Bulletin of the American Physical Society
50th Annual Meeting of the Division of Plasma Physics
Volume 53, Number 14
Monday–Friday, November 17–21, 2008; Dallas, Texas
Session CI1: Magnetic Confinement I |
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Chair: Piero Martin, Consorzio RFX - Associazione EURATOM-ENEA sulla fusione, Padova (Italy) Room: Landmark A |
Monday, November 17, 2008 2:00PM - 2:30PM |
CI1.00001: Observation of an impurity hole in a plasma with an ion internal transport barrier in the Large Helical Device Invited Speaker: Simultaneous achievement of high ion temperature and low impurity concentration is crucial for the high fusion triple product, because impurities cause dilution of the fueling particle. In the Large Helical Device (LHD), the internal transport barrier (ITB) in the ion temperature transport appears after the onset of negative neutral beam injection (160-180keV beam energy) to the relatively low density hydrogen target plasma sustained by positive neutral beam injection (40keV beam energy), where the electron temperature is comparable to the ion temperature. Since the particle diffusion in the plasma with an ITB is relatively low because of the suppression of turbulence, the sign of the convection, which appears as an off-diagonal term in the transport matrix, becomes important in the ITB plasma. The radial profile of carbon becomes hollow during the ITB phase and can be extremely hollow (denoted as `impurity hole') and the central carbon density drops to 0.3{\%} of plasma density when the ion temperature gradient becomes large. The transport analysis gives a low diffusion of 0.1-0.2 m$^{2}$/s and the outward convection velocity of 2 m/s at the half of minor radius, which is in contrast to the tendency in tokamak plasmas for the impurity density to increase due to an inward convection and low diffusion in the ITB region. This experimental result in LHD contradicts to the neoclassical prediction where the negative electric field and an inward convection are predicted because the ion temperature is much larger than the electron temperature by a factor of two in this ITB plasma. The outward convection of the impurity transport in the ITB plasma is considered to be beneficial for future fusion relevant plasmas. [Preview Abstract] |
Monday, November 17, 2008 2:30PM - 3:00PM |
CI1.00002: Improvement of the magnetic configuration in the RFP through successive bifurcations Invited Speaker: An ohmic RFP must have a helical deformation, which according to MHD simulations may be of either of two types connected through a bifurcation: stationary with a single helicity (SH) or fluctuating in time with a multiple helicity. Experiments in RFX- mod, the largest RFP, show that the plasma reaches Quasi-SH (QSH) states that come closer to SH when the last closed flux surface is made more toroidally symmetric, and by increasing plasma current. The SH configuration corresponds to a helical equilibrium, like the stellarator, but it is ruled by magnetic self-organization. In QSH, magnetic chaos partially degrades ideal SH. QSH may be of either of two types connected through a saddle-node bifurcation: with or without an island. Resilience of QSH states to chaos increases when the magnetic island is suppressed. We report the experimental observation of separatrix suppression in a QSH state, with the improvement of the electron temperature profile and of core confinement. RFX-mod has been coming steadily closer to this state by improving the magnetic boundary through feedback control of MHD modes, and by increasing the current up to the present record value of 1.6 MA. The spontaneous occurrence of this state, dubbed Single Helical Axis (SHAx), is observed at plasma current $I >$ 1 MA if the ratio between the amplitude of the dominant and of the secondary modes exceeds a threshold. The thermal helical structure, that covers up to 30 $\%$ of the volume, reaches temperatures in 1 keV range. Steep $T_e$ gradients build in the core, with values of R/L$_{Te} \sim$20-30, comparable to those of tokamak electron transport barriers. SHAx states provide the best RFX-mod electron temperature and confinement. Simulations of test particle transport confirm the improvement. This result paves the way for further improvements, towards a RFP with confinement suitable for a reactor. [Preview Abstract] |
Monday, November 17, 2008 3:00PM - 3:30PM |
CI1.00003: Optimizing Stability, Transport, and Divertor Operation Through Plasma Shaping for Steady-state Scenario Development in DIII-D Invited Speaker: Recent studies on DIII-D have elucidated key aspects of the dependence of stability, confinement, and density control on the plasma magnetic configuration, leading to noninductive operation (i.e., total inductive flux change $\approx\,$0) for $>$1$\,$s with pressure 30\% above the free boundary limit. Achieving fully noninductive operation requires high $\beta$, good confinement, and density control through divertor operation. Plasma geometry affects all of these. Ideal MHD modeling of the $n=1,2,$ and 3 external kink stability suggests it may be optimized by adjusting the shape parameter known as squareness ($\zeta$). Experiments confirm stability varies strongly with $\zeta$ in qualitative agreement with the modeling. Optimization of $n=1$ stability also seems to raise pedestal stability. Adjusting $\zeta$ above and below the midplane independently allows for small changes in the magnetic divertor balance about a double-null (DN) configuration. Energy confinement is found to be sensitive to this balance, with 20\% higher confinement observed in a balanced DN compared to a slightly unbalanced case. However, adequate density control requires a small imbalance to achieve densities necessary for efficient external noninductive current drive. The best density control (20\%-30\% below the balanced DN case) is obtained with a slight imbalance toward the divertor opposite the ion grad(B) drift direction. Consistency of modeling with these observed effects requires inclusion of $E\times B$ drifts in the divertor. Simultaneous optimization has been applied to achieve noninductive current fractions near 1 for over 1$\,$s with $\beta_N \sim\,$3.5-3.7, bootstrap fraction $>$65\%, and good confinement. [Preview Abstract] |
