54 research outputs found
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Progress Towards High Performance, Steady-state Spherical Torus
Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} 500 kA. In parallel, start-up using radio-frequency current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor
Plasma temperature rise toward the plasma-facing surface
Detailed measurements of axial electron temperature, Te, profiles in the presheath region were carried out using a Langmuir probe and the line intensity ratio technique for both He I (728.1 nm/706.5 nm) and Be II (467.3 nm/313.1 nm). The results show that Te increases toward the material surface, which contradicts the standard picture that Te is constant along the magnetic field in the sheath-limited regime. While no target bias voltage, Vb, dependence is seen, the Te rise becomes more prominent with decreasing neutral pressure. Similarly, the ion temperature, Ti, evaluated from Doppler broadening of a He II line emission at 468.6 nm is found to increase toward the surface, but also does not depend on Vb. Possible mechanisms of the Te and Ti rise as well as validity of the line intensity ratio technique near the material surface are discussed
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The charge exchange recombination diagnostic system on the DIII-D tokamak
The charge exchange recombination (CER) diagnostic system on the DIII-D tokamak is used to make spatially and temporally resolved measurements of the ion temperature and toroidal and poloidal rotation velocities. This is performed through visible spectroscopic measurements of the Doppler broadened and Doppler shifted HE II 468.6 nm, the CVI 529.1 nm, and the BV 494.5 nm spectral lines which have been excited by charge exchange recombination interactions between the fully stripped ions and the neutral atoms from the heating beams. The plasma viewing optics comprises 32 viewing chords spanning a typical plasma minor radius of 63 cm across the midplane, of which 15 spatial chords span 4.2 cm at the plasma edge just within the separatrix and provide a chord-to-chord spatial resolution of 0.3 cm. Fast camera readout electronics can provide a temporal resolution of 260 {mu}s per time slice, but the effective minimum integration time, at present, is 1 ms which is limited by the detected photon flux from the plasma and the decay times of the phosphors used on the multichannel plate image intensifiers. Significant changes in the edge plasma radial electric field at the L-H transition have been observed, as determined from the CER measurements, and these results are being extensively compared to theories which consider the effects of sheared electric fields on plasma turbulence. 13 refs., 10 figs
Beryllium deposition on International Thermonuclear Experimental Reactor first mirrors : layer morphology and influence on mirror reflectivity
Metallic mirrors will be essential components of the optical diagnostic systems in the International Thermonuclear Experimental Reactor (ITER). Reliability of these systems may be affected by mirror reflectivity changes induced by erosion and/or deposition of impurities (carbon, beryllium). The present study aims to assess the effect of beryllium (Be) deposition on the reflectivity of metallic mirrors and to collect data on the optical quality of these layers in terms of morphology, roughness, etc. Mirrors from molybdenum and copper were exposed in the PISCES-B linear plasma device to collect eroded material from graphite and beryllium targets exposed to beryllium-seeded deuterium plasma. After exposure, relative reflectivity of the mirrors was measured and different surface analysis techniques were used to investigate the properties of the deposited layers. Be layers formed in PISCES-B exhibit high levels of porosity which makes the reflectivity of the Be layers much lower than the reflectivity of pure Be. It is found that if Be deposition occurs on ITER first mirrors, the reflectivity of the coated mirrors will strongly depend on the layer morphology, which in turn depends on the deposition conditions. (C) 2007 American Institute of Physics
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Liquid Lithium Limiter Experiments in CDX-U
Recent experiments in the Current Drive Experiment-Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, B{sub toroidal} = 2 kG, I{sub P} = 100 kA, T{sub e}(0) = 100 eV, n{sub e}(0) {approx} 5 x 10{sup 19} m{sup -3}) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm{sup 2} (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium-limited discharges are consistent with Z{sub effective} < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced
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