Session GO6: NSTX and Other Spherical Torus

Overview of Results and Analysis from the National Spherical Torus Experiment

NSTX research targets predictive understanding of plasma energy confinement, long-pulse stability, and first wall heat flux handling needed for a Fusion Nuclear Science Facility and ITER. Collisionality can unify confinement trends of lithiated and unlithiated plasmas. Reduced high-k turbulence and thermal transport are correlated with increased ExB shear. BES measurements show that the pedestal turbulence poloidal correlation length increases at higher $n_{e}$, $\nabla n_{e}$, and decreases at higher $T_{i}$, $\nabla T_{e}$. Plasma characteristics (e.g. increased \textit{$\tau $}$_{E})$ change nearly continuously with increasing Li wall conditioning and ELMs stabilize by density and pressure profile alteration at r/a $>$ 80{\%}. RWM analysis shows stabilizing collisional dissipation is reduced at lower \textit{$\nu $}, but stabilizing resonant kinetic effects are enhanced. Disruption precursor analysis shows 99{\%} of disruptions can be predicted within $\sim $10ms, with an 8{\%} false positive rate. Halo currents can be toroidally asymmetric and can rotate at 0.5-2 kHz. Low frequency $n$=1 global kinks cause fast ion redistribution consistent with reduced CAE stability. The snowflake divertor configuration has greatly reduced peak divertor heat fluxes between, and during Type I ELMs. Coaxial Helicity Injection has produced plasmas with desired low density and inductance $\sim $0.35.   

Intrinsic Rotation and Torque in NSTX Ohmic H-mode Plasmas

Intrinsic rotation and torque in the co-Ip direction have been observed and investigated in NSTX ohmic plasmas, by utilizing passive views of CHERS diagnostics. Particularly the focus was placed on ohmic L-H transition to minimize the effects by other momentum exchange and sources. The analysis showed that the NTV torque by intrinsic error fields is also ignorable in ohmic plasmas due to the weak plasma response and the high collisionality. The increase of the intrinsic rotation in the edge is well correlated with ion temperature gradient change, compared with much weaker correlations with electron temperature or density gradient change. This is consistent with a corresponding theory of residual stress, and the measured rotation and torque could be directly compared with the theoretical prediction using the diffusivity as a free parameter. However, an uncertainty on the order of diamagnetic rotation exists in many places across measurement and theory, as well be discussed in details in the presentation. This work was supported by the US DOE Contract {\#}DE-AC02-09CH11466.   

Structure, Amplitude and Identification of CAEs and GAEs in NSTX

Detailed measurements of high frequency Alfv\'{e}n eigenmode (AE) amplitude and structure have been obtained in a high power (6 MW), beam-heated H-mode NSTX plasma (shot 141398) using a recently upgraded array of 16 fixed-frequency quadrature reflectometers. The observed modes are individually identified as either compressional (CAE) or global (GAE) AEs by comparison of their frequency and measured toroidal mode numbers with local Alfv\'{e}n dispersion relations. High frequency AEs --- driven by Doppler-shifted cyclotron resonance with beam ions --- correlate with enhanced electron thermal transport. These kinds of modes have historically been identified variously as compressional (CAE) or global (GAE) Alfv\'{e}n eigenmodes, but the identification has not proven conclusive. Identification is essential to understanding the extent of their effect, since the two types of modes have very different effects on resonant particle orbits. The modes identified as CAEs have higher frequencies and smaller toroidal mode numbers than the GAEs. Also, they are strongly core localized, in contrast with the GAEs, which also peak toward the plasma center but have much broader radial extent. The measurements are compared with results from the HYbrid and MHD simulation code, HYM.   

