44 research outputs found
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Electron heat transport in improved confinement discharges in DIII-D
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
Exploration of the equilibrium operating space for NSTX-Upgrade
This paper explores a range of high-performance equilibrium scenarios available in the NSTX-Upgrade device [J.E. Menard, submitted for publication to Nuclear Fusion]. NSTX-Upgrade is a substantial upgrade to the existing NSTX device [M. Ono, et al., Nuclear Fusion 40, 557 (2000)], with significantly higher toroidal field and solenoid capabilities, and three additional neutral beam sources with significantly larger current drive efficiency. Equilibria are computed with freeboundary TRANSP, allowing a self consistent calculation of the non-inductive current drive sources, the plasma equilibrium, and poloidal field coil current, using the realistic device geometry. The thermal profiles are taken from a variety of existing NSTX discharges, and different assumptions for the thermal confinement scalings are utilized. The no-wall and idealwall n=1 stability limits are computed with the DCON code. The central and minimum safety factors are quite sensitive to many parameters: they generally increases with large outer plasmawall gaps and higher density, but can have either trend with the confinement enhancement factor. In scenarios with strong central beam current drive, the inclusion of non-classical fast ion diffusion raises qmin, decreases the pressure peaking, and generally improves the global stability, at the expense of a reduction in the non-inductive current drive fraction; cases with less beam current drive are largely insensitive to additional fast ion diffusion. The non-inductive current level is quite sensitive to the underlying confinement and profile assumptions. For instance, for BT=1.0 T and Pinj=12.6 MW, the non-inductive current level varies from 875 kA with ITER-98y,2 thermal confinement scaling and narrow thermal profiles to 1325 kA for an ST specific scaling expression and broad profiles. This sensitivity should facilitate the determination of the correct scaling of transport with current and field to use for future fully non-inductive ST devices. Scenarios are presented which can be sustained for 8-10 seconds, or (20-30)τCR, at βN=3.8-4.5, facilitating, for instance, the study of disruption avoidance for very long pulse. Scenarios have been documented which can operate with βT~25% and equilibrated qmin>1. The value of qmin can be controlled at either fixed non-inductive fraction of 100% or fixed plasma current, by varying which beam sources are used, opening the possibility for feedback qmin control. In terms of quantities like collisionality, neutron emission, non-inductive fraction, or stored energy, these scenarios represent a significant performance extension compared to NSTX and other present spherical torii
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RI-mode investigations in the DIII-D tokamak with neon and argon induced radiating mantles
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
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Designing a VH-mode core/L-mode edge discharge
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
Benchmark of quasi-linear models against gyrokinetic single scale simulations in deuterium and tritium plasmas for a JET high beta hybrid discharge
A benchmark of the reduced quasi-linear models QuaLiKiz and TGLF with GENE gyrokinetic simulations has been performed for parameters corresponding to a JET high performance hybrid pulse in deuterium. Given the importance of the study of such advanced scenarios in view of ITER and DEMO operations, the dependence of the transport on the ion isotope mass has also been assessed, by repeating the benchmark changing the ion isotope to tritium. TGLF agrees better with GENE on the linear spectra and the flux levels. However, concerning the isotope dependence, only QuaLiKiz reproduces the GENE radial trend of a basically gyro-Bohm (gB) scaling at inner radii and instead anti-gB at outer radii. The physics effects which are responsible of the antigB effect in GENE simulations have been singled out