Experimental and Theoretical Investigation of Synergy between Ion Bernstein and Lower Hybrid Waves in PBX-M

The synergistic behavior of lower hybrid and ion Bernstein waves on the Princeton Beta Experiment-Modified tokamak [Phys. Fluids B 2, 1271 (1990)] is experimentally studied using a 2-D hard X-ray camera. The hard X-ray bremsstrahlung emission from suprathermal electrons, generated with lower hybrid current drive, is enhanced during ion Bernstein wave power injection. This enhancement is observed in limited regions of space suggesting the formation of localized current channels. The effects on plasma electrons during combined application of these two types of waves are theoretically investigated using a quasilinear model. The numerical code simultaneously solves the 3-D (R, Z, {Phi}) toroidal wave equation for the electric field (in the WKBJ approximation) and the Fokker-Planck equation for the distribution function in two dimensions (v{sub parallel}, v{sub perpendicular}) with an added quasilinear diffusion coefficient. The radial profile of the non-inductively generated current density, the transmitted power traces and the total power damping curve are calculated. The beneficial effects of a combined utilization of ion Bernstein and lower hybrid waves on the current drive are emphasized. The numerical results are compared with the experimental observations

Fusion In The Era Of Burning Plasma Studies: Workforce Planning For 2004-2014

This is the final report of a panel set up by the U.S. Department of Energy (DOE) Fusion Energy Sciences Advisory Committee (FESAC) in response to a charge letter from Dr. Raymond Orbach (Appendix A), asking FESAC to addressed the issue of workforce development in the U.S. fusion program. This report, submitted to FESAC March 29, 2004 and subsequently approved by them (Appendix B), presents FESAC\u27s response to that charge

An Overview of Science Education and Outreach Activities at the Princeton Plasma Physics Laboratory

The U. S. Department of Energy's Princeton Plasma Physics Laboratory (PPPL) has an energetic science education program and outreach effort. This overview describes the components of the programs and evaluates the changes that have occurred in this effort during the last several years. Efforts have been expanded to reach more students, as well as the public in general. The primary goal is to inform the public regarding the fusion and plasma research at PPPL and to excite students so that they can appreciate science and technology. A student's interest in science can be raised by tours, summer research experiences, in-classroom presentations, plasma expos, teacher workshops and web-based materials. The ultimate result of this effort is a better-informed public, as well as an increase in the number of women and minorities who choose science as a vocation. Measuring the results is difficult, but current metrics are reviewed. The science education and outreach programs are supported by a de dicated core group of individuals and supplemented by other members of the PPPL staff and consultants who perform various outreach and educational activities

Diode laser excited optogalvanic spectroscopy of glow discharges

The development of diode-laser-excited isotopically-selective optogalvanic spectroscopy (OGS) of uranium metal, oxide and fluoride in a glow discharge (GD) is presented. The technique is useful for determining isotopic ratios of {sup 235}U/({sup 235}U + {sup 238}U) in the above samples. The precision and accuracy of this determination is evaluated, and a study of experimental parameters pertaining to optimization of he measurement is discussed. Application of the GD-OGS to other f-transition elements is also described

Spherical Torus Plasma Interactions with Large-area Liquid Lithium Surfaces in CDX-U

The Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego. Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance

Testing of Liquid Lithium Limiters in CDX-U

Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described