56 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
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Enhanced confinement discharges in DIII-D with neon and argon induced radiation
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
ELM triggering conditions for the integrated modeling of H-mode plasmas
Recent advances in the integrated modeling of ELMy H-mode plasmas are
presented. A model for the H-mode pedestal and for the triggering of ELMs
predicts the height, width, and shape of the H-mode pedestal and the frequency
and width of ELMs. Formation of the pedestal and the L-H transition is the
direct result of ExB flow shear suppression of anomalous transport. The
periodic ELM crashes are triggered by either the ballooning or peeling MHD
instabilities. The BALOO, DCON, and ELITE ideal MHD stability codes are used to
derive a new parametric expression for the peeling-ballooning threshold. The
new dependence for the peeling-ballooning threshold is implemented in the ASTRA
transport code. Results of integrated modeling of DIII-D like discharges are
presented and compared with experimental observations. The results from the
ideal MHD stability codes are compared with results from the resistive MHD
stability code NIMROD.Comment: 12th International Congress on Plasma Physics, 25-29 October 2004,
Nice (France
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
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