Session CO3: NSTX and General Spherical Tokamak

Overview of the 2007 NSTX experimental campaign*

The 2007 experimental campaign covered a wide range of scientific topics. Measurements of the turbulent electron fluctuation spectrum were made in many different plasma conditions including H-mode plasmas, L-mode plasmas, and plasmas with internal transport barriers using the microwave scattering system which spans a range in wavenumber to $k_\perp \rho_e \sim 1$ . RWM feedback control has been explored with and without correction of time-varying error fields, allowing mode stabilization and plasma rotation sustainment throughout the discharge. The upgraded lithium evaporator was used extensively and was observed to raise the electron temperature in H-mode plasmas. Improved mode conversion of Electron Bernstein wave emission was observed in high elongation H-mode plasmas after lithium evaporation. A new Alfv\'en wave was observed, the $\beta$-induced Alfv\'en eigenmode, and $\beta$ suppression of Alfv\'en cascades was also observed. Deuterium puffing has been used to produce partial divertor detachment in highly shaped plasmas without affecting plasma performance. A record electron temperature of 4.7keV was achieved using HHFW heating. Plasmas with high elongation $\kappa \sim 2.6$ have been maintained with high non-inductive current fraction $f_{NI} \sim 0.55$. These and other results will be summarized. *This work was supported by DoE contract No. DE-AC02-76CH03073   

Experimental investigation of turbulent fluctuations with the scale of collisionless skin depth in NSTX plasmas

Various numerical simulations support the conjecture that the ubiquitous problem of anomalous electron transport in tokamaks may arise from a turbulence driven by the electron temperature gradient, with large radial structures on the scale of the collisionless skin depth (\textit{$\delta $}$_{s})$. In this paper, we present results from measurements of turbulent fluctuations in the National Spherical Torus Experiment (NSTX), where the low level of ion transport anomaly makes plasma conditions uniquely suitable for the study of electron transport. Plasma density fluctuations are measured with coherent scattering of electromagnetic waves using a novel scattering geometry where the probing beam propagates obliquely to the magnetic field and the radial resolution of measured signals is greatly improved by the toroidal curvature of magnetic field lines. The onset of a broadband turbulence with a radial scale of $\sim $ \textit{$\delta $}$_{s}$ is observed during electron heating with high harmonic fast waves (HHFW). Numerical calculations are underway to identify the nature of observed fluctuations.   

Internal transport barriers in NSTX reversed-shear plasmas

Simultaneous internal transport barriers in both ion and electron channels have been observed in reversed shear ($dq/dr< 0$) discharges on NSTX. While these ITBs can be observed in discharges with only neutral beam heating, a HHFW (high-harmonic fast wave) RF power scan was performed at constant beam input power to scan the electron temperature gradient. Measurements show that the electron and ion transport barriers can be at different minor radii, suggesting different mechanisms for the suppression of ion and electron turbulence. Examination of the roles of magnetic and velocity shears with respect to ion and electron transport is possible through the use of CHERS and the newly upgraded 16 channel NSTX MSE diagnostic which now provides full coverage from the outboard plasma edge to past the magnetic axis with 3-4 cm resolution. Additional measurements from these well diagnosed plasmas, including Thomson scattering, X-ray diodes, reflectometry, and high-k fluctuations, will be examined in addition to results from TRANSP and GS2 linear simulations.   

Beta Scaling and Momentum Transport Studies in NSTX

Experiments have been carried out in NSTX to study both the beta scaling of confinement and momentum transport. Beta scaling studies were carried out at fixed collisionality and normalized electron gyroradius both in highly shaped plasmas ($\kappa $=2.1, $\delta $=0.8) and in weakly-shaped plasmas ($\kappa $=1.8-1.9, $\delta $=0.4). In the highly shaped plasmas, which exhibited small ELMS in the range of beta from 7 to 20{\%}, no degradation of energy confinement was observed. In the more weakly shaped plasmas, the character and impact of ELMs changed markedly from low to high beta, leading to a severe degradation of confinement as beta increased. Momentum diffusivity in NSTX does not scale with ion thermal diffusivity, as at conventional aspect ratio, possibly due to suppression of ITG modes due to high ExB shear. Perturbative momentum transport studies, using non-resonant n=3 magnetic braking of the plasma, have been carried out, and these indicate momentum confinement times that are a factor of two to three greater than the energy confinement time, as well as significant inward momentum pinch velocities. This work is supported by United States DOE contract DE-AC02-76CH03073.   

