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Doyle, E J

Publications and source records attributed to Doyle, E J.

12 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Electron profile stiffness and critical gradient studies

Electron profile stiffness was studied in DIII-D L-mode discharges by systematically varying the heat flux in a narrow region with electron cyclotron heating and measuring the local change produced in {nabla}T{sub e}. Electron stiffness was found to slowly increase with toroidal rotation velocity. A critical inverse temperature gradient scale length 1/L{sub C} {approx} 3 m{sup -1} was identified at {rho}=0.6 and found to be independent of rotation. Both the heat pulse diffusivity and the power balance diffusivity, the latter determined by integrating the measured dependence of the heat pulse diffusivity on -{nabla}T{sub e}, were fit reasonably well by a model containing a critical inverse temperature gradient scale length and varying linearly with 1/L{sub T} above the threshold.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2012

Changes in particle transport as a result of resonant magnetic perturbations in DIII-D

In this paper, we introduce the first direct perturbed particle transport measurements in resonant magnetic perturbation (RMP) H-mode plasmas. The perturbed particle transport increases as a result of application of RMP deep into the core. In the core, a large reduction in E Multiplication-Sign B shear to a value below the linear growth rate, in conjunction with increasing density fluctuations, is consistent with an increase in turbulent particle transport. In the edge, the changes in turbulent particle transport are less obvious. There is a clear correlation between the linear growth rates and the density fluctuations measured at different scales, but it is uncertain which is the cause and which is the consequence.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Contributions of Electron Cyclotron Waves to Performance in Advanced Regimes on DIII-D

High-power electron cyclotron (EC) waves are used to increase performance in several Advanced Tokamak (AT) regimes on DIII-D where there is a simultaneous need for high noninductive current and high beta. In the Quiescent High-confinement mode (QH-mode), a direct measurement of the electron cyclotron current drive (ECCD) profile is made using modulation techniques, and a trapped electron mode (TEM) dominated regime with core T{sub e}>T{sub i} is created. In the 'highq{sub min}' AT scenario, ECCD provides part of the off-axis noninductive current and helps to produce a tearing stable equilibrium. In the hybrid regime, strong central current drive from EC waves and other sources increases the noninductive current fraction to {approx_equal}100%. Surprisingly, the core safety factor remains above unity, meaning good alignment between the current drive profile and the desired plasma current profile is not necessary in this scenario.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Multi-field/multi-scale turbulence response to electron cyclotron heating of DIII-D ohmic plasmas

The multi-field/multi-scale core ({rho}{approx} 0.5-0.8) turbulence response to electron cyclotron heating (ECH) of DIII-D Ohmic plasmas is reported for the first time. Long wavelength (low-k) electron temperature (T-tilde{sub e}/T{sub e}) and high-k density turbulence levels (n-tilde{sub e}/n{sub e}) are observed to strongly increase during ECH. In contrast, low-k and intermediate-k n-tilde{sub e}/n{sub e} showed little change, whereas the cross-phase between local low-k electron temperature and density fluctuations ({alpha}{sub n{sub eT{sub e}}}) was significantly modified. The increase in the electron thermal diffusivity determined from power balance is consistent with the increased turbulent transport correlated with the measured increases in low-k T-tilde{sub e}/T{sub e} and high-k n-tilde{sub e}/n{sub e}. Linear stability analysis using the trapped gyro-Landau fluid (TGLF) model indicates an enhanced growth rate for electron modes [e.g., trapped electron mode (TEM)] at low-k consistent with the observed modifications in T-tilde{sub e}/T{sub e} and {alpha}{sub n{sub eT{sub e}}}. TGLF also predicts an increase in high-k electron mode growth rates for normalized wavenumbers k{sub {theta}}{rho}{sub s} > 7, where electron temperature gradient (ETG) modes exist, which is consistent with the observed increase in high-kn-tilde{sub e}/n{sub e} turbulence.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Advances in validating gyrokinetic turbulence models against L- and H-mode plasmas

Robust validation of predictive turbulent transport models requires quantitative comparisons to experimental measurements at multiple levels, over a range of physically relevant conditions. Toward this end, a series of carefully designed validation experiments has been performed on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to obtain comprehensive multifield, multipoint, multiwavenumber fluctuation measurements and their scalings with key dimensionless parameters. The results of two representative validation studies are presented: an elongation scaling study performed in beam heated L-mode discharges and an electron heating power scan performed in quiescent H-mode (QH-mode) discharges. A 50% increase in the elongation {kappa} is observed to lead to a {approx}50% increase in energy confinement time {tau}{sub e} and accompanying decrease in fluctuation levels, qualitatively consistent with a priori theoretical predictions and nonlinear GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] simulations. However, these simulations exhibit clear quantitative differences from experiment in the predicted magnitudes and trends with radius of turbulent fluxes and fluctuation levels which cannot be fully accounted for by uncertainties due to transport stiffness. In the QH-mode study, local nonlinear GYRO simulations that neglect fast ion effects show a similar proportional response to the applied electron cyclotron heating as the experiment, but overpredict the magnitudes of transport and fluctuation levels by a factor of 10 or more. Possible sources of this overprediction, namely nonlocal effects and self-consistent fast beam ions, are identified and discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2011

