Session CO3 - NSTX and Other ST.
ORAL session, Monday afternoon, November 15
Room 204/205, SCC

[CO3.001] Plasma Studies at High Normalized Current in the \sc Pegasus Experiment

Operations on the \sc Pegasus Toroidal Experiment are focusing on the exploration of MHD stability limits at high normalized current, high \beta_t, and near-unity aspect ratio. Past plasma behavior was dominated by large resistive MHD modes and power supply limits. Recent upgrades will allow access to the I_p/I_tf > 1 regime by suppressing the resistive MHD. Tools to vary the q-profile and suppress the MHD activity include significantly increased poloidal field control, increased, time-variable toroidal field, and loop voltage programmability. Modeling with the DCON code has demonstrated that stable equilibria exist with I_p/I_tf approaching 3. A complete set of magnetics diagnostics, combined with simple plasma parameter measurements and eventually an imaging SXR camera system for poloidal flux surface mapping, provides data for reasonably accurate equilibrium reconstructions. The present experimental campaign is focused on the suppression of the internal tearing modes to allow access to the external kink limits at high I_N.

[CO3.002] Extending high-beta operation in NSTX through plasma shaping and early H-mode transitions

The National Spherical Torus Experiment (NSTX) resumed operation in February 2004 and completed in July a program of experiments which expanded its operating space and contributed to resolving issues both for the design of future ST devices and in toroidal magnetic confinement physics. Improvements in plasma control have led to the achievement of higher plasma elongation, \kappa = 2.6 and normalized current, I_N = I_p/aB_T = 6.9 MA/m \cdot T. By optimizing the current ramp-up, gas puffing and neutral beam power waveforms, reliable early transitions to the H-mode have been obtained, which improved confinement and broadened the pressure profiles, permitting the achievement of very high beta, \beta_T = 34%, determined by EFIT from external magnetic and kinetic profile data and corroborated by analysis with TRANSP. These improvements have led to the achievement of 1MA pulses with 1s duration that were not limited by the available transformer flux. First measurements have also been obtained with the MSE diagnostic. The features of the NSTX operating space and characteristics of the high-\beta plasmas will be discussed. This work supported by DOE Contract Number DE-AC02-76CH03073

[CO3.003] Confinement in high-\beta, NB-heated H- and L-mode plasmas in NSTX

The National Spherical Torus Experiment (NSTX) has completed a program of experiments which spanned significant ranges for the dominant engineering variables in the scaling of energy confinement in conventional tokamaks. The experiments included well controlled scans in H-mode plasmas of the plasma current and neutral beam heating power. The energy confinement time has been determined by equilibrium analysis based on external magnetic measurements and confirmed for a subset of discharges by analysis of the kinetic measurements with the TRANSP code. The variation in the global energy confinement time with current and heating power matches quite closely the dependences in the ITER-98pb(y,2) H-mode scaling both in the single-parameter scans and regression to the overall database. Nevertheless, for neutral beam heated H-mode plasmas, the global confinement exceeds the H-mode scaling predictions by a factor up to 1.6. In L-mode plasmas the confinement often exceeds the H-mode scaling predictions as well. These trends and the results of the TRANSP analysis will be presented.

[CO3.004] Electron Transport in NSTX

The high harmonic fast wave heated plasmas in NSTX are ideal plasmas for studying electron confinement regimes. Energy confinement times twice that of the ITER-L97 confinement law are produced at high electron plasma pressure-to-magnetic pressure ratios. Peaked density and temperature L-mode type profiles showing good particle confinement with low eta-e values as well as electron ITB discharges in low or weakly reversed magnetic shear profiles are analyzed for transport using standard drift wave transport models. The ETG mode gives the best interpretation of the heat flux versus gradient relationship. Although the trapped electron fraction is high, the trapped electron modes are suppressed by the high plasma pressure and high MHD alpha-factor that produces favorable bounce averaged guiding center drifts among other stabilizing effects. We show nonlinear simulations with specified heating power and edge losses that explain the inverse cascade in the ETG spectrum and produce thermal diffusivity profiles similar to the TRANSP power balance profiles.

Work supported by the Department of Energy Basic Energy Sciences.

[CO3.005] High Performance Regime with Small ELMs in NSTX

A high performance regime with small edge-localized modes (ELMs) has been observed in the National Spherical Torus Experiment (NSTX). The ELMS (“type V”) are consistent with high bootstrap current operation with line average density approaching Greenwald scaling, and provide enhanced edge particle transport compared with ELM-free discharges. Each individual ELM has no measurable impact on stored energy, The ELM perturbation is observed via ultra-soft X-rays (USXR) to typically originate near the lower divertor region and propagate with a poloidal component toward the outer midplane and often to the top of the machine. An electromagnetic signature is observed before the USXR perturbation on separate poloidal and toroidal Mirnov arrays, indicating a toroidal mode number of n=1, propagation counter to the plasma current, and persistence for up to two toroidal transit times. These type V ELMs are distinct from both type I and type III ELMs, which have also been observed in NSTX, and can co-exist with type I ELMs. When the type V ELMS are present, there is no evidence of an edge harmonic oscillation as observed in DIII-D quiescent H-mode discharges, or of an edge quasi-coherent mode observed in the PDX forced density rise scenario and the C-MOD Enhanced D-alpha H-mode scenario.

