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Impurity feedback control for enhanced divertor and edge radiation in DIII-D discharges

Description: Long pulse and steady state fusion ignition devices will require a significant radiated power fraction to minimize heat flux to, and sputtering of, the first wall. While impurity gases have been proposed to enhance radiation, precise control of impurity gas injection is essential to achieve an adequate radiative power fraction while maintaining good energy confinement and low central impurity concentration. We report here the first experiments in the DIII-D tokamak using feedback control of the rate of impurity gas injection. These experiments were carried out with active divertor pumping using the in-situ DIII-D cryopump. The radiated power fraction was controlled by sensing either UN edge line radiation (Ne{sup +7}) or mantle radiation from selected bolometer channels and using the DIII-D digital plasma control system to calculate radiated power real-time and generate an error signal to control an impurity gas injector valve.
Date: October 1, 1996
Creator: Jackson, G.L.; Staebler, G.M. & Allen, S.L.
Partner: UNT Libraries Government Documents Department

Electron heat transport in improved confinement discharges in DIII-D

Description: In DIII-D tokamak plasmas with an internal transport barrier (ITB), the comparison of gyrokinetic linear stability (GKS) predictions with experiments in both low and strong negative magnetic shear plasmas provide improved understanding for electron thermal transport within the plasma. Within a limited region just inside the ITB, the electron temperature gradient (ETG) modes appear to control the electron temperature gradient and, consequently, the electron thermal transport. The increase in the electron temperature gradient with more strongly negative magnetic shear is consistent with the increase in the ETG mode marginal gradient. Closer to the magnetic axis the T{sub e} profile flattens and the ETG modes are predicted to be stable. With additional core electron heating, FIR scattering measurements near the axis show the presence of high k fluctuations (12 cm{sup {minus}1}), rotating in the electron diamagnetic drift direction. This turbulence could impact electron transport and possibly also ion transport. Thermal diffusivities for electrons, and to a lesser degree ions, increase. The ETG mode can exist at this wavenumber, but it is computed to be robustly stable near the axis. Consequently, in the plasmas the authors have examined, calculations of drift wave linear stability do not explain the observed transport near the axis in plasmas with or without additional electron heating, and there are probably other processes controlling transport in this region.
Date: January 1, 1999
Creator: Stallard, B.W.; Greenfield, C.M. & Staebler, G.M.
Partner: UNT Libraries Government Documents Department

Behavior of electron and ion transport in discharges with an internal transport barrier in the DIII-D tokamak

Description: The authors report results of experiments to further determine the underlying physics behind the formation and development of internal transport barriers (ITB) in the DIII-D tokamak. The initial ITB formation occurs when the neutral beam heating power exceeds a threshold value during the early stages of the current ramp in low-density discharges. This region of reduced transport, made accessible by suppression of long-wavelength turbulence by sheared flows, is most evident in the ion temperature and impurity rotation profiles. In some cases, reduced transport is also observed in the electron temperature and density profiles. If the power is near the threshold, the barrier remains stationary and enclosed only a small fraction of the plasma volume. If, however, the power is increased, the transport barrier expands to encompass a larger fraction of the plasma volume. The dynamic behavior of the transport barrier during the growth phase exhibits rapid transport events that are associated with both broadening of the profiles and reductions in turbulence and associated transport. In some, but not all, cases, these events are correlated with the safety factor q passing through integer values. The final state following this evolution is a plasma exhibiting ion thermal transport at or below neoclassical levels. Typically, the electron thermal transport remains anomalously high. Recent experimental results are reported in which rf electron heating was applied to plasmas with an ion ITB, thereby increasing both the electron and ion transport. Although the results are partially in agreement with the usual {rvec E} x {rvec B} shear suppression hypothesis, the results still leave questions that must be addressed in future experiments.
Date: December 1, 1998
Creator: Greenfield, C. M.; Staebler, G. M. & Rettig, C. L.
Partner: UNT Libraries Government Documents Department

Electron thermal transport in enhanced core confinement regimes

Description: The cause of the anomalous electron thermal transport in a region of suppressed ion thermal transport is investigated using a comprehensive gyrokinetic stability code. Analysis of a DIII-D negative central shear discharge with additional fastwave electron heating is presented. It is found that the electron heating excites the electron temperature gradient mode (ETG). The enhanced electron thermal transport from power balance analysis is consistent with the increased growth rate for the ETG mode. The ion thermal transport barrier is observed to retreat towards the plasma center during the fastwave heating (FW). Transport modeling with self-consistent E x B velocity shear reproduces this effect for on-axis electron heating. The same transport model predicts that off-axis electron heating can extend the region of reduced transport outward.
Date: July 1998
Creator: Staebler, G. M.; Waltz, R. E. & Greenfield, C. M.
Partner: UNT Libraries Government Documents Department