Investigation of CAE/GAE-induced electron thermal transport for NSTX-U

High values of core electron thermal transport (several 10's m$^2$/s) resulting in flat core temperature profiles in high power H-mode spherical tokamak plasmas remain unexplained. One hypothesis is high frequency fast ion modes GAE/CAEs driven by strong, super-Alfv\'{e}nic neutral beam heating can increase electron thermal transport in the core and cause the observed high $\chi _{e}$ [D. Stutman, \textit{et al.}, Phys. Rev. Lett., \textbf{102}, 115002 ( 2009)]. Additional work using ORBIT modeling and ad-hoc models of the fast ion modes demonstrates that the interaction of multiple modes with different m/n numbers at a single location can induce stochastic transport of the electrons, and can also exhibit a strong scaling of transport with mode amplitude [N.N. Gorelenkov, \textit{et al.}, Nucl. Fusion, \textbf{50}, 084012 (2010)]. This modeling work is extended to include measured mode structure, and modes located at multiple radial locations, as observed in experimental discharges, to improve the comparison to measured $\chi _{e}$, and also extended to NSTX-U parameters and plasma conditions to predict core electron thermal transport in the upgraded device.   

Low-Z Impurity Transport Analysis by Transient Gas Puff Experiments

It is important to consider the possibility of core impurity accumulations in future tokamak reactors. Experiments have been performed that focus on injections of methane and helium into MAST beam heated, L- and H-mode discharges at a range of plasma currents. Results from a model integrating the transport equation for impurities (with diffusion coefficient, D, and convective velocity, V) and density measurements derived from the active charge exchange signal are presented. Through each scenario scan, there is a region of interest focussed around normalized minor radius (r/a) of 0.6. At this radial point, the L- to H-mode scan at constant current causes a change in sign of the impurity density peaking factor (V/D) from negative (inward velocity pinch) to positive (outward velocity pinch) respectively for both carbon and helium. As each scenario feels an inward edge pinch, there is a resulting accumulation of impurities near the plasma edge in H-mode. This sign change in V/D is also observed for carbon during the L-mode current scan. However this is less evident for helium where V/D remains negative in both scenarios. The features around this region are currently being investigated using the GS2 gyrokinetic code.   

Toroidal asymmetry of divertor heat deposition in NSTX

2-D heat flux data calculated by a 3-D heat conduction solver allowed for the evaluation of peak heat flux (q$_{peak})$ and heat flux width ($\lambda _{q})$ for each toroidal angle, which generates a toroidal array of q$_{peak}$ and $\lambda _{q}$ at each time slice. Then the toroidal degree of asymmetry (DoA) of q$_{peak}$ and $\lambda _{q}$ as a function of time was defined as DoA(q$_{peak})=\sigma _{qpeak}$/mean(q$_{peak})$ and DoA($\lambda _{q})=\sigma _{\lambda q}$ /mean($\lambda _{q})$, where $\sigma $ is the standard deviation of q$_{peak}$ and $\lambda _{q}$ over data in the toroidal array. In case of ELMs and 3-D field application, the helical heat deposition produces additional scatter of data around mean values to the background scatter level without these events and it raises DoA for both q$_{peak}$ and $\lambda _{q}$. Both values of DoA(q$_{peak})$ and DoA($\lambda _{q})$ are highest at the ELM peak times, with DoA(q$_{peak})$ up to $\sim $0.9 and DoA($\lambda _{q})$ up to $\sim $0.3 for typical type-III ELMs, while they become lower toward the later stage of the inter-ELM period, \textit{eg}, DoA(q$_{peak})\sim $0.15 and DoA($\lambda _{q})\sim $0.05. The correlation between DoA(q$_{peak})$ and DoA($\lambda _{q})$ is the strongest at the ELM peak times and becomes weaker later in the ELM cycle.   

Divertor ion temperature measurements on MAST by retarding field energy analyser

The ion temperature (T$_{i}$) at the divertor determines the heat flux and erosion due to plasma surface interactions. However, there are few measurements of T$_{i}$ at the divertor compared to the electron temperature (T$_{e}$) due to the relative complexity of the measurement. Divertor T$_{i}$ measurements have been made using a retarding field energy analyser at the lower outer divertor of MAST in L-mode, inter-ELM H-mode and during ELMs. These measurements are compared to the T$_{e}$ from Langmuir probes and the heat flux from IR thermography. The sweeping of the strike point means that T$_{i}$ profiles can be produced. In a range of L-mode discharges with I$_{P}$ = 400 \--- 900 kA, T$_{i}$ $\sim$ T$_{e}$, with T$_{i}$ = 3 \--- 15 eV, however in inter-ELM H-mode T$_{i}$/T$_{e}$ can be up to 3. In inter-ELM H-mode a dependence of T$_{i}$/T$_{e}$ on collisionality has been seen. First measurements of T$_{i}$ at the target during ELMs will be presented either measuring the ELM averaged ion temperature using a slow sweep of the discriminator voltage or using a fast (50$\mu$s) sweeping mode. The effect of parallel Mach flows on T$_{i}$ measured at the divertor will be considered for the discharges presented in this work.   

