Fusion Nuclear Science Facilities and Pilot Plants Based on the Spherical Tokamak

Dataset · Sat Oct 01 00:00:00 EDT 2016

DOI:https://doi.org/10.11578/1366722· OSTI ID:1366722

Menard, J.E. ^[1]; Brown, T. ^[2]; El-Guebaly, L. ^[2]; Boyer, M. ^[2]; Canik, J. ^[2]; Colling, B. ^[2]; Raman, R. ^[2]; Wang, Z. ^[2]; Zhai, Y. ^[2]; Buxton, P. ^[2]; Covele, B. ^[2]; D'Angelo, C. ^[2]; Davis, A. ^[2]; Gerhardt, S. ^[2]; Gryaznevich, M. ^[2]; Harb, M. ^[2]; Hender, T.C. ^[2]; Kaye, S. ^[2]; Kingham, D. ^[2]; Kotschenreuther, M. ^[2] ; Mahajan, S. ^[2]; Maingi, R. ^[2]; Marriott, E. ^[2]; Meier, E.T. ^[2]; Mynsberge, L. ^[2]; Neumeyer, C. ^[2]; Ono, M. ^[2]; Park, J.-K. ^[2]; Sabbagh, S.A. ^[2]; Soukhanovskii, V. ^[2]; Valanju, P. ^[2]; Woolley, R. ^[2]

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1. Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Princeton Plasma Physics Laboratory
2. Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

A Fusion Nuclear Science Facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to 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 approximately 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 greater than or equal to 1.7m, and a smaller R0=1m ST device has TBR approximately 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.

• Research Organization: Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States)
• Sponsoring Organization: USDOE Office of Science (SC), Fusion Energy Sciences (FES) (SC-24)
• DOE Contract Number: AC02-09CH11466
• OSTI ID: 1366722
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
Fusion nuclear science facility Pilot plant Spherical tokamak Tritium breeding Negative neutral beams Super-X divertor High-temperature superconductors