Monday, November 17, 2008 3:30PM - 4:00PM |
CI1.00004: Improved confinement at high current in the MST RFP Invited Speaker: Inductive current profile control has proven to be a robust means of reducing global magnetic tearing fluctuations and improving both particle and energy confinement in MST and other reversed-field pinches. The improved confinement has been maintained as these studies have been extended to higher plasma current. The electron stored energy in these plasmas increases with current; electron temperature increases from 0.6 keV to 2 keV as current increases from 0.2 MA to 0.5 MA. This is the largest $T_{e}$ yet achieved in the ohmically-heated RFP and is achieved in addition to measurements of sustained 1 keV ion temperatures, indicating that ion confinement is also improved. The global energy confinement time in these plasmas is about 12 ms, a modest improvement over the 10 ms confinement time at low current. The corresponding global thermal diffusivity ($a^{2}$/4\textit{$\tau $}$_{E})$ is about 5 m$^{2}$/s. Measurement of the x-ray spectrum from 2 keV to 100 keV, combined with Fokker-Planck analysis, indicates that energetic electrons are well-confined in these high current plasmas, with a velocity-independent diffusion coefficient which is inconsistent with transport by magnetic fluctuations. Thus, with current profile control we obtain favorable confinement of thermal electrons, thermal ions, and energetic electrons at 0.5 MA. Past results indicate that energetic ions (produced by neutral beam injection) are also well-confined. The improved confinement at high current occurs simultaneously with a plasma beta (volume-average pressure/surface-average magnetic pressure) of 10{\%}. This is less than the peak beta of 26{\%} achieved in low-current pellet-fueled MST plasmas, but beta is power-limited in these higher temperature, lower density plasmas. [Preview Abstract] |
Monday, November 17, 2008 4:00PM - 4:30PM |
CI1.00005: Integrated plasma control extending the Advance Tokamak regime in JT-60U Invited Speaker: In order to realize the economical fusion reactor, high confinement (H$_{98}$ factor), high normalized beta ($\beta _{N})$, high bootstrap current fraction (f$_{BS})$, i.e. the Advanced Tokamak (AT) plasma must be sustained. In the recent experimental campaigns from November in 2007 to August in 2008, the operational regime and pulse lengths of AT plasmas has been significantly extended and the various control techniques toward steady state both in the core plasmas and in the boundary plasmas were steadily improved in JT-60U. The optimization of the beam heating profile for sustaining ITB and the enhanced wall conditioning successfully extended the high $\beta _{N}$ $\sim $ 2.6 for 28 seconds (25 seconds for H$_{98}$ $\mathbin{\lower.3ex\hbox{$\buildrel>\over {\smash{\scriptstyle=}\vphantom{_x}}$}} $ 1) in the positive shear (PS) plasma without increase in particle recycling level in the divertor. Because of high G-factor ($\beta _{N}$ H$_{98}$/ q$_{95}^{2})$ of 0.25, this plasma is relevant for ITER hybrid operation scenario. While the reversed shear (RS) plasma with high f$_{BS}$ and high H$_{98}$ factor accompanied with the strong ITB is attracting for the ITER advanced operation scenario and DEMO, the safety factor q$_{95} \quad <$ 8 has not been accessible for f$_{BS} \quad \mathbin{\lower.3ex\hbox{$\buildrel>\over {\smash{\scriptstyle=}\vphantom{_x}}$}} $ 0.7 due to low beta limit in the previous campaigns. In this experimental campaign, the b$_{N}$ limit is significantly improved and $\beta _{N} \quad \sim $ 2.7 and f$_{BS} \quad \sim $ 0.9 was achieved at q$_{95} \quad \sim $ 5.3, by utilizing large volume configuration close to the conductive wall for stabilization of RWM. The real-time control for the power exhaust to the divertor was intensely investigated. Total radiation fraction of P$_{rad}$ / P$_{heat}$ = 0.8-0.9, was maintained continuously up to 13 seconds with H$_{98}$ = 0.77-0.84 by utilizing radiation feedback for Ar gas seeding. [Preview Abstract] |
Monday, November 17, 2008 4:30PM - 5:00PM |
CI1.00006: Improved Confinement During Magnetic Levitation in LDX Invited Speaker: We report improved particle confinement in the Levitated Dipole Experiment (\urllink{LDX}{http://www.psfc.mit.edu/ldx/}) when the high-field superconducting dipole is magnetically levitated. Magnet levitation eliminates power and particle losses to mechanical supports and causes radial transport processes to determine the profiles of the confined plasma. Initial LDX experiments used multiple-frequency electron cyclotron resonance heating (ECRH) to produce quasi-stationary discharges with stable high-beta energetic trapped electrons when the superconducting dipole was mechanically supported\footnote{D. T. Garnier, et al., Phys. Plasmas \textbf{13}, 056111 (2006).}. When the mechanical supports are fully retracted and the dipole is magnetically levitated, the pressure increases and becomes more isotropic, and the plasma density is seen to increase by 2 to 5 as compared with supported operation. Variations of the microwave heating power, power deposition locations, and neutral fueling rates are used to investigate plasma confinement and profile evolution. Density profile measurements were obtained with a multi-chord interferometer, and under certain circumstances these show a rearrangement of the density profile that results in a highly peaked profile with equal number of particles per flux tube. Such a density profile is the expected stationary state that accompanies the strongly peaked pressure profiles of active magnetospheres and is also the very favorable, centrally peaked profile required for fusion applications. Low frequency fluctuations are seen during rapid profile evolution, but the fluctuations are reduced during this stationary state. Finally, we report excellent technical operation\footnote{D. T. Garnier, et al., Fusion Eng and Design \textbf{81}, 2371 (2006).} as evidenced by (1) accurate position control of the levitated dipole magnet, and (2) the enhanced float time and reduced cryostat warming during magnetic levitation. [Preview Abstract] |
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