The Effects of Contamination by Residual Gases in NSTX on Deuterium Uptake and Retention in Li Films

Lithium-conditioned plasma-facing components (PFCs) have improved plasma performance by reducing the recycling of hydrogenic species; however, a quantitative understanding of the adsorption and retention of deuterium by lithium-conditioned materials is needed to optimize the performance of Li-PFCs, especially for long duration discharges anticipated in NSTX-U. We report on results from UHV surface science experiments using X-ray photoelectron spectroscopy, Auger electron spectroscopy, and thermal desorption spectroscopy. Oxygen uptake curves of solid lithium and lithium films on TZM before and after contamination by residual O$_2$, H$_2$O, CO, and air will be presented. Further work is planned for single crystal Mo(100) substrates that will distinguish the effects of impurities and grain boundaries in TZM. An ECR plasma source will be used to irradiate these substrates with deuterium neutrals and ions, and data for deuterium uptake on lithium-coated TZM and Mo(100) with exposure to the residual gases will be presented.   

Resistive Wall Mode Physics and Control to Sustain High Normalized Beta in NSTX

High bootstrap current fraction and efficient fusion production needed in continuously operating spherical torus fusion devices require plasmas with low plasma internal inductance, $l_{i}$, and high beta. Present research in NSTX focuses on greater understanding and verification of kinetic resistive wall mode (RWM) stabilization physics and analysis of improved active control techniques that have reduced disruptions in these plasmas. MISK code calculations indicate that the largest stabilizing kinetic effect comes from resonance between the mode and the precession motion of trapped thermal ions. The stabilizing effect of energetic particles depends on their anisotropic distribution, which also modifies the pressure-driven destabilization term. Long-pulse plasmas have reached \textit{$\beta $}$_{N}$/$l_{i} >$ 13. A positive and counter-intuitive result is that the greatest disruption probability does not occur at the highest \textit{$\beta $}$_{N}$, or \textit{$\beta $}$_{N}$/$l_{i}$, but at lower values closer to the $n = 1$ no-wall stability limit. The result can be understood by evaluating kinetic RWM stability for time-varying plasma rotation and equilibrium profiles, and is further examined by resonant field amplification evolution in kinetically stabilized plasmas at high \textit{$\beta $}$_{N}$.   

Progress in Non-solenoidal Startup via Local Helicity Injection in the Pegasus Experiment

The operating space for localized helicity injection for non-solenoidal startup is constrained by helicity input and dissipation rates and a geometric limit on plasma current set by Taylor relaxation. To test the understanding of dissipation mechanisms during helicity-driven startup, the helicity injection startup and growth is being expanded to $\sim $0.3 MA plasma currents and longer pulse lengths on the Pegasus experiment. Following initiation via active current sources, passive electrodes can be used to grow discharges for relatively long pulse lengths. Bursts of MHD activity are observed during helicity injection, and correlate with rapid equilibrium changes, including inward motion of the magnetic axis, redistribution of the toroidal current, and strong ion heating with ion temperatures $\sim $1 keV observed. The plasma arc injector impedance and the associated helicity injection rate appear to be constrained by a double-layer space charge limit at low currents and by the Alfv\'{e}n-Lawson limit for strong electron beams at high currents. Additions to the experiment include an expanded poloidal field coil system for added plasma control, new divertor coils, new plasma gun-electrode injector assemblies, expanded gas fueling techniques, and eventually a doubling of the toroidal field.   

