To model a Molten Salt Reactor (MSR) core, we apply a Multiphysics coupling scheme between the finite volume Computational Fluid Dynamics (CFD) code OpenFOAM and the Monte Carlo based neutronics code Serpent. The scheme employs the Serpent Multiphysics interface, which allows for high fidelity coupling to OpenFOAM by directly passing variable fields between the two codes. We simulate a graphite-moderated channel type MSR and compare the simulation results to data available on the Molten Salt Reactor Experiment (MSRE). Specifically, fuel and graphite temperature profiles and fuel velocity fields are derived for steady state operation and compared to the results of model calculations performed at the Oak Ridge National Laboratory (ORNL). A simple transient scenario of a step reactivity insertion is also modeled and the feedback of the system is evaluated and compared to experimental results.
|Title of host publication||Proceedings of the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics|
|Publisher||American Nuclear Society|
|Publication status||Published - 2019|
|Event||18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics - Portland, United States|
Duration: 18 Aug 2019 → 23 Aug 2019
Conference number: 18
|Conference||18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics|
|Period||18/08/2019 → 23/08/2019|
Nalbandyan, A., Klinkby, E. B., Lauritzen, B., Groth-Jensen, J., & Steyn, R. (2019). Coupled Neutronics/Thermal Hydraulics Assessment of Graphite Moderated Molten Salt Reactors. In Proceedings of the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (pp. 5342-5355). American Nuclear Society.