Session NO4: NSTX, Pegasus, and International Tokamaks

Progress in Nonsolenoidal Plasma Startup Using Point-Source Helicity Injection in the \textsc{Pegasus} ST

Compact, high-current plasma guns are employed as DC helicity injectors on the \textsc{Pegasus} ST. This startup technique has produced $I_p \sim 0.17$ MA to date, consistent with helicity balance and Taylor relaxation constraints. Once a tokamak-like plasma is formed, passive electrodes can act as helicity injectors for further growth. This may provide additional control of the edge current density, the Taylor relaxation limit, and provide a higher helicity input rate. Ion heating to $T_i \sim 0.5$ keV is observed in the strongly-reconnecting, helicity-driven phase. Efficient handoff from helicity to inductive drive requires the buildup of core current density. Following helicity formation, OH-driven plasmas are MHD-quiescent and sustained above 0.20 MA, apparently due to increased core magnetic shear. Outstanding issues for prediction to larger fusion experiments include: structure of the edge current density; impedance of the injected current channel; impurity behavior; and the behavior of the confinement and helicity dissipation rate as $I_p $ and $T_e $ increase.   

Recent results from the National Spherical Torus Experiment

Upgrades to NSTX for the 2010 campaign include the installation of its Liquid Lithium Divertor (LLD) for improved divertor pumping to control edge collisionality. The LLD forms an annular ring in the lower outer divertor and consists of four plates, coated on their plasma facing surface with a porous molybdenum layer to hold lithium, which can be heated to $\sim$300$^{\circ}$C. Several new diagnostics have been installed to investigate the LLD, including thermocouples, Langmuir probes, spectroscopy and fast cameras. The scaling of the peak heat flux and profile width with toroidal field and plasma current has been measured with a two-color IR camera capable of measuring on ELM timescales. For transport and turbulence studies, a new beam-emission spectroscopy (BES) diagnostic was installed and commissioned extending the wavelength range of turbulence diagnostics and complementing the existing high-k scattering diagnostic. The BES diagnostic has also provided the first measurements in H-mode plasmas of the amplitude and radial profile of TAEs during avalanches, complementing data from a new 16 channel reflectometer array.   

Operational Characteristics of Liquid Lithium Divertor in NSTX

Lithium coatings on plasma-facing components (PFC's) have resulted in improved plasma performance on NSTX in deuterium H-mode plasmas with neutral beam heating.$^{ }$Salient results included improved electron confinement and ELM suppression. In CDX-U, the use of lithium-coated PFC's and a large-area liquid lithium limiter resulted in a six-fold increase in global energy confinement time. A Liquid Lithium Divertor (LLD) has been installed in NSTX for the 2010 run campaign. The LLD PFC consists of a thin film of lithium on a temperature-controlled substrate to keep the lithium liquefied between shots, and handle heat loads during plasmas. This capability was demonstrated when the LLD withstood a strike point on its surface during discharges with up to 4 MW of neutral beam heating.   

Core impurity reduction using divertor D$_2$ injection in lithium-conditioned H-mode discharges in NSTX

The application of lithium evaporative coatings in NSTX resulted in improved confinement and MHD stability. However, the routine achievement of ELM-free regimes caused core impurity accumulation with $Z_{eff}$ (due to carbon) increasing up to 3-4 and core $P_{rad}$ (due to metals) ramping up to several MW. Strategies for impurity reduction are thus essential for NSTX ELM-free discharges. In a dedicated experiment, small divertor D$_2$ injections (3-7 Torr-l) were used in high triangularity, NBI heated, H-mode discharges resulting in the reduction in core $n_C$ and $Z_{eff}$ by up to 30$\%$ without confinement degradation. The purpose of this analysis is to identify the contributions from underlying effects such as the reduction of divertor carbon and metal sputtering, changes in the SOL transport, improved impurity compression, or neoclassical edge convection due to increased neutral pressure. Supported by the U.S. DOE under Contracts DE-AC52-07NA27344, DE-AC02-09CH11466, DE-AC05-00OR22725, DE-FG02-08ER54989, and W-7405-ENG-36.   

