Fusion Nuclear Science Facilities and Pilot Plants Based on
the Spherical Tokamak
Authors: J.E. Menard, T. Brown, L.
El-Guebaly, M. Boyer, J. Canik, B. Colling, R. Raman, Z. Wang, Y.
Zhai, P. Buxton, B. Covele, C. D'Angelo, A. Davis, S. Gerhardt, M.
Gryaznevich, M. Harb, T.C. Hender, S. Kaye, D. Kingham, M.
Kotschenreuther, S. Mahajan, R. Maingi, E. Marriott, E.T. Meier,
L. Mynsberge, C. Neumeyer, M. Ono, J-K. Park, S.A. Sabbagh, V.
Soukhanovskii, P. Valanju, R. Woolley
Abstract: A Fusion Nuclear Science Facility
(FNSF) could play an important role in the development of fusion
energy by providing the nuclear environment needed develop
fusion materials and components. The
spherical torus/tokamak (ST) is a leading
candidate for an FNSF due to
its potentially high neutron wall loading and modular
configuration. A key consideration for the choice of FNSF
configuration is the range of achievable missions as a
function of device size. Possible missions
include: providing high neutron wall loading and fluence,
demonstrating tritium self-sufficiency, and
demonstrating electrical self-sufficiency. All of
these missions must also be compatible with a viable
divertor, first-wall, and blanket solution.
ST-FNSF configurations have been developed simultaneously
incorporating for the first time: (1) a blanket
system capable of tritium breeding ratio
TBR ≈ 1, (2) a poloidal field coil set supporting high
elongation and triangularity for a range of internal
inductance and normalized beta values
consistent with NSTX/NSTX-U previous/planned
operation, (3) a long-legged
divertor analogous to the
MAST-U divertor which substantially
reduces projected peak
divertor heat-flux and has all
outboard poloidal field coils outside the vacuum
chamber and superconducting to reduce power consumption, and
(4) a vertical maintenance scheme in
which blanket structures and the
centerstack can be removed independently. Progress in these
ST-FNSF missions vs. configuration studies including
dependence on plasma major radius
R0 for a range 1m
to 2.2m are described. In particular,
it is found the threshold major
radius for TBR = 1 is R0
≥1.7m, and a smaller R0 =1m ST
device has TBR ≈ 0.9 which is below unity
but substantially reduces T consumption
relative to not breeding.
Calculations of neutral beam heating and current drive for
non-inductive ramp-up and sustainment are described. An
A=2, R0 = 3m device incorporating
high-temperature superconductor toroidal field
coil magnets capable of high neutron
fluence and both tritium and
electrical self-sufficiency is also presented following
systematic aspect ratio studies.
Submitted to: Nuclear Fusion
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