[en] One of the important goals in Columbia's HBT-EP tokamak program is the improvement in the stability of tokamak plasmas by controlling the bulk plasma flow relative to the conducting wall. The method for active plasma flow control in HBT-EP is the application of oscillating resonant magnetic perturbations to oppose the velocity of magnetic islands at the q=2 surface. Real time (10 kHz) feedback control without inserting a material probe necessitates the use of an optical toroidal rotation measurement whose data is available during the shot. This is being accomplished in a novel way by seeding the deuterium plasma with 5%--10% helium and measuring the Doppler shift of the chord-integrated emission of the HeII (n=4→3) line at 4686 Aa. Since the electron temperature is expected to be about 30 eV at the q=2 surface, helium is not fully stripped. The shift in wavelength is calculated by measuring the change in intensity as the line moves across the passband of an interference filter that varies linearly. Filters with less than 0.2% variation from a perfectly linear slope have been obtained. Fluctuations in the plasma emission are removed by having two detectors observe the same volume of plasma. This is achieved by splitting the optical view with a 'Y' composed of randomized optical fibers. One detector views the plasma through a filter whose passband has a negative slope and the other channel views it through a positive-slope filter. Systematic differences such as detector sensitivity, amplifier gain, fiber losses, etc. are compensated by normalizing each signal to the signal at a particular reference time. The ratio of signals at two different times does not depend on any detector or circuit characteristic that remains constant. The Doppler shift, relative to the reference time, is a function only of the slope of the filter's transmission. The Doppler shift at the HeII impurity emisson line is 0.25 Aa for a toroidal rotation of 3 km/s, and the slope of the filter passband is 8%--10% per Aa, resulting in a 4% variation in signal level relative to the other channel

[en] The high-speed rotation diagnostic developed for Columbia's HBT-EP tokamak requires a high quantum efficiency, very low drift detector/amplifier combination. An updated version of the circuit developed originally for the beam emission spectroscopy experiment on TFTR is being used. A low dark current (2 nA at 15 V bias), low input source capacitance (2 pF) FFD-040 N-type Si photodiode is operated in photoconductive mode. It has a quantum efficiency of 40% at the 468.6 nm (He II line that is being observed). A low-noise field-effect transistor (InterFET IFN152 with e_Na=1.2 nV/√Hz) is used to reduce the noise in the transimpedance preamplifier (A250 AMPTEK op-amp) and a very high speed (unity-gain bandwidth=200 MHz) voltage feedback amplifier (LM7171) is used to restore the frequency response up to 100 kHz. This type of detector/amplifier is photon-noise limited at this bandwidth for incident light with a power of >∼2 nW. The circuit has been optimized using SIMETRIX 4.0 SPICE software and a prototype circuit has been tested successfully. Though photomultipliers and avalanche photodiodes can detect much lower light levels, for light levels >2 nW and a 10 kHz bandwidth, this detector/amplifier combination is more sensitive because of the absence of excess (internally generated) noise

[en] The Electron Diffusion Gauge (EDG) is a Malmberg trap configuration used to investigate the application of pure electron plasmas to the measurement of background gas pressure. To form a useful gauge, the rates of different types of transport in EDG must be properly ordered and understood. The dependence of the asymmetry transport rate on plasma parameters has been determined experimentally. The relaxation of the plasma profile to a meta-equilibrium state during asymmetry transport has also been observed. A local model of the asymmetry flux consistent with the observations is presented. Implications for pressure measurements are summarized

[en] The TITAN facility at TRIUMF is a series of ion traps designed for precision mass spectrometry on rare isotopes. The combination of an on-line electron beam ion trap charge breeder with a Penning trap enables measurements with radioactive ions in high charge states. The use of highly charged ions (HCI) can yield a significant gain in mass precision and mass resolving power. However, the benefits of high charge states are mitigated since the charge breeding deteriorates the beam quality. To achieve suitable beam properties and access the full potential of Penning trap mass spectrometry with HCI a cooler Penning trap (CPET) for electron cooling of highly charged radioisotopes is being developed. In this device short-lived HCI will be sympathetically cooled by a co-trapped electron plasma prior to mass measurement. For electron plasma generation electrons are injected from an off-axis electron gun placed in the fringe field of CPET’s solenoid magnet. We report on the development of an electron gun design that is adapted to the operation in lateral magnetic fields and has enabled efficient and robust electron plasma formation in CPET.

