AUTHOR=Yilmaz Seda , Romano Paul K. , Chierici Lorenzo , Knudsen Erik B. , Shriwise Patrick C. TITLE=CAD and constructive solid geometry modeling of the Molten Salt Reactor Experiment with OpenMC JOURNAL=Frontiers in Nuclear Engineering VOLUME=3 YEAR=2024 URL=https://www.frontiersin.org/journals/nuclear-engineering/articles/10.3389/fnuen.2024.1385478 DOI=10.3389/fnuen.2024.1385478 ISSN=2813-3412 ABSTRACT=
In this study, we present a detailed comparison of two independently developed models of the Molten Salt Reactor Experiment (MSRE) for Monte Carlo particle transport simulations: the constructive solid geometry (CSG) model that was developed in support of the MSRE benchmark in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and a CAD model that was developed by Copenhagen Atomics. The original Serpent reference CSG model was first converted to OpenMC’s input format so that it could be systematically compared to the CAD model, which was already available as an OpenMC model, using the same Monte Carlo code. Results from simulations using the Serpent and OpenMC CSG models showed that