D. Hoffman

Operation of the NSTX Thomson scattering system

The National Spherical Torus Experiment (NSTX) multipoint Thomson scattering system has been in operation for nearly two years and provides routine Te(R,t) and ne(R,t) measurements. The laser beams from two 30 Hz Nd:YAG lasers are imaged by a spherical mirror onto 36 fiber-optic bundles. In the present configuration, the output ends of 20 of these bundles are instrumented with filter polychromators and avalanche photodiode detectors. In this article, we discuss the laser implementation and the installed collection optics. We follow with examples of raw and analyzed data, and close with some comments about calibration.

Multilayer mirror and foil filter AXUV diode arrays on CDX-U spherical torus

Recent upgrades to CDX-U spherical torus diagnostics include two 10-channel AXUV diode arrays. The multilayer mirror (MLM) array measures the λ150 O VI brightness profile in the poloidal plane using the Mo/B4C synthetic multilayer structures as dispersive elements. The foil filter array has a tangential view and is equipped with interchangeable clear aperture, beryllium and titanium filters. This allows measurements of radiated power, O VI or C V radial distributions, respectively. The O VI and C V emissivity and the radiated power profiles are highly peaked. A Neoclassical impurity accumulation mechanism is considered as an explanation. For radiated power measurements in the Te⩽100 eV plasmas, photon energy dependent corrections must be used in order to account for nonlinear AXUV sensitivity in the range Ephot⩽20 eV. The arrays are also used for characterization of resistive MHD phenomena, such as the low m modes, saw-tooth oscillations and internal reconnection events. Based on the successful operation ...

Diagnostics for liquid lithium experiments in CDX-U

A flowing liquid lithium first wall or divertor target could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls in fusion reactors. To investigate the interaction of a spherical torus plasma with liquid lithium limiters, large area divertor targets, and walls, discharges will be established in the Current Drive Experiment-Upgrade (CDX-U) where the plasma–wall interactions are dominated by liquid lithium surfaces. Among the unique CDX-U lithium diagnostics is a multilayer mirror (MLM) array, which will monitor the 13.5 nm LiIII line for core lithium concentrations. Additional spectroscopic diagnostics include a grazing incidence extreme ultraviolet (XUV) spectrometer (STRS) and a filterscope system to monitor Dα and various impurity lines local to the lithium limiter. Profile data will be obtained with a multichannel tangential bolometer and a multipoint Thomson scattering system configured to give enhanced edge resolution. Coupons on...

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 (UCSD). 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.

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, B toroidal = 2 kG, J P = 100 kA, T e (0) ∼ 100 eV, n e (0) ∼ 5 × 10 19 m -3 ) short-pulse ( < 25 ms) 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 2 . The tokamak edge plasma, when operated in contact with the lithium-filled tray, shows evidence of reduced impurities and recycling. The reduction in recycling and impurities is largest when the lithium is liquefied by heating to 250 °C. 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.

Observation of neoclassical impurity transport in Ohmically heated plasmas of CDX-U low aspect ratio tokamak

High ?, good confinement and stability properties of the low aspect ratio tokamaks, or spherical tori (ST), have been predicted theoretically and preliminarily confirmed in several large experiments recently. This paper reports on impurity transport experiments carried out in ohmically heated plasmas of the small spherical torus CDX-U with the aspect ratio of A1.5. Vacuum ultraviolet and soft x-ray multichannel spectroscopic diagnostics are used to measure intrinsic carbon, oxygen and radiated power radial brightness profiles in plasmas with Te(0)60?80?eV and ne(0)2?1013?cm?3. The measurements are performed in both magnetohydrodynamically dominated and quiescent phase of the plasmas. The properties of the observed low m/n modes, sawtooth oscillations, and ST-specific reconnection events are discussed in the context of particle transport. The measured impurity profiles are modelled using one-dimensional impurity transport code MIST and a collisional-radiative package CRMLIN. Impurity diffusion of 0.2?m2?s?1?D?0.6?m2?s?1 and convection velocity of v4?6?m?s?1 are inferred from the modelling. These transport coefficients are very close to the neoclassical theory predictions obtained with the FORCEBAL code, which uses analytical plasma viscosity expressions valid for an arbitrary aspect ratio geometry. Neoclassical analysis indicates that both carbon and oxygen are in the collisional regime, and the Pfirsch?Schluter flux is the major fraction of the impurity flux. The causes of the observed strong non-diffusive transport are discussed, and it is concluded that the ?ni/ni term, resulting from highly peaked ion density profile, makes the largest contribution to the inward pinch. Present analysis suggests that drift wave turbulence is reduced in CDX-U ohmically heated discharges within at?least r/a?0.4, however more refined measurements are needed to interpret the results in the framework of ST ion transport.

Effects of Large Area Liquid Lithium Limiters on Spherical Torus Plasmas

Use of a large-area liquid lithium surface as a first wall has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter.

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.

A toroidal liquid lithium limiter for CDX-U

Attention has focused recently on flowing liquid lithium as a first wall for a reactor because of its potentially attractive physics and engineering features. In order to test the suitability of liquid lithium as a plasma facing component, the Current Drive eXperiment - Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has recently installed a fully toroidal liquid lithium limiter. CDX-U is a compact (R = 34 cm, a = 22 cm, B/sub toroidal/ = 2 kG, I/sub p/ =100 kA, T/sub e/(O) /spl sim/ 100 eV, n/sub e/(0) /spl sim/ 5 /spl times/ 10/sup 19/ m/sup -3/ short-pulse (< 25 msec) spherical torus (ST) with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, is 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 (melting point = 180.5/spl deg/C) prior to the experiment. The total area of liquid lithium exposed to the plasma is approximately 2000 cm/sup 2/. The design of the limiter, modifications to CDX-U to accommodate in-vessel inventories of approximately 1 liter of liquid lithium, techniques for loading lithium onto the limiter, and other preparations will be described. CDX-U has previously been successfully operated with a smaller area cm/sup 2/) liquid lithium rail limiter. Diagnostics specific to lithium operations include multichord spectrometry of the 135 /spl Aring/ LiIII line in the core plasma, monitors for neutral lithium light at the lithium limiter, and a fast (10,000 frame per second) camera which monitors motion of the liquid during the discharge. First results of plasma operations with the toroidal liquid lithium limiter will also be given.

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.