[CO3.006] Solenoid free plasma start-up in NSTX by Coaxial Helicity Injection

A new method recently developed on HIT-II, referred to as transient CHI, has unambiguously demonstrated the presence of a robust closed flux equilibrium and central transformer volt-seconds savings when coupled to induction, showing that the CHI produced plasmas are not only compatible with conventionally produced inductive plasmas but that they also improve the performance of these inductive plasmas [R.Raman et al., Phys. Rev. Lett. 90, 075005 (2003)]. In this method the plasma is rapidly grown through the use of an appropriately sized capacitor bank. The initial static magnetic flux is chosen such that this rapidly expanding plasma has a tendency to detach from the lower divertor electrodes. Rapidly reducing the injector current during this phase eases the detachment process, which results in the production of a robust closed field line equilibrium. Results from these experiments on NSTX will be reported. DOE Contract No. DE-AC02-76CH03073 and DE-FG03-99ER54519 supports this work.

[CO3.007] Plasma formation without a central solenoid in a Spherical Tokamak

Obtaining the plasma current in any ST is a challenge as there is limited space for the central solenoid; indeed this is absent in most proposed ST fusion devices due to the high neutron flux, there being no space for an effective shield. A range of start-up techniques will be described, with special emphasis on the Merging-Compression (M-C) method of plasma formation used on START and MAST, and a `Double Null Merging' scheme whereby two initial plasmas formed in low-order nulls between Poloidal Field coils are merged. Results from M-C confirm that high temperature plasmas are obtained from merging. Results and/or modelling indicate that both techniques can produce hot initial plasmas of 0.5MA in MAST without use of the central solenoid, in readiness for subsequent ramp-up using high power NBI and the associated input of flux (significant in an ST) from the increasing vertical field

[CO3.008] Toroidal Rotation Damping Physics in NSTX

Spherical torus and tokamak experiments show that the combination of sufficient plasma rotation and a resistive wall can stabilize ideal kink/ballooning modes. However, toroidal rotation damping has been observed at sufficiently high plasma beta, eliminating passive stabilization. Rapid and global damping at high beta above the ideal MHD no-wall limit experimentally expedites mode locking and plasma beta collapse. ELMs lead to edge rotation damping that is typically insignificant. Resonant electromagnetic drag may explain rotation damping for resistive modes with small island width, while non-resonant neo-classical toroidal viscous drag is in quantitative agreement with measured global damping in resistive wall mode, 1/1 mode, and local damping in ELMing plasmas. A pair of non-axisymmetric coils was used to generate an n=1 magnetic field perturbation which allowed greater rotation control. Quantitative comparison is made between theory and experiment to determine the physics and parameter dependence of the damping evolution for the observed instabilities.

[CO3.009] Neoclassical tearing mode onset in NSTX

Neoclassical theory predicts well the time evolution of experimentally observed saturated tearing modes. The theory also predicts that many more modes are non-linearly unstable than are seen. The theory does less well in predicting which modes are either linearly unstable or triggered by other events in the plasma. Experiments have been performed in NSTX to identify the parametric dependences of NTM onset. In the first experiment, repetitive sawteeth were generated in plasmas with different beta and plasma current and the growth (damping) rate for the NTMs was determined. In the second. Similar, experiment, repetitive fast ion driven chirping modes (fishbones) were used as the NTM trigger. For all shots, the NTM parameters were evaluated using TRANSP, including the calculation of the neoclassical (bootstrap) current fraction. Diagnostics for mode detection included soft x-ray cameras and Mirnov coils.

[CO3.010] Edge/SOL characterization in the NSTX spherical tokamak

The edge/SOL of NSTX has been investigated with a high temporal (1 ?s) and spatial (1.5 mm) resolution probe array in L-mode and H-mode discharges. The decay lengths of the density and temperature profiles in the SOL in L-mode are 4 cm and 12 cm respectively, independently of power. In H-mode the profile decay lengths are reduced considerably to 1-2 cm. The profile decay lengths shorten significantly to ~2 cm in H-mode. Features of interest are: 1) The Te SOL profile is flat at \sim20 eV and highly fluctuating and, 2) the Te pedestal is shifted inwards from the Ne pedestal by \sim3 cm, indicating decoupled energy and particle transport. Strong intermittency in the edge/SOL has been observed and quantified using conditional averaging. UEDGE modeling is in progress.

