T. Gray

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 × 10 19 m −3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium pool limiter with an area of 2000 cm 2 (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.

Performance projections for the lithium tokamak experiment (LTX)

Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5-10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fuelling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4-6 ms), loop voltage (<0.5 V), and density for a typical CDX-U lithium discharge. The same transport model has also been used to project the performance of the LTX, in Ohmic operation, or with modest neutral beam injection (NBI). NBI in LTX, with a low recycling wall of liquid lithium, is predicted to result in core electron and ion temperatures of 1-2 keV, and energy confinement times in excess of 50 ms. Finally, the unique design features of LTX are summarized.

Low recycling and high power density handling physics in the Current Drive Experiment-Upgrade with lithium plasma-facing components

The Current Drive Experiment-Upgrade [T. Munsat, P. C. Efthimion, B. Jones, R. Kaita, R. Majeski, D. Stutman, and G. Taylor, Phys. Plasmas 9, 480 (2002)] spherical tokamak research program has focused on lithium as a large area plasma-facing component (PFC). The energy confinement times showed a sixfold or more improvement over discharges without lithium PFCs. This was an increase of up to a factor of 3 over ITER98P(y,1) scaling [ITER Physics Basis Editors, Nucl. Fusion 39, 2137 (1999)], and reflects the largest enhancement in confinement ever seen in Ohmic plasmas. Recycling coefficients of 0.3 or below were achieved, and they are the lowest to date in magnetically confined plasmas. The effectiveness of liquid lithium in redistributing heat loads at extremely high power densities was demonstrated with an electron beam, which was used to generate lithium coatings. When directed to a lithium reservoir, evaporation occurred only after the entire volume of lithium was raised to the evaporation temperature. T...

Effect of nonaxisymmetric magnetic perturbations on divertor heat and particle flux profiles in National Spherical Torus Experiment a)

Small, nonaxisymmetric magnetic perturbations generated by external coils have been found to break the axisymmetry of heat and particle flux deposition pattern in the divertor area in the National Spherical Torus Experiment (NSTX). This breaking by the applied 3-D field causes strike point splitting that is represented as local peaks and valleys in the divertor profiles. In case of n = 3 fields application, the broken toroidal symmetry of the divertor profile shows 120° of spatial periodicity while data for n = 1 fields provide a fully nonaxisymmetric heat and particle deposition. Field line tracing showed good agreement with the measured heat and particle flux profiles. Higher toroidal mode number (n = 3) of the applied perturbation produced more and finer striations in the divertor profiles than in the lower mode number (n = 1) case. Following the previous result of the intrinsic strike point splitting by the n = 3 error fields [Nucl. Fusion 50, 045010 (2010); J. Nucl. Mater. (2011), doi:10.1016/j.jnucm...

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 rece...

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 limi...