Non-inductive Plasma Start-up and Current Ramp-up in NSTX-U

Results from NSTX Transient Coaxial Helicity Injection (CHI) experiments have demonstrated generation of 300kA start-up currents, and when these discharges were coupled to induction they attained 1MA of plasma current consuming 65{\%} of the inductive flux of standard inductive-only discharges in NSTX. In addition, the CHI-initiated discharge has lower plasma density and a low normalized internal plasma inductance of 0.35, as needed for achieving advanced scenarios in NSTX-U. CHI will be used for generating the initial seed current for non-inductive current ramp-up and non-inductive current sustainment in NSTX-U. Improved positioning of the injector flux generating coils on NSTX-U substantially increases the CHI current generation potential in NSTX-U. Scenario modeling results using the TSC code for full non-inductive start-up using CHI and subsequent non-inductive current ramp-up using neutral beams and RF will be presented. This work supported by U.S. DOE Contracts DE-AC02-09CH11466 and DE-FG02-99ER54519 AM08.   

Fast wave power flow along SOL field lines in NSTX

On NSTX, a major loss of high-harmonic fast wave (HHFW) power can occur along open field lines passing in front of the antenna over the width of the scrape-off layer (SOL). Up to 60{\%} of the RF power can be lost and at least partially deposited in bright spirals on the divertor floor and ceiling [1,2]. The flow of HHFW power from the antenna region to the divertor is mostly aligned along the SOL magnetic field [3], which explains the pattern of heat deposition as measured with infrared (IR) cameras. By tracing field lines from the divertor back to the midplane, the IR data can be used to estimate the profile of HHFW power coupled to SOL field lines. We hypothesize that surface waves are being excited in the SOL, and these results should benchmark advanced simulations of the RF power deposition in the SOL (e.g., [4]). Minimizing this loss is critical optimal high-power long-pulse ICRF heating on ITER while guarding against excessive divertor erosion.\\[4pt] [1] J.C. Hosea \textit{et al}., \textit{AIP Conf Proceedings }\textbf{1187} (2009) 105. \\[0pt] [2] G. Taylor \textit{et al}., \textit{Phys. Plasmas} \textbf{17} (2010) 056114. \\[0pt] [3] R.J. Perkins \textit{et al.}, to appear in \textit{Phys. Rev. Lett}. \\[0pt] [4] D.L. Green \textit{et al., Phys. Rev. Lett. }\textbf{107 }(2011) 145001.   

Two-Dimensional Characterization of ELM Precursors in NSTX

Gas Puff Imaging (GPI) has been used to capture the two-dimensional evolution of Edge Localized Mode (ELM) precursors. Precursor events were observed preceding ELMs and ELM-induced H-L back transitions in radio frequency (RF) heated H-mode plasmas, and the growth of the precursor mode through the ELM filamentation was imaged in the plane perpendicular to the local B-field. Strong edge intensity modulations appeared to propagate in the electron diamagnetic direction while steadily drifting radially outward. Intensity fluctuations were observed at frequencies around 20 kHz and wavenumbers of 0.05-0.2 cm$^{-1}$. Edge intensity fluctuations are strongly correlated with magnetic signals from Mirnov coils, and toroidally distributed coils estimated toroidal mode numbers of n=5-10. Upon growing to a trigger point, precursor fluctuations were seen to form filamentary structures and move into the Scrape-Off Layer (SOL) explosively with radial velocities peaking at 8 km/s. Quantitatively similar precursors have been observed in Ohmic H-mode plasmas as well, though significantly fewer events are seen in the Ohmic cases and none were observed in similar near-threshold NBI shots studied.   

Overview of recent experimental results from the Rotamak-ST discharges

By adding an axial rod current $I_{z}$, rotamak can operate as an ultra-low aspect ratio spherical tokamak (A $\approx $ 1.1), in which the plasma current is driven in a steady state, non-inductive fashion by rotating magnetic field. Recent experiments on Prairie View (PV) Rotamak-ST are focusing on high elongation $k$, high toroidal field utilization factor ($I_{p}$/$I_{z})$, and high {\ss}$_{t}$ discharges. Recorded highest elongation $k \approx $ 4 and high value of $I_{p}$/$I_{z} \approx $ 2.5 have been achieved through actively controlling five magnetic shaping coils in typical rotamak-ST discharge, where there is no appearance of any MHD instability due to relatively high toroidal field. The combination of active plasma shape control and operation under low toroidal field allows PV rotamak to achieve both higher toroidal field utilization factor ($I_{p}$/$I_{z} \approx $ 4.0) and higher {\ss}$_{t}$ ($>$ 40{\%}) discharges; but in this regime, low-$m$ / $n$ = 1 kink mode has been observed.