Triggered confinement and pedestal temperature enhancement in NSTX H-mode discharges

We report progress in the extension of a high performance regime (``Enhanced Pedestal'' or EP H-mode) in NSTX discharges, where the pedestal temperature doubles and the energy confinement increases by 50{\%}, above and beyond the confinement enhancement from lithium wall coatings [1]. The spontaneous transition is triggered by a large edge-localized mode, either natural or externally triggered by 3-D fields; the EP H-mode itself is ELM-free. The transport barrier grows inward from the edge, with a doubling of both the pedestal pressure width and the spatial extent of steep radial electric field shear. While short EP H-mode phases were previously reported [2], an EP H-mode with duration up to three energy confinement times was recently observed. The normalized beta $\sim $ 6.5 is amongst the highest values sustained in NSTX. Moreover the non-inductive fraction $\sim $ 0.65 is the highest plasma current$\sim $0.9 MA in NSTX. Experiments are continuing for achievement of reproducible EP H-modes. \\[4pt] [1] H.W. Kugel, et. al.,Phys. Plasma \textbf{15} (2008) 056118. \\[0pt] [2] R. Maingi, et. al., J. Nucl. Mater. \textbf{390-391} (2009) 440.   

Effect of lithium wall conditioning on heat flux widths and SOL transport in NSTX

Peak heat fluxes in NSTX of up to 15 MW/m$^2$ have been measured on the divertor during high performance (I$_p$ = 1.2 MA, P$_{NBI}$ = 6 MW, $\delta~\sim$ 0.7) H-mode discharges. While the plasma facing components on NSTX consist of ATJ graphite, a program of lithium wall conditioning has been used in place of boronization. The effect on NSTX discharges has been to improve energy confinement, reduce ohmic flux consumption therefore extending the discharge lifetime, and the elimination of Edge Localized Modes (ELMs) when sufficient lithium is applied. However, when lithium wall conditioning is used, the heat flux footprint as measured by IR thermography contracts by 50-60\%. The implications for transport in the scrape-off layer (SOL) and the impact on divertor heat flux will be presented.   

Pedestal Scaling with Global Plasma Parameters in NSTX

A successful mode of operation of ITER will require establishing sufficiently high pedestal pressure during H-mode discharges as well as small or no ELMs. Recent observations from high aspect ratio tokamaks (DIIID, ASDEX, JT60) have shown a general trend in the pedestal pressure and density widths scaling with the pedestal poloidal beta ($\beta)$ to the power one half The low aspect ratio tokamak MAST reported similar scalings where the electron pedestal temperature width scales weakly with $\rho$* but correlates with $\beta ^{1/2}$ We report analysis of the NSTX pedestal pressure and density structure (height and width) during ELMy H-mode discharges. In this analysis, we focus on measurements of the pedestal structure as a function of plasma current and toroidal field. To approach the peeling-ballooning stability limit, and therefore the maximum achievable pedestal pressure, we target the intrinsic ELMs as opposed to those induced by resonant magnetic perturbations. Furthermore, to address the impact of edge turbulence on the pedestal structure during the evolution of an ELM, preliminarily turbulence measurements across multiple scales are discussed. Work supported by DOE contract DE-AC02-09CH11466.   

