Advances in the simulation of toroidal gyro-Landau fluid model turbulence

Waltz, R. E.; Kerbel, G. D.; Milovich, J. L.; Hammett, G. W.

doi:10.1063/1.871264

Cited by 200 publications

(173 citation statements)

References 10 publications

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“…It is adjusted to be narrow during ramp up and broadened for H-mode conditions at high auxiliary power input. The temperature profiles are determined from the assumed transport model, typically used is GLF23 [9] without rotation over most of the profile and the edge (where GLF23 is known not to be valid) by a scaled L-mode [10] model. The overall minimum thermal conductivity is set to neoclassical values.…”

Section: Scenario Control Evaluation (Sce) Modementioning

confidence: 99%

ITER shape controller and transport simulations

Casper

Meyer

Pearlstein

et al. 2008

Fusion Engineering and Design

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We currently use the CORSICA integrated modeling code for scenario studies for both the DIII-D and ITER experiments. In these simulations, free-or fixed-boundary equilibria are simultaneously converged with thermal evolution determined from transport models providing temperature and current density profiles. Using a combination of fixed boundary evolution followed by free-boundary calculation to determine the separatrix and coil currents. In the free-boundary calculation, we use the state-space controller representation with transport simulations to provide feedback modeling of shape, vertical stability and profile control. In addition to a tightly coupled calculation with simulator and controller imbedded inside CORSICA, we also use a remote procedure call interface to couple the CORSICA nonlinear plasma simulations to the controller environments developed within the Mathworks Matlab/Simulink environment. We present transport simulations using full shape and vertical stability control with evolution of the temperature profiles to provide simulations of the ITER controller and plasma response.

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Section: Scenario Control Evaluation (Sce) Modementioning

confidence: 99%

ITER shape controller and transport simulations

Casper

Meyer

Pearlstein

et al. 2008

Fusion Engineering and Design

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“…The uncertainties in the fitting coefficients (listed in the caption of figure 13) are large because the small amount of data and the size of the measurement uncertainties limits the capability to separate the scaling with the pressure profile shape from the scaling with q core and q 95 . In addition, discharge-to-discharge differences in the details of the pressure profile shape are not included in equation (4). An F-test [36] showed that each of the fitting coefficients improves the fit by a statistically significant amount.…”

Section: Scaling Of F Bsmentioning

confidence: 99%

“…Conversely, the q profile is strongly dependent on J BS when f BS is large, and in a steady-state tokamak, f BS must be large in order to minimize the required external current drive power [3]. The temperature and density gradients depend on the transport coefficients, which also depend on q as well as the magnetic shear [4,5].…”

Section: Introductionmentioning

confidence: 99%

Optimization of the safety factor profile for high noninductive current fraction discharges in DIII-D

et al. 2011

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Abstract. In order to assess the optimum q profile for discharges in DIII-D with 100% of the current driven noninductively (f NI = 1), the self-consistent response of the plasma profiles to changes in the q profile was studied in high f NI , high N discharges through a scan of q min and q 95 at two values of N . As expected, both the bootstrap current fraction, f BS , and f NI increased with q 95 . The temperature and density profiles were found to broaden as either q min or N is increased.A consequence is that f BS does not continue to increase at the highest values of q min . A scaling function that depends on q min , q 95 , and the peaking factor for the thermal pressure was found to represent well the f BS / N inferred from the experimental profiles. The changes in the shapes of the 2 density and temperature profiles as N is increased modify the bootstrap current density (J BS ) profile from peaked close to the axis to relatively flat in the region between the axis and the H-mode pedestal. Therefore, significant externally-driven current density in the region inside the H-mode pedestal is required in addition to J BS in order to match the profiles of the noninductive current density (J NI ) to the desired total current density (J). In this experiment, the additional current density was provided mostly by neutral beam current drive with the neutral-beam-driven current fraction 40%-90% of f BS . The profiles of J NI and J were most similar at q min 1.35-1.65, q 95 6.8, where f BS is also maximum, establishing this q profile as the optimal choice for f NI = 1 operation in DIII-D with the existing set of external current drive sources.3

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“…37 Hence, important open qualitative questions remain, even before the quantitative ones can be addressed about the detailed dependence of both e and ˙e on s and q: Does the increased negative central shear affect the barrier at s = 0 indirectly through a modification of its gradient sЈ there, or does it act to reduce the local energy diffusivity? [38][39][40] In the latter case, does the dynamical evolution ͑˙e͒ depend on s, resulting in a delayed but ultimately efficient diffusivity quench in the center? If the barrier instead does remain localized at s = 0, as advocated by the so-called radial-gap or zero-shear-gap theories, 41 what parameters govern its width?…”

Section: -5mentioning

confidence: 99%

High-bootstrap, noninductively sustained electron internal transport barriers in the Tokamak à Configuration Variable

Coda¹,

Goodman²,

Henderson³

et al. 2005

Physics of Plasmas

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Important ingredients of the advanced-tokamak route to fusion have been explored in depth in the Tokamak à Configuration Variable ͓F. Hofmann, J. B. Lister, M. Anton et al., Plasma Phys. Controlled Fusion 36, B277 ͑1994͔͒ over the past two years. Using a uniquely powerful and flexible electron-cyclotron resonance heating ͑ECRH͒ system as the primary actuator, fully noninductive, steady-state electron internal transport barrier discharges have been generated with an electron-energy confinement time up to five times longer than in L mode, poloidal ␤ up to 2.4, and bootstrap fraction up to 75%. Interpretative transport modeling confirms that the safety factor profile is nonmonotonic in these discharges. The formation of the barrier is a discrete event resulting in rapid and localized confinement improvement consistent with the time and location of magnetic-shear reversal. In steady state, however, the confinement quality appears to depend on the current gradient in a broader negative-shear region enclosed by the barrier, improving with increasing shear: in particular, the width and depth of the barrier can be controlled and finely tuned, along a magnetohydrodynamic-stable path, by manipulating the current profile with ECRH ͑six independently steerable 0.45 MW launchers͒. The crucial role of the current profile has been clearly demonstrated by applying small Ohmic current perturbations which dramatically alter the properties of the barrier, enhancing or reducing the confinement with negative and positive current, respectively, with negligible Ohmic heating. These results are in agreement with theoretical estimates: first-principle-based numerical simulations of microinstability dynamics and turbulence-driven transport predict a substantial suppression of turbulence and anomalous energy diffusivity near the location of the minimum in the safety factor.

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Advances in the simulation of toroidal gyro-Landau fluid model turbulence

Cited by 200 publications

References 10 publications

ITER shape controller and transport simulations

ITER shape controller and transport simulations

Optimization of the safety factor profile for high noninductive current fraction discharges in DIII-D

High-bootstrap, noninductively sustained electron internal transport barriers in the Tokamak à Configuration Variable

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