Abstract
Very low aspect ratio (Rç/a ~ 1.2) tokamaks utilizing a demountable, normal conducting center
leg to carry the toroidal field current are estimated to lead to compact (R0 + a ~ 5.4 m) and economically
competitive fusion power plants, given certain key plasma properties. Calculations are carried out
to determine if these properties can be obtained for such low Rçla, using tokamak physics models
recently improved by data from START. High toroidal field utilization (/p//lfc ~ 1-2) and strong plasma
"shaping" S ( s l^q^/aB^ — 100-200 MA-nT1 -T"1) can be obtained using only two pairs of outboard
poloidal field coils. The fraction of the scrape-off layer (SOL) connected to the inboard side of such
plasmas diminishes as RJa is reduced, while the remainder is diverted. Order-unity toroidal average
betas (<j3t> = 2/^0<p>/fi^ - 0.5-1) which are ballooning stable are obtained having a second-regime
core of low magnetic shear and high q0 (5 > q0 > 3), and a first-regime periphery of high magnetic
shear and high edge <fy (22 > q^ > 11). Large self-driven currents (I^f/Ip < 0.7), well aligned with
a hollow plasma current in these equilibria, are calculated for e/3j < 0.9. For marginal MHD stability
of the outboard SOL, it is shown that die minimum pressure e-folding length Xp scales approximately
according to («¡7j/i?)(Z,c/,B)2, /i¡ and 7j being the edge ion density and temperature, respectively, and Lc
being the connected field line length. A large outboard SOL pressure e-folding thickness can therefore
exist for low RQ and B^ configurations of very low RJa (Xp ~ AS0L/3 ~ 10 cm in the case of reactors).
This reduces the divertor heat flux expected in compact fusion systems of high power density. Progress
in these topics will help ensure that very low Rçla tokamaks are effective vehicles for fusion power as
well as blanket testing.
