Liquid Lithium Limiter Experiments in CDX-U

Abstract

Abstract. Recent experiments in the Current Drive eXperiment -Upgrade 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, B toroidal = 2 kG, I P =100 kA, T e (0)~100 eV, n e (0)~ 5 × 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray 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 limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium limited discharges are consistent with Z effective <1.2 (compared to 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.

Recent progress on spherical torus research

“…1 and 2. [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15][16][17][18] Those ST facilities operated before 2000, including the START device, are covered in the previous review/overview papers. 34,35 It also should be noted that most of the ST facilities are located at universities with a strong student education mission.…”

“…The CDX-U was the first fusion device to test the effectiveness of large free surface liquid lithium as a plasma facing component. 6 The CDX-U facility was upgraded significantly in 2009 to become the Lithium Tokamak Experiment (LTX). 7 The projected LTX device and plasma parameters are R 0 ¼ 40 cm, a ¼ 26 cm, B T0 3.4 kG, and I p $ 0.4 MA with > 100 ms current flat top.…”

“…It is often quoted that the fusion energy gain factor Q scales as H [5][6][7] , where H is the confinement improvement factor compared to the standard H-mode. This would then imply that one would need to know the H-factor to within 10% in order to have sufficient confidence of predicting Q within a factor of 2.…”

“…1 and 2), there are a large number of significant ST results which are unique to STs, while at the same time reaffirming many common physics features with conventional tokamaks. [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15][16][17][18] As a member of the tokamak family, 19 ST research contributes to advancing conventional tokamaks such as ITER 20,21 by providing data that extends into a unique plasma and device parameter space. Since ITER represents a significant extrapolation from present day tokamaks, ST research provides a different set of tokamak-related data for the improvement of predictive capabilities by providing leverage over a wide range of parameter space.…”

Particle Control and Plasma Performance in the Lithium Tokamak Experiment (LTX)

“…In March 2012 the LTX OH system was reconfigured to produce longer discharges (30-35 msec, twice the discharge duration on the predecessor to LTX, the Current Drive eXperimentUpgrade, CDX-U), 9 with reduced loop voltage, in preparation for liquid lithium operation. Peak plasma current during 2012 was typically 40 -50 kA for 30 -35 msec discharges.…”