Current Drive Experiment-Upgrade Research Articles

Experiments with liquid metal walls: Status of the lithium tokamak experiment

Abstarct Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The lithium tokamak experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the current drive experiment-upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in ohmically heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy (CHERS). Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions.

Extremely low recycling and high power density handling in CDX-U lithium experiments

The mission of the Current Drive eXperiment-Upgrade (CDX-U) spherical tokamak is to investigate lithium as a plasma-facing component (PFC). The latest CDX-U experiments used a combination of a toroidal liquid lithium limiter and lithium wall coatings applied between plasma shots. Recycling coefficients for these plasmas were deduced to be 30% or below, and are the lowest ever observed in magnetically-confined plasmas. The corresponding energy confinement times showed nearly a factor of six improvement over discharges without lithium PFC’s. An electron beam (e-beam) for evaporating lithium from the toroidal limiter was one of the techniques used to create lithium wall coatings in CDX-U. The evaporation was not localized to the e-beam spot, but occurred only after the entire volume of lithium in toroidal limiter was liquefied. This demonstration of the ability of lithium to handle high heat loads can have significant consequences for PFC’s in future burning plasma devices.

Magnetic probe response function calibrations for plasma equilibrium reconstructions of CDX-U

A novel response function calibration technique has been developed to account for time-dependent nonaxisymmetric eddy currents near magnetic sensors in toroidal magnetic confinement devices. The response function technique provides a means to cross calibrate against all available external field coil systems to calculate the absolute sensitivity of each magnetic field sensor, even when induced eddy currents are present in the vacuum vessel wall. The response function information derived in the calibration process can be used in equilibrium reconstructions to separate plasma signals from signals due to externally produced eddy currents at magnetic field sensor locations, without invoking localized wall current distribution details. The response function technique was used for the first ever equilibrium reconstructions of spherical torus plasmas, when applied to the Current Drive Experiment-Upgrade (CDX-U) device. In conjunction with the equilibrium and stability code (ESC), equilibria were obtained for recent CDX-U experiments with lithium plasma-facing components. A description of the CDX-U magnetic sensor configuration and the response function calibration technique will be presented along with examples of resulting plasma equilibrium for CDX-U lithium wall operations.

Fast gas injection as a diagnostic technique for particle confinement time measurements

The determination of the effective particle confinement time (τp*), i.e., the particle confinement time normalized to recycling coefficient, is difficult when its value is long compared to the discharge duration in magnetically confined plasmas. Recent experiments on the current drive experiment upgrade (CDX-U) spherical torus have successfully achieved a significant reduction in recycling with large-area liquid lithium plasma-facing surfaces. The low recycling walls result in an increase in particle pumping and make it possible to measure τp* in short duration plasmas. Measurements of τp* are made using a supersonic gas injector which is closely coupled to plasma. A fast gas pulse is emitted from the supersonic gas injector, after which the density decay is measured using a microwave interferometer. The design of the supersonic gas injector and its configuration on CDX-U will be presented. The results of this technique will be shown as applied to the study of the effects of a liquid lithium toroidal limiter and evaporative lithium coatings on overall plasma density and τp*.

Recent liquid lithium limiter experiments in CDX-U

Recent experiments in the Current Drive Experiment-Upgrade (CDX-U) 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, Btoroidal = 2 kG, IP = 100 kA, Te(0)∼ 100 eV, ne(0) ∼ 5 × 1019 m−3) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium pool limiter with an area of 2000 cm2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium pool limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5–8 increase in gas fuelling to achieve a comparable density, indicating that recycling is strongly reduced. Modelling of the discharges demonstrated that the lithium limited discharges are consistent with Zeffective < 1.2 (compared with 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.

Effects of large area liquid lithium limiters on spherical torus plasmas

Use of a large-area liquid lithium surface as a limiter 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 been 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.

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, I 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.

CDX-U operation with a large area liquid lithium limiter

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, I P=100 kA, T e(0)∼100 eV, n e(0)∼5×10 19 m −3) short-pulse (<25 ms) spherical torus 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. Surface coatings are evident on part of the lithium. Despite the surface coatings, tokamak discharges operated in contact with the lithium-filled tray show evidence of reduced impurities and recycling. The reduction in recycling is largest when the lithium is liquefied.

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–lithium interaction in the CDX-U spherical torus

Results on the interaction between plasma in the current drive experiment-upgrade (CDX-U) spherical torus and a liquid lithium limiter are reported. It is observed that macroscopic lithium droplets detach from the limiter head and fall towards the plasma core. However, no disruptions occurred during these discharges despite the fact that relatively large-scale blobs are observed entering the confined plasma. A multi-tip Langmuir probe measures the edge plasma properties. It is found that the average density and temperature and their fluctuations are unaffected by the presence of lithium within experimental error.

Transient transport experiments in the current-drive experiment upgrade spherical torus

Electron transport has been measured in the Current-Drive Experiment Upgrade (CDX-U) (T. Jones, Ph.D. thesis, Princeton University, 1995) using two separate perturbative techniques. Sawteeth at the q=1 radius (r/a∼0.15) induced outward-propagating heat pulses, providing time-of-flight information leading to a determination of χe as a function of radius. Gas modulation at the plasma edge introduced inward-propagating cold pulses, providing a complementary time-of-flight based χe profile measurement. This work represents the first localized measurement of χe in a spherical torus. Core (r/a&amp;lt;1/3) χe values from the sawtooth study are 1–2 m2/s, and from the gas modulation study are 1–6 m2/s, increasing by an order of magnitude or more outside of the core region. Furthermore, the χe profile exhibits a sharp transition near r/a=1/3. Spectral and profile analyses of the soft x-rays, scanning interferometer, and edge probe data show no evidence of a significant magnetic island causing the high χe region. Comparisons are performed to several theoretical models, with measured χe≈5–10× neoclassical estimates in the core.

