Selection of Fuel Isotope Composition in Heating Injectors of the FNS-ST Compact Fusion Neutron Source

Abstract

For the FNS-ST compact neutron source, the dependence of the neutron yield on the tritium content in the bulk plasma is analyzed for the operation of the heating injectors with different isotope compositions of the neutral beams. Self-consistent simulations of the FNS-ST operating regimes are performed using the SOLPS4.3 and ASTRA codes for different densities of the bulk plasma and diffusion coefficients. The FC-FNS code is used to calculate the required fluxes of the fuel components into the plasma provided by different injection systems: the pellet injectors and the neutral beams. In simulations, the plasma density is varied in the range ne = (7–10) × 1019 m−3, and the ratio of the diffusivity to the heat conductivity in the range D/χe = 0.2–0.6. For the scenarios with the D + T or D beams, in the window of the operating parameters, the maximum possible fractions of tritium in the bulk plasma are calculated, and the corresponding neutron yields are obtained. For the regimes with the maximum neutron yield (4.5–5.5) × 1017 s−1, the accumulation of tritium at the site (up to 550 g) is calculated for different heating beams.