49th Annual Meeting of the Division of Plasma Physics
Volume 52, Number 11
Monday–Friday, November 12–16, 2007;
Orlando, Florida
Session BI1: Pedestal, SOL and Divertor
9:30 AM–12:30 PM,
Monday, November 12, 2007
Rosen Centre Hotel
Room: Junior Ballroom
Chair: James Terry, Massachusetts Institute of Technology
Abstract ID: BAPS.2007.DPP.BI1.5
Abstract: BI1.00005 : Divertor Heat Flux Reduction and Detachment in the National Spherical Torus eXperiment.*
11:30 AM–12:00 PM
Preview Abstract
Abstract
Author:
Vsevolod Soukhanovskii
(Lawrence Livermore National Laboratory)
Steady-state handling of the heat flux is a critical divertor issue for both
the International Thermonuclear Experimental Reactor and spherical torus
(ST) devices. Because of an inherently compact divertor, it was thought that
ST-based devices might not be able to fully utilize radiative and
dissipative divertor techniques based on induced power and momentum loss.
However, initial experiments conducted in the National Spherical Torus
Experiment in an open geometry horizontal carbon plate divertor using 0.8 MA
2-6 MW NBI-heated lower single null H-mode plasmas at the lower end of
elongations $\kappa $=1.8-2.4 and triangularities $\delta $=0.45-0.75
demonstrated that high divertor peak heat fluxes, up to 6-10 MW/ m$^{2}$,
could be reduced by 50-75{\%} using a high-recycling radiative divertor
regime with D$_{2}$ injection. Furthermore, similar reduction was obtained
with a partially detached divertor (PDD) at high D$_{2}$ injection rates,
however, it was accompanied by an X-point MARFE that quickly led to
confinement degradation. Another approach takes advantage of the ST relation
between strong shaping and high performance, and utilizes the poloidal
magnetic flux expansion in the divertor region. Up to 60 {\%} reduction in
divertor peak heat flux was achieved at similar levels of scrape-off layer
power by varying plasma shaping and thereby increasing the outer strike
point (OSP) poloidal flux expansion from 4-6 to 18-22. In recent experiments
conducted in highly-shaped 1.0-1.2 MA 6 MW NBI heated H-mode plasmas with
divertor D$_{2}$ injection at rates up to 10$^{22}$ s$^{-1}$, a PDD regime
with OSP peak heat flux 0.5-1.5 MW/m$^{2}$ was obtained without noticeable
confinement degradation. Calculations based on a two point scrape-off layer
model with parameterized power and momentum losses show that the short
parallel connection length at the OSP sets the upper limit on the radiative
exhaust channel, and both the impurity radiation and large momentum sink
achievable only at high divertor neutral pressures are required for
detachment.
*Supported by US DOE under W-7405-Eng-48.
To cite this abstract, use the following reference: http://meetings.aps.org/link/BAPS.2007.DPP.BI1.5