quantity of gas is injected, the neutrals inhibit runaway electrons. A physical model of radiative cooling is developed and validated against DIII-D experiments. The model shows that gas jet mitigation, including runaway suppression, extrapolates favorably to burning plasmas where disruption damage will be more severe. Initial results of real-time disruption detection triggering gas jet injection for mitigation are shown.
Date: October 1, 2002
Creator: WHYTE, DG; JERNIGAN, TC; HUMPHREYS, DA; HYATT, AW; LASNIER, CJ; PARKS, PB et al.
Partner: UNT Libraries Government Documents Department

STATIONARY HIGH-PERFORMANCE DISCHARGES IN THE DII-D TOKAMAK

Description: 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{sub 95} > 4). A figure of merit for fusion gain ({beta}{sub N}H{sub 89}/q{sub 95}{sup 2}) has been maintained at values corresponding to ! = 10 operation in a burning plasma for > 6 s or 36{tau}{sub E} and 2{tau}{sub R}. The key element is the relaxation of the current profile to a stationary state with q{sub min} > 1. In the absence of sawteeth and fishbones, stable operation has been achieved up to the estimated no-wall {beta} 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.
Date: November 1, 2002
Creator: LUCE,TC; WADE,MR; FERRON,JR; HYATT,AW; KELLMAN,AG; KINSEY,JE et al.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: June 1, 1999
Creator: Phillips, J.C.; Greene, K.L.; Hyatt, A.W.; McHarg, B.B. Jr. & Penaflor, B.G.
Partner: UNT Libraries Government Documents Department

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

Description: 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 {micro}s, 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 {micro}m 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{sup 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 {beta} disruptions will also be presented.
Date: December 1, 1995
Creator: Lee, R.L.; Hyatt, A.W.; Kellman, A.G.; Taylor, P.L. & Lasnier, C.J.
Partner: UNT Libraries Government Documents Department

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

Description: OAK-B135 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 geometry for the ECCD launch, the second harmonic resonance 2f{sub ce} and the q=2 surface are shown. 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.
Date: June 1, 2003
Creator: LA HAYE,RJ; LUCE,TC; PETTY,CC; HUMPHREYS,DA; HYATT,AW; PERKINS,FW et al.
Partner: UNT Libraries Government Documents Department

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

Description: 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{sub 95} > 4). A figure of merit for fusion gain {Beta}{sub N}H{sub 89}/q{sub 95}{sup 2} has been maintained at values corresponding to Q = 10 operation in a burning plasma for > 6 s or 36 {tau}{sub E} and 2 {tau}{sub R}. The key element is the relaxation of the current profile to a stationary state with q{sub min} > 1, which allows stable operation up to the no-wall ideal {beta} 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.
Date: November 1, 2002
Creator: LUCE,TC; WADE,MR; FERRON,JR; HYATT,AW; KELLMAN,AG; KINSEY,JE et al.
Partner: UNT Libraries Government Documents Department

CHANGES IN PARTICLE PUMPING DUE TO VARIATION IN MAGNETIC BALANCE NEAR DOUBLE-NULL IN DIII-D

Description: OAK-B135 The authors report on a recent experiment examining how changes in the divertor magnetic balance affect the rate that particles can be pumped at the divertor targets. They find that both the edge density of the core plasma and divertor recycling play important roles in properly interpreting this pumping result. Previous studies on DIII-D have identified several important differences between double-null (DN) and single-null (SN) divertor operation. Small variations in the magnetic balance near-DN have large effects on both the power- and particle loadings at the divertor targets. These most likely result from an interplay between the plasma geometry and ion particle drifts, e.g., ''B x {del}B'' and ''E x B'' drifts. Other studies have shown that changes in magnetic balance affect the core plasma and where ELMs strike the vessel. In this paper, they examine how variations in the magnetic balance impact the rate at which particles are removed from the core plasma via pumping.
Date: July 1, 2003
Creator: PETRIE,TW; WATKINS,JG; ALLEN,SL; BROOKS,NH; FENSTERMACHER,ME; FERRON,JR et al.
Partner: UNT Libraries Government Documents Department

TRANSFORMERLESS OPERATION OF DIII-D WITH HIGH BOOTSTRAP FRACTION

Description: OAK-B135 The authors have initiated an experimental program to address some of the questions associated with operation of a tokamak with high bootstrap current fraction under high performance conditions, without assistance from a transformer. In these discharges they have maintained stationary (or slowly improving) conditions for > 2.2 s at {beta}{sub N} {approx} {beta}{sub p} {approx} 2.8. Significant current overdrive, with dI/dt > 50 kA/s and zero or negative voltage, is sustained for over 0.7 s. The overdrive condition is usually ended with the appearance of MHD activity, which alters the profiles and reduces the bootstrap current. Characteristically these plasmas have 65%-80% bootstrap current, 25%-30% NBCD, and 5%-10% ECCD. Fully noninductive operation is essential for steady-state tokamaks. For efficient operation, the bootstrap current fraction must be close to 100%, allowing for a small additional ({approx} 10%) external current drive capability to be used for control. In such plasmas the current and pressure profiles are rightly coupled because J(r) is entirely determined by p(r) (or more accurately by the kinetic profiles). The pressure gradient in turn is determined by transport coefficients which depend on the poloidal field profile.
Date: July 1, 2003
Creator: POLITZER,PA; HYATT,AW; LUCE,TC; MAHDAVI,MA; MURAKAMI,M; PERKINS,FW et al.
Partner: UNT Libraries Government Documents Department