RI-mode investigations in the DIII-D tokamak with neon and argon induced radiating mantles

Description: The RI-mode regime, with high radiating power fractions from 0.5 to 0.9, energy confinement enhancements, H{sub 89P}, over ITER89-P L-mode scaling greater than 1.6, and operation at or above the Greenwald density limit (n{sub GW}) is an attractive operating scenario for future fusion burning plasma devices. The TEXTOR tokamak has demonstrated this scenario in a limiter device with steady state conditions, {Delta}t{sub RI-mode}/{tau}{sub E} > 100. Studies have been initiated on the DIII-D tokamak with the goals of: (a) extending these results to a larger non circular machine (providing size and shape scaling), (b) investigating the underlying physical mechanisms of RI-mode with a complementary diagnostic set to that on TEXTOR, and (c) using non-intrinsic impurities, e.g., neon and argon, to obtain high performance diverted discharges, ({beta}{sub N}H{sub 89P} > 6) in support of the DIII-D advanced tokamak (AT) program, where {beta}{sub N} = {beta}{sub T}/(I{sub p}/aB{sub T}) and {beta}{sub T}, I{sub p}, a, and B{sub T} are toroidal beta (in %), plasma current (MA), minor radius (m), and toroidal magnetic field (T) respectively. The authors define P{sub radLCFS} as the radiated power inside the LCFS and note that nearly all of this radiation occurs in the mantle region 0.6 < {rho} < 1.0, i.e., P{sub mantle} {approx} P{sub radLCFS}. Three types of DIII-D discharges where mantle radiation plays a significant role are discussed in this paper: (i) ELMing H-mode puff and pump, (ii) limiter L-mode, and (III) high performance.
Date: July 1, 1998
Creator: Jackson, G.L.; Staebler, G.M. & Murakami, M.
Partner: UNT Libraries Government Documents Department

Enhanced confinement discharges in DIII-D with neon and argon induced radiation

Description: Enhanced energy confinement in discharges with impurity induced radiating power fractions, P{sub rad}/P{sub in}, from 50--100% have been observed in the DIII-D tokamak, with neon and argon gas puffing. These radiating mantle enhanced confinement discharges have been obtained in the DIII-D tokamak under a variety of conditions: diverted and limited configurations with both an H-mode and L-mode edge. Confinement enhancements as high as the ELM free H-mode scaling relation have been obtained with impurity gas puffing, although operation at the highest densities is transient. Similarities and differences between these DIII-D discharges and RI-mode discharges obtained in the TEXTOR tokamak are discussed.
Date: August 1, 1998
Creator: Jackson, G.L.; Staebler, G.M. & Murakami, M.
Partner: UNT Libraries Government Documents Department

Core turbulence and transport reduction in DIII-D discharges with weak or negative magnetic shear

Description: Core turbulence fluctuation levels have been suppressed in DIII-D discharges with weak or negative magnetic shear (NCS) near the magnetic axis. In some weak magnetic shear discharges the ion thermal transport has been reduced to neoclassical levels throughout the whole plasma. The cause of the transport reduction is investigated by calculating the stability of toroidal drift waves, i.e., ion temperature gradient modes (ITG) and trapped electron modes (TE), with a comprehensive gyrokinetic linear stability code. It is found that the ITG modes and TE modes are stabilized by ExB velocity shear. The ExB velocity shear is primarily responsible for the spontaneous growth of a region of suppressed ion thermal transport. Surprisingly, the negative magnetic shear and Shafranov shift are only weak stabilizing influences for the ITG and TE modes in the DIII-D cases studied. Negative magnetic shear does eliminate the ideal magnetohydrodynamic ballooning mode instability which is a necessary access criteria for these improved core confinement regimes. Dilution of the thermal ions by fast ions from the heating beams and hot ions compared to electrons are found to be important stabilizing influences in the core.
Date: June 1, 1997
Creator: Staebler, G.M.; Greenfield, C.M. & Schissel, D.P.
Partner: UNT Libraries Government Documents Department