Gyroradius-Scale Ion Gradients in NSTX

The spherical torus geometry is well suited to compare experimental profiles on a fundamental transport scale. The low-field region in the low-aspect ratio NSTX along with a high spatial resolution charge exchange spectroscopy system make gyroradius-scale ion measurements possible. Ion temperature and density gradients with gyroradius-scale widths have been measured in some NSTX plasmas. Measured ion temperature changes of $\Delta T_i = 250-500$ eV are observed over $\Delta r \sim 1$ cm with total Larmor radius $\sim 1$ cm. Minimum gradient widths comparable to the larger banana orbit width would normally be expected from ``random walk'' arguments of neoclassical theory. Since radial electric field ($E_r$) gradients also vary on a gyro-radius scale, distortion of the gyro orbit is expected to cause orbit shrinking or orbit expansion [1], depending on the sign of $E_r$. Orbit shrinking removes the apparent discrepancy in the gradient width. \newline \newline [1] K. C. Shaing, A. Y. Aydemir, R. D. Hazeltine, Phys. Plasmas 5, 3680 (1998).   

Active Resistive Wall Mode Feedback with Expanded Sensors in NSTX

The resistive wall mode (RWM) active stabilization system on NSTX is expanded to include sensors measuring radial and poloidal mode components both above and below the plasma midplane. Various combinations of these control sensors are used to determine the effect on feedback performance. Plasma rotation and profile variation is generated by non-resonant magnetic braking using an applied n = 3 field configuration. Variation of the relative difference between the measured n = 1 RWM phase and the applied control field phase demonstrates both positive and negative feedback. Poloidal deformation of the mode observed during feedback at low plasma rotation [1] is examined with feedback using sensors both above and below the device midplane. Amplitude modulation of the measured n = 1 RWM sensor signal, thought to be resonant field amplification, can appear in both radial and poloidal sensors below the computed ideal MHD no-wall beta limit as determined by DCON stability analysis. The frequency of this modulation decreases as the RWM becomes unstable. \newline [1] S.A. Sabbagh, R. Bell, J.E. Menard, et al., \textit{Phys. Rev. Lett.} \textbf{97}, 045004 (2006).   

Toroidal Alfven Eigenmode Avalanches on the National Spherical Torus Experiment

Experiments on the National Spherical Torus Experiment have found the fast ion beta threshold for excitation of Toroidal Alfven Eigenmodes (TAE). A further increase in beam heating power is seen to push the TAE into a repetitive cycle of increasingly stronger bursts, each cycle culminating in a large, multi-mode burst and a drop in the neutron rate of approximately 10\%. These strong bursts are identified as TAE avalanches [Nucl. Fusion 35 (1995) 1661]. In such an avalanche, the fast-ion phase-space islands describing the orbits of fast ions trapped in the TAE wave field have reached such an amplitude that islands from multiple modes overlap, leading to greatly enhanced transport of the fast ions, and a concomitant increase in the drive for the modes. Transport of fast ions in ITER is expected to be through a similar multi-mode interaction. Fast ion transport will be studied with NOVA and ORBIT, benchmarked on mode amplitudes measured with a multi-channel reflectometer array.   

Dependence of the L-H power threshold on magnetic balance and heating method in NSTX

H-mode access is a critical issue for next step devices, such as the International Thermonuclear Experimental Reactor (ITER), which is projected to have a modest heating power margin over the projected L-H power threshold (PLH). The importance of a second X-point in setting the value of PLH has been clarified in recent experiments on several tokamaks. Specifically a reduction of PLH was observed when the magnetic configuration was changed from single null (SN) to double null (DN) in the MAST, NSTX, and ASDEX-Upgrade devices [1]. Motivated by these results, detailed PLH studies on NSTX have compared discharges with neutral beam and rf heating, as a function of drsep. Similar PLH values and edge parameters are observed with the two heating methods in the same magnetic configuration, with PLH $\sim$ 0.6 MW lowest in DN and increasing to $\sim$ 1.1 MW and 2-4 MW in lower-SN and upper-SN configurations respectively (ion grad-B-drift towards lower X-point). The evolution of the experimental profiles of parameters in L-mode before the L/H transition will be compared with simulations using the XGC code (C.S. Chang). \newline [1] MEYER, H. et al., Nucl. Fusion 46 (2006) 64.   

Overview of Transient CHI Plasma Start-up in NSTX and HIT-II

The capability for solenoid-free current generation provides greater flexibility in a tokamak reactor design by allowing access to lower aspect ratio configurations. NSTX is testing the method of transient coaxial helicity injection as a way of generating this current. This current is produced by discharging a capacitor across the lower divertor plates in the presence of toroidal and poloidal magnetic fields. This causes an expanding plasma to detach from the lower divertor plates resulting in the formation of a closed flux equilibrium. The method was originally developed in the HIT-II ST at the Univ. of Washington, where up to 100 kA of closed flux current was generated. Coupling this to induction resulted in CHI started discharges to consistently out-perform what was possible by induction alone, producing nearly 300 kA of current using only 52 mWb of solenoid flux. Recently application of this method on NSTX has resulted in the production of record levels of non-inductively generated closed flux current ($>$160 kA). Initial tests on NSTX have allowed coupling this current to induction from the central solenoid. \textit{This work supported by U.S. DOE Contracts {\#} DE-AC02-76CH03073and }\textit{DE-FG02-99ER54519 AM08}   