Edge Stability of Small-ELM Regimes in NSTX

NSTX has observed low-frequency (f $<$ 10 kHz) unstable modes with characteristics similar to the edge harmonic oscillation (EHO) coincident with transition to a small-ELM regime where the ELMs reduce the plasma stored energy by less than 1\%. Soft x-ray emission indicates that these modes are localized just inside the density pedestal. Microwave reflectometry shows density fluctuations in the pedestal at the mode frequency. Toroidal Mirnov analysis indicates n=1 for the lowest order mode, which rotates at the plasma rotation frequency, with higher harmonics observed simultaneously in some cases. Stability analysis during the observed modes indicates instability to n=1-4 with unstable mode eigenfunctions peaked near the plasma edge. Transition to this regime is associated with a downward biased plasma ($\delta$$_r$$^s$$^e$$^p$ $<$ -5 mm) and increased edge collisionality ($\nu$$^*$$_p$$_e$$_d$ $>$ 1). Increased pedestal pressure and bootstrap current are observed in the small-ELM regime, with the peak in the bootstrap current moved inward from $\Psi$$_N$ = 0.96 to 0.94.   

Initial density fluctuation measurements from the NSTX Beam Emission Spectroscopy diagnostic system*

Density fluctuation measurements on the ion gyroscale have been obtained on NSTX with a newly commissioned beam emission spectroscopy (BES) diagnostic system. The BES system measures red-shifted $D_\alpha$ emission near 660~nm from deuterium neutral beams with high throughput optics and high efficiency detectors. The system presently employs 16 detection channels arranged in radial and poloidal arrays, and an expansion to 32 channels is planned. Radial arrays can measure fluctuations from $r/a\approx0.1$ to beyond the last closed flux surface and resolve fluctuations with $k_\perp\rho_i\leq1.5$. Initial BES measurements reveal broadband turbulence and coherent modes below 300~kHz for $r/a\geq0.4$. The broadband turbulence appears in high gradient regions and increases at H-L transitions. The frequency characteristics of the coherent modes correlate with Alfv\'{e}n/energetic particle modes in Mirnov probe measurements, but some coherent modes appear in BES measurements only. *Supported by the U.S. Department of Energy under Contract Nos. DE-FG02-89ER53296, DE-AC02-09CH11466 and DE-SC0001288.   

Comparison of Aspect Ratio Effects on Neoclassical Tearing Modes Between DIII-D and NSTX

Experimental data is analyzed in which $m/n=2/1$ neoclassical tearing modes self-stabilized; this ``marginal point" is valuable for evaluating the relative importance of the terms in the balanced Modified Rutherford equation. DIII-D and NSTX have similar cross-sectional area and shape except for the large difference in aspect ratio. The aspect ratio effects for NTMs explicitly occur in the MRE in the destabilizing helically perturbed bootstrap current term, in the small island stabilizing effects, and in the stabilizing curvature term. The marginal island width on NSTX at $q_{95}\sim 8$ is about three times the ion banana width. This agrees with the $q_{95}\sim 4$ data on DIII-D but is a larger multiple than for DIII-D at $q_{95}\sim 7$. The balance in the MRE indicates that while the stabilizing effect of the curvature term in DIII-D is negligible, it is important in NSTX. The temporal behavior of the mode suggests NSTX operates closer to marginal classical tearing stability, but benefits from the stabilizing effect of curvature.   

Results of the NSTX Control Experiments

New control implementations and its effect on plasma performance on the NSTX are summarized. The control algorithm for the NSTX system has been tuned for performance and four new controllers namely, upper/lower inner/outer strike point (SP) controllers, were installed. An offline system identification of the plasma response to the control inputs was performed and several control improvements were identified. An online automatic relay-feedback PID tuning algorithm was implemented which has the advantage of tuning the controller in a single shot and more accurately. The PID controller for the SP was tuned by employing the Ziegler-Nichols method. The resulting SP controller was successfully employed to control the particle and heat flux on the Liquid Lithium Divertor. This capability enabled achieving the ``snowflake'' divertor configuration for the first time in NSTX. The new control algorithm lead to better disturbance rejection (more stable against ELMs) with longer pulse length compared to the uncontrolled shots. Work supported by U.S. DOE Contract DE-AC02-09CH11466.   

