ICONE23-1659 AN INVESTIGATION INTO THE FEASIBILITY OF UTILIZING THORIUM-BASED FUEL CYCLES IN A SMALL FLUORIDE-SALT-COOLED HIGH-TEMPERATURE REACTOR

Abstract

The objective of this study is to investigate the potential use of thorium-based fuels, ThO2-UO2 and ThO2-PuO2 in particular, in a small (125 MWth) fluoride-salt-cooled high-temperature reactor (FHR) with lithium-beryllium fluoride salt (Li2Be4F) coolant. This study investigates thorium as an alternative fuel because it offers several potential benefits including enhanced proliferation resistance, lower waste radiotoxicity, and a higher abundance in nature compared to uranium. The deterministic lattice physics code WIMS was used to find the k∞ values and reactivity (fuel Doppler and coolant temperature) coefficients and to carry out fuel depletion. The results were verified using the Monte Carlo code SERPENT. For the analysis of homogeneous ThO2-UO2 fuels, calculations are performed over a range of thorium volume fractions and uranium enrichments (up to 20 wt.%) that can achieve the same total burnup of 5.07 MWd/kg (with acceptable reactivity coefficients) as a 100% UO2 4.69 wt.% enriched fuel reference case. The maximum thorium volume fraction that can be utilised to obtain same burnup is 0.668 at 20 wt.%. In ThO2-PuO2 analysis, reactor grade Pu with an isotopic vector taken from typical light water reactor (LWR) discharge fuel with initial 4.5 wt.% enrichment, 50 MWd/kg burnup is used. The volumetric proportions of ThO2 and PuO2 are varied and the impact on total burnup, plutonium incineration rates and reactivity coefficients investigated. To achieve the same total burnup, the thorium volume fraction is estimated to be ∼0.86.