Implications of TAE modes for the design of ITER

A simple mixing-length estimate of diffusion of alphas particles by toroidicity-induced shear Alfven eigenmodes (TAE) is used, in zero and one-dimensional models, to evaluate the importance of diffusion of meeting ignition requirements for ITER and other next-generation burning plasma experiments. It is found that, depending on a number of assumptions, diffusion could reduce that effectiveness of alpha heating in the core as much as an order of magnitude. However, the effect would be less if only alphas resonant with the Alfven waves diffuse. Also, in the Appendix it is argued that the mixing length diffusion formula, though qualitatively reasonable, may be an over estimate. 12 refs., 7 figs., 1 tab

Estimates if population inversion for deep-UV transitions in Kr-like Y,Zr,Nb and Mo in a high-current reflex discharge

Kr-like ions are good candidates for FUV lasing since they can be produced in plasmas quite easily. We present results from a spectroscopic investigation of Y IV emission from a high current density, cold cathode reflex discharge. The Y II to Y V emission is recorded in the 200-3000 {angstrom} range using photometrically calibrated spectrometers, while the emission of trace aluminum ions serves for plasma diagnostics. The intensities of the Y IV 4d - 5p and 5s - 5p transitions strongly increase relative to lines from Y II and Y III with increasing plasma current. The spectra studied here are obtained at a current density of 1.75 A/cm{sup 2}. Experimental Y IV intensity ratios spanning several excited configurations are compared with collisional radiative predictions of the HULLAC atomic physics package. Good agreement is found for the measured and predicted ratios of 4p{sup 5}5p to 4p{sup 5}5s level populations per statistical weight. Finally, the response of the Kr-like system to a fast, transient excitation pulse is examined using the RADEX code. Large transient gains are predicted for several 5s - 5p transitions in Y IV, Zr V, Nb VI and Mo VII

Helium transport and exhaust studies in enhanced confinement regimes in DIII-D

A better understanding of helium transport in the plasma core and edge in enhanced confinement regimes is now emerging from recent experimental studies on DIII-D. Overall, the results are encouraging. Significant helium exhaust ({tau}*{sub He}/{tau}{sub E} {approximately} 11) has been obtained in a diverted, ELMing H-mode plasma simultaneous with a central source of helium. Detailed analysis of the helium profile evolution indicates that the exhaust rate is limited by the exhaust efficiency of the pump ({approximately}5%) and not by the intrinsic helium transport properties of the plasma. Perturbative helium transport studies using gas puffing have shown that D{sub He}/X{sub eff}{approximately}1 in all confinement regimes studied to date (including H-mode and VH-mode). Furthermore, there is no evidence of preferential accumulation of helium in any of these regimes. However, measurements in the core and pumping plenum show a significant dilution of helium as it flows from the plasma core to the pumping plenum. Such dilution could be the limiting factor in the overall removal rate of helium in a reactor system

Plasma Performance Improvements with Liquid Lithium Limiters in CDX-U

The use of flowing liquid lithium as a first wall for a reactor has potentially attractive physics and engineering features. The Current Drive experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has begun experiments with a fully toroidal liquid lithium limiter. CDX-U is a compact [R = 34 cm, a = 22 cm, Btoroidal = 2 kG, IP =100 kA, T(subscript)e(0) {approx} 100 eV, n(subscript)e(0) {approx} 5 x 10{sup 19} m-3] short-pulse (<25 msec) spherical tokamak with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, can be filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium prior to the experiment. The total area of the tray is approximately 2000 cm{sup 2}. The tokamak edge plasma, when operated in contact with the lithium-filled tray, shows evidence of reduced impurities and recycling. The reduction in re cycling and impurities is largest when the lithium is liquefied by heating to 250 degrees Celsius. Discharges which are limited by the liquid lithium tray show evidence of performance enhancement. Radiated power is reduced and there is spectroscopic evidence for increases in the core electron temperature. Furthermore, the use of a liquid lithium limiter reduces the need for conditioning discharges prior to high current operation. The future development path for liquid lithium limiter systems in CDX-U is also discussed

Recent Progress on the National Spherical Torus Experiment (NSTX)

Recent upgrades to the NSTX facility have led to improved plasma performance. Using 5MW of neutral beam injection, plasmas with toroidal {beta}{sub T} (= 2{micro}{sub 0}<p>/B{sub T}{sup 2} where B{sub T} is the vacuum toroidal field at the plasma geometric center) > 30% have been achieved with normalized {beta}{sub N} (= {beta}{sub T}aB{sub I}/I{sub p}) {approx} 6% {center_dot} m {center_dot} T/MA.. The highest {beta} discharge exceeded the calculated no-wall {beta} limit for several wall times. The stored energy has reached 390kJ at higher toroidal field (0.55T) corresponding to {beta}{sub T} {approx} 20% and {beta}{sub N} = 5.4. Long pulse ({approx}1s) high {beta}{sub p} ({approx}1.5) discharges have also been obtained at higher {beta}{sub {phi}} (0.5T) with up to 6MW NBI power. The highest energy confinement times, up to 120ms, were observed during H-mode operation which is now routine. Confinement times of {approx}1.5 times ITER98pby2 for several {tau}{sub E} are observed during both H-Mode and non-H-Mode discharges. Calculations indicate that many NSTX discharges have very good ion confinement, approaching neoclassical levels. High Harmonic Fast Wave current drive has been demonstrated by comparing discharges with waves launched parallel and anti-parallel to the plasma current