skip to main content

SciTech ConnectSciTech Connect

This content will become publicly available on August 16, 2017

Title: Fusion nuclear science facilities and pilot plants based on the spherical tokamak

Here, 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 ≈ 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 versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is $${{R}_{0}}\geqslant 1.7$$ m, and a smaller R 0 = 1 m 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, R = 3 m 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.
Authors:
 [1] ;  [1] ;  [2] ;  [1] ;  [3] ;  [4] ;  [5] ;  [1] ;  [1] ;  [6] ;  [7] ;  [2] ;  [2] ;  [1] ;  [6] ;  [2] ;  [4] ;  [1] ;  [6] ;  [7] more »;  [7] ;  [1] ;  [2] ;  [8] ;  [2] ;  [1] ;  [1] ;  [1] ;  [9] ;  [10] ;  [7] ;  [1] « less
  1. Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
  2. Univ. of Wisconsin, Madison, WI (United States)
  3. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  4. Culham Science Centre, Oxfordshire (United Kingdom)
  5. Univ. of Washington, Seattle, WA (United States)
  6. Tokamak Energy Ltd., Oxfordshire (United Kingdom)
  7. Univ. of Texas at Austin, Austin, TX (United States)
  8. College of William and Mary, Williamsburg, VA (United States); Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
  9. Columbia Univ., New York, NY (United States)
  10. Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
Publication Date:
OSTI Identifier:
1335165
Report Number(s):
5280
Journal ID: ISSN 0029-5515
Grant/Contract Number:
EP/I501045; AC02-09CH11466
Type:
Accepted Manuscript
Journal Name:
Nuclear Fusion
Additional Journal Information:
Journal Volume: 56; Journal Issue: 10; Journal ID: ISSN 0029-5515
Publisher:
IOP Science
Research Org:
Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States)
Sponsoring Org:
USDOE Office of Science (SC), Fusion Energy Sciences (FES) (SC-24)
Country of Publication:
United States
Language:
English
Subject:
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