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

TLDR

Fusion nuclear science facilities (FNSFs) are essential for developing fusion materials, and the spherical tokamak is a leading candidate due to its high neutron loading, modular design, and ability to support missions such as tritium self‑sufficiency and electrical self‑sufficiency while meeting divertor, first‑wall, and blanket requirements. ST‑FNSF configurations combine a tritium‑breeding blanket (TBR≈1), high‑elongation poloidal field coils, a long‑legged divertor with outboard coils, and a vertical maintenance scheme, and are evaluated across major radii 1–2.2 m to assess mission feasibility and heating/current drive strategies. The study finds that a major radius of at least 1.7 m is required for TBR = 1, while a 1 m device achieves TBR ≈ 0.9, reducing tritium consumption, and larger devices (A = 2, R = 3 m) can meet tritium and electrical self‑sufficiency with high‑temperature superconductor TF coils.

Abstract

Abstract 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 <?CDATA ${{R}_{0}}\geqslant 1.7$ ?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mstyle displaystyle="false"> <mml:mstyle displaystyle="false"> <mml:msub> <mml:mrow> <mml:mi>R</mml:mi> </mml:mrow> <mml:mrow> <mml:mn>0</mml:mn> </mml:mrow> </mml:msub> </mml:mstyle> <mml:mo>⩾</mml:mo> <mml:mn>1.7</mml:mn> </mml:mstyle> </mml:math> 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 0 = 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.