Non-Solenoidal Startup of the Ultra-Low Aspect Ratio Pegasus ST

Pegasus is an extremely low aspect ratio tokamak exploring quasi-spherical, high-pressure plasmas and developing plasma formation and control techniques for future ST/tokamak applications. Non-inductive startup has been demonstrated using washer gun current sources in the lower divertor region. The injected current initially follows the helical structure of the crossed toroidal and vertical fields, but at extremely low field applied field (B$_{V} \quad \approx $ .005 T, B$_{T} \quad \approx $ .01 T) relaxation to a tokamak-like plasma occurs. Toroidal currents of 50 kA have been driven by less than 4 kA injected current. I$_{N }>_{ }$12 MA/m-T is observed in these plasmas. Compatibility with ohmic operation has been demonstrated by successful hand-off of gun-produced plasmas to ohmic current drive. A mid-plane gun system, easily implemented on any machine with mid-plane port access, is presently under development. The formation of non-inductive plasmas and hand-off to ohmic induction has been demonstrated using this system. Current drive experiments using PF induction of a midplane seed plasma are underway.   

Recent EBW Emission Results on NSTX

NSTX high beta plasmas operate in the overdense regime, allowing the electrostatic electron Bernstein wave (EBW) to propagate and be strongly absorbed and emitted at the electron cyclotron resonances. As such, EBWs may enable local electron temperature measurements and provide local electron heating and current drive. For these applications, efficient coupling between EBWs and electromagnetic waves outside the plasma is needed. Thermal EBW emission, measured on NSTX with two remotely steered, quad-ridged antennas, has been used to determine the EBW transmission efficiency for a wide range of plasma conditions. The antennas collect fundamental (8-18GHz), second and third (18-40 GHz) EBW emission via the oblique B-X-O mode conversion process. Recent H-mode results show that fundamental and second harmonic EBW transmission efficiencies $>$30\% are observed for certain edge conditions. Experimental results from this diagnostic and comparisons to modeling will be presented.   

Recent results from MAST spherical tokamak

Following installation of a new PINI source, the MAST spherical tokamak has been operating with up to 3.8 MW of NBI heating. The scaling of confinement could hence be determined over a greater range of power and current and integrated analysis of diagnostic data has facilitated transport analysis. First particle confinement data from pellet fueled plasmas has also been obtained. Density peaking in L-mode has been found to scale with normalised current density as in conventional tokamaks. Off-axis NBI current drive has been studied in extreme SND discharges. A new 28 GHz RF system has generated and sustained 33 kA of plasma current without the solenoid and EBW assisted start-up has also been demonstrated. Internal coils have been installed for ELM control and to excite TAE modes and study their damping. The structure and evolution of the edge radial E-field has been measured with a new edge spectroscopy system and first measurements of core density turbulence are available from a BES system. Further results on ELM structure, SOL filaments and the divertor plasma have also been obtained using improved edge diagnostics.   

Investigation of transient phenomena on MAST using high resolution Thomson scattering

The MAST tokamak is equipped with high spatial resolution Ruby laser and high time resolution Nd:YAG laser TS diagnostics. The Nd:YAG lasers are viewed by two separate sets of optics. One of these lens systems views the core region and measures at spatial resolution of 2.5-4cm and the other examines the plasma edge with 1cm resolution. This newly installed edge system has already produced a number of important results. In H-mode and L-mode filaments have been observed using laser time separations of 1-20$\mu $s. The high spatial resolution has allowed determination of the evolution of the outboard pressure pedestal, which plays a critical role in determining plasma stability. The variable time separation between lasers has also been exploited to study pellet deposition and retention in the plasma. A major upgrade to the core Nd:YAG system is now being planned. It is proposed to replace the current four lasers with a combined sampling rate of 200Hz at 1.0J with eight lasers with a combined sampling rate of 240Hz at 1.6J. The increase in laser energy together with new optics will allow the system to sample at high spatial resolution.   

Lithium Loaded Target Plate for driving NSTX toward high performance

Following CDX-U, the NSTX device in PPPL is on its way to the new plasma confinement and stability regimes when the pumping lithium surface will provide a high temperature plasma edge. Both ion and electron gradient turbulence is expected to be suppressed in this regime, while the finite edge current density at the separatrix will stabilize ELMs. So far, NSTX has made only a modest step in this direction using the LITER evaporators, which did improve the boundary conditions for the plasma but did not provide pumping of the plasma particles. Nevertheless, even with such modest implementation of the idea of the LiWalls a significant improvement of the confinement as well as ELM stabilization became evident on NSTX. The real step toward the high performance requires transition to a target divertor surface with molten lithium. The only option for NSTX, which was not designed for lithium pumping, is the Lithium Loaded Target Plate (LLTP), with inertial cooling by a copper substrate (separated from lithium by a thin stainless steel or other material foil). The talk describes the reference LiWall regime in NSTX with a 6 fold increase of the confinement time compared to its current value. It also describes the consistency of LLTP with heat and particle extraction from the plasma.