Fusion pilot plants (FPPs) will require tritium self-sufficiency, which is achieved through the breeding blanket. The liquid immersion blanket (LIB) concept employing liquid breeders has been shown to reduce complexity and costs, but the most popular candidate for LIBs, FLiBe, contains highly toxic beryllium. In order to attain tritium self-sufficiency without the drawbacks of high toxicity, lithium-chloride lithium-fluoride (ClLiF) molten salt is suggested as an alternative liquid breeding candidate.

This work analyzes the viability of ClLiF from a neutronics perspective using the OpenMC transport code. Simulations with a simple, ideal blanket neutronics model with no first wall or structural materials were carried out and revealed that ClLiF enriched in 37Cl is competitive with FLiBe in terms of both the tritium breeding ratio (TBR) and energy multiplication . Next, a scan across salt temperatures, neutron multiplier materials, neutron multiplier thicknesses, LiCl fractions, 37Cl enrichments, and 6Li enrichment was conducted to identify the parameters that improve ClLiF performance.

These improved parameters were then applied to a more realistic model of a compact, toroidal reactor with a first wall and structural materials. The results from this model demonstrated that a blanket made up of ClLiF, enriched in 37Cl, achieved a TBR greater than that of FLiBe, but had a reduced energy multiplication unless a thicker external beryllium layer was introduced. Last, the effects of nuclear data and density uncertainties on the TBR and were quantified, and uncertainties in 35Cl nuclear data resulted in the greatest source of uncertainty in the calculation of the TBR and . However, a new evaluation of 35Cl cross sections by Los Alamos National Laboratory with lower uncertainty led to greater TBRs and ME’s than those calculated using the ENDF/B-VIII.0 and TENDL-2019 libraries.