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

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.

聚变核科学装置（Fusion Nuclear Science Facility, FNSF）可为聚变材料与组件的研发提供所需的核环境，在聚变能源发展进程中发挥关键作用。球型环面托卡马克（Spherical Torus/Tokamak, ST）因具备潜在的高中子壁负载与模块化构型，成为FNSF的主流候选方案之一。选择FNSF构型的核心考量因素之一，是其可实现的任务随装置尺寸变化的范围。其潜在任务包括：提供高中子壁负载与中子注量、验证氚自持能力，以及验证电能自持能力。所有上述任务均需与可行的偏滤器、第一壁及包层方案兼容。

本次研究首次同步开发了ST-FNSF构型，其整合了四大核心设计：（1）氚增殖比（Tritium Breeding Ratio, TBR）约为1的包层系统；（2）极向场线圈组，可在与国家球形环面实验装置（National Spherical Torus Experiment, NSTX）及其过往/规划运行一致的内部电感与归一化beta值范围内，支持高拉长比与高三角比构型；（3）类似兆安球形托卡马克升级装置（Mega Ampere Spherical Tokamak Upgrade, MAST-U）的长腿型偏滤器，可大幅降低预估的偏滤器峰值热流密度，且所有外侧极向场线圈均布置于真空室外侧并采用超导设计以降低功耗；（4）垂直维护方案，可独立拆卸包层结构与中心堆芯。

本文阐述了ST-FNSF任务随构型的研究进展，包括等离子体大半径R0在1米至2.2米范围内的依赖特性。研究发现，实现TBR=1的临界大半径为R0≥1.7米；而R0=1米的小型ST装置的TBR约为0.9，虽未达到1，但相比不进行氚增殖的方案可大幅减少氚消耗。

本文还介绍了用于非感应启动与维持的中性束加热及电流驱动计算结果。通过系统的纵横比研究，本文还提出了一款A=2、R0=3米的装置方案，其搭载采用高温超导体的环向场线圈磁体，可实现高中子注量并同时满足氚与电能自持需求。