Designing a VH-mode core/L-mode edge discharge

Description: An operating mode with a very high confinement core like the VH-mode but a very low power flow to the divertor plates and low edge particle confinement like an L-mode would be beneficial. For a large tokamak like the proposed ITER, the power density at the separatrix is not that far above the scaled H-mode power threshold so not much of the power can be radiated inside of the separatrix without causing a return to L-mode. The thicker scrape-off layer of an L-mode increases the radiating volume of the scrape-off layer and helps shield impurities from the core. This is especially important if the first wall is metallic. In this paper an H-mode transport model based on E x B velocity shear suppression of turbulence will be used to show that it is possible to have a strongly radiating mantle near the separatrix, which keeps the edge in L-mode, while having a VH-mode core with a broad region of suppressed turbulence. The existing results of enhanced L-mode confinement during impurity injection on a number of tokamaks will be surveyed. The operating conditions which will most likely result in the further improvement of the core confinement by control of the heating, fueling, and torque profiles will be identified.
Date: December 1, 1995
Creator: Staebler, G.M.; Hinton, F.L. & Wiley, J.C.
Partner: UNT Libraries Government Documents Department

TRANSPORT FROM OVERLAPPING ELECTRON AND ION DRIFTWAVE INSTABILITIES

Description: The electron temperature gradient (ETG) mode is a likely contributor to electron thermal transport in tokamaks. The ETG modes are dominantly unstable for poloidal wavelengths shorter than the ion gyroradius (high-k) where the ion response is adiabatic. Thus, they do not directly produce ion thermal or momentum transport or particle transport. Two potential mechanisms whereby ETG modes could produce transport in these channels are explored in this paper: a nonlinear coupling between high-k ETG modes and ions at low-k and a direct coupling when ETG modes and ion temperature gradient (ITG) modes are unstable in overlapping wavenumber ranges. It will be shown that the particle and momentum transport required to match experiment is small compared to the ETG driven electron thermal transport. Even quasilinearly ETG modes can produce ion transport if the ITG and ETG modes are both unstable at low-k. The implications of this for transport will be explored at the quasilinear level. A new gyro-Landau-fluid (GLF) closure model has been constructed in order to build a transport model which can include the coupling between electron and ion modes including trapped particles. The first growth rate spectra from this model will be shown to give an accurate approximation to the kinetic linear growth rates of drift-ballooning modes in tokamaks.
Date: July 2, 2004
Creator: STAEBLER,G.M; KINSEY,J.E & WALTZ,R.E
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

H-mode threshold power scaling and the {gradient}B drift effect

Description: One of the largest influences on the H-mode power threshold (P{sub TH}) is the direction of the ion {gradient}B drift relative to the X-point location, where factors of 2--3 increase in P{sub TH} are observed for the ion {gradient}B drift away from the X-point. It is proposed that the threshold power scaling observed in single-null configurations with the ion {gradient}B drift toward the X-point location (P{sub TH} {approximately} nB, where n is the plasma density, and B is the toroidal field) is due to the scaling of the magnitude of the {gradient}B drift effect. Hinton and later Hinton and Stebler have modeled this effect as neoclassical cross field fluxes of both heat and particles driven by poloidal temperature gradients on the open field lines in the scrape-off layer (SOL). The {gradient}B drift effect influences the power threshold by affecting the edge conditions needed for the L-H transition. It is not essential for the L-H transition itself since transitions are observed with either direction of B. Predictions of this model include saturation of the B scaling of P{sub TH} at high field, 1/B scaling of P{sub TH} with reverse B, and no B scaling of P{sub TH} in balanced double-null configurations. This last prediction is consistent with the observed scaling of p{sub TH} in double-null plasma sin DIII-D.
Date: June 1, 1997
Creator: Carlstrom, T.N.; Burrell, K.H.; Groebner, R.J. & Staebler, G.M.
Partner: UNT Libraries Government Documents Department

Effects of particle fueling and plasma wall interactions on DIII-D discharges

Description: DIII-D has successfully operated with an all-graphite first wall, including the first observations of VH-mode without boronization. A major goal of this, and other recent upgrades, was to control impurity influxes and hydrogenic fueling. Graphite tiles were carefully pre-conditioned, first by ex situ preparation and then by baking and helium glow conditioning. No deuterium or hydrogen was used until tokamak operation commenced. With the all graphite wall, both impurity and deuterium influxes during tokamak discharges were lower than previous boronized discharges; central nickel impurity line radiation, NiXXV and NiXXVI, was an order of magnitude lower than previous discharges during the ELM free beam heated phase. The effect of reduced particle fueling on plasma performance, particularly H- and VH-mode discharges, will be presented.
Date: November 1, 1994
Creator: Jackson, G.L.; Baker, D.R.; Holtrop, K.L.; Staebler, G.M.; West, W.P.; Maingi, R. et al.
Partner: UNT Libraries Government Documents Department

