Author/Authors :
Kugel، نويسنده , , H.W. and Bell، نويسنده , , M. and Berzak، نويسنده , , L. and Brooks، نويسنده , , A. and Ellis، نويسنده , , R. and Gerhardt، نويسنده , , S. and Harjes، نويسنده , , H. and Kaita، نويسنده , , R. and Kallman، نويسنده , , J. and Maingi، نويسنده , , R. and Majeski، نويسنده , , R. and Mansfield، نويسنده , , D. and Menard، نويسنده , , J. and Nygren، نويسنده , , R.E. and Soukhanovskii، نويسنده , , V. and Stotler، نويسنده , , D. and Wakeland، نويسنده , , P. and Zakharov، نويسنده , , L.E.، نويسنده ,
Abstract :
Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW ∼ 1), to enable ne scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., ne/nGW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.