Study of Aspect Ratio Effects on Kinetic MHD Instabilities in NSTX and DIII-D

Description: We report general observations of kinetic instabilities on the low aspect-ratio National Spherical Torus Experiment (NSTX) and describe explicit aspect ratio scaling studies of kinetic instabilities using both the NSTX and the DIII-D tokamak. The NSTX and the DIII-D tokamak are nearly ideal for such experiments, having a factor of two difference in major radius but otherwise similar parameters. We also introduce new theoretical work on the physics of kinetic ballooning modes (KBM), toroidal Alfven eigenmodes (TAE), and compressional Alfven eigenmodes (CAE) with applications to NSTX.
Date: October 21, 2004
Creator: Fredrickson, E.D.; Heidbrink, W.W.; Cheng, C.Z.; Gorelenkov, N.N.; Belova, E.; Hyatt, A.W. et al.
Partner: UNT Libraries Government Documents Department

ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

Description: A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response.
Date: October 1, 2002
Creator: HUMPHREYS,DA; FERRON,JR; GAROFALO,AM; HYATT,AW; JERNIGAN,TC; JOHNSON,RD et al.
Partner: UNT Libraries Government Documents Department

PHYSICS PROCESSES IN DISRUPTION MITIGATION USING MASSIVE NOBLE GAS INJECTION

Description: Methods for detecting imminent disruptions and mitigating disruption effects using massive injection of noble gases (He, Ne, or Ar) have been demonstrated on the DIII-D tokamak [1]. A jet of high injected gas density (> 10{sup 24} m{sup -3}) and pressure (> 20 kPa) penetrates the target plasma at the gas sound speed ({approx}300-500 m/s) and increases the atom/ion content of the plasma by a factor of > 50 in several milliseconds. UV line radiation from the impurity species distributes the plasma energy uniformly on the first wall, reducing the thermal load to the divertor by a factor of 10. Runaway electrons are almost completely eliminated by the large density of free and bound electrons supplied by the gas injection. The small vertical plasma displacement before current quench and high ratio of current decay rate to vertical growth rate result in a 75% reduction in peak halo current amplitude and attendant forces.
Date: July 1, 2002
Creator: HUMPHREYS, D.A.; WHYTE, D.G.; JERNIGAN, T.C.; T.E.EVANS; GRAY, D.S.; HOLLMANN, E.M. et al.
Partner: UNT Libraries Government Documents Department

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

Description: Changes in the divertor magnetic balance in DIII-D H-mode plasmas affects core, edge, and divertor plasma behavior. Both the pedestal density n{sub e,PED} and plasma stored energy W{sub 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.
Date: June 1, 2002
Creator: PETRIE, T.W.; WATKINS, J.G.; BAYLOR, L.R.; BROOKS, N.H.; FENSTERMACHER, M.E.; HYATT, A.W. et al.
Partner: UNT Libraries Government Documents Department

A Comparison of Plasma Performance Between Single-Null and Double-Null Configurations During Elming H-Mode

Description: Tokamak plasma performance generally improves with increased shaping of the plasma cross section, such as higher elongation and higher triangularity. The stronger shaping, especially higher triangularity, leads to changes in the magnetic topology of the divertor. Because there are engineering and divertor physics issues associated with changes in the details of the divertor flux geometry, especially as the configuration transitions from a single-null (SN) divertor to a marginally balanced double-null (DN) divertor, we have undertaken a systematic evaluation of the plasma characteristics as the magnetic geometry is varied, particularly with respect to (1) energy confinement, (2) the response of the plasma to deuterium gas fueling, (3) the operational density range for the ELMing H-mode, and (4) heat flux sharing by the diverters. To quantify the degree of divertor imbalance (or equivalently, to what degree the shape is double-null or single-null), we define a parameter DRSEP. DRSEP is taken as the radial distance between the upper divertor separatrix and the lower divertor separatrix, as determined at the outboard midplane. For example, if DRSEP=O, the configuration is a magnetically balanced DN; if DRSEP = +1.0 cm, the divertor configuration is biased toward the upper divertor. Three examples are shown in Fig. 1. In the following discussions, VB drift is directed toward the lower divertor.
Date: July 1, 1999
Creator: Petrie, T.W.; Fenstermacher, M.E.; Allen, S.L.; Carlstrom, T.N.; Gohil, P.; Groebner, R.J. et al.
Partner: UNT Libraries Government Documents Department