Stability of a radiative mantle in ITER

Description: We report results of a study to evaluate the efficacy of various impurities for heat dispersal by a radiative mantle and radiative divertor(including SOL). We have derived a stability criterion for the mantle radiation which favors low Z impurities and low ratios of edge to core thermal conductivities. Since on the other hand the relative strength of boundary line radiation to core bremsstrahlung favors high Z impurities, we find that for the ITER physics phase argon is the best gaseous impurity for mantle radiation. For the engineering phase of ITER, more detailed analysis is needed to select between krypton and argon.
Date: December 1, 1996
Creator: Mahdavi, M.A.; Staebler, G.M.; Wood, R.D.; Whyte, D.G. & West, W.P.
Partner: UNT Libraries Government Documents Department

DIII-D Advanced Tokamak Research Overview

Description: This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously {beta}{sub N}H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues.
Date: December 1, 1999
Creator: Chan, V.S.; Greenfield, C.M.; Lao, L.L.; Luce, T.C.; Petty, C.C. & Staebler, G.M.
Partner: UNT Libraries Government Documents Department

ARIES-ST STUDIES REPORT FOR THE PERIOD JANUARY 1, 1998 THROUGH DECEMBER 31, 1998

Description: During 1998, the General Atomics (GA) ARIES-Spherical Torus (ST) team examined several critical issues related to the physics performance of the ARIES-ST design, and a number of suggestions were made concerning possible improvements in performance. These included specification of a reference plasma equilibrium, optimization about the reference equilibrium to achieve higher beta limits, examination of three possible schemes for plasma initiation, development of a detailed scenario for ramp-up of the plasma current and pressure to its full, final operating values, an assessment of the requirement for electron confinement, and several suggestions for divertor heat flux reduction. The reference equilibrium was generated using the TOQ code, with the specification of a 100%, self-consistent bootstrap current. The equilibrium has {beta} = 51%, 10% below the stability limit (a margin specified by the ARIES-ST study). In addition, a series of intermediate equilibria were defined, corresponding to the ramp-up scenario discussed. A study of the influence of shaping on ARIES-ST performance indicates that significant improvement in both kink and ballooning stability can be obtained by modest changes in the squareness of the plasma. In test equilibria the ballooning beta limit is increased from 58% to 67%. Also the maximum allowable plasma-wall separation for kink stability can be increased by 30%. Three schemes were examined for noninductive plasma initiation. These are helicity injection (HICD), electron cyclotron heating (ECH)-assisted startup, and inductive startup using only the external equilibrium coils. HICD startup experiments have been done on the HIT and CDX devices. ECH-assisted startup has been demonstrated on CDX-U and DIII-D. External coil initiation is based on calculations for a proposed DIII-D experiment. In all cases, plasma initiation and preparation of an approximately 0.3 MA plasma for ARIES-ST appears entirely feasible.
Date: April 1, 1999
Creator: CHAN, V.S.; LAO, L.L.; LIN-LIU, Y.R.; MILLER, R.L.; PETRIE, T.W.; POLITZER, P.A. et al.
Partner: UNT Libraries Government Documents Department

TRANSPORT STUDIES IN DIII-D WITH MODULATED ECH

Description: Experiments have been performed where the T{sub e} profile stiffness was tested at several spatial locations by varying the ECH resonance location. Propagation of the pulses was Fourier analyzed and compared to simulations based on several transport models. The plasma appears to be near the critical T{sub e} gradient for ETG modes and marginally stable to ITG modes. However, the local T{sub e} response to a locally applied heat pulse does not indicate a nonlinear, critical gradient model where T{sub e} is clipped when trying to rise above a critical gradient. The response can be simply understood as the plasma integrating the ECH power, producing an increase in T{sub e} which equilibrates to a new local level with an exponential time constant representing the local confinement time.
Date: July 1, 2002
Creator: DeBOO, J.C.; AUSTIN, M.E.; BRAVENEC, R.V.; KINSEY, J.E; LOHR, J.; LUCE, T.C. et al.
Partner: UNT Libraries Government Documents Department

