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

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

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

Monte Carlo impurity transport modeling in the DIII-D transport

A description of the carbon transport and sputtering physics contained in the Monte Carlo Impurity (MCI) transport code is given. Examples of statistically significant carbon transport pathways are examined using MCI`s unique tracking visualizer and a mechanism for enhanced carbon accumulation on the high field side of the divertor chamber is discussed. Comparisons between carbon emissions calculated with MCI and those measured in the DIII-D tokamak are described. Good qualitative agreement is found between 2D carbon emission patterns calculated with MCI and experimentally measured carbon patterns. While uncertainties in the sputtering physics, atomic data, and transport models have made quantitative comparisons with experiments more difficult, recent results using a physics based model for physical and chemical sputtering has yielded simulations with about 50% of the total carbon radiation measured in the divertor. These results and plans for future improvement in the physics models and atomic data are discussed

Quantitative comparisons between experimentally measured 2-D carbon radiation and Monte Carlo impurity (MCI) code simulations

Experimentally measured carbon line emissions and total radiated power distributions from the DIII-D divertor and Scrape-Off Layer (SOL) are compared to those calculated with the Monte Carlo Impurity (MCI) model. A UEDGE background plasma is used in MCI with the Roth and Garcia-Rosales (RG-R) chemical sputtering model and/or one of six physical sputtering models. While results from these simulations do not reproduce all of the features seen in the experimentally measured radiation patterns, the total radiated power calculated in MCI is in relatively good agreement with that measured by the DIII-D bolometric system when the Smith78 physical sputtering model is coupled to RG-R chemical sputtering in an unaltered UEDGE plasma. Alternatively, MCI simulations done with UEDGE background ion temperatures along the divertor target plates adjusted to better match those measured in the experiment resulted in three physical sputtering models which when coupled to the RG-R model gave a total radiated power that was within 10% of measured value