Free-Boundary Modeling of NSTX Plasmas

We have implemented a new capability within the SWIM framework to facilitate detailed free-boundary transport-timescale modeling of NSTX and other tokamaks and comparison with experimental data. The SWIM framework provides a convenient method to bring together state-of-the-art tokamak physics codes and code packages. In these simulations, we make use of the experimental coil currents in the free-boundary Tokamak Simulation Code (TSC) to simulate the discharge current ramp-up, flattop, and in some cases, the rapid disruptive termination. The Monte Carlo neutral beam code NUBEAM is used to calculate neutral beam heating and current drive. A new option, using TRXPL, allows us to import the density and/or pressure profiles from a previous TRANSP run into the simulation and to use these instead of the predictive profiles computed by TSC. This provides a convenient way to effectively decouple the errors made in the prediction of these profiles from those made in the evolution of the current profile, while still using the bootstrap and NUBEAM current drive terms. Applications include a detailed benchmarking exercise between TSC and Free-Boundary-PTRANSP, and comparison of predicted disruption halo-currents with experimental measurements.   

High-Harmonic Fast Wave (HHFW) Heating Results on NSTX

This talk will present recent experimental and modeling results from NSTX HHFW research. HHFW heating of low current (200 - 400 kA) plasmas has resulted in a transition to a high bootstrap current fraction, H-mode regime needed for solenoid-free ramp-up. Coupling of HHFW power to NBI H-mode plasmas has been improved with lithium wall conditioning [1], although significant rf power is measured to flow to the divertor, particularly at longer launch wavelengths. Modeling results for H-mode discharges that use a combination of HHFW and NBI heating predict a strong competition between direct electron heating and fast-ion acceleration. A double-feed upgrade of the HHFW antenna in 2009 did not improve the stand off voltage by as much as predicted and appears to be limited by RF currents induced on the antenna surface. However, the stand off voltage limit can be increased with sufficient antenna conditioning. [1] G. Taylor, et al., Phys. Plasmas 17, 056114 (2010).   

Identification of the background plasma damping mechanisms of antenna-driven toroidal Alfven eigenmodes of medium n on JET

In tokamak burning plasma experiments such as ITER, it is expected that Alfv\'en eigenmode (AE) instabilities of, typically, medium and high toroidal mode number $n$ will be triggered by populations of energetic ions, such as $\alpha$-particles. The stability of this specific class of AEs is studied experimentally in the Joint European Torus by observing the plasma response to antenna-driven frequency-sweeping perturbations at the plasma edge. During the 2008/9 experimental campaigns, the complete set of the new antennas was operated and medium-$n$ AEs were excited under various plasma conditions. A big collection of damping rate measurements of, mainly, toroidal AEs (TAEs) has been obtained following the technical optimization of the diagnostic. A subset of these measurements are compared to different plasma models, as implemented in the codes LEMan and CASTOR, allowing the identification of the background plasma damping mechanisms that come into play.   

Recent Progress on Long Pulse Divertor Operation on EAST

Significant progress has been made in EAST on both physics and engineering fronts toward steady state operations. Long pulse diverted discharges well over 60 seconds have been achieved in EAST by lower hybrid current drive with advanced wall conditioning and active divertor pumping. A series of divertor physics experiments have been carried out to assess both SN and DN divertor performance in EAST under Ohmic and L-mode plasma conditions. Dedicated field reversal experiments have also been conducted to investigate divertor asymmetries, and contributions from various drifts are assessed using the SOLPS code. To actively control power and particle fluxes to divertor target plates, gas puffing with Ar and N$_{2}$ has been explored in EAST to preferentially reduce peak heat fluxes near outer strike points and mitigate in-out divertor asymmetries. Methane injection at various divertor locations (inner divertor, outer divertor and private region) has also been explored to quantitatively assess divertor screening for intrinsic carbon impurities. Further efforts will be dedicated to active feedback control of gas puffing rates to maintain the divertor plasma in partial detachment conditions, which is essential for achieving higher power, long pulse operations.