[CO3.011] Edge Turbulence Imaging During L-H Transitions in NSTX

High speed movies of the radial vs. poloidal structure of edge turbulence during L-H and H-L transitions in NSTX will be shown and discussed. These images were made in D-alpha light at up to 250,000 frames/sec with a 300 frame, 64x64 pixel ultra-high speed PSI-5 camera . Transitions from L-mode to H-mode appear as a continuous evolution from a turbulent state to a quiescent state over about 0.1 ms, apparently without any new spatial features or flows. Transitions from H-L generally appear as high-n poloidal perturbations which evolve into radially moving blobs. Analysis of this data is being done to extract local velocity fields to estimate shear and zonal flows. Comparison of these images with data from the UCLA reflectometers will be discussed. Movies of ELMs and other MHD effects will also be shown if time allows.

[CO3.012] High Harmonic Fast Wave Heating on NSTX- I

Previous experiments on NSTX have demonstrated that HHFW heats electrons and drives current as expected from theory. Some anomalies have been observed, including edge ion heating and apparent reduced heating and current drive under some conditions. A series of detailed experiments to quantify these observations has been undertaken. The edge heating is believed to be due to parametric decay of the applied fast wave into ion Bernstein waves. Peaks in the frequency spectrum of an edge langmuir probe corresponding to these waves have been observed. The number of peaks and their magnitude increases with rf power. The incremental energy confinement time and percent of rf power absorbed as well as the radial power deposition profile for a variety of plasmas and antenna phasing has been explored utilizing amplitude modulated rf power waveforms. Differences in these quantities between deuterium and helium plasmas and with varying applied wave spectra have been found and will be compared to macroscopic plasma parameters and theory.

[CO3.013] High Harmonic Fast Wave Heating on NSTX- II

HHFW experiments have been carried out in both RF-heated plasmas and RF+NBI-heated plasmas. In RF-only plasmas, the heating efficiency was greatest for 14 m^-1, decreased as the wavenumber was decreased to 7 m^-1, and was minimal for 3 m^-1 operation. The decrease in heating efficiency follows increasingly centralized power deposition. H-mode was obtained for double-null diverted (DND) plasmas for symmetric spectra at both 14 m^-1 and 7 m^-1. Current drive measurements were obtained from the differences in loop voltages for co- and counter-current directions for the same plasma conditions. The normalized CD efficiency obtained at wavenumbers of ±7 m^-1 in DND He plasmas was similar to those previously obtained in D and He lower single-null diverted plasmas at lower densities. CD attempted at –3 m^-1 (after preheating the plasma to Te(0) = 2 keV with 3 MW at 7 m^-1) was unsuccessful due to rapid collapse of the temperature and stored energy. A fraction of the HHFW power is absorbed by fast particles during combined operation with NBI.

[CO3.014] Progress in EBW Current Drive Research Towards Enabling Sustained High Beta, Solenoid-Free Operation on NSTX

Off-axis rf-driven current can play a critical role in sustaining high beta spherical torus (ST) plasmas without a central solenoid. Numerical modeling of electron Bernstein wave current drive (EBWCD) for a beta \sim 40% ST plasma predicts efficient, EBWCD at r/a > 0.5 where the large trapped electron fraction provides favorable conditions for Ohkawa EBWCD. Normalized EBWCD efficiency increases with r/a and is a factor of two higher at r/a = 0.7 than has been obtained with ECCD near the axis of large aspect ratio Tokamaks. Calculations predict 3-4 MW of 28 GHz RF power would maintain the \sim 100 kA off-axis current needed to stabilize a beta \sim 40% plasma. Modeling shows that efficient coupling to EBWs can be readily accomplished using near-circularly polarized electromagnetic waves via an ``O-X-B" oblique launch to the confining magnetic field. Initial measurements of EBW emission support pursuing this approach.

[CO3.015] Ideal MHD stability of the Component Test Facility and Culham Spherical Tokamak Power Plant

As well as complementing the physics base for ITER, theory and progress in Spherical Tokamak (ST) experiments have prompted design studies into next step ST devices such as the Component Test Facility (CTF) [1], and ST power plants such as the 3GW (fusion) / 1GW (electric) Culham ST Power Plant (STPP) [2]. In this study, low toroidal wave number (n=0,1,2,3) and ballooning (n=\infty) ideal MHD stability of the CTF and STPP concepts is addressed: for low n modes scans of growth rate with conformal wall radius and on-axis safety factor q_0 are performed, whilst for ballooning modes the proximity to the marginally stable pressure-driven boundaries assessed. Stable equilibria are reported in both configurations, consolidating the viability of the ST line. Time permitting, the effects of toroidal flow will be qualitatively investigated, and the role of fast particles discussed.

[1] H. R. Wilson et al, EPS Conf. Plasma Phs., P4-196. 2004. [2] H. R. Wilson et al, in-press, Nuc. Fus. 2004.

Part C of program listing