Integrated modelling of steady-state scenarios and heating and current drive mixes for ITER

Recent progress on ITER steady-state (SS) scenario modelling by the ITPA-IOS group is reviewed. Code-to-code benchmarks as the IOS group's common activities for the two SS scenarios (weak shear scenario and internal transport barrier scenario) are discussed in terms of transport, kinetic profiles, and heating and current drive (CD) sources using various transport codes. Weak magnetic shear scenarios integrate the plasma core and edge by combining a theory-based transport model (GLF23) with scaled experimental boundary profiles. The edge profiles (at normalized radius rho = 0.8-1.0) are adopted from an edge-localized mode-averaged analysis of a DIII-D ITER demonstration discharge. A fully noninductive SS scenario is achieved with fusion gain Q = 4.3, noninductive fraction f(NI) = 100%, bootstrap current fraction f(BS) = 63% and normalized beta beta(N) = 2.7 at plasma current I(p) = 8MA and toroidal field B(T) = 5.3 T using ITER day-1 heating and CD capability. Substantial uncertainties come from outside the radius of setting the boundary conditions (rho = 0.8). The present simulation assumed that beta(N)(rho) at the top of the pedestal (rho = 0.91) is about 25% above the peeling-ballooning threshold. ITER will have a challenge to achieve the boundary, considering different operating conditions (T(e)/T(i) approximate to 1 and density peaking). Overall, the experimentally scaled edge is an optimistic side of the prediction. A number of SS scenarios with different heating and CD mixes in a wide range of conditions were explored by exploiting the weak-shear steady-state solution procedure with the GLF23 transport model and the scaled experimental edge. The results are also presented in the operation space for DT neutron power versus stationary burn pulse duration with assumed poloidal flux availability at the beginning of stationary burn, indicating that the long pulse operation goal (3000s) at I(p) = 9 MA is possible. Source calculations in these simulations have been revised for electron cyclotron current drive including parallel momentum conservation effects and for neutral beam current drive with finite orbit and magnetic pitch effects.

Murakami, Masanori↗
DOE OSTI2011

ITER Test Blanket Module Error Field Simulation Experiments at DIII-D

Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Delta upsilon/upsilon similar to 60% via non-resonant braking. Changes to global Delta n/n, Delta beta/beta and Delta H(98)/H(98) were similar to 3 times smaller. These effects are stronger at higher beta. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L-and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.

Schaffer, M J↗
DOE OSTI2011

Balancing Current Drive and Heating in DIII-D High Noninductive Current Fraction Discharges Through Choice of the Toroidal Field

In order to maintain stationary values of the stored energy and the plasma current in a tokamak discharge with all of the current driven noninductively, the sum of the alpha-heating power and the power required to provide externally driven current must be equal to the power required to maintain the pressure against transport losses. In a study of high noninductive current fraction discharges in the DIII-D tokamak, it is shown that in the case of present-day tokamaks with no alpha-heating, adjustment of the toroidal field strength (B(T)) is a tool to obtain this balance between the required current drive and heating powers with other easily modifiable discharge parameters (beta(N), q(95), discharge shape, n(e)) fixed at values chosen to satisfy specific constraints. With all of the external power sources providing both heating and current drive, and beta(N) and q(95) fixed, the fraction of externally driven current scales with B(T) with little change in the bootstrap current fraction, thus allowing the noninductive current fraction to be adjusted.

Ferron, J R↗
DOE OSTI2011

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

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/{alpha}B were reproduced, with the size reduced by a factor of 3.7, while matching key performance targets for {beta}{sub N} and H{sub 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 ({nu}*{sub e} {approx} 0.1). Target values for {beta}{sub N} and H{sub 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{sub e} and T{sub 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 {approx} 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{sub NI}) on q{sub 95}, as found in the experimental I{sub p} scan, indicating that optimization of the q profile is critical to simultaneously achieving the f{sub NI} = 1 and Q = 5 goals. Extended integrated modeling is being developed to improve capability for ITER projection by including the experimental observations of density peaking, ELM characteristics, NTM suppression and coupled core-edge-SOL transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