Nuclear data uncertainty challenges in Molten Salt Reactor safeguards

摘要

This work presented a depletion analysis of an LEU-fueled molten salt reactor based on the MSDR design using the Monte Carlo code Serpent 2 and several sequences within the SCALE package. Multi-group results showed differences up to 1000 pcm in k-eff compared to continuous energy, indicating limitations in the 252-group cross sections or in the lattice cell treatment in the SCALE model. Although initial criticality results between continuous energy KENO and Serpent were in agreement, the methods showed some increasing differences with burnup in k-eff, ~235U, and Pu concentations. The effect of nuclear data uncertainties on plutonium concentration was estimated using Sampler, which showed uncertainties up to 2% in ~239Pu content, resulting to an amount of 29 kg when ap-plied to the core inventory. From a safeguards perspective, an uncertainty of 0.55% for ~239Pu in the MSDR model would be equivalent to 1 significant quantity. This presents a challenge for any nuclear material accounting strategy for MSRs, and alternative metrics should be sought to identify potential diver-sion. Future studies will also investigate the effect of online fuel reprocessing using the new feed and removal capabilities of both SCALE and Serpent.