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

Description: Core transport barriers can be reliably formed in DIII-D by tailoring the evolution of the current density profile. This paper reports studies of the relative role of magnetic and ExB shear in creating core transport barriers in the DIII-D tokamak and considers the detailed dynamics of the barrier formation. The core barriers seen in DIII-D negative shear discharges form in a stepwise fashion during the initial current ramp. The reasons for the stepwise formation is not known; these steps do not correlate with integer values of q(O) or minimum q. The data from DIII-D is consistent with previous results that negative magnetic shear facilitates the formation of core transport barriers in the ion channel but is not necessary. However, strongly negative magnetic shear does allow formation of transport barriers in particle, electron thermal, ion thermal and angular momentum transport channels. Shots with strong negative magnetic shear have produced the steepest ion temperature and toroidal rotation profiles seen yet in DIII-D. In addition, the ExB shearing rates seen in these shots exceed the previous DIII-D record value by a factor of four.
Date: December 31, 1997
Creator: Burrell, K.H.; Greenfield, C.M.; Lao, L.L.; Staebler, G.M.; Austin, M.E.; Rice, B.W. et al.
Partner: UNT Libraries Government Documents Department

MECHANISMS FOR ELECTRON TRANSPORT BARRIER FORMATION IN THE DIII-D TOKAMAK

Description: The E x B shear stabilization paradigm explains much of the phenomenology of ion thermal transport in tokamaks. Behavior in the electron channel, however, has continued to challenge our understanding. Recent experiments in DIII-D and elsewhere produce regions where electron thermal transport is almost completely eliminated with intense, localized, direct electron heating. Simulations of DIII-D discharges identify {alpha}-stabilization, local magnetic shear stabilization due to the Shafranov shift, as the dominant turbulence reduction mechanism in these experiments and may point the way toward regimes with simultaneous electron and ion internal transport barriers.
Date: February 1, 2001
Creator: GREENFIELD, C.M.; PRATER, R.; STAEBLER, G.M.; KINSEY, J.E.; BURRELL, K.H.; BOO, J.C. De et al.
Partner: UNT Libraries Government Documents Department

OVERVIEW OF H-MODE PEDESTAL RESEARCH ON DIII-D

Description: Developing an understanding of the processes that control the H-mode transport barrier is motivated by the significant impact this small region (typically &lt;2% of the minor radius) can have on overall plasma performance. Conditions at the inner edge of the H-mode transport barrier can strongly influence the overall energy confinement, and the maximum density, and therefore fusion power, that can be achieved with the typically flat H-mode density profiles [1,2]. The ELM instability, which usually regulates the pressure gradient in the H-mode edge, can result in large power loads to, and erosion of, the divertor targets in a reactor scale device [3]. The goal of H-mode pedestal research at DIII-D is to: (1) develop a physics based model that would allow prediction of the conditions at the top of the H-mode pedestal, (2) develop an understanding of processes which control Type I ELM effects in the core and divertor, and (3) explore alternatives to the Type I ELM regime.
Date: July 1, 2001
Creator: OSBORNE, T.H.; BURRELL, K.H.; CARLSTROM, T.N.; CHU, M.S.; DOYLE, E.J.; FERRON, J.R. et al.
Partner: UNT Libraries Government Documents Department

Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

Description: The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads to a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.
Date: October 7, 2003
Creator: Jardin, S.C.; Kessel, C.E.; Mau, T.K.; Miller, R.L.; Najmabadi, F.; Chan, V.S. et al.
Partner: UNT Libraries Government Documents Department

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

Description: Shear in the E x B flow velocity can stabilize turbulent transport [1], and so it is of interest to understand the physics behind electric field generation and modification in the tokamak. In DIII-D the core radial electric field in many regimes is generated by flow velocities driven by momentum input from neutral beam injection (NBI). In a variety of conditions it is observed that direct electron heating is accompanied by a reduction in the NBI driven toroidal rotation velocity, U{sub {phi}}, and the ion temperature, T{sub i}, primarily in the core, {rho} &lt;0.5 (where {rho} is a radial coordinate of the normalized toroidal flux). This electron heating can be done with either electron cyclotron heating (ECH) or fast wave electron heating (FWEH). Both can be accompanied by the reduction in U{sub {phi}} and T{sub i} [2-4]. Details of the parallel wavenumber (k//) spectrum of the launched rf do not seem to be important in either case for the effect to exist. Reductions are observed for EC waves launched with nonzero k// for current drive or launched radially with k//=0; and for FWEH with waves directed either co or counter, using the DIII-D four strap antennas [5], This universality indicates that increased electron temperature, T{sub e}, is increasing ion momentum and thermal transport, at least in the parameter regimes of these experiments. It is also possible that nonambipolar transport of resonantly heated particles is playing a role. To date, the great majority of the DIII-D experiments have been conducted with the rf target discharges driven by co-injected NBI.
Date: July 1, 1999
Creator: deGrassie, J.S.; Greenfield, C.M.; Baker, D.R.; Burrell, K.H.; Lin-Liu, Y.R.; Lohr, J. et al.
Partner: UNT Libraries Government Documents Department