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 the inner wall of the CDX-U vacuum vessel will be used for surface analysis. A 10 000 frame per second fast visible camera and an IR camera will also be available.

High-harmonic fast magnetosonic wave coupling, propagation, and heating in a spherical torus plasma

A novel rotatable two-strap antenna has been installed in the current drive experiment upgrade (CDX-U) [T. Jones, Ph.D. thesis, Princeton University (1995)] in order to investigate high-harmonic fast wave coupling, propagation, and electron heating as a function of strap angle and strap phasing in a spherical torus plasma. Radio-frequency-driven sheath effects are found to fit antenna loading trends at very low power and become negligible above a few kilowatts. At sufficiently high power, the measured coupling efficiency as a function of strap angle is found to agree favorably with cold plasma wave theory. Far-forward microwave scattering from wave-induced density fluctuations in the plasma core tracks the predicted fast wave loading as the antenna is rotated. Signs of electron heating during rf power injection have been observed in CDX-U with central Thomson scattering, impurity ion spectroscopy, and Langmuir probes. While these initial results appear promising, damping of the fast wave on thermal ions at high ion-cyclotron-harmonic number may compete with electron damping at sufficiently high ion β—possibly resulting in a significantly reduced current drive efficiency and production of a fast ion population. Preliminary results from ray-tracing calculations which include these ion damping effects are presented.

Low velocity boron micro-pellet injector for edge and core impurity transport measurements

A simple low velocity boron micro-pellet injector has been under development for Current Drive Experiment Upgrade (CDX-U) spherical torus edge and core impurity transport measurements, and wall conditioning. The injector consists of 16 barrels on a rotatable turret. Each barrel can be loaded with boron powder particles of diameters ranging from 1 to 40 μm diameter in amounts ranging from less than 0.25 mg to more than 2 mg. A selected barrel is manually rotated into firing position using a vacuum precision rotary/linear motion feedthrough. A piezoelectric valve gas feed system triggered by CDX-U discharge timing is used to control H2 or D2 propellant gas at a cylinder pressure of 5.8×10−3 Pa (40 psi) or less. The injector barrel-to-CDX-U plasma edge distance is 0.47 m. Initial low mass injections of neutral boron beams were performed into CDX-U plasmas at a velocity of 23 m/s. Measurements were obtained with a filtered gated charge coupled device TV camera, bolometry, visible spectroscopy, and ultrasoft x-ray diagnostics. This work is in support of the present CDX-U research program and possible applications on National Spherical Torus Experiment.

New electron cyclotron emission diagnostic for measurement of temperature based upon the electron Bernstein wave

Most magnetically confined plasma devices cannot take advantage of standard electron cyclotron emission (ECE) diagnostics to measure temperature. They either operate at high density relative to their magnetic field (e.g., ωp≫Ωc in spherical tokamaks) or they do not have sufficient density and temperature to reach the blackbody condition (τ&amp;gt;2). The standard ECE technique measures the electromagnetic waves emanating from the plasma. Here we propose to measure electron Bernstein waves (EBW) to ascertain the local electron temperature in these plasmas. The optical thickness of EBW is extremely high because it is an electrostatic wave with a large ki. For example, the National Spherical Torus Experiment (NSTX) will have an optical thickness τ≈3000 and CDX-U will have τ≈300. One can reach the blackbody condition with a plasma density ≈1011 cm−3 and Te≈1 eV. This makes it attractive to most plasma devices. The serious issue with using EBW is the wave accessibility for the emission measurement. Simple accessibility arguments indicate the wave may be accessible by either direct coupling or mode conversion through an extremely narrow layer (≈1–2 mm). EBW experiments on the Current Drive Experiment-Upgrade (CDX-U) will test the accessibility properties of the spherical tokamak configuration.

Investigation of the formation of a fully pressure-driven tokamak*

A noninductive current drive concept, based on internal pressure-driven currents in a low-aspect-ratio toroidal geometry, has been demonstrated on the Current Drive Experiment Upgrade (CDX-U) [Forest et al., Phys. Rev. Lett. 68, 3559 (1992)] and further tested on DIII-D [in Plasma Physics and Controlled Nuclear Fusion Research, 1986, Proceedings of the 11th International Conference, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159]. For both experiments, electron cyclotron power provided the necessary heating to breakdown and maintain a plasma with high-βp and low collisionality (εβp∼1, ν*≤1). A poloidal vacuum field similar to a simple magnetic mirror is superimposed on a much stronger toroidal field to provide the initial confinement for a hot, trapped electron species. With application of electron cyclotron heating (ECH), toroidal currents spontaneously flow within the plasma and increase with applied ECH power. The direction of the generated current is independent of the toroidal field direction and depends only on the direction of the poloidal field, scaling inversely with magnitude of the later. On both CDX-U and DIII-D, these currents were large enough that stationary closed flux surfaces were observed to form with no additional Ohmic heating. The existence of such equilibria provides further evidence for the existence of some type of bootstrap current. Equilibrium reconstructions show the resulting plasma exhibits properties similar to more conventional tokamaks, including a peaked current density profile which implies some form of current on axis or nonclassical current transport.