RECIRCULATING FLOW ANALYSES IN A LEAD-COOLED SMALL MODULAR REACTOR

IVAN K. UMEZU; DARIO M. GODINO; DAMIAN E. RAMAJO; CLAUBIA PEREIRA; CLARYSSON A. SILVA; ANTONELLA L. COSTA

Belo Horizonte - Brasil

Congreso; Semana Nacional de Enngenharia Nuclear e da Energia e Ciências das Radiações - VI SENCIR; 2022

Universidade Federal de Minas Gerais e Departamento de Engenheria Nuclear

Computational Fluid Dynamics (CFD) has been used to advance studies on thermal hydraulics of innovative nuclear reactors, especially those cooled by liquid metal. Thus, to demonstrate the capabilities of using CFD codes in the Department of Nuclear Engineering (Departamento de Engenharia Nuclear – DEN) at the Universidade Federal de Minas Gerais (UFMG), this work presents an analysis of the flow patterns in a lead-cooled reactor under steady-state operating conditions using CFD simulations. To carry out the studies, the SEALER (Swedish Advanced Lead Reactor), a Small Modular Reactor (SMR), was chosen mainly due to the available reference data in the literature and for being one of the most relevant liquid-metal cooled reactors recently proposed. For such, the CFD model was based on published data, including geometry and operating conditions. The core region was divided into fuel, control, reflector, and shielding assemblies, modelled as porous regions and porous jumps, using the Darcy-Forchheimer model, according to their respective pressure drop estimations. The axial and radial thermal power distributions were considered and applied as volumetric heat sources in the fuel assemblies. The liquid lead coolant was modeled with temperature-dependent thermophysical properties. The domain was limited to the core barrel region and reduced to a symmetric 1/4 of the whole. Thus, symmetry boundary conditions were applied to side faces and free fluid surfaces were modeled as slip walls. The case was run on ANSYS Fluent 2019 R3. The results, though representative of a steady state condition, were based on transient calculations, due to the natural convection and geometry implications in computation stability. In conclusion, by employing CFD in this study, detailed coolant flow patterns could be analyzed, which would not be possible if traditional thermal-hydraulic system codes were used. The results for coolant temperature and velocity fields were compared to the available publications and were in accordance with them. These preliminary results point to the future expansion of the analysis domain and geometrical detailing and endorse DEN’s technical development in the field of CFD applied to advanced nuclear systems