Status of Advanced Tokamak Scenario Modeling with Off-Axis Electron Cyclotron Current Drive in DIII-D

Description: The status of modeling work focused on developing the advanced tokamak scenarios in DIII-D is discussed. The objectives of the work are two-fold: (1) to develop AT scenarios with ECCD using time-dependent transport simulations, coupled with heating and current drive models, consistent with MHD equilibrium and stability; and (2) to use time-dependent simulations to help plan experiments and to understand the key physics involved. Time-dependent simulations based on transport coefficients derived from experimentally achieved target discharges are used to perform AT scenario modeling. The modeling indicates off-axis ECCD with approximately 3 MW absorbed power can maintain high-performance discharges with q{sub min} &gt; 1 for 5 to 10 s. The resultant equilibria are calculated to be stable to n = 1 pressure driven modes. The plasma is well into the second stability regime for high-n ballooning modes over a large part of the plasma volume. The role of continuous localized ECCD is studied for stabilizing m/n = 2/1 tearing modes. The progress towards validating current drive and transport models, consistent with experimental results, and developing self-consistent, integrated high performance AT scenarios is discussed.
Date: December 1, 1999
Creator: Murakami, M.; St.John, H.E.; Casper, T.A.; Chu, M.S.; DeBoo, J.C.; Greenfield, C.M. et al.
Partner: UNT Libraries Government Documents Department

Advanced Tokamak Scenario Modeling with Off-Axix ECH in DIII-D

Description: Time-dependent simulations with transport coefficients derived from experimentally achieved discharges are used to explore the capability of off-axis electron cyclotron current drive (ECCD) to control hollow current profiles in negative central shear discharges. Assuming these transport coefficients remain unchanged at higher EC power levels, the simulation results show that high confinement, high normalized beta and high bootstrap fraction can be achieved with EC power expected to be available in the near future in the DIII-D tokamak.
Date: July 1, 1999
Creator: Murakami, M.; Casper, T.A.; Lao, L.L.; St. John, H.E.; Deboo, J.C.; Greenfield, C.M. et al.
Partner: UNT Libraries Government Documents Department

RECENT EXPERIMENTAL STUDIES OF EDGE AND INTERNAL TRANSPORT BARRIERS IN THE DIII-D TOKAMAK

Description: Results from recent experiments on the DIII-D tokamak have revealed many important details on transport barriers at the plasma edge and in the plasma core. These experiments include: (a) the formation of the H-mode edge barrier directly by pellet injection; (b) the formation of a quiescent H-mode edge barrier (QH-mode) which is free from edge localized modes (ELMs), but which still exhibits good density and radiative power control; (c) the formation of multiple transport barriers, such as the quiescent double barrier (QDB) which combines a internal transport barrier with the quiescent H-mode edge barrier. Results from the pellet-induced H-mode experiments indicate that: (a) the edge temperature (electron or ion) is not a critical parameter for the formation of the H-mode barrier, (b) pellet injection leads to an increased gradient in the radial electric field, E{sub r}, at the plasma edge; (c) the experimentally determined edge parameters at barrier transition are well below the predictions of several theories on the formation of the H-mode barrier, (d) pellet injection can lower the threshold power required to form the H-mode barrier. The quiescent H-mode barrier exhibits good density control as the result of continuous magnetohydrodynamic (MHD) activity at the plasma edge called the edge harmonic oscillation (EHO). The EHO enhances the edge particle transport while maintaining a good energy transport barrier. The ability to produce multiple barriers in the QDB regime has led to long duration, high performance plasmas with {beta}{sub NH{sub 89}} values of 7 for up to 10 times the confinement time. Density profile control in the plasma core of QDB plasmas has been demonstrated using on-axis ECH.
Date: August 1, 2002
Creator: GOHIL, P.; BAYLOR, L.R.; BURRELL, K.H.; CASPER, T.A.; DOYLE, E.J.; GREENFIELD, C.M. et al.
Partner: UNT Libraries Government Documents Department