[en] One of the key challenges for the conventional divertor tokamak is the plasma wall interaction interface and its materials. High divertor heat loads and material erosion in the spherical torus (ST) are of particular concern because of the compact divertor, and as a result, small plasma-wetted surfaces. The implications of the toroidal plasma physics at low aspect ratio and high β for edge energy and particle transport, properties of the scrape-off layer (SOL) and divertor are being studied on the National Spherical Torus Experiment (NSTX)--a medium size ST (R = 0.85 m, a = 0.67 m, A ≅ 1.27, β_t < 32 %, β_N < 5 %). NSTX operates routinely with stationary outer target plate peak heat loads up to 6 MW/m² for up to 1 s in the 6 MW NBI heated H-mode regime with type I, III, V ELMs , with the largest peak heat flux measured to date q_out = 10 MW/m² A detached divertor is an effective heat flux mitigation technique which has been developed in large aspect ratio tokamaks. Heat flux at the plate is reduced in the detached divertor through volumetric momentum and energy dissipative processes--the ion-neutral elastic collisions, recombination and radiative cooling

[en] The National Spherical Torus Experiment (NSTX) produces plasmas, with toroidal aspect ratio as low as 1.25 and plasma currents up to 1.5 MA, which can be heated by up to 6 MW High-Harmonic Fast Waves and up to 7 MW of deuterium Neutral Beam Injection. With these capabilities, NSTX has already made considerable progress in advancing the scientific understanding of high performance plasmas needed for low-aspect-ratio reactor concepts and for ITER. In transport and turbulence research on NSTX, the role of magnetic shear is being elucidated in discharges in which electron energy transport barriers are observed. Scaling studies indicate a weaker dependence on plasma current than at conventional aspect ratio and a significant dependence on toroidal field (B_T).

[en] Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4–6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width λ_q was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5–1 MW/m² (from 4–7 MW/m² in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P_rad in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

[en] Experiments in the National Spherical Torus Experiment (NSTX) have shown beneficial effects on the performance of divertor plasmas as a result of applying lithium coatings on the graphite and carbon-fiber-composite plasma-facing components. These coatings have mostly been applied by a pair of lithium evaporators mounted at the top of the vacuum vessel which inject collimated streams of lithium vapor toward the lower divertor. In neutral beam injection (NBI)-heated deuterium H-mode plasmas run immediately after the application of lithium, performance modifications included decreases in the plasma density, particularly in the edge, and inductive flux consumption, and increases in the electron and ion temperatures and the energy confinement time. Reductions in the number and amplitude of edge-localized modes (ELMs) were observed, including complete ELM suppression for periods of up to 1.2 s, apparently as a result of altering the stability of the edge. However, in the plasmas where ELMs were suppressed, there was a significant secular increase in the effective ion charge Z_eff and the radiated power as a result of increases in the carbon and medium-Z metallic impurities, although not of lithium itself which remained at a very low level in the plasma core, <0.1%. The impurity buildup could be inhibited by repetitively triggering ELMs with the application of brief pulses of an n = 3 radial field perturbation. The reduction in the edge density by lithium also inhibited parasitic losses through the scrape-off-layer of ICRF power coupled to the plasma, enabling the waves to heat electrons in the core of H-mode plasmas produced by NBI. Lithium has also been introduced by injecting a stream of chemically stabilized, fine lithium powder directly into the scrape-off-layer of NBI-heated plasmas. The lithium was ionized in the SOL and appeared to flow along the magnetic field to the divertor plates. This method of coating produced similar effects to the evaporated lithium but at lower amounts.

[en] The application of nonaxisymmetric magnetic fields is shown to destabilize edge-localized modes (ELMs) during otherwise ELM-free periods of discharges in the National Spherical Torus Experiment (NSTX). Profile analysis shows the applied fields increased the temperature and pressure gradients, decreasing edge stability. This robust effect was exploited for a new form of ELM control: the triggering of ELMs at will in high performance H mode plasmas enabled by lithium conditioning, yielding high time-averaged energy confinement with reduced core impurity density and radiated power.

[en] An overarching goal of magnetic fusion research is the integration of steady state operation with high fusion power density, high plasma β, good thermal and fast particle confinement, and manageable heat and particle fluxes to reactor internal components. NSTX has made significant progress in integrating and understanding the interplay between these competing elements. Sustained high elongation up to 2.5 and H-mode transitions during the I_p ramp-up have increased β_p and reduced l_i at high current resulting in I_p flat-top durations exceeding 0.8s for I_p>0.8MA. These shape and profile changes delay the onset of deleterious global MHD activity yielding β_N values >4.5 and β_T∼20% maintained for several current diffusion times. Higher ∫_N discharges operating above the non-wall limit are sustained via rotational stabilization of the RWM. H-mode confinement scaling factors relative to H98(y,2) span the range 1±0.4 for B_T>4kG and show a stron (Nearly linear) residual scaling with B_T. Power balance analysis indicates the electron thermal transport dominates the loss power in beam-heated H_mode discharges, but the core χ_e can be significantly reduced through current profile modification consistent with reversed magnetic shear. Small ELM regimes have been obtained in high performance plasmas on NSTX, but the ELM type and associated pedestal energy loss are found to depend sensitively on the boundary elongation, magnetic balance, and edge collisionality. NPA data and TRANSP analysis suggest resonant interactions with mid-radius tearing modes may lead to large fast-ion transport. The associated fast-ion diffusion and/or loss likely impact(s) both the driven current and power deposition profiles from NBI heating. Results from experiments to initiate the plasma without the ohmic solenoid and integrated scenario with the TSC code will also be described. (Author)